Plastic solidification of radioactive wastes

International Nuclear Information System (INIS)

Moriyama, Noboru

1981-01-01

Over 20 years have elapsed after the start of nuclear power development, and the nuclear power generation in Japan now exceeds the level of 10,000 MW. In order to meet the energy demands, the problem of the treatment and disposal of radioactive wastes produced in nuclear power stations must be solved. The purpose of the plastic solidification of such wastes is to immobilize the contained radionuclides, same as other solidification methods, to provide the first barrier against their move into the environment. The following matters are described: the nuclear power generation in Japan, the radioactive wastes from LWR plants, the position of plastic solidification, the status of plastic solidification in overseas countries and in Japan, the solidification process for radioactive wastes with polyethylene, and the properties of solidified products, and the leachability of radionuclides in asphalt solids. (J.P.N.)

Plastic solidification method for radioactive waste

International Nuclear Information System (INIS)

Tomita, Toshihide; Inakuma, Masahiko.

1992-01-01

Condensed liquid wastes in radioactive wastes are formed by mixing and condensing several kinds of liquid wastes such as liquid wastes upon regeneration of ion exchange resins, floor draining liquid wastes and equipment draining liquid wastes. Accordingly, various materials are contained, and it is found that polymerization reaction of plastics is inhibited especially when reductive material, such as sodium nitrite is present. Then, in the present invention, upon mixing thermosetting resins to radioactive wastes containing reducing materials, an alkaline material is admixed to an unstaturated polyester resin. This can inactivate the terminal groups of unsaturated polyester chain, to prevent the dissociation of the reducing agent such as sodium nitrite. Further, if an unsaturated polyester resin of low acid value and a polymerization initiator for high temperature are used in addition to the alkaline material, the effect is further enhanced, thereby enabling to obtain a strong plastic solidification products. (T.M.)

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

International Nuclear Information System (INIS)

Solbrig, Charles W.; Bateman, Kenneth J.

2010-01-01

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

Solidification of ion exchange resin wastes

International Nuclear Information System (INIS)

1982-08-01

Solidification media investigated included portland type I, portland type III and high alumina cements, a proprietary gypsum-based polymer modified cement, and a vinyl ester-styrene thermosetting plastic. Samples formulated with hydraulic cement were analyzed to investigate the effects of resin type, resin loading, waste-to-cement ratio, and water-to-cement ratio. The solidification of cation resin wastes with portland cement was characterized by excessive swelling and cracking of waste forms, both after curing and during immersion testing. Mixed bed resin waste formulations were limited by their cation component. Additives to improve the mechanical properties of portland cement-ion exchange resin waste forms were evaluated. High alumina cement formulations dislayed a resistance to deterioration of mechanical integrity during immersion testing, thus providing a significant advantage over portland cements for the solidification of resin wastes. Properties of cement-ion exchange resin waste forms were examined. An experiment was conducted to study the leachability of 137 Cs, 85 Sr, and 60 Co from resins modified in portland type III and high alumina cements. The cumulative 137 Cs fraction release was at least an order of magnitude greater than that of either 85 Sr or 60 Co. Release rates of 137 Cs in high alumina cement were greater than those in portland III cement by a factor of two.Compressive strength and leach testing were conducted for resin wastes solidified with polymer-modified gypsum based cement. 137 Cs, 85 Sr, and 60 Co fraction releases were about one, two and three orders of magnitude higher, respectively, than in equivalent portland type III cement formulations. As much as 28.6 wt % dry ion exchange resin was successfully solidified using vinyl ester-styrene compared with a maximum of 25 wt % in both portland and gypsum-based cement

Radioactive waste solidification material

International Nuclear Information System (INIS)

Nishihara, Yukio; Wakuta, Kuniharu; Ishizaki, Kanjiro; Koyanagi, Naoaki; Sakamoto, Hiroyuki; Uchida, Ikuo.

1992-01-01

The present invention concerns a radioactive waste solidification material containing vermiculite cement used for a vacuum packing type waste processing device, which contains no residue of calcium hydroxide in cement solidification products. No residue of calcium hydroxide means, for example, that peak of Ca(OH) 2 is not recognized in an X ray diffraction device. With such procedures, since calcium sulfoaluminate clinker and Portland cement themselves exhibit water hardening property, and slugs exhibit hydration activity from the early stage, the cement exhibits quick-hardening property, has great extension of long term strength, further, has no shrinking property, less dry- shrinkage, excellent durability, less causing damages such as cracks and peeling as processing products of radioactive wastes, enabling to attain highly safe solidification product. (T.M.)

Treatment methods for radioactive mixed wastes in commercial low-level wastes: technical considerations

International Nuclear Information System (INIS)

MacKenzie, D.R.; Kempf, C.R.

1986-01-01

Treatment options for the management of three generic categories of radioactive mixed waste in commercial low-level wastes (LLW) have been identified and evaluated. These wastes were characterized as part of a BNL study in which LLW generators were surveyed for information on potential chemical hazards in their wastes. The general treatment options available for mixed wastes are destruction, immobilization, and reclamation. Solidification, absorption, incineration, acid digestion, wet-air oxidation, distillation, liquid-liquid wastes. Containment, segregation, decontamination, and solidification or containment of residues, have been considered for lead metal wastes which have themselves been contaminated and are not used for purposes of waste disposal shielding, packaging, or containment. For chromium-containing wastes, solidification, incineration, wet-air oxidation, acid digestion, and containment have been considered. For each of these wastes, the management option evaluation has included an assessment of testing appropriate to determine the effect of the option on both the radiological and potential chemical hazards present

Treatment methods for radioactive mixed wastes in commercial low-level wastes: technical considerations

Energy Technology Data Exchange (ETDEWEB)

MacKenzie, D.R.; Kempf, C.R.

1986-01-01

Treatment options for the management of three generic categories of radioactive mixed waste in commercial low-level wastes (LLW) have been identified and evaluated. These wastes were characterized as part of a BNL study in which LLW generators were surveyed for information on potential chemical hazards in their wastes. The general treatment options available for mixed wastes are destruction, immobilization, and reclamation. Solidification, absorption, incineration, acid digestion, wet-air oxidation, distillation, liquid-liquid wastes. Containment, segregation, decontamination, and solidification or containment of residues, have been considered for lead metal wastes which have themselves been contaminated and are not used for purposes of waste disposal shielding, packaging, or containment. For chromium-containing wastes, solidification, incineration, wet-air oxidation, acid digestion, and containment have been considered. For each of these wastes, the management option evaluation has included an assessment of testing appropriate to determine the effect of the option on both the radiological and potential chemical hazards present.

Characteristics of Cement Solidification of Metal Hydroxide Waste

Directory of Open Access Journals (Sweden)

Dae-Seo Koo

2017-02-01

Full Text Available To perform the permanent disposal of metal hydroxide waste from electro-kinetic decontamination, it is necessary to secure the technology for its solidification. The integrity tests on the fabricated solidification should also meet the criteria of the Korea Radioactive Waste Agency. We carried out the solidification of metal hydroxide waste using cement solidification. The integrity tests such as the compressive strength, immersion, leach, and irradiation tests on the fabricated cement solidifications were performed. It was also confirmed that these requirements of the criteria of Korea Radioactive Waste Agency on these cement solidifications were met. The microstructures of all the cement solidifications were analyzed and discussed.

Characteristics of cement solidification of metal hydroxide waste

Energy Technology Data Exchange (ETDEWEB)

Koo, Dae Seo; Sung, Hyun Hee; Kim, Seung Soo; Kim, Gye Nam; Choi, Jong Won [Dept. of Decontemination Decommission Technology Development, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2017-02-15

To perform the permanent disposal of metal hydroxide waste from electro-kinetic decontamination, it is necessary to secure the technology for its solidification. The integrity tests on the fabricated solidification should also meet the criteria of the Korea Radioactive Waste Agency. We carried out the solidification of metal hydroxide waste using cement solidification. The integrity tests such as the compressive strength, immersion, leach, and irradiation tests on the fabricated cement solidifications were performed. It was also confirmed that these requirements of the criteria of Korea Radioactive Waste Agency on these cement solidifications were met. The microstructures of all the cement solidifications were analyzed and discussed.

Microwave solidification development for Rocky Flats waste

Energy Technology Data Exchange (ETDEWEB)

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

1994-04-01

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

Microwave solidification development for Rocky Flats waste

International Nuclear Information System (INIS)

Dixon, D.; Erle, R.; Eschen, V.

1994-04-01

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

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

International Nuclear Information System (INIS)

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

1994-03-01

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

''New ' technology of solidification of liquid radioactive waste'

International Nuclear Information System (INIS)

Sytyl, V.A.; Svistova, L.M.; Spiridonova, V.P.

1998-01-01

It is generally accepted that the best method of processing of radioactive waste is its solidification and then storage. At present time, three methods of solidification of radioactive waste are widely used in the world: cementation, bituminous grouting and vitrification. But they do not solve the problem of ecologically processing of waste because of different disadvantages. General disadvantages are: low state of filling, difficulties in solidification of the crystalline hydrated forms of radioactive waste; particular sphere of application and economical difficulties while processing the great volume of waste. In connection with it the urgent necessity is emerging: to develop less expensive and ecologically more reliable technology of solidification of radioactive waste. A new method of solidification is presented with its technical schema. (N.C.)

Treatment methods for radioactive mixed wastes in commercial low-level wastes - technical considerations

International Nuclear Information System (INIS)

MacKenzie, D.R.; Kempf, C.R.

1986-01-01

Treatment options for the management of three generic categories of radioactive mixed waste in commercial low-level wastes (LLW) have been identified and evaluated. These wastes were characterized as part of a BNL study in which LLW generators were surveyed for information on potential chemical hazards in their wastes. The general treatment options available for mixed wastes are destruction, immobilization, and reclamation. Solidification, absorption, incineration, acid digestion, wet-air oxidation, distillation, liquid-liquid solvent extraction, and specific chemical destruction techniques have been considered for organic liquid wastes. Containment, segregation, decontamination, and solidification or containment of residues, have been considered for lead metal wastes which have themselves been contaminated and are not used for purposes of waste disposal shielding, packaging, or containment. For chromium-containing wastes, solidification, incineration, wet-air oxidation, acid digestion, and containment have been considered. Fore each of these wastes, the management option evaluation has included an assessment of testing appropriate to determine the effect of the option on both the radiological and potential chemical hazards present

Treatability study of absorbent polymer waste form for mixed waste treatment

International Nuclear Information System (INIS)

Herrmann, S. D.; Lehto, M. A.; Stewart, N. A.; Croft, A. D.; Kern, P. W.

2000-01-01

A treatability study was performed to develop and characterize an absorbent polymer waste form for application to low level (LLW) and mixed low level (MLLW) aqueous wastes at Argonne National Laboratory-West (ANL-W). In this study absorbent polymers proved effective at immobilizing aqueous liquid wastes in order to meet Land Disposal Restrictions for subsurface waste disposal. Treatment of aqueous waste with absorbent polymers provides an alternative to liquid waste solidification via high-shear mixing with clays and cements. Significant advantages of absorbent polymer use over clays and cements include ease of operations and waste volume minimization. Absorbent polymers do not require high-shear mixing as do clays and cements. Granulated absorbent polymer is poured into aqueous solutions and forms a gel which passes the paint filter test as a non-liquid. Pouring versus mixing of a solidification agent not only eliminates the need for a mixing station, but also lessens exposure to personnel and the potential for spread of contamination from treatment of radioactive wastes. Waste minimization is achieved as significantly less mass addition and volume increase is required of and results from absorbent polymer use than that of clays and cements. Operational ease and waste minimization translate into overall cost savings for LLW and MLLW treatment

Solidification of low-level waste - a dilemma for the small user

International Nuclear Information System (INIS)

Harris, S.; Gilmore, A.

1980-01-01

The requirement that radioactive waste for sea disposal must be solidified by the originator is discussed. Attempts to solidify small quantities of radioactive waste such as contaminated oils and labelled benzyopyrene with other solvents are described. Encapsulation media tested were concrete and interior and exterior grade Polyfilla (a plaster and cellulose based filler). Problems were presented by the difficulty of mixing the materials and by the maximum uptake of solvents while still allowing solidification. In all cases a soft crumbling material resulted. It is concluded that solidification processing on a small scale does not make economic or scientific sense and that if solidification is necessary it would be better carried out as a national operation by collecting liquids from users. (U.K.)

Sulfur polymer stabilization/solidification (SPSS) treatment of mixed waste mercury recovered from environmental restoration activities at BNL

International Nuclear Information System (INIS)

Kalb, P.; Adams, J.; Milian, L.

2001-01-01

Over 1,140 yd 3 of radioactively contaminated soil containing toxic mercury (Hg) and several liters of mixed-waste elemental mercury were generated during a Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) removal action at Brookhaven National Laboratory (BNL). The US Department of Energy's (DOE) Office of Science and Technology Mixed Waste Focus Area (DOE MWFA) is sponsoring a comparison of several technologies that may be used to treat these wastes and similar wastes at BNL and other sites across the DOE complex. This report describes work conducted at BNL on the application and pilot-scale demonstration of the newly developed Sulfur Polymer Stabilization/Solidification (SPSS) process for treatment of contaminated mixed-waste soils containing high concentrations (approximately 5,000 mg/L) of mercury and liquid elemental mercury. BNL's SPSS (patent pending) process chemically stabilizes the mercury to reduce vapor pressure and leachability and physically encapsulates the waste in a solid matrix to eliminate dispersion and provide long-term durability. Two 55-gallon drums of mixed-waste soil containing high concentrations of mercury and about 62 kg of radioactive contaminated elemental mercury were successfully treated. Waste loadings of 60 wt% soil were achieved without resulting in any increase in waste volume, while elemental mercury was solidified at a waste loading of 33 wt% mercury. Toxicity Characteristic Leaching Procedure (TCLP) analyses indicate the final waste form products pass current Environmental Protection Agency (EPA) allowable TCLP concentrations as well as the more stringent proposed Universal Treatment Standards. Mass balance measurements show that 99.7% of the mercury treated was successfully retained within the waste form, while only 0.3% was captured in the off gas system

Nuclear waste solidification

Science.gov (United States)

Bjorklund, William J.

1977-01-01

High level liquid waste solidification is achieved on a continuous basis by atomizing the liquid waste and introducing the atomized liquid waste into a reaction chamber including a fluidized, heated inert bed to effect calcination of the atomized waste and removal of the calcined waste by overflow removal and by attrition and elutriation from the reaction chamber, and feeding additional inert bed particles to the fluidized bed to maintain the inert bed composition.

Nuclear waste solidification

International Nuclear Information System (INIS)

Bjorklund, W.J.

1977-01-01

High level liquid waste solidification is achieved on a continuous basis by atomizing the liquid waste and introducing the atomized liquid waste into a reaction chamber including a fluidized, heated inert bed to effect calcination of the atomized waste and removal of the calcined waste by overflow removal and by attrition and elutriation from the reaction chamber, and feeding additional inert bed particles to the fluidized bed to maintain the inert bed composition

Glass-solidification method for high level radioactive liquid waste

International Nuclear Information System (INIS)

Kawamura, Kazuhiro; Kometani, Masayuki; Sasage, Ken-ichi.

1996-01-01

High level liquid wastes are removed with precipitates mainly comprising Mo and Zr, thereafter, the high level liquid wastes are mixed with a glass raw material comprising a composition having a B 2 O 3 /SiO 2 ratio of not less than 0.41, a ZnO/Li 2 O ratio of not less than 1.00, and an Al 2 O 3 /Li 2 O ratio of not less than 2.58, and they are melted and solidified into glass-solidification products. The liquid waste content in the glass-solidification products can be increased up to about 45% by using the glass raw material having such a predetermined composition. In addition, deposition of a yellow phase does not occur, and a leaching rate identical with that in a conventional case can be maintained. (T.M.)

Treatment of radioactive mixed wastes in commercial low-level wastes

International Nuclear Information System (INIS)

Kempf, C.R.; MacKenzie, D.R.

1985-01-01

Management options for three generic categories of radioactive mixed waste in commercial low-level wastes have been identified and evaluated. These wastes were characterized as part of a BNL study in which a large number of generators were surveyed for information on potentially hazardous low-level wastes. The general management targets adopted for mixed wastes are immobilization, destruction, and reclamation. It is possible that these targets may not be practical for some wastes, and for these, goals of stabilization or reduction of hazard are addressed. Solidification, absorption, incineration, acid digestion, segregation, and substitution have been considered for organic liquid wastes. Containment, segregation, and decontamination and re-use have been considered for lead metal wastes which have themselves been contaminated and are not used for purposes of waste disposal shielding, packaging, or containment. For chromium-containing wastes, solidification, incineration, containment, substitution, chemical reduction, and biological removal have been considered. For each of these wastes, the management option evaluation has necessarily included assessment/estimation of the effect of the treatment on both the radiological and potential chemical hazards present. 10 refs

Sodalite-type radioactive waste solidification product and method of synthesizing the same

International Nuclear Information System (INIS)

Koyama, Masashi; Yoshida, Takumasa.

1995-01-01

Radioactive waste solidification products formed by solidifying radioactive wastes comprising halides such as chlorides of alkali metal elements, alkaline earth metal elements, rare earth elements, noble metal elements generated upon dry-type reprocessing of nuclear fuels and separation of dry-type high level liquid wastes, are solidified to stable products by incorporating radioactive wastes in the form of halides into a cavity of sodalite condensation cage of aluminosilicates having three dimensional skeleton structure. Alternatively, NaOH, Al 2 O 3 , SiO 2 are mixed and heated to 600 to 900degC to form an intermediate reaction products, and then the reaction products are mixed with the halides and heated to form sodalite-type radioactive water solidification products. Thus, the halides in fission products can be held by the three dimensional skeleton structure similar with that of sodalite which is a sort of natural minerals containing chlorides, thereby enabling to solidify them stably. (N.H.)

Sulfur polymer stabilization/solidification (SPSS) treatment of mixed waste mercury recovered from environmental restoration activities at BNL

Energy Technology Data Exchange (ETDEWEB)

Kalb, P.; Adams, J.; Milian, L.

2001-01-29

Over 1,140 yd{sup 3} of radioactively contaminated soil containing toxic mercury (Hg) and several liters of mixed-waste elemental mercury were generated during a Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) removal action at Brookhaven National Laboratory (BNL). The US Department of Energy's (DOE) Office of Science and Technology Mixed Waste Focus Area (DOE MWFA) is sponsoring a comparison of several technologies that may be used to treat these wastes and similar wastes at BNL and other sites across the DOE complex. This report describes work conducted at BNL on the application and pilot-scale demonstration of the newly developed Sulfur Polymer Stabilization/Solidification (SPSS) process for treatment of contaminated mixed-waste soils containing high concentrations ({approximately} 5,000 mg/L) of mercury and liquid elemental mercury. BNL's SPSS (patent pending) process chemically stabilizes the mercury to reduce vapor pressure and leachability and physically encapsulates the waste in a solid matrix to eliminate dispersion and provide long-term durability. Two 55-gallon drums of mixed-waste soil containing high concentrations of mercury and about 62 kg of radioactive contaminated elemental mercury were successfully treated. Waste loadings of 60 wt% soil were achieved without resulting in any increase in waste volume, while elemental mercury was solidified at a waste loading of 33 wt% mercury. Toxicity Characteristic Leaching Procedure (TCLP) analyses indicate the final waste form products pass current Environmental Protection Agency (EPA) allowable TCLP concentrations as well as the more stringent proposed Universal Treatment Standards. Mass balance measurements show that 99.7% of the mercury treated was successfully retained within the waste form, while only 0.3% was captured in the off gas system.

Solidification of liquid concentrate and solid waste generated as by-products of the liquid radwaste treatment systems in light-water reactors

International Nuclear Information System (INIS)

Neilson, R.M. Jr.; Colombo, P.

1977-01-01

The treatment of liquid concentrate and solid waste produced in light-water reactors as by-products of liquid radwaste treatment systems consists of five basic operations: waste collection, waste pretreatment, solidification agent handling, mixing/packaging (solidification) and waste package handling. This paper will concern itself primarily with the solidification operation, however, the other operations enumerated as well as the types of wastes treated and their origins will be briefly described, especially with regards to their effects on solidification. During solidification, liquid concentrate and solid wastes are incorporated with a solidification agent to form a monolithic, free-standing solid. The basic solidification agent types either currently used in the United States or proposed for use include absorbants, hydraulic cement, urea-formaldehyde, other polymer systems, and bitumen. The operation, formulations and limitations of these agents as used for radwaste solidification will be discussed. Properties relevant to the evaluation of solidified waste forms will be identified and relative comparisons made for wastes solidified by various processes

Polyethylene solidification of low-level wastes

International Nuclear Information System (INIS)

Kalb, P.D.; Colombo, P.

1985-02-01

This topical report describes the results of an investigation on the solidification of low-level radioactive waste in polyethylene. Waste streams selected for this study included those which result from advanced volume reduction technologies (dry evaporator concentrate salts and incinerator ash) and those which remain problematic for solidification using contemporary agents (ion exchange resins). Four types of commercially available low-density polyethylenes were employed which encompass a range of processing and property characteristics. Process development studies were conducted to ascertain optimal process control parameters for successful solidification. Maximum waste loadings were determined for each waste and polyethylene type. Property evaluation testing was performed on laboratory-scale specimens to assess the potential behavior of actual waste forms in a disposal environment. Waste form property tests included water immersion, deformation under compressive load, thermal cycling and radionuclide leaching. Recommended waste loadings of 70 wt % sodium sulfate, 50 wt % boric acid, 40 wt % incinerator ash, and 30 wt % ion exchange resins, which are based on process control and waste form performance considerations are reported. 37 refs., 33 figs., 22 tabs

Solidification as low cost technology prior to land filling of industrial hazardous waste sludge.

Science.gov (United States)

El-Sebaie, O; Ahmed, M; Ramadan, M

2000-01-01

The aim of this study is to stabilize and solidify two different treated industrial hazardous waste sludges, which were selected from factories situated close to Alexandria. They were selected to ensure their safe transportation and landfill disposal by reducing their potential leaching of hazardous elements, which represent significant threat to the environment, especially the quality of underground water. The selected waste sludges have been characterized. Ordinary Portland Cement (OPC), Cement Kiln Dust (CKD) from Alexandria Portland Cement Company, and Calcium Sulphate as a by-product from the dye industry were used as potential solidification additives to treat the selected treated waste sludges from tanning and dyes industry. Waste sludges as well as the solidified wastes have been leach-tested, using the General Acid Neutralization Capacity (GANC) procedure. Concentration of concerning metals in the leachates was determined to assess changes in the mobility of major contaminants. The treated tannery waste sludge has an acid neutralization capacity much higher than that of the treated dyes waste sludge. Experiment results demonstrated the industrial waste sludge solidification mix designs, and presented the reduction of contaminant leaching from two types of waste sludges. The main advantages of solidification are that it is simple and low cost processing which includes readily available low cost solidification additives that will convert industrial hazardous waste sludges into inert materials.

Low level waste solidification practice in Japan

International Nuclear Information System (INIS)

Sakata, S.; Kuribayashi, H.; Kono, Y.

1981-01-01

Both sea dumping and land isolation are planned to be accomplished for low level waste disposal in Japan. The conceptual design of land isolation facilities has been completed, and site selection will presently get underway. With respect to ocean dumping, safety surveys are being performed along the lines of the London Dumping Convention and the Revised Definitions and Recommendations of the IAEA, and the review of Japanese regulations and applicable criteria is being expedited. This paper discusses the present approach to waste solidification practices in Japan. It reports that the bitumen solidification process and the plastic solidification process are being increasingly used in Japan. Despite higher investment costs, both processes have advantages in operating cost, and are comparable to the cement solidification process in overall costs

Methodology to remediate a mixed waste site

International Nuclear Information System (INIS)

Berry, J.B.

1994-08-01

In response to the need for a comprehensive and consistent approach to the complex issue of mixed waste management, a generalized methodology for remediation of a mixed waste site has been developed. The methodology is based on requirements set forth in the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) and the Resource Conservation and Recovery Act (RCRA) and incorporates ''lessons learned'' from process design, remediation methodologies, and remediation projects. The methodology is applied to the treatment of 32,000 drums of mixed waste sludge at the Oak Ridge K-25 Site. Process technology options are developed and evaluated, first with regard to meeting system requirements and then with regard to CERCLA performance criteria. The following process technology options are investigated: (1) no action, (2) separation of hazardous and radioactive species, (3) dewatering, (4) drying, and (5) solidification/stabilization. The first two options were eliminated from detailed consideration because they did not meet the system requirements. A quantitative evaluation clearly showed that, based on system constraints and project objectives, either dewatering or drying the mixed waste sludge was superior to the solidification/stabilization process option. The ultimate choice between the drying and the dewatering options will be made on the basis of a technical evaluation of the relative merits of proposals submitted by potential subcontractors

Methodology to remediate a mixed waste site

Energy Technology Data Exchange (ETDEWEB)

Berry, J.B.

1994-08-01

In response to the need for a comprehensive and consistent approach to the complex issue of mixed waste management, a generalized methodology for remediation of a mixed waste site has been developed. The methodology is based on requirements set forth in the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) and the Resource Conservation and Recovery Act (RCRA) and incorporates ``lessons learned`` from process design, remediation methodologies, and remediation projects. The methodology is applied to the treatment of 32,000 drums of mixed waste sludge at the Oak Ridge K-25 Site. Process technology options are developed and evaluated, first with regard to meeting system requirements and then with regard to CERCLA performance criteria. The following process technology options are investigated: (1) no action, (2) separation of hazardous and radioactive species, (3) dewatering, (4) drying, and (5) solidification/stabilization. The first two options were eliminated from detailed consideration because they did not meet the system requirements. A quantitative evaluation clearly showed that, based on system constraints and project objectives, either dewatering or drying the mixed waste sludge was superior to the solidification/stabilization process option. The ultimate choice between the drying and the dewatering options will be made on the basis of a technical evaluation of the relative merits of proposals submitted by potential subcontractors.

Low-level radwaste solidification

International Nuclear Information System (INIS)

Naughton, M.D.; Miller, C.C.; Nelson, R.A.; Tucker, R.F.

1983-01-01

This paper reports on a study of ''Advanced Low-Level Radioactive Waste Treatment Systems'' conducted under an EPRI contract. The object of the study is to identify advanced lowlevel radwaste treatment systems that are commercially available or are expected to be in the near future. The current state-ofthe-art in radwaste solidification technology is presented. Related processing technologies, such as the compaction of dry active waste (DAW), containers available for radwaste disposal, and the regulatory aspects of radwaste transportation and solidification, are described. The chemical and physical properties of the currently acceptable solidification agents, as identified in the Barnwell radwaste burial site license, are examined. The solidification agents investigated are hydraulic cements, thermoplastic polymers, and thermosetting polymers. It is concluded that solidification processes are complex and depend not only on the chemical and physical properties of the binder material and the waste, but also on how these materials are mixed

Toxic and hazardous waste disposal. Volume 1. Processes for stabilization/solidification

International Nuclear Information System (INIS)

Pojasek, R.B.

1979-01-01

Processes for the stabilization and/or solidification of toxic, hazardous, and radioactive wastes are reviewed. The types of wastes classified as hazardous are defined. The following processes for the solidification of hazardous wastes are described: lime-based techniques; thermoplastic techniques; organic polymer techniques; and encapsulation. The following processes for the solidification of high-level radioactive wastes are described: calcination; glassification; and ceramics. The solidification of low-level radioactive wastes with asphalt, cement, and polymeric materials is also discussed. Other topics covered include: the use of an extruder/evaporator to stabilize and solidify hazardous wastes; effect disposal of fine coal refuse and flue gas desulfurization slurries using Calcilox additive stabilization; the Terra-Tite Process; the Petrifix Process; the SFT Terra-Crete Process; Sealosafe Process; Chemfix Process; and options for disposal of sulfur oxide wastes

Solidification of low-level wastes by inorganic binder

International Nuclear Information System (INIS)

Sasaki, M.T.; Shimojo, M.; Suzuki, K.; Kajikawa, A.; Karasawa, Y.

1995-01-01

The use of an alkali activated slag binder has been studied for solidification and stabilization of low-level wastes in nuclear power stations and spent fuel processing facilities. The activated slag effectively formed waste products having good physical properties with high waste loading for sodium sulfate, sodium nitrate, calcium pyrophosphate/phosphate and spent ion-exchange resins. Moreover, the results of the study suggest the slag has the ability to become a common inorganic binder for the solidification of various radioactive wastes. This paper also describes the fixation of radionuclides by the activated slag binder

Microwave solidification project overview

Energy Technology Data Exchange (ETDEWEB)

Sprenger, G.

1993-01-01

The Rocky Flats Plant Microwave Solidification Project has application potential to the Mixed Waste Treatment Project and the The Mixed Waste Integrated Program. The technical areas being addressed include (1) waste destruction and stabilization; (2) final waste form; and (3) front-end waste handling and feed preparation. This document covers need for such a program; technology description; significance; regulatory requirements; and accomplishments to date. A list of significant reports published under this project is included.

Microwave solidification project overview

International Nuclear Information System (INIS)

Sprenger, G.

1993-01-01

The Rocky Flats Plant Microwave Solidification Project has application potential to the Mixed Waste Treatment Project and the The Mixed Waste Integrated Program. The technical areas being addressed include (1) waste destruction and stabilization; (2) final waste form; and (3) front-end waste handling and feed preparation. This document covers need for such a program; technology description; significance; regulatory requirements; and accomplishments to date. A list of significant reports published under this project is included

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

Solidification of radioactive waste in a cement/lime mixture

International Nuclear Information System (INIS)

Zhou, H.; Colombo, P.

1984-01-01

The suitability of a cement/lime mixture for use as a solidification agent for different types of wastes was investigated. This work includes studies directed towards determining the wasted/binder compositional field over which successful solidification occurs with various wastes and the measurement of some of the waste from properties relevant to evaluating the potential for the release of radionuclides to the environment. In this study, four types of low-level radioactive wastes were simulated for incorporation into a cement/lime mixture. These were boric acid waste, sodium sulfate wastes, aion exchange resins and incinerator ash. 7 references, 3 figures, 2 tables

SULFUR POLYMER STABILIZATION/SOLIDIFICATION (SPSS) TREATABILITY OF SIMULATED MIXED-WASTE MERCURY CONTAMINATED SLUDGE.

Energy Technology Data Exchange (ETDEWEB)

ADAMA, J.W.; BOWERMAN, B.S.; KALB, P.D.

2002-10-01

The Environmental Protection Agency (EPA) is currently seeking to validate technologies that can directly treat radioactively contaminated high mercury (Hg) subcategory wastes without removing the mercury from the waste. The Sulfur Polymer Stabilization/Solidification (SPSS) process developed at Brookhaven National Laboratory is one of several candidate technologies capable of successfully treating various Hg waste streams. To supplement previously supplied data on treatment of soils, EPA needs additional data concerning stabilization of high Hg subcategory waste sludges. To this end, a 5000 ppm sludge surrogate, containing approximately 50 wt% water, was successfully treated by pilot-scale SPSS processing. In two process runs, 85 and 95 wt% of water was recovered from the sludge during processing. At waste loadings of 30 wt% dry sludge, the treated waste form had no detectable mercury (<10 ppb) in TCLP leachates. Data gathered from the demonstration of treatment of this sludge will provide EPA with information to support revisions to current treatment requirements for high Hg subcategory wastes.

SULFUR POLYMER STABILIZATION/SOLIDIFICATION (SPSS) TREATABILITY OF SIMULATED MIXED-WASTE MERCURY CONTAMINATED SLUDGE

Energy Technology Data Exchange (ETDEWEB)

Adams, J. W.; Bowerman, B. S.; Kalb, P. D.

2002-02-25

The Environmental Protection Agency (EPA) is currently evaluating alternative treatment standards for radioactively contaminated high mercury (Hg) subcategory wastes, which do not require the removal of mercury from the waste. The Sulfur Polymer Stabilization/Solidification (SPSS) process developed at Brookhaven National Laboratory is one of several candidate technologies capable of successfully treating various Hg waste streams. To supplement previously supplied data on treatment of soils, EPA needed additional data concerning stabilization of high Hg subcategory waste sludges. To this end, a 5000 ppm sludge surrogate, containing approximately 50 wt% water, was successfully treated by pilot-scale SPSS processing. In two process runs, 85 and 95 wt% of water was recovered from the sludge during processing. At waste loadings of 46 wt% (30 wt% dry) sludge, the treated waste form had no detectable mercury (<10 ppb) in TCLP leachates. Data gathered from the demonstration of treatment of this sludge will provide the EPA with information to support revisions to current treatment requirements for high Hg subcategory wastes.

SULFUR POLYMER STABILIZATION/SOLIDIFICATION (SPSS) TREATABILITY OF SIMULATED MIXED-WASTE MERCURY CONTAMINATED SLUDGE

International Nuclear Information System (INIS)

ADAMA, J.W.; BOWERMAN, B.S.; KALB, P.D.

2002-01-01

The Environmental Protection Agency (EPA) is currently seeking to validate technologies that can directly treat radioactively contaminated high mercury (Hg) subcategory wastes without removing the mercury from the waste. The Sulfur Polymer Stabilization/Solidification (SPSS) process developed at Brookhaven National Laboratory is one of several candidate technologies capable of successfully treating various Hg waste streams. To supplement previously supplied data on treatment of soils, EPA needs additional data concerning stabilization of high Hg subcategory waste sludges. To this end, a 5000 ppm sludge surrogate, containing approximately 50 wt% water, was successfully treated by pilot-scale SPSS processing. In two process runs, 85 and 95 wt% of water was recovered from the sludge during processing. At waste loadings of 30 wt% dry sludge, the treated waste form had no detectable mercury (<10 ppb) in TCLP leachates. Data gathered from the demonstration of treatment of this sludge will provide EPA with information to support revisions to current treatment requirements for high Hg subcategory wastes

SULFUR POLYMER STABILIZATION/SOLIDIFICATION (SPSS) TREATABILITY OF SIMULATED MIXED-WASTE MERCURY CONTAMINATED SLUDGE

International Nuclear Information System (INIS)

Adams, J. W.; Bowerman, B. S.; Kalb, P. D.

2002-01-01

The Environmental Protection Agency (EPA) is currently evaluating alternative treatment standards for radioactively contaminated high mercury (Hg) subcategory wastes, which do not require the removal of mercury from the waste. The Sulfur Polymer Stabilization/Solidification (SPSS) process developed at Brookhaven National Laboratory is one of several candidate technologies capable of successfully treating various Hg waste streams. To supplement previously supplied data on treatment of soils, EPA needed additional data concerning stabilization of high Hg subcategory waste sludges. To this end, a 5000 ppm sludge surrogate, containing approximately 50 wt% water, was successfully treated by pilot-scale SPSS processing. In two process runs, 85 and 95 wt% of water was recovered from the sludge during processing. At waste loadings of 46 wt% (30 wt% dry) sludge, the treated waste form had no detectable mercury (<10 ppb) in TCLP leachates. Data gathered from the demonstration of treatment of this sludge will provide the EPA with information to support revisions to current treatment requirements for high Hg subcategory wastes

Solidification process for toxic and hazardous wastes. Second part: Cement solidification matrices

International Nuclear Information System (INIS)

Donato, A.; Arcuri, L.; Dotti, M.; Pace, A.; Pietrelli, L.; Ricci, G.; Basta, M.; Cali, V.; Pagliai, V.

1989-05-01

This paper reports the second part of a general study carried out at the Nuclear Fuel Division aiming at verifying the possible application of the radioactive waste solidification processes to industrial hazardous wastes (RTN). The cement solidification of several RTN types has been taken into consideration, both from the technical and from the economic point of view. After a short examination of the Italian juridical and economical situation in the field, which demonstrates the need of the RTN solidification, the origin and characteristics of the RTN considered in the study and directly provided by the producing industries are reviewed. The laboratory experimental results of the cementation of RTN produced by gold manufacturing industries and by galvanic industries are reported. The cementation process can be considered a very effective mean for reducing both the RTN management costs and the environmental impact of RTN disposal. (author)

Solidification of ash from incineration of low-level radioactive waste

International Nuclear Information System (INIS)

Roberson, W.A.; Albenesius, E.L.; Becker, G.W.

1983-01-01

The safe disposal of both high-level and low-level radioactive waste is a problem of increasing national attention. A full-scale incineration and solidification process to dispose of suspect-level and low-level beta-gamma contaminated combustible waste is being demonstrated at the Savannah River Plant (SRP) and Savannah River Laboratory (SRL). The stabilized wasteform generated by the process will meet or exceed all future anticipated requirements for improved disposal of low-level waste. The incineration process has been evaluated at SRL using nonradioactive wastes, and is presently being started up in SRP to process suspect-level radioactive wastes. A cement solidification process for incineration products is currently being evaluated by SRL, and will be included with the incineration process in SRP during the winter of 1984. The GEM alumnus author conducted research in a related disposal solidification program during the GEM-sponsored summer internship, and upon completion of the Masters program, received full-time responsibility for developing the incineration products solidification process

Plastic solidification system for radioactive waste

International Nuclear Information System (INIS)

Kani, Jiro; Irie, Hiromitsu; Obu, Etsuji; Nakayama, Yasuyuki; Matsuura, Hiroyuki.

1979-01-01

The establishment of a new solidification system is an important theme for recent radioactive-waste disposal systems. The conditions required of new systems are: (1) the volume of the solidified product to be reduced, and (2) the property of the solidified product to be superior to the conventional ones. In the plastic solidification system developed by Toshiba, the waste is first dried and then solidified with thermosetting resin. It has been confirmed that the property of the plastic solidified product is superior to that of the cement-or bitumen-solidified product. Investigation from various phases is being carried on for the application of this method to commercial plants. (author)

Centralized cement solidification technique for low-level radioactive wastes

International Nuclear Information System (INIS)

Matsuda, Masami; Nishi, Takashi; Izumida, Tatsuo; Tsuchiya, Hiroyuki.

1996-01-01

A centralized cement solidification system has been developed to enable a single facility to solidify such low-level radioactive wastes as liquid waste, spent ion exchange resin, incineration ash, and miscellaneous solid wastes. Since the system uses newly developed high-performance cement, waste loading is raised and deterioration of waste forms after land burial prevented. This paper describes the centralized cement solidification system and the features of the high-performance cement. Results of full-scale pilot plant tests are also shown from the viewpoint of industrial applicability. (author)

Test procedures for polyester immobilized salt-containing surrogate mixed wastes

International Nuclear Information System (INIS)

Biyani, R.K.; Hendrickson, D.W.

1997-01-01

These test procedures are written to meet the procedural needs of the Test Plan for immobilization of salt containing surrogate mixed waste using polymer resins, HNF-SD-RE-TP-026 and to ensure adequacy of conduct and collection of samples and data. This testing will demonstrate the use of four different polyester vinyl ester resins in the solidification of surrogate liquid and dry wastes, similar to some mixed wastes generated by DOE operations

Stabilization of mixed waste at the Idaho National Engineering Laboratory

International Nuclear Information System (INIS)

Boehmer, A.M.; Gillins, R.L.; Larsen, M.M.

1989-01-01

EG and G Idaho, Inc. has initiated a program to develop safe, efficient, cost-effective treatment methods for the stabilization of some of the hazardous and mixed wastes generated at the Idaho National Engineering Laboratory. Laboratory-scale testing has shown that extraction procedure toxic wastes can be successfully stabilized by solidification, using various binders to produce nontoxic, stable waste forms for safe, long-term disposal as either landfill waste or low-level radioactive waste, depending upon the radioactivity content. This paper presents the results of drum-scale solidification testing conducted on hazardous, low-level incinerator flyash generated at the Waste Experimental Reduction Facility. The drum-scale test program was conducted to verify that laboratory-scale results could be successfully adapted into a production operation

Ordinary Portland Cement matrix for solidification of cellulosic protective clothes hazardous wastes

International Nuclear Information System (INIS)

Shatta, H.A.; Saleh, H.M.

2006-01-01

The used cellulosic protective clothes constitutes considerable fraction of the hazardous and radioactive wastes accumulated during the practical daily life. The direct solidification of these wastes with ordinary Portland cement resulted in waste forms having undesired characters, therefore, it is recommended to immobilize the secondary waste solutions coming from the oxidative degradation of the used protective clothes waste simulates rather than direct imbedding. IR analyses, X-ray diffraction and thermal characteristics for products of both direct encapsulation of the waste and the cementation of its degradation products were performed to evaluate the properties of the final waste cemented form before their disposal. Based on the results reached from X-ray diffraction, IR spectrograms and thermal analyses reports, it could be stated that no detectable changes in hydration and curing coarse of ordinary Portland cement when mixing the residual secondary waste solution resulting from the oxidative degradation of the used protective clothes waste simulate compared with mixing cement with water and in reverse with imbedding the unprocessed waste in cement matrix

Solidification of high-level radioactive wastes. Final report

International Nuclear Information System (INIS)

1979-06-01

A panel on waste solidification was formed at the request of the Nuclear Regulatory Commission to study the scientific and technological problems associated with the conversion of liquid and semiliquid high-level radioactive wastes into a stable form suitable for transportation and disposition. Conclusions reached and recommendations made are as follows. Many solid forms described in this report could meet standards as stringent as those currently applied to the handling, storage, and transportation of spent fuel assemblies. Solid waste forms should be selected only in the context of the total radioactive waste management system. Many solid forms are likely to be satisfactory for use in an appropriately designed system, The current United States policy of deferring the reprocessing of commercial reactor fuel provides additional time for R and D solidification technology for this class of wastes. Defense wastes which are relatively low in radioactivity and thermal power density can best be solidified by low-temperature processes. For solidification of fresh commercial wastes that are high in specific activity and thermal power density, the Panel recommends that, in addition to glass, the use of fully-crystalline ceramics and metal-matrix forms be actively considered. Preliminary analysis of the characteristics of spent fuel pins indicates that they may be eligible for consideration as a waste form. Because the differences in potential health hazards to the public resulting from the use of various solid form and disposal options are likely to be small, the Panel concludes that cost, reliability, and health hazards to operating personnel will be major considerations in choosing among the options that can meet safety requiremens. The Panel recommends that responsibility for all radioactive waste management operations (including solidification R and D) should be centralized

Current high-level waste solidification technology

International Nuclear Information System (INIS)

Bonner, W.F.; Ross, W.A.

1976-01-01

Technology has been developed in the U.S. and abroad for solidification of high-level waste from nuclear power production. Several processes have been demonstrated with actual radioactive waste and are now being prepared for use in the commercial nuclear industry. Conversion of the waste to a glass form is favored because of its high degree of nondispersibility and safety

Solidification of intermediate level liquid waste - ILLW, CEMEX waste form qualification

International Nuclear Information System (INIS)

D'Andrea, V.; Guerra, M.; Pancotti, F.; Maio, V.

2015-01-01

In the Sogin EUREX Facility about 125 m 3 of intermediate level radioactive waste and about 113 m 3 of low level radioactive waste, produced during the re-processing of MTR and CANDU fuel, are stored. Solidification of these wastes is planned in order to fulfill the specific requirements established by the Safety Authority, taking into account the criteria set up in a Technical Guide on the issue of radioactive waste management. The design of a cementation plant (CEMEX) of all liquid radioactive wastes is currently ongoing. The process requires that the liquid waste is neutralized with NaOH (NaOH 19 M) and metered into 440 liter drum together with the cement, while the mixture is stirred by a lost paddle ('in drum mixing process'). The qualification of the Waste Form consists of all the activities demonstrating that the final cemented product has the minimum requirements (mechanical, chemical and physical characteristics) compliant with all the subsequent management phases: long-term interim storage, transport and long-term disposal of the waste. All tests performed to qualify the conditioning process for immobilizing first extraction cycle (MTR and CANDU) and second extraction cycle liquid wastes, gave results in compliance with the minimum requirements established for disposal

Development of a geopolymer solidification method for radioactive wastes by compression molding and heat curing

International Nuclear Information System (INIS)

Shimoda, Chiaki; Matsuyama, Kanae; Okabe, Hirofumi; Kaneko, Masaaki; Miyamoto, Shinya

2017-01-01

Geopolymer solidification is a good method for managing waste because of it is inexpensive as compared with vitrification and has a reduced risk of hydrogen generation. In general, when geopolymers are made, water is added to the geopolymer raw materials, and then the slurry is mixed, poured into a mold, and cured. However, it is difficult to control the reaction because, depending on the types of materials, the viscosity can immediately increase after mixing. Slurries of geopolymers easily attach to the agitating wing of the mixer and easily clog the plumbing during transportation. Moreover, during long-term storage of solidified wastes containing concentrated radionuclides in a sealed container without vents, the hydrogen concentration in the container increases over time. Therefore, a simple method using as little water as possible is needed. In this work, geopolymer solidification by compression molding was studied. As compared with the usual methods, it provides a simple and stable method for preparing waste for long-term storage. From investigations performed before and after solidification by compression molding, it was shown that the crystal structure changed. From this result, it was concluded that the geopolymer reaction proceeded during compression molding. This method (1) reduces the energy needed for drying, (2) has good workability, (3) reduces the overall volume, and (4) reduces hydrogen generation. (author)

Some techniques for the solidification of radioactive wastes in concrete

International Nuclear Information System (INIS)

Colombo, P.; Neilson, R. Jr.

1976-06-01

Some techniques for the solidification of radioactive wastes in concrete are discussed. The sources, storage, volume reduction, and solidification of liquid wastes at Brookhaven National Laboratory (BNL) using the cement-vermiculite process is described. Solid waste treatment, shipping containers, and off-site shipments of solid wastes at BNL are also considered. The properties of low-heat-generating, high-level wastes, simulating those in storage at the Savannah River Plant (SRP), solidified in concrete were determined. Polymer impregnation was found to further decrease the leachability and improve the durability of these concrete waste forms

Solidification of ion exchange resin wastes in hydraulic cement

International Nuclear Information System (INIS)

Neilson, R.M. Jr.; Kalb, P.; Fuhrmann, M.; Colombo, P.

1982-01-01

Work has been conducted to investigate the solidification of ion exchange resin wastes with portland cements. These efforts have been directed toward the development of acceptable formulations for the solidification of ion exchange resin wastes and the characterization of the resultant waste forms. This paper describes formulation development work and defines acceptable formulations in terms of ternary phase compositional diagrams. The effects of cement type, resin type, resin loading, waste/cement ratio and water/cement ratio are described. The leachability of unsolidified and solidified resin waste forms and its relationship to full-scale waste form behavior is discussed. Gamma irradiation was found to improve waste form integrity, apparently as a result of increased resin crosslinking. Modifications to improve waste form integrity are described. 3 tables

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

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

INEL studies concerning solidification of low-level waste in cement

International Nuclear Information System (INIS)

Mandler, J.W.

1989-01-01

The Idaho National Engineering Laboratory (INEL) has performed numerous studies addressing issues concerning the solidification of low-level radioactive waste in cement. These studies have been performed for both the Nuclear Regulatory Commission (NRC) and the Department of Energy (DOE). This short presentation will only outline the major topics addressed in some of these studies, present a few conclusions, and identify some of the technical concerns we have. More details of the work and pertinent results will be given in the Working Group sessions. The topics that have been addressed at the INEL which are relevant to this Workshop include (1) solidification of ion-exchange resins and evaporator waste in cement at commercial nuclear power plants, (2) leachability and compressive strength of power plant waste solidified in cement, (3) suggested guidelines for preparation of a solid waste process control program (PCP), (4) cement solidification of EPICOR-II resin wastes, and (5) performance testing of cement-solidified EPICOR-II resin wastes

Low-level waste cement solidification design, installation, and start-up

International Nuclear Information System (INIS)

Jezek, G.R.

1988-08-01

This report describes the design, installation, and start-up activities of the Cement Solidification System (CSS) at the West Valley Demonstration Project (WVDP), West Valley, New York. The CSS, designed to operate within an existing process cell, automatically and remotely solidifies low-level nuclear waste by mixing it with Portland Type I cement. The qualified waste form mixture is placed into square, 270-litre (71-gallon) metal drums. The drums have an integral polyethylene liner to protect the carbon-steel material from potential corrosion. The CSS produces drums at a continuous operation rate of four drums per hour. All system processing data is monitored by a computerized Data Acquisition System (DAS). 6 figs

Improved cement solidification of low and intermediate level radioactive wastes

International Nuclear Information System (INIS)

1993-01-01

Cementation was the first and is still the most widely applied technique for the conditioning of low and intermediate level radioactive wastes. Compared with other solidification techniques, cementation is relatively simple and inexpensive. However, the quality of the final cemented waste forms depends very much on the composition of the waste and the type of cement used. Different kinds of cement are used for different kinds of waste and the compatibility of a specific waste with a specific cement type should always be carefully evaluated. Cementation technology is continuously being developed in order to improve the characteristics of cemented waste in accordance with the increasing requirements for quality of the final solidified waste. Various kinds of additives and chemicals are used to improve the cemented waste forms in order to meet all safety requirements. This report is meant mainly for engineers and designers, to provide an explanation of the chemistry of cementation systems and to facilitate the choice of solidification agents and processing equipment. It reviews recent developments in cementation technology for improving the quality of cemented waste forms and provides a brief description of the various cement solidification processes in use. Refs, figs and tabs

Processing and solidification of Savannah River Plant high-level waste

International Nuclear Information System (INIS)

Kelley, J.A.

1981-01-01

The entire flowsheet for processing and solidification of Savannah River Plant (SRP) high-level wastes has been demonstrated. A new small-scale integrated pilot plant is operating with actual radioactive wastes, and large-scale equipment is being demonstrated with nonradioactive simulated wastes. Design of a full-scale waste solidification plant is in progress. Plant construction is expected to begin in 1983, and startup is anticipated in 1988. The plant will poduce about 500 cans of glass per year with each can containing about 1.5 tons of glass

Techniques for the solidification of high-level wastes

International Nuclear Information System (INIS)

1977-01-01

The problem of the long-term management of the high-level wastes from the reprocessing of irradiated nuclear fuel is receiving world-wide attention. While the majority of the waste solutions from the reprocessing of commercial fuels are currently being stored in stainless-steel tanks, increasing effort is being devoted to developing technology for the conversion of these wastes into solids. A number of full-scale solidification facilities are expected to come into operation in the next decade. The object of this report is to survey and compare all the work currently in progress on the techniques available for the solidification of high-level wastes. It will examine the high-level liquid wastes arising from the various processes currently under development or in operation, the advantages and disadvantages of each process for different types and quantities of waste solutions, the stages of development, the scale-up potential and flexibility of the processes

Polymer solidification national program

International Nuclear Information System (INIS)

Kalb, P.D.; Colombo, P.

1993-04-01

Brookhaven National Laboratory (BNL) has developed several new and innovative polymer processes for the solidification of low-level radioactive, hazardous and mixed wastes streams. Polyethylene and modified sulfur cement solidification technologies have undergone steady, gradual development at BNL over the past nine years. During this time they have progressed through each of the stages necessary for logical technology maturation: from process conception, parameter optimization, waste form testing, evaluation of long-term durability, economic analysis, and scale-up feasibility. This technology development represents a significant investment which can potentially provide DOE with both short- and long-term savings

Development of stable solidification methods for toxic lead oxide in radioactive wastes

International Nuclear Information System (INIS)

Hitoshi Mimura; Shingo Ikeda; Yuichi Niibori

2009-01-01

The objective of this study is to develop the advanced solidification methods for the toxic lead oxide contained in radioactive wastes and to examine their chemical durability in terms of leachability and surface alteration; the solidification characteristics and leachability for the following three kinds of solidified products immobilizing lead were examined, and the experimental results were summarized as follows. (a) Mineral solidified products: A-zeolite or fly ash (FA) was used as a binder, and NaAlO 2 and Na 2 SiO 3 were mixed as additives. The leachability of lead ions in pure water was considerably lowered by the heat treatment at higher temperature (1,000 degree C), and the concentration of lead ions leached was under criterion value of 0.3 mg/l. The products prepared by mixing A-zeolite and fly ash also had low leachability under 0.3 mg/l even in the saturated Ca(OH)2 solution. (b) Melted solidified products: A-zeolite or fly ash was used as a binder and glass-forming reagents of B 2 O 3 and NaH 2 PO 4 were used as additives. The XRD peaks assigned PbO were not observed in all products. The products for the mixtures of FA:NaH 2 PO 4 :PbO (2:2:1 and 3:1:1) had low leachability under criterion value in both leachants of deionized water and saturated Ca(OH) 2 solution. (c) Phosphate ceramics products: the chemically bonded phosphate ceramics were produced by using MgKPO 4 , MgHPO 4 , Zr(HPO 4 ) 2 , potassium iron phosphate and sodium iron phosphate, and FA was used as additives. In particular, by using MgHPO 4 , the leachability of the products was lowered less than 0.3 mg/l in both leachants. The phosphate ceramics products and melted solidified products are favorable as the waste solid forms immobilizing lead. In particular, novel ceramics products have advantages in the simple solidification procedure similarly to the cement products. As for mineral solidification, natural zeolites and FA as binder also useful from the viewpoint of cost efficiency

Evaluation of process alternatives for solidification of the West Valley high-level liquid wastes

International Nuclear Information System (INIS)

Holton, L.K.; Larson, D.E.

1982-01-01

The Department of Energy (DOE) established the West Valley Solidification Project (WVSP) in 1980. The project purpose is to demonstrate removal and solidification of the high-level liquid wastes (HLLW) presently stored in tanks at the Western New York Nuclear Service Center (WNYNSC), West Valley, New York. As part of this effort, the Pacific Northwest Laboratory (PNL) conducted a study to evaluate process alternatives for solidifcation of the WNYNSC wastes. Two process approaches for waste handling before solidification, together with solidification processes for four terminal and four interim waste forms, were considered. The first waste-handling approach, designated the salt/sludge separation process, involves separating the bulk of the nonradioactive nuclear waste constituents from the radioactive waste constituents, and the second waste-handling approach, designated the combined-waste process, involves no waste segregation prior to solidification. The processes were evaluated on the bases of their (1) readiness for plant startup by 1987, (2) relative technical merits, and (3) process cost. The study has shown that, based on these criteria, the salt/sludge separation process with a borosilicate glass waste form is preferred when producing a terminal waste form. It was also concluded that if an interim waste form is to be used, the preferred approach would be the combined waste process with a fused-salt waste form

Solidification/stabilization of fly and bottom ash from medical waste incineration facility.

Science.gov (United States)

Anastasiadou, Kalliopi; Christopoulos, Konstantinos; Mousios, Epameinontas; Gidarakos, Evangelos

2012-03-15

In the present work, the stabilization/solidification of fly and bottom ash generated from incinerated hospital waste was studied. The objectives of the solidification/stabilization treatment were therefore to reduce the leachability of the heavy metals present in these materials so as to permit their disposal in a sanitary landfill requiring only a lower degree of environmental protection. Another objective of the applied treatment was to increase the mechanical characteristics of the bottom ash using different amounts of Ordinary Portland Cement (OPC) as a binder. The solidified matrix showed that the cement is able to immobilize the heavy metals found in fly and bottom ash. The TCLP leachates of the untreated fly ash contain high concentrations of Zn (13.2 mg/l) and Pb (5.21 mg/l), and lesser amounts of Cr, Fe, Ni, Cu, Cd and Ba. Cement-based solidification exhibited a compressive strength of 0.55-16.12 MPa. The strength decreased as the percentage of cement loading was reduced; the compressive strength was 2.52-12.7 MPa for 60% cement mixed with 40% fly ash and 6.62-16.12 MPa for a mixture of 60% cement and 40% bottom ash. The compressive strength reduced to 0.55-1.30 MPa when 30% cement was mixed with 70% fly ash, and to 0.90-7.95 MPa when 30% cement was mixed with 70% bottom ash, respectively. Copyright © 2011 Elsevier B.V. All rights reserved.

Development of high-level waste solidification technology 1

Energy Technology Data Exchange (ETDEWEB)

Kim, Joon Hyung; Kim, Hwan Young; Kim, In Tae [and others

1999-02-01

Spent nuclear fuel contains useful nuclides as valuable resource materials for energy, heat and catalyst. High-level wastes (HLW) are expected to be generated from the R and D activities and reuse processes. It is necessary to develop vitrification or advanced solidification technologies for the safe long-term management of high level wastes. As a first step to establish HLW vitrification technology, characterization of HLWs that would arise at KAERI site, glass melting experiments with a lab-scale high frequency induction melter, and fabrication and property evaluation of base-glass made of used HEPA filter media and additives were performed. Basic study on the fabrication and characterization of candidate ceramic waste form (Synroc) was also carried out. These HLW solidification technologies would be directly useful for carrying out the R and Ds on the nuclear fuel cycle and waste management. (author). 70 refs., 29 tabs., 35 figs.

The solidification of aluminum production waste in geopolymer matrix

Czech Academy of Sciences Publication Activity Database

Perná, Ivana; Hanzlíček, Tomáš

2014-01-01

Roč. 84, DEC 1 (2014), s. 657-662 ISSN 0959-6526 Institutional support: RVO:67985891 Keywords : aluminum waste * solidification * recycling * geopolymer Subject RIV: DM - Solid Waste and Recycling Impact factor: 3.844, year: 2014

Study of plastic solidification process on solid radioactive waste treatment

International Nuclear Information System (INIS)

Jing Weiguan; Zhang Yinsheng; Qian Wenju

1994-01-01

Comparisons between the plastic solidification conditions of incinerated ash and waste cation resin by using thermosetting plastic polyvinyl chloride (PVC), polystyrene (PS) and polyethylene (PE), and identified physico-chemical properties and irradiation resistance of solidified products were presented. These solidified products have passed through different tests as compression strength, leachability, durability, stability, permeability and irradiation resistance (10 6 Gy) etc. The result showed that the solidified products possessed stable properties and met the storage requirement. The waste tube of radioimmunoassay, being used as solidification medium to contain incinerated ash, had good mechanical properties and satisfactory volume reduction. This process may develop a new way for disposal solid radioactive waste by means of re-using waste

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

Utilization of coal fly ash in solidification of liquid radioactive waste from research reactor.

Science.gov (United States)

Osmanlioglu, Ahmet Erdal

2014-05-01

In this study, the potential utilization of fly ash was investigated as an additive in solidification process of radioactive waste sludge from research reactor. Coal formations include various percentages of natural radioactive elements; therefore, coal fly ash includes various levels of radioactivity. For this reason, fly ashes have to be evaluated for potential environmental implications in case of further usage in any construction material. But for use in solidification of radioactive sludge, the radiological effects of fly ash are in the range of radioactive waste management limits. The results show that fly ash has a strong fixing capacity for radioactive isotopes. Specimens with addition of 5-15% fly ash to concrete was observed to be sufficient to achieve the target compressive strength of 20 MPa required for near-surface disposal. An optimum mixture comprising 15% fly ash, 35% cement, and 50% radioactive waste sludge could provide the solidification required for long-term storage and disposal. The codisposal of radioactive fly ash with radioactive sludge by solidification decreases the usage of cement in solidification process. By this method, radioactive fly ash can become a valuable additive instead of industrial waste. This study supports the utilization of fly ash in industry and the solidification of radioactive waste in the nuclear industry.

Performance of cement solidification with barium for high activity liquid waste including sulphate

International Nuclear Information System (INIS)

Waki, Toshikazu; Yamada, Motoyuki; Horikawa, Yoshihiko; Kaneko, Masaaki; Saso, Michitaka; Haruguchi, Yoshiko; Yamashita, Yu; Sakai, Hitoshi

2009-01-01

The target liquid waste to be solidified is generated from PWR primary loop spent resin treatment with sulphate acid, so, its main constituent is sodium sulphate and the activity of this liquid is relatively high. Waste form of this liquid waste is considered to be a candidate for the subsurface disposal. The disposed waste including sulphate is anticipated to rise a concentration of sulphate ion in the ground water around the disposal facility and it may cause degradation of materials such as cement and bentonite layer and comprise the disposal facility. There could be two approaches to avoid this problem, the strong design of the disposal facility and the minimization of sulphaste ion migration from the solidified waste. In this study, the latter approach was examined. In order to keep the low concentration of sulphate ion in the ground water, it is effective to make barium sulphate by adding barium compound into the liquid waste in solidification. However, adding equivalent amount of barium compound with sulphate ion causes difficulty of mixing, because production of barium sulphate causes high viscosity. In this study, mixing condition after and before adding cement into the liquid waste was estimated. The mixing condition was set with consideration to keep anion concentration low in the ground water and of mixing easily enough in practical operation. Long term leaching behavior of the simulated solidified waste was also analyzed by PHREEQC. And the concentration of the constitution affected to the disposal facility was estimated be low enough in the ground water. (author)

Solidification of low-level radioactive wastes in masonry cement

International Nuclear Information System (INIS)

Zhou, H.; Colombo, P.

1987-03-01

Portland cements are widely used as solidification agents for low-level radioactive wastes. However, it is known that boric acid wastes, as generated at pressurized water reactors (PWR's) are difficult to solidify using ordinary portland cements. Waste containing as little as 5 wt % boric acid inhibits the curing of the cement. For this purpose, the suitability of masonry cement was investigated. Masonry cement, in the US consists of 50 wt % slaked lime (CaOH 2 ) and 50 wt % of portland type I cement. Addition of boric acid in molar concentrations equal to or less than the molar concentration of the alkali in the cement eliminates any inhibiting effects. Accordingly, 15 wt % boric acid can be satisfactorily incorporated into masonry cement. The suitability of masonry cement for the solidification of sodium sulfate wastes produced at boiling water reactors (BWR's) was also investigated. It was observed that although sodium sulfate - masonry cement waste forms containing as much as 40 wt % Na 2 SO 4 can be prepared, waste forms with more than 7 wt % sodium sulfate undergo catastrophic failure when exposed to an aqueous environment. It was determined by x-ray diffraction that in the presence of water, the sulfate reacts with hydrated calcium aluminate to form calcium aluminum sulfate hydrate (ettringite). This reaction involves a volume increase resulting in failure of the waste form. Formulation data were identified to maximize volumetric efficiency for the solidification of boric acid and sodium sulfate wastes. Measurement of some of the waste form properties relevant to evaluating the potential for the release of radionuclides to the environment included leachability, compression strengths and chemical interactions between the waste components and masonry cement. 15 refs., 19 figs., 9 tabs

Stabilization/Solidification Remediation Method for Contaminated Soil: A Review

Science.gov (United States)

Tajudin, S. A. A.; Azmi, M. A. M.; Nabila, A. T. A.

2016-07-01

Stabilization/Solidification (S/S) is typically a process that involves a mixing of waste with binders to reduce the volume of contaminant leachability by means of physical and chemical characteristics to convert waste in the environment that goes to landfill or others possibly channels. Stabilization is attempts to reduce the solubility or chemical reactivity of the waste by changing the physical and chemical properties. While, solidification attempt to convert the waste into easily handled solids with low hazardous level. These two processes are often discussed together since they have a similar purpose of improvement than containment of potential pollutants in treated wastes. The primary objective of this review is to investigate the materials used as a binder in Stabilization/Solidification (S/S) method as well as the ability of these binders to remediate the contaminated soils especially by heavy metals.

Interim solidification of SRP waste with silica, bentonite, or phosphoric acid

International Nuclear Information System (INIS)

Thompson, G.H.

1976-03-01

One option for interim waste management at the Savannah River Plant is in-tank solidification of the liquid waste solutions. This would reduce the mobility of these highly radioactive solutions until techniques for their long-term immobilization and storage are developed and implemented. Interim treatments must permit eventual retrieval of waste and subsequent incorporation into a high-integrity form. This study demonstrated the solidification of simulated alkaline waste solutions by reaction with silica, bentonite, and phosphoric acid. Alkaline waste can be solidified by reaction with silica gel, silica flour, or sodium silicate solution. Solidified products containing waste salt can be retrieved by slurrying with water. Alkaline supernate (solution in equilibrium with alkaline sludge in SRP waste tanks) can be solidified by reaction with bentonite to form cancrinite powder. The solidified waste can be retrieved by slurrying with water. Alkaline supernate can be solidified by partial evaporation and reaction with phosphoric acid. Water is incorporated into hydrated complexes of trisodium phosphate. The product is soluble, but actual plant waste would not solidify completely because of decay heat. Reaction of simulated alkaline waste solutions with silica gel, silica flour, or bentonite increases the volume by a factor of approximately 6 over that of evaporated waste; reaction with phosphoric acid results in a volume 1.5 times that of evaporated waste. At present, the best method for in-tank solidification is by evaporation, a method that contributes no additional solids to the waste and does not compromise any waste management options

NPP radioactive waste processing and solidification

International Nuclear Information System (INIS)

Nikiforov, A.S.; Polyakov, A.S.; Zakharova, K.P.

1983-01-01

The problems of proce-sing NPP intermediate level- and low-level liquid radioactive wastes (LRW) are considered. Various methods are compared of LWR solidification on the base of bituminization, cement grouting and inclusion into synthetic resins. It is concluded that the considered methods ensure radioactive radionuclides effluents into open hydronetwork at the level below the sanitary, standards

Results from five years of treatability studies using hydraulic binders to stabilize low-level mixed waste at the INEL

International Nuclear Information System (INIS)

Gering, K.L.; Schwendiman, G.L.

1997-01-01

This paper summarizes work involving bench-scale solidification of nonincinerable, land disposal restricted low-level mixed waste. Waste forms included liquids, sludges, and solids; treatment techniques included hydraulic systems (Portland cement with and without additives), proprietary commercial formulations, and sulphur polymer cement. Solidification was performed to immobilize hazardous heavy metals (including mercury, lead, chromium, and cadmium), and volatile and semivolatile organic compounds. Pretreatment options for mixed wastes are discussed, using a decision tree based on the form of mixed waste and the type of hazardous constituents. Hundreds of small concrete monoliths were formed for a variety of waste types. The experimental parameters used for the hydraulic concrete systems include the ratio of waste to dry binder (Portland cement, proprietary materials, etc.), the total percentage of water in concrete, and the amount of concrete additives. The only parameter that was used for the sulfur polymer-based monoliths is ratio of waste to binder. Optimum concrete formulations or open-quotes recipesclose quotes for a given type of waste were derived through this study, as based on results from the Toxicity Characteristic Leaching Procedure analyses and a free liquids test. Overall results indicate that high waste loadings in the concrete can be achieved while the monolithic mass maintains excellent resistance to leaching of heavy metals. In our study the waste loadings in the concrete generally fell within the range of 0.5 to 2.0 kg mixed waste per kg dry binder. Likewise, the most favorable amount of water in concrete, which is highly dependent upon the concrete constituents, was determined to be generally within the range of 300 to 330 g/kg (30-33% by weight). The results of this bench-scale study will find applicability at facilities where mixed or hazardous waste solidification is a planned or ongoing activity. 19 refs., 1 fig., 5 tabs

Using cement, lignite fly ash and baghouse filter waste for solidification of chromium electroplating treatment sludge

Directory of Open Access Journals (Sweden)

Wantawin, C.

2004-02-01

Full Text Available The objective of the study is to use baghouse filter waste as a binder mixed with cement and lignite fly ash to solidify sludge from chromium electroplating wastewater treatment. To save cost of solidification, reducing cement in binder and increasing sludge in the cube were focused on. Minimum percent cement in binder of 20 for solidification of chromium sludge was found when controlling lignite fly ash to baghouse filter waste at the ratio of 30:70, sludge to binder ratio of 0.5, water to mixer ratio of 0.3 and curing time of 7 days. Increase of sludge to binder ratio from 0.5 to 0.75 and 1 resulted in increase in the minimum percent cement in binder up to 30 percent in both ratios. With the minimum percent cement in binder, the calculated cement to sludge ratios for samples with sludge to binder ratios of 0.5, 0.75 and 1 were 0.4, 0.4 and 0.3 respectively. Leaching chromium and compressive strength of the samples with these ratios could achieve the solidified waste standard by the Ministry of Industry. For solidification of chromium sludge at sludge to binder ratio of 1, the lowest cost binder ratio of cement to lignite fly ash and baghouse filter waste in this study was 30:21:49. The cost of binder in this ratio was 718 baht per ton dry sludge.

Alternative method of solidification for low-level class a radioactive waste

International Nuclear Information System (INIS)

Mayo, K.S.

1988-01-01

New solidification media have been developed that exhibit excellent spatial efficiency over the entire range of virtually all Class A liquid wastes. These new media are being used to incorporate from 41 to 48 gallons of liquid radioactive waste in a 55-gallon drum. To date, wastes processing at nuclear power plants and facilities include oils, evaporator bottoms, sludges, and ion-exchanges resins as well as combinations of these waste streams. This paper comparatively discusses the performance of solidification agents known as AQUASET TM and PETROSET TM with other currently available agents. It presents key advantages of using the AQUASET and PETROSET media over other media. These advantages include improvements in packaging efficiency, leachability, and repeatability

Handling 78,000 drums of mixed-waste sludge

International Nuclear Information System (INIS)

Berry, J.B.; Harrington, E.S.; Mattus, A.J.

1991-01-01

The Oak Ridge Gaseous Diffusion Plant (now known as the Oak Ridge K-25 Site) closed two mixed-waste surface impoundments by removing the sludge and contaminated pond-bottom clay and attempting to process it into durable, nonleachable, concrete monoliths. Interim, controlled, above-ground storage included delisting the stabilized sludge from hazardous to nonhazardous and disposing of the delisted monoliths as Class 1 radioactive waste. Because of schedule constraints and process design and control deficiencies, ∼46,000 drums of material in various stages of solidification and ∼32,000 barrels of unprocessed sludge are stored. The abandoned treatment facility still contains ∼16,000 gal of raw sludge. Such storage of mixed waste does not comply with the Resource Conservation and Recovery Act (RCRA) guidelines. This paper describes actions that are under way to bring the storage of ∼78,000 drums of mixed waste into compliance with RCRA. Remediation of this problem by treatment to meet regulatory requirements is the focus of the discussion. 3 refs., 2 figs., 4 tabs

A perspective of hazardous waste and mixed waste treatment technology at the Savannah River Site

International Nuclear Information System (INIS)

England, J.L.; Venkatesh, S.; Bailey, L.L.; Langton, C.A.; Hay, M.S.; Stevens, C.B.; Carroll, S.J.

1991-01-01

Treatment technologies for the preparation and treatment of heavy metal mixed wastes, contaminated soils, and mixed mercury wastes are being considered at the Savannah River Site (SRS), a DOE nuclear material processing facility operated by Westinghouse Savannah River Company (WSRC). The proposed treatment technologies to be included at the Hazardous Waste/Mixed Waste Treatment Building at SRS are based on the regulatory requirements, projected waste volumes, existing technology, cost effectiveness, and project schedule. Waste sorting and size reduction are the initial step in the treatment process. After sorting/size reduction the wastes would go to the next applicable treatment module. For solid heavy metal mixed wastes the proposed treatment is macroencapsulation using a thermoplastic polymer. This process reduces the leachability of hazardous constituents from the waste and allows easy verification of the coating integrity. Stabilization and solidification in a cement matrix will treat a wide variety of wastes (i.e. soils, decontamination water). Some pretreatments may be required (i.e. Ph adjustment) before stabilization. Other pretreatments such as soil washing can reduce the amount of waste to be stabilized. Radioactive contaminated mercury waste at the SRS comes in numerous forms (i.e. process equipment, soils, and lab waste) with the required treatment of high mercury wastes being roasting/retorting and recovery. Any unrecyclable radioactive contaminated elemental mercury would be amalgamated, utilizing a batch system, before disposal

Analysis of capital and operating costs associated with high level waste solidification processes

International Nuclear Information System (INIS)

Heckman, R.A.; Kniazewycz, B.G.

1978-03-01

An analysis was performed to evaluate the sensitivity of annual operating costs and capital costs of waste solidification processes to various parameters defined by the requirements of a proposed Federal waste repository. Five process methods and waste forms examined were: salt cake, spray calcine, fluidized bed calcine, borosilicate glass, and supercalcine multibarrier. Differential cost estimates of the annual operating and maintenance costs and the capital costs for the five HLW solidification alternates were developed

Proceedings of the workshop on radioactive, hazardous, and/or mixed waste sludge management

International Nuclear Information System (INIS)

Lomenick, T.F.

1992-01-01

A workshop sponsored by the US Department of Energy (DOE) Field Office, Oak Ridge, was held on December 4--6, 1990, in Knoxville, Tennessee. The primary objective of the workshop was the exchange of information, experiences, solutions, and future plans of DOE and its prime contractors who are engaged in work on the packaging, grouting, storage, and transport of waste sludges. In addition, the group met with industrial participants in an open forum to discuss problems and needs in the management of these wastes and to learn of possible industrial experiences, approaches, and solutions, including demonstrations of potential tools and techniques. Topics discussed include the following: mixed waste sludge issue at the K-25 site; processing saltstone from waste streams at the Savannah River Plant; the Hanford Grout Treatment Facility; treatment of pond sludge at the Rocky Flats Plant; cement solidification of low-level radioactive sludge at the West Valley Demonstration Project; studies on the solidification of low-level radioactive wastes in cement at INEL; cement solidification systems at Los Alamos National Laboratory; emergency avoidance solidification campaign at ORNL; diffusion plant sludge storage problems at the Portsmouth Gaseous Diffusion Plant; the proposed fixation of sludge in cement at the feed materials production center; regulatory aspects of sludge management; and delisting efforts for K-1407-C pond sludges. Individual projects are processed separately for the data bases

Survey of agents and techniques applicable to the solidification of low-level radioactive wastes

International Nuclear Information System (INIS)

Fuhrmann, M.; Neilson, R.M. Jr.; Colombo, P.

1981-12-01

A review of the various solidification agents and techniques that are currently available or potentially applicable for the solidification of low-level radioactive wastes is presented. An overview of the types and quantities of low-level wastes produced is presented. Descriptions of waste form matrix materials, the wastes types for which they have been or may be applied and available information concerning relevant waste form properties and characteristics follow. Also included are descriptions of the processing techniques themselves with an emphasis on those operating parameters which impact upon waste form properties. The solidification agents considered in this survey include: hydraulic cements, thermoplastic materials, thermosetting polymers, glasses, synthetic minerals and composite materials. This survey is part of a program supported by the United States Department of Energy's Low-Level Waste Management Program (LLWMP). This work provides input into LLWMP efforts to develop and compile information relevant to the treatment and processing of low-level wastes and their disposal by shallow land burial

Survey of agents and techniques applicable to the solidification of low-level radioactive wastes

Energy Technology Data Exchange (ETDEWEB)

Fuhrmann, M.; Neilson, R.M. Jr.; Colombo, P.

1981-12-01

A review of the various solidification agents and techniques that are currently available or potentially applicable for the solidification of low-level radioactive wastes is presented. An overview of the types and quantities of low-level wastes produced is presented. Descriptions of waste form matrix materials, the wastes types for which they have been or may be applied and available information concerning relevant waste form properties and characteristics follow. Also included are descriptions of the processing techniques themselves with an emphasis on those operating parameters which impact upon waste form properties. The solidification agents considered in this survey include: hydraulic cements, thermoplastic materials, thermosetting polymers, glasses, synthetic minerals and composite materials. This survey is part of a program supported by the United States Department of Energy's Low-Level Waste Management Program (LLWMP). This work provides input into LLWMP efforts to develop and compile information relevant to the treatment and processing of low-level wastes and their disposal by shallow land burial.

Species redistribution during solidification of nuclear fuel waste metal castings

Energy Technology Data Exchange (ETDEWEB)

Naterer, G F; Schneider, G E [Waterloo Univ., ON (Canada)

1994-12-31

An enthalpy-based finite element model and a binary system species redistribution model are developed and applied to problems associated with solidification of nuclear fuel waste metal castings. Minimal casting defects such as inhomogeneous solute segregation and cracks are required to prevent container corrosion and radionuclide release. The control-volume-based model accounts for equilibrium solidification for low cooling rates and negligible solid state diffusion for high cooling rates as well as intermediate conditions. Test problems involving nuclear fuel waste castings are investigated and correct limiting cases of species redistribution are observed. (author). 11 refs., 1 tab., 13 figs.

A comparison of solidification media for the stabilization of low- level radioactive wastes

International Nuclear Information System (INIS)

Cowgill, M.G.

1991-10-01

When requirements exist to stabilize low-level radioactive waste (LLW) prior to disposal, efforts to achieve this stability often center on the mixing of the waste with a solidification medium. Although historically the medium of choice has been based on the use of portland cement as the binder material, several other options have been developed and subsequently implemented. These include thermoplastic polymers, thermosetting polymers and gypsum. No one medium has thus far been successful in providing stability to all forms of LLW. The characteristics and attributes of these different binder materials are reviewed and compared. The aspects examined include availability of information, limitations to use, sensitivity to process or waste chemistry changes, radionuclide retention ability, modeling of radionuclide release processes, ease and safety of use, and relative costs

Mobile concrete solidification systems for power reactor waste

International Nuclear Information System (INIS)

Tchemitcheff, E.; Bordas, Y.

1990-01-01

In late 1988 SGN received an order from Electricite de France (EDF) for the construction of a mobile concrete solidification system to process secondary system resins generated by the P'4 and N4 series PWR power plants in France. This order was placed in view of SGN's experience with low- and medium-level radioactive waste treatment and conditioning over a period of almost 20 years. In addition to the construction of fixed waste processing facilities using more conventional technologies, SGN has been involved in application of the mobile system concept to the bituminization process in the United States, which led to the construction and commissioning of two transportable systems in collaboration with its American licensee US Ecology. It has also conducted large-scale R ampersand D on LLW/MLW concrete solidification, particularly for ion exchange resins. 5 figs

SULFUR POLYMER STABILIZATION/SOLIDIFICATION (SPSS) TREATABILITY OF LOS ALAMOS NATIONAL LABORATORY MERCURY WASTE

International Nuclear Information System (INIS)

ADAMS, J.W.; KALB, P.D.

2001-01-01

Brookhaven National Laboratory's Sulfur Polymer Stabilization/Solidification (SPSS) process was used to treat approximately 90kg of elemental mercury mixed waste from Los Alamos National Laboratory. Treatment was carried out in a series of eight batches using a 1 ft(sup 3) pilot-scale mixer, where mercury loading in each batch was 33.3 weight percent. Although leach performance is currently not regulated for amalgamated elemental mercury (Hg) mixed waste, Toxicity Characteristic Leach Procedure (TCLP) testing of SPSS treated elemental mercury waste indicates that leachability is readily reduced to below the TCLP limit of 200 ppb (regulatory requirement following treatment by retort for wastes containingandgt; 260 ppb Hg), and with process optimization, to levels less than the stringent Universal Treatment Standard (UTS) limit of 25 ppb that is applied to waste containingandlt; 260 ppm Hg. In addition, mercury-contaminated debris, consisting of primary glass and plastic containers, as well as assorted mercury thermometers, switches, and labware, was first reacted with SPSS components to stabilize the mercury contamination, then macroencapsulated in the molten SPSS product. This treatment was done by vigorous agitation of the sulfur polymer powder and the comminuted debris. Larger plastic and metal containers were reacted to stabilize internal mercury contamination, and then filled with molten sulfur polymer to encapsulate the treated product

Method of processing radioactive liquid wastes by solidification with cement

International Nuclear Information System (INIS)

Yasumura, Keijiro; Matsuura, Hiroyuki.

1975-01-01

Object: To subject radioactive liquid wastes to a cement solidification treatment after heating and drying it by a thin film scrape-off drier to render it into the form of power, and then molding it into pellets for the treatment. Structure: Radioactive liquid wastes discharged from a nuclear power plant or nuclear reactor are supplied through a storage tank into a thin film scrape-off drier. In the drier, the radioactive liquid wastes are heated to separate the liquid, and the residue is taken out as dry powder from the scrape-off apparatus. The powder obtained in this way is molded into pellets of a desired form. These pellets are then packed in a drum can or similar container, into which cement paste is then poured for solidification. (Moriyama, K.)

Immobilization of wet solid wastes at nuclear power plants

International Nuclear Information System (INIS)

Neilson, R.M. Jr.

1977-01-01

Wet solid wastes are classified into four basic types: spent resins, filter sludges, evaporator concentrates, and miscellaneous liquids. Although the immobilization of wet solid wastes is primarily concerned with the incorporation of the waste with a solidification agent, there are a number of other discrete operations or subsystems involved in the treatment of these wastes that may affect the immobilized waste product. The immobilization process may be broken down into five basic operations: waste collection, waste pretreatment, solidification agent handling, mixing/packaging, and waste package handling. The properties of the waste forms that are ultimately shipped from the reactor site are primarily influenced by the methods utilized during the waste collection, waste pretreatment and mixing/packaging operations. The mixing/packaging (solidification) operation is perhaps the most important stage of the immobilization process. The basic solidification agent types are: absorbants, hydraulic cement, urea-formaldehyde, bitumen, and other polymer systems

Physicochemical characterization of solidification agents used and products formed with radioactive wastes at LWR nuclear power plants

International Nuclear Information System (INIS)

Kibbey, A.H.; Godbee, H.W.

1978-01-01

Solidification of evaporator concentrates, filter sludges, and spent ion exchange resins used in LWR streams is discussed. The introduction of solidification agents to immobilize these sludges and resins can increase the volume of these wastes by a factor of slightly over 1 to greater than 2, depending on the binder chosen. The agents and methods used or proposed for use in solidification of LWR power plant wastes are generally suitable for treating most of the other-than-high-level wastes generated throughout the entire fuel cycle. Among the solidification agents most commonly used or suggested for use are the inorganic cements and organic plastics, which are listed and compared. A summary of considerations important in choosing a solidification agent is presented tabularly

Solidification of radioactive wastes with inorganic binders (literature survey)

International Nuclear Information System (INIS)

Rudolph, G.; Koester, R.

A survey is provided on solidification of radioactive waste solutions, sludges and tritium waste water through cement and other inorganic binders. A general survey of the possibilities described in the literature is followed by a somewhat more detailed description of the work carried on at four research establishments in the United States, Oak Ridge National Laboratory, Savannah River Laboratory, Brookhaven National Laboratory, and Atlantic Richfield Hanford Company, supplemented by personal information. Additional sections describe the experiences with various types of cement and the possibilities for improvement of solidification products through preliminary fixation of the toxic nuclides (transformation into insoluble products or absorption); there is a further possibility of post-treatment through polymer impregnation. Finally, definition and determination of leachability are provided and some results compiled. 74 references, 7 figures, 5 tables

Small-scale demonstration of high-level radioactive waste processing and solidification using actual SRP waste

International Nuclear Information System (INIS)

Okeson, J.K.; Galloway, R.M.; Wilhite, E.L.; Woolsey, G.B.; Ferguson, R.B.

1980-01-01

A small-scale demonstration of the high-level radioactive waste solidification process by vitrification in borosilicate glass is being conducted using 5-6 liter batches of actual waste. Equipment performance and processing characteristics of the various unit operations in the process are reported and, where appropriate, are compared to large-scale results obtained with synthetic waste

Production-scale LLW and RMW solidification system operational testing at Argonne National Laboratory-East (ANL-E)

International Nuclear Information System (INIS)

Wescott, J.; Wagh, A.; Singh, D.; Nelson, R.; No, H.

1997-01-01

Argonne National Laboratory-East (ANL-E) has begun production-scale testing of a low-level waste and radioactive mixed waste solidification system. This system will be used to treat low-level and mixed radioactive waste to meet land burial requirements. The system can use any of several types of solidification media, including a chemically bonded phosphate ceramic developed by ANL-E scientists. The final waste product will consist of a solidified mass in a standard 208-liter drum. The system uses commercial equipment and incorporates several unique process control features to ensure proper treatment. This paper will discuss the waste types requiring treatment, the system configuration, and operation results for these waste streams

Radioactive waste processing device

International Nuclear Information System (INIS)

Ikeda, Takashi; Funabashi, Kiyomi; Chino, Koichi.

1992-01-01

In a waste processing device for solidifying, pellets formed by condensing radioactive liquid wastes generated from a nuclear power plant, by using a solidification agent, sodium chloride, sodium hydroxide or sodium nitrate is mixed upon solidification. In particular, since sodium sulfate in a resin regenerating liquid wastes absorbs water in the cement upon cement solidification, and increases the volume by expansion, there is a worry of breaking the cement solidification products. This reaction can be prevented by the addition of sodium chloride and the like. Accordingly, integrity of the solidification products can be maintained for a long period of time. (T.M.)

Handling 78,000 drums of mixed-waste sludge

International Nuclear Information System (INIS)

Berry, J.B.; Gilliam, T.M.; Harrington, E.S.; Youngblood, E.L.; Baer, M.B.

1991-01-01

The Oak Ridge Gaseous Diffusion Plant (now know as the Oak Ridge K-25 Site) prepared two mixed-waste surface impoundments for closure by removing the sludge and contaminated pond-bottom clay and attempting to process it into durable, nonleachable, concrete monoliths. Interim, controlled, above-ground storage of the stabilized waste was planned until final disposition. The strategy for disposal included delisting the stabilized pond sludge from hazardous to nonhazardous and disposing of the delisted monoliths as radioactive waste. Because of schedule constraints and process design and control deficiencies, ∼46,000 drums of material in various stages of solidification and ∼32,000 drums of unprocessed sludge are presently being stored. In addition, the abandoned treatment facility still contains ∼16,000 gal of raw sludge. Such conditions do not comply with the requirements set forth by the Resource Conservation and Recovery Act (RCRA) for the storage of listed waste. Various steps are being taken to bring the storage of ∼78,000 drums of mixed waste into compliance with RCRA. This paper (1) reviews the current situation, (2) discusses the plan for remediation of regulatory noncompliances, including decanting liquid from stabilized waste and dewatering untreated waste, and (3) provides an assessment of alternative raw-waste treatment processes. 1 ref., 6 figs., 2 tabs

SULFUR POLYMER STABILIZATION/SOLIDIFICATION (SPSS) TREATABILITY OF LOS ALAMOS NATIONAL LABORATORY MERCURY WASTE.

Energy Technology Data Exchange (ETDEWEB)

ADAMS,J.W.; KALB,P.D.

2001-11-01

Brookhaven National Laboratory's Sulfur Polymer Stabilization/Solidification (SPSS) process was used to treat approximately 90kg of elemental mercury mixed waste from Los Alamos National Laboratory. Treatment was carried out in a series of eight batches using a 1 ft{sup 3} pilot-scale mixer, where mercury loading in each batch was 33.3 weight percent. Although leach performance is currently not regulated for amalgamated elemental mercury (Hg) mixed waste, Toxicity Characteristic Leach Procedure (TCLP) testing of SPSS treated elemental mercury waste indicates that leachability is readily reduced to below the TCLP limit of 200 ppb (regulatory requirement following treatment by retort for wastes containing > 260 ppb Hg), and with process optimization, to levels less than the stringent Universal Treatment Standard (UTS) limit of 25 ppb that is applied to waste containing < 260 ppm Hg. In addition, mercury-contaminated debris, consisting of primary glass and plastic containers, as well as assorted mercury thermometers, switches, and labware, was first reacted with SPSS components to stabilize the mercury contamination, then macroencapsulated in the molten SPSS product. This treatment was done by vigorous agitation of the sulfur polymer powder and the comminuted debris. Larger plastic and metal containers were reacted to stabilize internal mercury contamination, and then filled with molten sulfur polymer to encapsulate the treated product.

Solidification of commercial and defense low-level radioactive waste in polyethylene

International Nuclear Information System (INIS)

Franz, E.M.; Heiser, L.H.; Colombo, P.

1987-08-01

A process was developed for the solidification of salt wastes, incinerator ash and ion-exchange resins in polyethylene. Of the salt wastes, sodium sulfate and boric acid are representative of the wastes produced at commercial nuclear facilities while sodium nitrate in a typical high-volume waste generated at defense-related facilities. Ease of processibility and high loading efficiencies were obtained through the use of low-density polyethylene with melt indices ranging from 2.0 to 55.0 g/minute. The process utilized a commercially available single-screw extruder to incorporate the wastes into the polyethylene at about 120 0 C to produce a homogeneous mixture. Although present studies utilize dry wastes, wet wastes can also be processed using vented extruders of the type used commercially for the bitumen solidification process. Tests were performed on the waste forms to determine leachability and mechanical properties. To confirm the compatibility of polyethylene and nitrate salt waste at elevated temperatures, the self-ignition temperatures were measured and a differential scanning calorimeter was used to characterize the thermal behavior of oxidizing compounds contained in the simulated waste, as well as the real Savannah River Plant waste. No exothermic reactions were observed over the temperature range studied from 50 0 C to 400 0 C. 18 refs., 7 figs., 8 tabs

Development of sodium disposal technology. Experiment of sodium compound solidification process

International Nuclear Information System (INIS)

Matsumoto, Toshiyuki; Ohura, Masato; Yatoh, Yasuo

2007-07-01

A large amount of sodium containing radioactive waste will come up at the time of final shutdown/decommission of FBR plant. The radioactive waste is managed as solid state material in a closed can in Japan. As for the sodium, there is no established method to convert the radioactive sodium to solid waste. Further, the sodium is highly reactive. Thus, it is recommended to convert the sodium to a stable substance before the solidification process. One of the stabilizing methods is conversion of sodium into sodium hydroxide solution. These stabilization and solidification processes should be safe, economical, and efficient. In order to develop such sodium disposal technology, nonradioactive sodium was used and a basic experiment was performed. Waste-fluid Slag Solidification method was employed as the solidification process of sodium hydroxide solution. Experimental parameters were mixing ratio of the sodium hydroxide and the slag solidification material, temperature and concentration of the sodium hydroxide. The best parameters were obtained to achieve the maximum filling ratio of the sodium hydroxide under a condition of enough high compressive strength of the solidified waste. In a beaker level test, the solidified waste was kept in a long term and it was shown that there was no change of appearance, density, and also the compressive strength was kept at a target value. In a real scale test, homogeneous profiles of the density and the compressive strength were obtained. The compressive strength was higher than the target value. It was shown that the Waste-fluid Slag Solidification method can be applied to the solidification process of the sodium hydroxide solution, which was produced by the stabilization process. (author)

Treatment of Petroleum Drill Cuttings Using Stabilization/Solidification Method by Cement and Modified Clay Mixes

Directory of Open Access Journals (Sweden)

Soroush Ghasemi

2017-04-01

Full Text Available High organic content in petroleum drill cuttings is a substantial obstacle which hinders cement hydration and subsequently decreases the clean-up efficiency of the stabilization/solidification (S/S process. In this study, a modified clayey soil (montmorillonite with low to moderate polarity was used as an additive to cement. Because of its high adsorption capacity, the clay is capable of mitigating the destructive role of organic materials and preventing their interference with the hydration process. Mixes containing different ratios of cement, waste and modified clay were prepared and tested for their mechanical and chemical characteristics. Total petroleum hydrocarbons (TPH and Pb content of the samples were analyzed as well. For this purpose, the mixes were subjected to unconfined compressive strength (UCS and toxicity characteristic leaching procedure (TCLP tests. The results indicated that the specimens with 28-day curing time at a cement/waste ratio of 25% or higher (w/w and 10% modified clay (w/w met the Environmental Protection Agency (EPA criterion for compressive strength. Moreover, a reduction of 94% in the leaching of TPH was observed with the specimens undergoing the TCLP with a cement/waste ratio of 30% (w/w and a clay/waste ratio of 30% (w/w. Finally, the specimens with 30% cement/waste and 10% clay/waste ratios showed the least concentration (6.14% of leached Pb.

Demonstration of GTS Duratek Process for Stabilizing Mercury Contaminated (<260 ppm) Mixed Wastes. Mixed Waste Focus Area. OST Reference No. 2409

International Nuclear Information System (INIS)

1999-01-01

Mercury-contaminated wastes in many forms are present at various U. S. Department of Energy (DOE) sites. At least 26 different DOE sites have this type of mixed low-level waste in their storage facilities, totaling approximately 6,000 m 3 . Mercury contamination in the wastes at DOE sites presents a challenge because it exists in various forms, such as soil, sludges, and debris, as well as in different chemical species of mercury. Stabilization is of interest for radioactively contaminated mercury waste (<260 ppm Hg) because of its success with particular wastes, such as soils, and its promise of applicability to a broad range of wastes. However, stabilization methods must be proven to be adequate to meet treatment standards. They must also be proven feasible in terms of economics, operability, and safety. This report summarizes the findings from a stabilization technology demonstration conducted by GTS Duratek, Inc. Phase I of the study involved receipt and repackaging of the material, followed by preparations for waste tracking. Phase II examined the bench-scale performance of grouting at two different loadings of waste to grouted mass. Phase III demonstrated in-drum mixing and solidification using repackaged drums of sludge. Phase IV initially intended to ship final residues to Envirocare for disposal. The key results of the demonstration are as follows: (1) Solidification tests were performed at low and high waste loading, resulting in stabilization of mercury to meet the Universal Treatment Standard of 0.025 mg/L at the low loading and for two of the three runs at the high loading. The third high-loading run had a Toxicity Characteristic Leaching Procedure (TCLP) of 0.0314 mg/L. (2) Full-drum stabilization using the low loading formula was demonstrated. (3) Organic compound levels were discovered to be higher than originally reported, including the presence of some pesticides. Levels of some radionuclides were much higher than initially reported. (4

The cement solidification systems at LANL

International Nuclear Information System (INIS)

Veazey, G.W.

1990-01-01

There are two major cement solidification systems at Los Alamos National Laboratory. Both are focused primarily around treating waste from the evaporator at TA-55, the Plutonium Processing Facility. The evaporator receives the liquid waste stream from TA-55's nitric acid-based, aqueous-processing operations and concentrates the majority of the radionuclides in the evaporator bottoms solution. This is sent to the TA-55 cementation system. The evaporator distillate is sent to the TA-50 facility, where the radionuclides are precipitated and then cemented. Both systems treat TRU-level waste, and so are operated according to the criteria for WIPP-destined waste, but they differ in both cement type and mixing method. The TA-55 systems uses Envirostone, a gypsum-based cement and in-drum prop mixing; the TA-50 systems uses Portland cement and drum tumbling for mixing

Method of solidifying radioactive wastes with plastics

International Nuclear Information System (INIS)

Matsuura, Hiroyuki; Yasumura, Keijiro; Minami, Yuji; Tomita, Toshihide

1980-01-01

Purpose: To prevent solidification of solidifying agents in the mixer by conducting the mixing process for the solidifying agents and the radioactive wastes at a temperature below the initiation point for the solidification of the agents thereby separating the mixing process from the solidification-integration process. Method: Catalyst such as cobalt naphthenate is charged into an unsaturated polyester resin in a mixer previously cooled, for example, to -10 0 C. They are well mixed with radioactive wastes and the mixture in the mixer is charged in a radioactive waste storage container. The temperature of the mixture, although kept at a low temperature initially, gradually increases to an ambient temperature whereby curing reaction is promoted and the reaction is completed about one day after to provide firm plastic solidification products. This can prevent the solidification of the solidifying agents in the mixer to thereby improve the circumstance's safety. (Kawakami, Y.)

Solidification of problem wastes: Annual progress report, October 1985-September 1986

International Nuclear Information System (INIS)

Franz, E.M.; Heiser, J.H. III; Colombo, P.

1987-02-01

This report describes initial work on the development of solidification systems for sodium nitrate waste and compacted waste. Sodium nitrate waste has been solidified in three types of materials: polyethylene, polyester-styrene (PES), and latex cement. Evaluations of the properties of the waste form, such as the ANS 16.1 leaching test, water immersion test and compressive strength measurements were performed on the waste forms containing various amounts of sodium nitrate. 9 refs., 9 figs., 7 tabs

Melting, solidification, remelting, and separation of glass and metal

International Nuclear Information System (INIS)

Ebadian, M.A.; Xin, R.C.; Liu, Y.Z.

1998-01-01

Several high-temperature vitrification technologies have been developed for the treatment of a wide range of mixed waste types in both the low-level waste and transuranic (TRU) mixed waste categories currently in storage at DOE sites throughout the nation. The products of these processes are an oxide slag phase and a reduced metal phase. The metal phase has the potential to be recycled within the DOE Complex. Enhanced slag/metal separation methods are needed to support these processes. This research project involves an experimental investigation of the melting, solidification, remelting, and separation of glass and metal and the development of an efficient separation technology. The ultimate goal of this project is to find an efficient way to separate the slag phase from the metal phase in the molten state. This two-year project commenced in October 1995 (FY96). In the first fiscal year, the following tasks were accomplished: (1) A literature review and an assessment of the baseline glass and metal separation technologies were performed. The results indicated that the baseline technology yields a high percentage of glass in the metal phase, requiring further separation. (2) The main melting and solidification system setup was established. A number of melting and solidification tests were conducted. (3) Temperature distribution, solidification patterns, and flow field in the molten metal pool were simulated numerically for the solidification processes of molten aluminum and iron steel. (4) Initial designs of the laboratory-scale DCS and CS technologies were also completed. The principal demonstration separation units were constructed. (5) An application for a patent for an innovative liquid-liquid separation technology was submitted and is pending

Cadarache LOR (liquides organiques radioactifs) treatment by a solidification process using NOCHAR polymers

International Nuclear Information System (INIS)

Vaudey, Claire-Emilie; Renou, Sebastien; Kelley, Dennis; Cochaud, Chantal; Serrano, Roger

2013-01-01

In France, two options can be considered to handle the Very Low Level Waste (VLLW) and the Low Level Waste (LLW). The first one is the incineration at CENTRACO facility and the second one is the disposal at ANDRA sites. The waste acceptance in these two channels is dependent upon the adequacy between the waste characteristics (physical chemistry and radiological) and the channel specifications. If the waste characteristics and the channel specifications (presence of significant quantities of halogens, complexing agents, organic components... or/and high activity limits) are incompatible, an alternative solution have to be identify. It consists of a waste pre-treatment process. For Cadarache LOR (Liquides Organiques Radioactifs) waste streams, two radioactive scintillation cocktails have to be treated. They are composed of a mix of organic liquids and water: for the first one, 19 % of organic compounds (xylene, mesitylene, diphenyloxazole, TBP...) and 86.9 % of water, and for the second one, 23 % of organic compounds (TBP...) and 77 % of water. They contain halogens (chlorine and fluorine), complexants agents (nitrate, sulphate, oxalate and formate) and have got αβγ spectra with mass activities equal to some 100 Bq/g. Therefore, tritium is also present. As a consequence, in order for storage acceptance at the ANDRA site, it is necessary to pre-treat the waste. An adequate solution seems to be a solidification process using NOCHAR polymers. Indeed, NOCHAR polymers correspond to an important variety of products applied to the treatment of radioactive aqueous and organic liquids (solvent, oil, solvent/oil mixing ...) and sludge through a mechanical and chemical solidification process. For Cadarache LOR, N910 and N960 respectively dedicated to the organic and aqueous liquids solidification are considered. With the N910, the organic waste solidification occurs in two steps. As the organic liquid travels moves through the polymer strands, the strands swell and

Solidification and performance of cement doped with phenol

International Nuclear Information System (INIS)

Vipulanandan, C.; Krishnan, S.

1991-01-01

Treating mixed hazardous wastes using the solidification/stabilization technology is becoming a critical element in waste management planning. The effect of phenol, a primary constituent in many hazardous wastes, on the setting and solidification process of Type I Portland cement was evaluated. The leachability of phenol from solidified cement matrix (TCLP test) and changes in mechanical properties were studied after curing times up to 28 days. The changes in cement hydration products due to phenol were studied using the X-ray diffraction (XRD) powder technique. Results show that phenol interferes with initial cement hydration by reducing the formation of calcium hydroxide and also reduces the compressive strength of cement. A simple model has been proposed to quantify the phenol leached from the cement matrix during the leachate test

Economic analysis of a volume reduction/polyethylene solidification system for low-level radioactive wastes

International Nuclear Information System (INIS)

Kalb, P.D.; Colombo, P.

1985-01-01

A study was conducted at Brookhaven National Laboratory to determine the economic feasibility of a fluidized bed volume reduction/polyethylene solidification system for low-level radioactive wastes. These results are compared with the ''null'' alternative of no volume reduction and solidification of aqueous waste streams in hydraulic cement. The economic analysis employed a levelized revenue requirement (LRR) technique conducted over a ten year period. An interactive computer program was written to conduct the LRR calculations. Both of the treatment/solidification options were considered for a number of scenarios including type of plant (BWR or PWR) and transportation distance to the disposal site. If current trends in the escalation rates of cost components continue, the volume reduction/polyethylene solidification option will be cost effective for both BWRs and PWRs. Data indicate that a minimum net annual savings of $0.8 million per year (for a PWR shipping its waste 750 miles) and a maximum net annual savings of $9 million per year (for a BWR shipping its waste 2500 miles) can be achieved. A sensitivity analysis was performed for the burial cost escalation rate, which indicated that variation of this factor will impact the total levelized revenue requirement. The burial cost escalation rate which yields a break-even condition was determined for each scenario considered. 11 refs., 8 figs., 39 tabs

Properties and solidification of decontamination wastes

International Nuclear Information System (INIS)

Davis, M.S.; Piciulo, P.L.; Bowerman, B.S.; Adams, J.W.; Milian, L.

1983-01-01

LWRs will require one or more chemical decontaminations to achieve their designed lifetimes. Primary system decontamination is designed to lower radiation fields in areas where plant maintenance personnel must work. Chemical decontamination methods are either hard (concentrated chemicals, approximately 5 to 25 weight percent) or soft (dilute chemicals less than 1 percent by weight). These methods may have different chemical reagents, some tailor-made to the crud composition and many methods are and will be proprietary. One factor common to most commercially available processes is the presence of organic acids and chelates. These types of organic reagents are known to enhance the migration of radionuclides after disposal in a shallow land burial site. The NRC sponsors two programs at Brookhaven National Laboratory that are concerned with the management of decontamination wastes which will be generated by the full system decontamination of LWRs. These two programs focus on potential methods for degrading or converting decontamination wastes to more acceptable forms prior to disposal and the impact of disposing of solidified decontamination wastes. The results of the solidification of simulated decontamination resin wastes will be presented. Recent results on combustion of simulated decontamintion wastes will be described and procedures for evaluating the release of decontamination reagents from solidified wastes will be summarized

Stabilization and Solidification of Nitric Acid Effluent Waste at Y-12

Energy Technology Data Exchange (ETDEWEB)

Singh, Dileep [Argonne National Lab. (ANL), Argonne, IL (United States); Lorenzo-Martin, Cinta [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-12-16

Consolidated Nuclear Security, LLC (CNS) at the Y-12 plant is investigating approaches for the treatment (stabilization and solidification) of a nitric acid waste effluent that contains uranium. Because the pH of the waste stream is 1-2, it is a difficult waste stream to treat and stabilize by a standard cement-based process. Alternative waste forms are being considered. In this regard, Ceramicrete technology, developed at Argonne National Laboratory, is being explored as an option to solidify and stabilize the nitric acid effluent wastes.

Influence of non-technical policies on choices of waste solidification technologies

International Nuclear Information System (INIS)

Trubatch, S.L.

1987-01-01

This paper describes and discusses non-technical policy considerations which may improperly influence decisions on the solidification of low-level radioactive wastes (''LLW''). These policy considerations are contained principally in several State and Federal statutes which regulate various aspects of LLW disposal. One policy consideration in particular, the unqualified bias in favor of volume reduction, is shown to present a substantial potential for leading to technically suboptimal decisions on the appropriate processes for solidifying LLW. To avoid the unintended skewing of technical decisions by non-technical policy considerations, certain current policies may need to be revised to ensure that the choices of waste treatment, including decisions on solidification, are based primarily on reasonable assurance of adequate protection of public health and safety. This goal may be realized in part by basing any disposal fee structure on more than just LLW volume to include consideration of the waste's activity and its difficulty of confinement

Treatability studies for polyethylene encapsulation of INEL low-level mixed wastes. Final report

International Nuclear Information System (INIS)

Lageraaen, P.R.; Patel, B.R.; Kalb, P.D.; Adams, J.W.

1995-10-01

Treatability studies for polyethylene encapsulation of Idaho National Engineering Laboratory (INEL) low-level mixed wastes were conducted at Brookhaven National Laboratory. The treatability work, which included thermal screening and/or processibility testing, was performed on priority candidate wastes identified by INEL to determine the applicability of polyethylene encapsulation for the solidification and stabilization of these mixed wastes. The candidate wastes selected for this preliminary study were Eutectic Salts, Ion Exchange Resins, Activated Carbons, Freon Contaminated Rags, TAN TURCO Decon 4502, ICPP Sodium Bearing Liquid Waste, and HTRE-3 Acid Spill Clean-up. Thermal screening was conducted for some of these wastes to determine the thermal stability of the wastes under expected pretreatment and processing conditions. Processibility testing to determine whether the wastes were amenable to extrusion processing included monitoring feed consistency, extruder output consistency, waste production homogeneity, and waste form performance. Processing parameters were not optimized within the scope of this study. However, based on the treatability results, polyethylene encapsulation does appear applicable as a primary or secondary treatment for most of these wastes

Process control of Low and Intermediate-level radioactive wastes solidification

International Nuclear Information System (INIS)

1993-01-01

Safety guidelines issued by the Spanish Council of Nuclear Safety (CSN) with basic criteria which must be adopted for the control of the Process for wastes solidification, establishing, in addition, a series of protocols and basic contents to assist the elaboration of Process Control Programs

Cement solidification of spent ion exchange resins produced by the nuclear industry

International Nuclear Information System (INIS)

Jaouen, C.; Vigreux, B.

1988-01-01

Cement solidification technology has been applied to spent ion exchange resins for many years in countries throughout the world (at reactors, research centers and spent fuel reprocessing plants). Changing specifications for storage of radioactive waste have, however, confronted the operators of such facilities with a number of problems. Problems related both to the cement solidification process (water/cement/resin interactions and chemical interactions) and to its utilization (mixing, process control, variable feed composition, etc.) have often led waste producers to prefer other, polymer-based processes, which are very expensive and virtually incompatible with water. This paper discusses research on cement solidification of ion exchange resins since 1983 and the development of application technologies adapted to nuclear service conditions and stringent finished product quality requirements

Sandia solidification process: a broad range aqueous waste solidification method

International Nuclear Information System (INIS)

Lynch, R.W.; Dosch, R.G.; Kenna, B.T.; Johnstone, J.K.; Nowak, E.J.

1976-01-01

New ion-exchange materials of the hydrous oxide type were developed for solidifying aqueous radioactive wastes. These materials have the general formula M[M'/sub x/O/sub y/H/sub z/]/sub n/, where M is an exchangeable cation of charge +n and M' may be Ti; Nb; Zr, or Ta. Affinities for polyvalent cations were found to be very high and ion-exchange capacities large (e.g., 4.0--4.5 meq/g for NaTi 2 O 5 H depending on moisture content). The effectiveness of the exchangers for solidifying high-level waste resulting from reprocessing light-water reactor fuel was demonstrated in small-scale tests. Used in conjunction with anion exchange resin, these materials reduced test solution radioactivity from approximately 0.2 Ci/ml to as low as approximately 2 nCi/ml. The residual radioactivity was almost exclusively due to 106 Ru and total α-activity was only a few pCi/ml. Alternative methods of consolidating the solidified waste were evaluated using nonradioactive simulants. Best results were obtained by pressure-sintering which yielded essentially fully dense ceramics, e.g., titanate/titania ceramics with bulk density as high as 4.7 g/cm 3 , waste oxide content as high as 1.2 g/cm 3 , and leach resistance comparable to good borosilicate glass. Based on the above results, a baseline process for solidifying high-level waste was defined and approximate economic analyses indicated costs were not prohibitive. Additional tests have demonstrated that, if desired, operating conditions could be modified to allow recovery of radiocesium (and perhaps other isotopes) during solidification of the remaining constituents of high-level waste. Preliminary tests have also shown that these materials offer promise for treating tank-stored neutralized wastes

Integral solution of equiaxed solidification with an interface kinetics model for nuclear waste management

International Nuclear Information System (INIS)

Naterer, G.F.

1996-01-01

In this paper, a one-dimensional analysis of energy and species transport during binary dendritic solidification is presented and compared to experimental results. The paper's objective is a continuation of previous studies of solidification control for the waste management of nuclear materials in the underground disposal concept. In the present analysis, interface kinetics at the solid - liquid interface accounts for recalescent thermal behaviour during solidification. The theoretical results were compared to available experimental results and the agreement appears fair although some discrepancies have been attributed to uncertainties with thermophysical properties. (author)

Performance evaluation of rotating pump jet mixing of radioactive wastes in Hanford Tanks 241-AP-102 and -104

International Nuclear Information System (INIS)

Onishi, Y.; Recknagle, K.P.

1998-07-01

The purpose of this study was to confirm the adequacy of a single mixer pump to fully mix the wastes that will be stored in Tanks 241-AP-102 and -104. These Hanford double-shell tanks (DSTs) will be used as staging tanks to receive low-activity wastes from other Hanford storage tanks and, in turn, will supply the wastes to private waste vitrification facilities for eventual solidification. The TEMPEST computer code was applied to Tanks AP-102 and -104 to simulate waste mixing generated by the 60-ft/s rotating jets and to determine the effectiveness of the single rotating pump to mix the waste. TEMPEST simulates flow and mass/heat transport and chemical reactions (equilibrium and kinetic reactions) coupled together. Section 2 describes the pump jet mixing conditions the authors evaluated, the modeling cases, and their parameters. Section 3 reports model applications and assessment results. The summary and conclusions are presented in Section 4, and cited references are listed in Section 5

Testing and evaluation of alternative process systems for immobilizing radioactive mixed particulate waste in cement

International Nuclear Information System (INIS)

Weingardt, K.M.; Weber, J.R.

1994-03-01

Radioactive and Hazardous Mixed Wastes have accumulated at the Department of Energy (DOE) Hanford Site in south-central Washington State. Ongoing operations and planned facilities at Hanford will also contribute to this waste stream. To meet the Resource Conservation and Recovery Act (RCRA) Land Disposal Restrictions most of this waste will need to be treated to permit disposal. In general this treatment will need to include stabilization/solidification either as a sole method or as part of a treatment train. A planned DOE facility, the Waste Receiving and Processing (WRAP) Module 2A, is scoped to provide this required treatment for containerized contact-handled (CH), mixed low-level waste (MLLW) at Hanford. An engineering development program has been conducted by Westinghouse Hanford Company (WHC) to select the best system for utilizing a cement based process in WRAP Module 2A. Three mixing processes were developed for analysis and testing; in-drum mixing, continuous mixing, and batch mixing. Some full scale tests were conducted and 55 gallon drums of solidified product were produced. These drums were core sampled and examined to evaluate mixing effectiveness. Total solids loading and the order of addition of waste and binder constituents were also varied. The highest confidence approach to meet the WRAP Module 2A waste immobilization system needs appears to be the out-of-drum batch mixing concept. This system is believed to offer the most flexibility and efficiency, given the highly variable and troublesome waste streams feeding the facility

Comparison of modified sulfur cement and hydraulic cement for encapsulation of radioactive and mixed wastes

International Nuclear Information System (INIS)

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

1990-01-01

The majority of solidification/stabilization systems for low-level radioactive waste (LLW) and mixed waste, both in the commercial sector and at Department of Energy (DOE) facilities, utilize hydraulic cement (such as portland cement) to encapsulate waste materials and yield a monolithic solid waste form for disposal. Because hydraulic cement requires a chemical hydration reaction for setting and hardening, it is subject to potential interactions between elements in the waste and binder that can retard or prevent solidification. A new and innovative process utilizing modified sulfur cement developed by the US Bureau of Mines has been applied at Brookhaven National Laboratory (BNL) for the encapsulation of many of these problem wastes. Modified sulfur cement is a thermoplastic material, and as such, it can be heated above its melting point, combined with dry waste products to form a homogeneous mixture, and cooled to form a monolithic solid product. Under sponsorship of the DOE, research and development efforts at BNL have successfully applied the modified sulfur cement process for treatment of a range of LLWs including sodium sulfate salts, boric acid salts, and incinerator bottom ash and for mixed waste contaminated incinerator fly ash. Process development studies were conducted to determine optimal waste loadings for each waste type. Property evaluation studies were conducted to test waste form behavior under disposal conditions by applying relevant performance testing criteria established by the Nuclear Regulatory Commission (for LLW) and the Environmental Protection Agency (for hazardous wastes). Based on both processing and performance considerations, significantly greater waste loadings were achieved using modified sulfur cement when compared with hydraulic cement. Technology demonstration of the modified sulfur cement encapsulation system using production-scale equipment is scheduled for FY 1991

Method of processing liquid waste containing fission product

International Nuclear Information System (INIS)

Funabashi, Kiyomi; Kawamura, Fumio; Matsuda, Masami; Komori, Itaru; Miura, Eiichi.

1988-01-01

Purpose: To prepare solidification products of low surface dose by removing cesium which is main radioactive nuclides from re-processing plants. Method: Liquid wastes containing a great amount of fission products are generated accompanying the reprocessing for spent nuclear fuels. After pH adjustment, the liquid wastes are sent to a concentrator to concentrate the dissolved ingredients. The concentrated liquid wastes are pumped to an adsorption tower in which radioactive cesium contributing much to the surface dose is removed. Then, the liquid wastes are sent by way of a surge tank to a mixing tank, in which they are mixed under stirring with solidifying agents such as cements. Then, the mixture is filled in a drum-can and solidified. According to this invention, since radioactive cesium is removed before solidification, it is possible to prepare solidification products at low surface dose and facilitate the handling of the solidification products. (Horiuchi, T.)

Immobilisation/solidification of hazardous toxic waste in cement matrices

Directory of Open Access Journals (Sweden)

Macías, A.

1999-06-01

Full Text Available Immobilization and solidification of polluting waste, introduced into the industrial sector more than 20 years ago, and throughout last 10 years is being the object of a growing interest for engineers and environment scientists, has become a remarkable standardized process for treatment and management of toxic and hazardous liquid wastes, with special to those containing toxic metals. Experimental monitorization of the behaviour of immobilized waste by solidification and stabilisation in life time safe deposits is not possible, reason why it is essential to develop models predicting adequately the behaviour of structures that have to undergo a range of conditions simulating the environment where they are to be exposed. Such models can be developed only if the basic physical and chemical properties of the system matrix/solidifying-waste are known. In this work immobilization/solidification systems are analyzed stressing out the formulation systems based on Portland cement. Finally, some examples of the results obtained from the study of interaction of specific species of wastes and fixation systems are presented.

La inmovilización y solidificación de residuos contaminantes, implantada en el sector comercial desde hace más de 20 años y que desde hace diez es objeto de creciente interés por parte de ingenieros y científicos medioambientales, se ha convertido en un proceso estandarizado único para el tratamiento y gestión de residuos tóxicos y peligrosos líquidos y, en especial, de los que contienen metales pesados. La monitorización experimental del comportamiento de un residuo inmovilizado por solidificación y estabilización en el tiempo de vida de un depósito de seguridad no es posible, por lo que es imprescindible desarrollar modelos que predigan satisfactoriamente el comportamiento del sistema bajo un rango representativo de condiciones del entorno de exposición. Tales modelos sólo pueden ser desarrollados si se

Application of sulfur concrete for solidification of radioactive wastes and building of repositories

International Nuclear Information System (INIS)

Cholerzynski, A.; Tomczak, W.; Switalski, J.

2000-01-01

The application of sulfur concrete as solidification material for radioactive wastes and as building material used in repositories have been presented. Their high shear strength, low level of leaching, and high radiation resistance decide of positive recommendation of such material for wide use in radioactive waste treatment processes and repositories building

Solidification process for toxic and hazardous wastes. Second part: Cement solidification matrices; Inertizzazione di rifiuti tossici e nocivi (RTN). Parte seconda: Inertizzazione in matrici cementizie

Energy Technology Data Exchange (ETDEWEB)

Donato, A; Arcuri, L; Dotti, M; Pace, A; Pietrelli, L; Ricci, G [ENEA - Dipartimento Ciclo del Combustibile, Centro Ricerche Energia, Casaccia (Italy); Basta, M; Cali, V; Pagliai, V [ENEA - Dipartimento Ciclo del Combustibile, Centro Ricerche Energia, Saluggia (Italy)

1989-05-15

This paper reports the second part of a general study carried out at the Nuclear Fuel Division aiming at verifying the possible application of the radioactive waste solidification processes to industrial hazardous wastes (RTN). The cement solidification of several RTN types has been taken into consideration, both from the technical and from the economic point of view. After a short examination of the Italian juridical and economical situation in the field, which demonstrates the need of the RTN solidification, the origin and characteristics of the RTN considered in the study and directly provided by the producing industries are reviewed. The laboratory experimental results of the cementation of RTN produced by gold manufacturing industries and by galvanic industries are reported. The cementation process can be considered a very effective mean for reducing both the RTN management costs and the environmental impact of RTN disposal. (author)

Defense waste solidification studies. Volume 2. Drawing supplement. Savannah River Plant, Project S-1780

International Nuclear Information System (INIS)

1977-01-01

Volume 2 contains the drawings prepared and used in scoping and estimating the Glass-Form Waste Solidification facilities and the alternative studies cited in the report, the Off-Site Shipping Case, the Decontaminated Salt Storage Case, and a revised Reference Plant (Concrete-Form Waste) Case

Method of cement-solidification of radioactive liquid wastes containing surfactant

Energy Technology Data Exchange (ETDEWEB)

Sugimoto, Y; Yusa, H

1979-04-10

Purpose: To provide the subject method comprising the steps of adjusting the concentration of the surfactant to a value less than the predetermined value even when the concentration of the surfactant is high, and rendering the uniaxial compression strength of the cement-solidification body into more than the defined fabrication reference value. Method: To radioactive liquid wastes there are applied means for boiling and heating liquid wastes by addition of sulfuric acid, means for cracking surfactants by the addition of oxidants and means for precipitating and arresting surfactants. After suppressing the hindrance of the cement hydration reaction by surfactants, the radioactive liquid wastes are cement-solidified. (Nakamura, S.).

Industrial wastes solidification and material recovery: prospectives in Italy. Prospettive dell'applicazione delle tecniche di inertizzazione

Energy Technology Data Exchange (ETDEWEB)

De Angelis, G; Balzano, S

1988-12-01

This paper focuses on state-of-the-art materials recovery techniques employed in the solidification/stabilization of industrial wastes. Particular consideration is given to the Italian situation. After a review, with reference to waste/matrix compatibility inherent problems, of the presently employed main encapsulation techniques (with matrices based on cement, lime, clay, thermoplastic materials, organic polymers, macroencapsulating compounds), attention is addressed to solidification systems which allow a recovery of the waste material as low-technology by-products. Regarding the most important industrial waste streams: thermoplastic refuse, incinerator ashes, chemical sludges, the paper reviews efforts devoted not only to their chemical fixation in order to fulfill the current land disposal requirements, but mainly to their employment for production of manufactured articles.

Recent advances in cement solidification of radioactive wastes

International Nuclear Information System (INIS)

Vigreux, B.; Jaouen, C.

1987-01-01

Advanced cement solidification processes and systems have been developed by SGN to meet changing requirements in radioactive waste processing and packaging and to avoid the difficulties often encountered in waste concreting on an industrial scale. SGN applies a strict development methodology to ensure integration of the most recent information on chemical behavior of solidified wastes plus compliance with the precise needs of waste producers and evolving regulatory requirements concerning waste package storage and disposal. Based on a hierarchical definition of objectives, this methodology was implemented following an overall study on radwaste concreting performed in 1983 and 1984 for Electricite de France (EdF), France's national electric power utility. It ensures that industrial and regulatory factors are fully considered from the start of development work. It also constrains development in the direction of true process optimization and guarantees compliance with defined objectives. The methodology has helped SGN develop concreting processes adapted to various types of radioactive waste. The most widely employed processes are first briefly described in this paper. It then presents continuous and batch systems using these processes, focusing on technological features chosen at a very early stage in development

Solidification Technologies for Radioactive and Chemical Liquid Waste Treatment - Final CRADA Report

International Nuclear Information System (INIS)

Castiglioni, Andrew J.; Gelis, Artem V.

2016-01-01

This project, organized under DOE/NNSA's Global Initiatives for Proliferation Prevention program, joined Russian and DOE scientists in developing more effective solidification and storage technologies for liquid radioactive waste. Several patent applications were filed by the Russian scientists (Russia only) and in 2012, the technology developed was approved by Russia's Federal State Unitary Enterprise RADON for application throughout Russia in cleaning up and disposing of radioactive waste.

Sulfur polymer cement, a new stabilization agent for mixed and low- level radioactive waste

International Nuclear Information System (INIS)

Darnell, G.R.

1991-01-01

Solidification and stabilization agents for radioactive, hazardous, and mixed wastes are failing to pass governmental tests at alarming rates. The Department of Energy's National Low-Level Waste Management Program funded testing of Sulfur Polymer Cement (SPC) by Brookhaven National Laboratory during the 1980s. Those tests and tests by the US Bureau of Mines (the original developer of SPC), universities, states, and the concrete industry have shown SPC to be superior to hydraulic cements in most cases. Superior in what wastes can be successfully combined and in the quantity of waste that can be combined and still pass the tests established by the US Environmental Protection Agency and the US Nuclear Regulatory Commission

Solidification of oils and organic liquids

International Nuclear Information System (INIS)

Clark, D.E.; Colombo, P.; Neilson, R.M. Jr.

1982-07-01

The suitability of selected solidification media for application in the disposal of low-level oil and other organic liquid wastes has been investigated. In the past, these low-level wastes (LLWs) have commonly been immobilized by sorption onto solid absorbents such as vermiculite or diatomaceous earth. Evolving regulations regarding the disposal of these materials encourage solidification. Solidification media which were studied include Portland type I cement; vermiculite plus Portland type I cement; Nuclear Technology Corporation's Nutek 380-cement process; emulsifier, Portland type I cement-sodium silicate; Delaware Custom Materiel's cement process; and the US Gypsum Company's Envirostone process. Waste forms have been evaluated as to their ability to reliably produce free standing monolithic solids which are homogeneous (macroscopically), contain < 1% free standing liquids by volume and pass a water immersion test. Solidified waste form specimens were also subjected to vibratory shock testing and flame testing. Simulated oil wastes can be solidified to acceptable solid specimens having volumetric waste loadings of less than 40 volume-%. However, simulated organic liquid wastes could not be solidified into acceptable waste forms above a volumetric loading factor of about 10 volume-% using the solidification agents studied

Solidification of acidic liquid waste from 99Mo isotope production

International Nuclear Information System (INIS)

Parsons, G.J.

2001-01-01

results in the solidification of the deammoniated product in stainless steel vessels designed for long term storage. The process was developed and commissioned through sequential steps. Initial testing was conducted on natural uranium nitrate based solutions followed by similar solutions with increasing levels of trace activity derived from the stored waste. The process was commissioned on stored liquid waste in 1999 and is now a routine operation. Initial processing through the concentration phase has been successful in removing 82-95% of the original liquor volume at a throughput rate of generally 4-4.5 L/h. The ammonia content in the acid waste had arisen principally from the addition of ammonia bearing condensate from the molybdenum extraction and initial purification process. This practice of combining these two liquid wastes is no longer continued but has resulted in an inventory of historical acid waste containing small concentrations of ammonia. A deammoniation process was developed to treat batches of concentrate before solidification. This processing step has been successful in reducing NH 3 -N to less than 10ppm under controlled conditions. Nitrogen oxides (NOx gasses) are a product of this chemical process and off gas is treated through a catalytic converter. Solidification to date has resulted in a product of 0.6-2.3% of the original liquor volume (or 1.7- 5.7% of the original solution weight). The solidification takes place in thick- walled once-use stainless steel vessels. The vessel is heated in a thermic oil bath with slow continuous feed of deammoniated concentrate and withdrawal of condensate. This phase is slower with throughput rates of around 1L/h decreasing to less than 0.5L/h as processing continues. When the required amount has been added to the vessel it is further heated, resulting in a product which solidifies on cooling. When this process is complete the connections to the vessel are removed and the vessel ports plugged. The vessel is then

Solidification of radioactive waste effluents

International Nuclear Information System (INIS)

Mergan, L.M.; Cordier, J.-P.

1981-01-01

A process and apparatus for solidifying radioactive waste liquid containing dissolved and/or suspended solids is disclosed. The process includes chemically treating for pH adjustment and precipitation of solids, concentrating solids with a thin-film evaporator to provide liquid concentrate containing about 50% solids, and drying the concentrate with a heated mixing apparatus. The heated mixing apparatus includes a heated wall and working means for shearing dried concentrate from internal surfaces and subdividing dry concentrate into dry, powdery particles. The working means includes a rotor and helical means for positively advancing the concentrate and resulting dry particles from inlet to outlet of the mixing apparatus. The dry particles may also be encapsulated in a matrix material. Entrained particles in the vapor stream from the evaporator and mixer are removed in an integral particle separator and the vapor is subsequently condensed and may be recycled upstream of the thin-film evaporator. A section of the mixer may be used for mixing dry particles with the matrix material in a continuous drying and mixing sequence. A section of the mixer also may be used for mixing the treating chemical with the waste liquid

Onset of solid state mantle convection and mixing during magma ocean solidification

Science.gov (United States)

Maurice, Maxime; Tosi, Nicola; Samuel, Henri; Plesa, Ana-Catalina; Hüttig, Christian; Breuer, Doris

2017-04-01

The fractional crystallization of a magma ocean can cause the formation of a compositional layering that can play a fundamental role for the subsequent long-term dynamics of the interior, for the evolution of geochemical reservoirs, and for surface tectonics. In order to assess to what extent primordial compositional heterogeneities generated by magma ocean solidification can be preserved, we investigate the solidification of a whole-mantle Martian magma ocean, and in particular the conditions that allow solid state convection to start mixing the mantle before solidification is completed. To this end, we performed 2-D numerical simulations in a cylindrical geometry. We treat the liquid magma ocean in a parametrized way while we self-consistently solve the conservation equations of thermochemical convection in the growing solid cumulates accounting for pressure-, temperature- and, where it applies, melt-dependent viscosity as well as parametrized yield stress to account for plastic yielding. By testing the effects of different cooling rates and convective vigor, we show that for a lifetime of the liquid magma ocean of 1 Myr or longer, the onset of solid state convection prior to complete mantle crystallization is likely and that a significant part of the compositional heterogeneities generated by fractionation can be erased by efficient mantle mixing.

Sulfur polymer cement stabilization of elemental mercury mixed waste

International Nuclear Information System (INIS)

Melamed, D.; Fuhrmann, M.; Kalb, P.; Patel, B.

1998-04-01

Elemental mercury, contaminated with radionuclides, is a problem throughout the Department of Energy (DOE) complex. This report describes the development and testing of a process to immobilize elemental mercury, contaminated with radionuclides, in a form that is non-dispersible, will meet EPA leaching criteria, and has low mercury vapor pressure. In this stabilization and solidification process (patent pending) elemental mercury is mixed with an excess of powdered sulfur polymer cement (SPC) and additives in a vessel and heated to ∼35 C, for several hours, until all of the mercury is converted into mercuric sulfide (HgS). Additional SPC is then added and the mixture raised to 135 C, resulting in a homogeneous molten liquid which is poured into a suitable mold where is cools and solidifies. The final stabilized and solidified waste forms were characterized by powder X-ray diffraction, as well as tested for leaching behavior and mercury vapor pressure. During this study the authors have processed the entire inventory of mixed mercury waste stored at Brookhaven National Laboratory (BNL)

OVERVIEW OF THE HISTORY, PRESENT STATUS, AND FUTURE DIRECTION OF SOLIDIFICATION/STABILIZATION TECHNOLOGIES FOR HAZARDOUS WASTE TREATMENT

Science.gov (United States)

Solidification/stabilization (S/S) technology processes are currently being utilized in the United States to treat inorganic and organic hazardous waste and radioactive waste. These wastes are generated from operating industry or have resulted from the uncontrolled management of ...

An innovative in-situ mixing technology and its applications in the waste remediation industry

International Nuclear Information System (INIS)

Toor, I.A.; Lanter, R.

1994-01-01

An innovative in-situ remediation technology has been developed for solidification and stabilization of hazardous wastes. The system incorporates a specially designed rotary mixing head attached to the boom of a long-reach backhoe or other dirt-moving equipment. A variety of mixing-head configurations are available to treat various types of wastes, ranging from oil sludge to very dry contaminated soils containing significant amounts of large aggregates and gravel. The system has been successfully applied in the field to remediate hazardous petroleum sludge, mine tailings, and steel mill process sediments containing heavy metals (e.g., chromium, arsenic, cadmium, and lead). A very elaborate quality assurance/quality control program was implemented to ensure minimum variation in additive concentration and thorough mixing. The mixing effectiveness and reagent injection capabilities of this unit have resulted in the in-situ treatment of listed hazardous wastes to below delisting thresholds at depths in excess of 15 ft. Applications of this unit are currently being reviewed for incorporating and mixing nutrients in a bioremediation process. The new technology provides a very economical means for treatment, with excellent product quality

Solidification of metal oxide from electrokinetic-electrodialytic decontamination

Energy Technology Data Exchange (ETDEWEB)

Koo, Daeseo; Park, Uk-Ryang; Kim, Gye-Nam; Kim, Seung-Soo; Moon, Jei-Kwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

Electrokinectic-electrodialytic decontamination technology reduced 80% of the concentration of the uranium soil waste to below the concentration of self-disposal. After conducting electrokinectic-electrodialytic decontamination, more than 10% of the remainder of radioactive waste from the cathodes of electrokinectic-electrodialytic equipment were produced. To dispose of such waste, it is necessary to solidify second radioactive waste owing to the requirements of radioactive waste from public corporations. In this study, a solidification experiment was carried out using a polymer. At first, a sampling of second radioactive waste was conducted. Then, second radioactive waste and a polymer were mixed. Third, the solidified state between the second radioactive waste and polymer was checked. In our next study, an experiment for the requirements of a public radioactive waste corporation will be conducted.

High-level waste solidification system for the Western New York Nuclear Service Center

International Nuclear Information System (INIS)

Carrell, J.R.; Holton, L.K.; Siemens, D.H.

1982-01-01

A preconceptual design for a waste conditioning and solidification system for the immobilization of the high-level liquid wastes (HLLW) stored at the Western New York Nuclear Service Center (WNYNSC), West Valley, New York was completed in 1981. The preconceptual design was conducted as part of the Department of Energy's (DOE) West Valley Demonstration Project, which requires a waste management demonstration at the WNYNSC. This paper summarizes the bases, assumptions, results and conclusions of the preconceptual design study

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

International Nuclear Information System (INIS)

Anon.

1994-01-01

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

SRTC criticality technical review: Nuclear Criticality Safety Evaluation 93-18 Uranium Solidification Facility's Waste Handling Facility

International Nuclear Information System (INIS)

Rathbun, R.

1993-01-01

Separate review of NMP-NCS-930058, open-quotes Nuclear Criticality Safety Evaluation 93-18 Uranium Solidification Facility's Waste Handling Facility (U), August 17, 1993,close quotes was requested of SRTC Applied Physics Group. The NCSE is a criticality assessment to determine waste container uranium limits in the Uranium Solidification Facility's Waste Handling Facility. The NCSE under review concludes that the NDA room remains in a critically safe configuration for all normal and single credible abnormal conditions. The ability to make this conclusion is highly dependent on array limitation and inclusion of physical barriers between 2x2x1 arrays of boxes containing materials contaminated with uranium. After a thorough review of the NCSE and independent calculations, this reviewer agrees with that conclusion

Low-temperature setting phosphate ceramics for stabilization of DOE problem low level mixed-waste: I. Material and waste form development

International Nuclear Information System (INIS)

Singh, D.; Wagh, A.; Knox, L.; Mayberry, J.

1994-03-01

Chemically bonded phosphate ceramics are proposed as candidates for solidification and stabilization of some of the open-quotes problemclose quotes DOE low-level mixed wastes at low-temperatures. Development of these materials is crucial for stabilization of waste streams which have volatile species and any use of high-temperature technology leads to generation of off-gas secondary waste streams. Several phosphates of Mg, Al, and Zr have been investigated as candidate materials. Monoliths of these phosphates were synthesized using chemical routes at room or slightly elevated temperatures. Detailed physical and chemical characterizations have been conducted on some of these phosphates to establish their durability. Magnesium ammonium phosphate has shown to possess excellent mechanical and as well chemical properties. These phosphates were also used to stabilize a surrogate ash waste with a loading ranging from 25-35 wt.%. Characterization of the final waste forms show that waste immobilization is due to both chemical stabilization and physical encapsulation of the surrogate waste which is desirable for waste immobilization

NOCHAR Polymers: An Aqueous and Organic Liquid Solidification Process for Cadarache LOR (Liquides Organiques Radioactifs) - 13195

International Nuclear Information System (INIS)

Vaudey, Claire-Emilie; Renou, Sebastien; Porco, Julien; Kelley, Dennis; Cochaud, Chantal; Serrano, Roger

2013-01-01

To handle the Very Low Level Waste (VLLW) and the Low Level Waste (LLW) in France, two options can be considered: the incineration at CENTRACO facility and the disposal facility on ANDRA sites. The waste acceptance in these radwaste routes is dependent upon the adequacy between the waste characteristics (physical chemistry and radiological) and the radwaste route specifications. If the waste characteristics are incompatible with the radwaste route specifications (presence of significant quantities of chlorine, fluorine, organic component etc or/and high activity limits), it is necessary to find an alternative solution that consists of a waste pre-treatment process. In the context of the problematic Cadarache LOR (Liquides Organiques Radioactifs) waste streams, two radioactive scintillation cocktails have to be treated. The first one is composed of organic liquids at 13.1 % (diphenyloxazol, mesitylene, TBP, xylene) and water at 86.9 %. The second one is composed of TBP at 8.6 % and water at 91.4 %. They contain chlorine, fluorine and sulphate and have got alpha/beta/gamma spectra with mass activities equal to some kBq.g -1 . Therefore, tritium is present and creates the second problematic waste stream. As a consequence, in order for disposal acceptance at the ANDRA site, it is necessary to pre-treat the waste. The NOCHAR polymers as an aqueous and organic liquid solidification process seem to be an adequate solution. Indeed, these polymers constitute an important variety of products applied to the treatment of radioactive aqueous and organic liquids (solvent, oil, solvent/oil mixing etc) and sludge through a mechanical and chemical solidification process. For Cadarache LOR, N910 and N960 respectively dedicated to the organic and aqueous liquids solidification are considered. With the N910, the organic waste solidification occurs in two steps. As the organic liquid travels moves through the polymer strands, the strands swell and immobilise the liquid. Then as the

Solidification of low and medium level wastes in bitumen at Barsebaeck nuclear power station

International Nuclear Information System (INIS)

Harfors, C.

1979-01-01

Operating experience is presented from 4 years of bitumen solidification of wastes coming from two boiling water reactors. Methods used to sample, analyse and document the wastes are described. Transport and storage methods without remote handling have been adopted. The risk of fire is discussed and a description is given of the measures taken for fire protection. (author)

Method of storing solidification products

International Nuclear Information System (INIS)

Tani, Yutaro.

1985-01-01

Purpose: To enable to efficiently and satisfactorily cool and store solidification products of liquid wastes generated from the reactor spent fuel reprocessing process by a simple facility. Method: Liquid wastes generated from the reactor spent fuel reprocessing process are caused to flow from the upper opening to the inside of a spherical canistor. The opening of the spherical canistor is welded with a lid by a remote control and the liquid wastes are tightly sealed within the spherical canistor as glass solidification products. Spherical canistors having the solidification products tightly sealed therein are sent into and stored in a hopper by the remote control. Further, a blower is driven upon storing to suck cooling air from the cooling air intake port to the inside of the hopper to absorb the decay heat of radioactive materials in the solidification products and the air is discharged from the duct and through the stack to the atmosphere. (Kawakami, Y.)

Mixed waste management options

International Nuclear Information System (INIS)

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

1992-01-01

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

Solidification of radioactive waste solutions by pelletization technique

International Nuclear Information System (INIS)

Akbar, A.H.; Koester, R.; Rudolph, G.

1980-04-01

A possible way of performing the cement fixation of radioactive wastes is the incorporation into cement pellets on a pan pelletizer, followed by embedding the pellets into an inactive cement matrix. This procedure is suitable for various types of waste, particularly for medium level liquid wastes, and can be used both at drum disposal and at in-situ solidification. This report describes some initial studies on the pelletization technique using a laboratory pelletizer. Formation and size of the pellets have been found to be determined by speed, angle, and load of the pan, ratio and mode of addition of the liquid and solid components, ect. Pellets in various compositions have been produced from cement and water or simulated waste solution, in some cases with the addition of bentonite for improving cesium retention. Some mechanical properties of the pellets such as fall height of fresh pellets, development of hardness (crush test), impact and abrasion resistance, have been determined. Some preliminary experiments were done on backfilling the void space between the pellets - about 40 per cent of the bulk volume - with cement grouts of appropriate compositions. (orig.) [de

Purification and solidification of reactor wastes at a Canadian nuclear generating station

International Nuclear Information System (INIS)

Buckley, L.P.; Burt, D.A.

1981-01-01

The study aimed at development and demonstration of volume reduction and solidification of CANDU reactor wastes has been underway at Chalk River Nuclear Laboratories in the Province of Ontario, Canada. The study comprises membrane separation processes, evaporator appraisal and immobilization of concentrated wastes in bitumen. This paper discusses the development work with a wiped-film evaporator and the successful completion of demonstration tests at Douglas Point Nuclear Generating Station. Heavy water from the moderator system was purified and wastes arising from pump bowl decontamination were immobilized in bitumen with the wiped-film evaporator that was used in the development tests at Chalk River

Mixed Waste Management Facility

International Nuclear Information System (INIS)

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

1993-08-01

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

Stabilization of high and low solids Consolidated Incinerator Facility (CIF) waste with super cement

International Nuclear Information System (INIS)

Walker, B.W.

2000-01-01

This report details solidification activities using selected Mixed Waste Focus Area technologies with the High and Low Solid waste streams. Ceramicrete and Super Cement technologies were chosen as the best possible replacement solidification candidates for the waste streams generated by the SRS incinerator from a list of several suggested Mixed Waste Focus Area technologies. These technologies were tested, evaluated, and compared to the current Portland cement technology being employed. Recommendation of a technology for replacement depends on waste form performance, process flexibility, process complexity, and cost of equipment and/or raw materials

Electric melting furnace for waste solidification

International Nuclear Information System (INIS)

Masaki, Toshio.

1990-01-01

To avoid electric troubles or reduction of waste processing performance even when platinum group elements are contained in wastes to be applied with glass solidification. For this purpose, a side electrode is disposed to the side wall of a melting vessel and a central electrode serving as a counter electrode is disposed about at the center inside the melting vessel. With such a constitution, if conductive materials are deposited at the bottom of the furnace or the bottom of the melting vessel, heating currents flow selectively between the side electrode and the central electrode. Accordingly, no electric currents flow through the conductive deposits thereby enabling to prevent abnormal heating in the bottom of the furnace. Further, heat generated by electric supply between the side electrode and the central electrode is supplied efficiently to raw material on the surface of the molten glass liquid to improve the processing performance. Further, disposition of the bottom electrode at the bottom of the furnace enables current supply between the central electrode and the bottom electrode to facilitate the temperature control for the molten glass in the furnace than in the conventional structure. (I.S.)

Cementation and solidification of miscellaneous mixed wastes at the Rocky Flats Environmental Technology Site

International Nuclear Information System (INIS)

Phillips, J.A.; Semones, G.B.

1995-01-01

The Rocky Flats Environmental Technology Site produces a variety of wastes which are amenable to micro-encapsulation in cement Portland cement is an inexpensive and readily available material for this application. The Waste Projects (WP) group at Rocky Flats evaluated cementation to determine its effectiveness in encapsulating several wastes. These included waste analytical laboratory solutions, incinerator ash, hydroxide precipitation sludge, and an acidic solution from the Delphi process (a chemical oxidation technology being evaluated as an alternative to incineration). WP prepared surrogate wastes and conducted designed experiments to optimize the cement formulation for the waste streams. These experiments used a Taguchi or factorial experimental design, interactions between the variables were also considered in the testing. Surrogate waste samples were spiked with various levels of each of six Resource Conservation and Recovery Act (RCRA) listed metals (Cd, Cr, Ba, Pb, Ni, and Ag), cemented using the optimized formulation, and analyzed for leach resistance using the Toxicity Characteristic Leaching Procedure (TCLP). The metal spike levels chosen were based on characterization data, and also based on an estimate of the highest levels of contaminants suspected in the waste. This paper includes laboratory test results for each waste studied. These include qualitative observations as well as quantitative data from TCLP analyses and environmental cycling studies. The results from these experiments show that cement stabilization of the different wastes can produce final waste forms which meet the current RCRA Land Disposal Restriction (LDR) requirements. Formulations that resulted in LDR compliant waste forms are provided. The volume increases associated with cementation are also lower than anticipated. Future work will include verification studies with actual mixed radioactive waste as well as additional formulation development studies on other waste streams

Method of solidifying powderous wastes

International Nuclear Information System (INIS)

Kakimoto, Akira; Miyake, Takashi; Sato, Shuichi; Inagaki, Yuzo.

1985-01-01

Purpose: To improve the properties of solidification products, in the case of solidifying powderous wastes with thermosetting resins. Method. A solvent for the solution of the thermosetting resin is admixed with the powderous wastes into a paste-like form prior to adding the resin to the wastes, which are then mixed with the resin solution. As the result, those solidification products having the specific gravity and the compression strength more excellent than those of the conventional ones, and much higher than the reference values can be obtained. (Kamimura, M.)

Mixed waste: Proceedings

International Nuclear Information System (INIS)

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

1993-01-01

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

Mixed waste: Proceedings

Energy Technology Data Exchange (ETDEWEB)

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

1993-12-31

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

Mixed Waste Integrated Program: A technology assessment for mercury-containing mixed wastes

International Nuclear Information System (INIS)

Perona, J.J.; Brown, C.H.

1993-03-01

The treatment of mixed wastes must meet US Environmental Protection Agency (EPA) standards for chemically hazardous species and also must provide adequate control of the radioactive species. The US Department of Energy (DOE) Office of Technology Development established the Mixed Waste Integrated Program (MWIP) to develop mixed-waste treatment technology in support of the Mixed Low-Level Waste Program. Many DOE mixed-waste streams contain mercury. This report is an assessment of current state-of-the-art technologies for mercury separations from solids, liquids, and gases. A total of 19 technologies were assessed. This project is funded through the Chemical-Physical Technology Support Group of the MWIP

Guidelines for mixed waste minimization

International Nuclear Information System (INIS)

Owens, C.

1992-02-01

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

Preliminary evaluation of alternative waste form solidification processes. Volume I. Identification of the processes

International Nuclear Information System (INIS)

Treat, R.L.; Nesbitt, J.F.; Blair, H.T.; Carter, J.G.; Gorton, P.S.; Partain, W.L.; Timmerman, C.L.

1980-04-01

This document contains preconceptual design data on 11 processes for the solidification and isolation of nuclear high-level liquid wastes (HLLW). The processes are: in-can glass melting (ICGM) process, joule-heated glass melting (JHGM) process, glass-ceramic (GC) process, marbles-in-lead (MIL) matrix process, supercalcine pellets-in-metal (SCPIM) matrix process, pyrolytic-carbon coated pellets-in-metal (PCCPIM) matrix process, supercalcine hot-isostatic-pressing (SCHIP) process, SYNROC hot-isostatic-pressing (SYNROC HIP) process, titanate process, concrete process, and cermet process. For the purposes of this study, it was assumed that each of the solidification processes is capable of handling similar amounts of HLLW generated in a production-sized fuel reprocessing plant. It was also assumed that each of the processes would be enclosed in a shielded canyon or cells within a waste facility located at the fuel reprocessing plant. Finally, it was assumed that all of the processes would be subject to the same set of regulations, codes and standards. Each of the solidification processes converts waste into forms that may be acceptable for geological disposal. Each process begins with the receipt of HLLW from the fuel reprocessing plant. In this study, it was assumed that the original composition of the HLLW would be the same for each process. The process ends when the different waste forms are enclosed in canisters or containers that are acceptable for interim storage. Overviews of each of the 11 processes and the bases used for their identification are presented in the first part of this report. Each process, including its equipment and its requirements, is covered in more detail in Appendices A through K. Pertinent information on the current state of the art and the research and development required for the implementation of each process are also noted in the appendices

Polymer Solidification Technology - Technical Issues and Challenges

International Nuclear Information System (INIS)

Jensen, Charles; Kim, Juyoul

2010-01-01

Many factors come into play, most of which are discovered and resolved only during full-scale solidification testing of each of the media commonly used in nuclear power plants. Each waste stream is unique, and must be addressed accordingly. This testing process is so difficult that Diversified's Vinyl Ester Styrene and Advanced Polymer Solidification are the only two approved processes in the United States today. This paper summarizes a few of the key obstacles that must be overcome to achieve a reliable, repeatable process for producing an approved Stable Class B and C waste form. Before other solidification and encapsulation technologies can be considered compliant with the requirements of a Stable waste form, the tests, calculations and reporting discussed above must be conducted for both the waste form and solidification process used to produce the waste form. Diversified's VERI TM and APS TM processes have gained acceptance in the UK. These processes have also been approved and gained acceptance in the U. S. because we have consistently overcome technical hurdles to produce a complaint product. Diversified Technologies processes are protected intellectual property. In specific instances, we have patents pending on key parts of our process technology

Solidification of low-level radioactive liquid waste using a cement-silicate process

International Nuclear Information System (INIS)

Grandlund, R.W.; Hayes, J.F.

1979-01-01

Extensive use has been made of silicate and Portland cement for the solidification of industrial waste and recently this method has been successfully used to solidify a variety of low level radioactive wastes. The types of wastes processed to date include fuel fabrication sludges, power reactor waste, decontamination solution, and university laboratory waste. The cement-silicate process produces a stable solid with a minimal increase in volume and the chemicals are relatively inexpensive and readily available. The method is adaptable to either batch or continuous processing and the equipment is simple. The solid has leaching characteristics similar to or better than plain Portland cement mixtures and the leaching can be further reduced by the use of ion-exchange additives. The cement-silicate process has been used to solidify waste containing high levels of boric acid, oils, and organic solvents. The experience of handling the various types of liquid waste with a cement-silicate system is described

Feasibility study of solidification for low-level liquid waste generated by sulfuric acid elution treatment of spent ion exchange resin

International Nuclear Information System (INIS)

Asano, Takashi; Kawasaki, Tooru; Higuchi, Natsuko; Horikawa, Yoshihiko

2007-01-01

Low-level liquid waste with relatively high levels of radioactivity is generated by the sulfuric acid elution treatment of spent ion exchange resin used in water purification systems of nuclear power plants. We studied cement-like solidification process for this type waste that contains a high concentration of sodium sulfate. For this type waste, it is important that the sulfate ion should not dissolve from the solid waste because it forms ettringite on reaction with minerals in the concrete, and this leads to cracking during repository storage. It is also preferable that the pH of pore water of the solid waste be low, because the bentonite of the repository changes in quality on exposure to alkaline solution. Our solidification process has two procedures: conversion into insoluble sulfate from sodium sulfate (CIS) and formation of low pH cement-like solid (FLS). In the CIS procedure, BaSO 4 precipitation occurs with addition of Ba(OH) 2 ·8H 2 O to the liquid waste when the Ba/SO 4 molar ratio > 1. In the FLS procedure, silica fume and blast furnace slag are added to the liquid wastes containing Ba S O 4 precipitate. The CIS reaction yield is over 98% and the pH of pore water of the solid waste is 11.5 or less. Therefore, we think that our solidification process is one of the best methods for treating liquid waste that contains a high concentration of sodium sulfate. (author)

Alternative waste management concept for medium and low level wastes by in-situ solidification

International Nuclear Information System (INIS)

Kraemer, R.

1982-01-01

Since 1976, a German R and D project has been carried out to find an alternative concept for the treatment and disposal of MLW and LLW arising mainly in the planned German reprocessing plant and other nuclear facilities (LWR, fuel fabrication, R and D establishments). The main feature of this concept is an in-situ solidification of preconditioned waste granules in large salt caverns located in the deep geological underground, thus avoiding such non-radioactive ballast as lost concrete shielding and container material. (orig./RW)

Liquid low-level waste (LLLW) solidification at Oak Ridge National Laboratory

International Nuclear Information System (INIS)

Schultz, R.M.; Monk, T.H.; duMont, S.P.; Helms, R.E.; Keigan, M.V.; Morris, M.I.

1987-01-01

In general, the presentation describes the disposal of liquid, low-level (radioactive) waste (LLLW) by the hydrofracture process at Oak Ridge National Laboratory until 1984, when it was shut down due to regulatory concerns and operational anomalies. As a result of this, about 400,000 gallons of concentrated LLLW and 50,000 gallons of transuranic waste-bearing sludges have accumulated in the active, double-contained tank system which is reaching its operational capacity. A major initiative to develop an alternative means of LLLW treatment and disposal was begun about two years ago. This presentation summarizes the implementation strategy of the most likely process options. The strategy is being developed in two phases; a near-term flowsheet and a long-term or reference flowsheet. First, reliable and fully demonstrated commercial, cement solidification systems are being assessed for execution of an initial 50,000 gallon campaign in 1988. Second, development is under way to determine viable sludge separation, LLLW decontamination and solidification alternatives. A flowsheet analysis and cost study is being conducted by a consultant to ensure proper consideration of process developments at other sites. It is estimated that, depending upon funding requirements, it could take up to six years to implement the reference flowsheet

Development and evaluation of polyethylene as solidification agent for low-level waste

International Nuclear Information System (INIS)

Franz, E.M.; Colombo, P.

1986-09-01

A polyethylene solidification process, using an extrusion system, has been developed for the immobilization of dry wastes resulting from volume reduction technologies. Ease of processibility and high packing efficiencies were obtained through the use of low-density polyethylene (0.917 to 0.924 g/cm 3 ) with melt indices from 2.0 to 55.0 g/10 min. Maximum waste loadings of 70 wt % sodium sulfate, 50 wt % boric acid, 40 wt % incinerator ash and 65 wt % ion exchnge were obtained. A series of tsts were conducted to assess the acceptability of polyethylene waste forms to meet the requirements of 10 CFR 61. Based on test results and process control considerations, optimal waste loadings of 70 wt % sodium sulfate, 50 wt % boric acid, 40 wt % incinerator ash and 30 wt % ion exchange resins are recommended

Method of solidifying radioactive wastes

International Nuclear Information System (INIS)

Fukazawa, Tetsuo; Ootsuka, Masaharu; Uetake, Naoto; Ozawa, Yoshihiro.

1984-01-01

Purpose: To prepare radioactive solidified wastes excellent in strength, heat resistance, weather-proof, water resistance, dampproof and low-leaching property. Method: A hardening material reactive with alkali silicates to form less soluble salts is used as a hardener for alkali silicates which are solidification filler for the radioactive wastes, and mixed with cement as a water absorbent and water to solidify the radioactive wastes. The hardening agent includes, for example, CaCO 3 , Ca(ClO 4 ) 2 , CaSiF 6 and CaSiO 3 . Further, in order to reduce the water content in the wastes and reduce the gap ratio in the solidification products, the hardener adding rate, cement adding rate and water content are selected adequately. As the result, solidification products can be prepared with no deposition of easily soluble salts to the surface thereof, with extremely low leaching of radioactive nucleides. (Kamimura, M.)

UJV line for research into radioactive wastes solidification

International Nuclear Information System (INIS)

Neumann, L.; Feist, I.; Kepak, F.; Nachmilner, L.; Napravnik, J.; Novak, M.; Pecak, V.; Vojtech, O.

1985-01-01

An experimental line with a capacity of 0.01 m 3 /h was developed and built for research of the solidification of liquid radioactive wastes at the Nuclear Research Institute. The line allows the research and pilot plant testing of processes based on vitrification but also on other procedures including calcination. It consists of a horizontal calciner, a resistance melting unit, a homogenization device for research into cementation of the calcinate, and equipment for the disposal of gaseous emissions. The facility is provided with a control console which allows remote control and the control of all basic operating parameters. The design of the line allows its eventual completion with other equipment. (Z.M.)

Mixed Waste Working Group report

International Nuclear Information System (INIS)

1993-01-01

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

Feasibility study of solidification for low-level liquid waste generated by sulfuric acid elution treatment of spent ion exchange resin

International Nuclear Information System (INIS)

Asano, Takashi; Kawasaki, Tooru; Higuchi, Natsuko; Horikawa, Yoshihiko

2008-01-01

We studied cement-like solidification process for low-level liquid waste with relatively high levels of radioactivity that contains a high concentration of sodium sulfate. For this type waste, it is important that the sulfate ion should not dissolve from the solid waste because it forms ettringite on reaction with minerals in the concrete of the planned repository, and this leads to cracking during repository storage. It is also preferable that the pH of the pore water of the solid waste be low, because the bentonite of the repository changes in quality on exposure to alkaline solution. Therefore, the present solidification process has two procedures: conversion into insoluble sulfate from sodium sulfate (CIS) and formation of low pH cement-like solid (FLS). In the CIS procedure, BaSO 4 precipitation occurs with addition of Ba(OH) 2 ·8H 2 O to the liquid waste. In the FLS procedure, silica fume and blast furnace slag are added to the liquid waste containing BaSO 4 precipitate. We show the range of appropriate Ba/SO 4 molar ratio is from 1.1 to 1.5 in the present solidification process by leaching tests for some kinds of solid waste samples. The CIS reaction yield is over 98% at a typical CIS condition, i.e. Ba/SO 4 molar ratio=1.3, reaction temperature=60 deg C, and time=3 hr. (author)

Solidification of high level liquid waste (HLLW) into ceramics by sintering process

International Nuclear Information System (INIS)

Masuda, Sumio; Oguino, Naohiko; Tsunoda, Naomi; O-oka, Kazuo; Ohta, Takao.

1979-01-01

One of the alternatives to vitrified solid which is acceptable and well characterized for storing radioactive HLLW with desirable long-term stability is ceramics. On the other hand, the solidification process of highly radioactive wastes should be simple and suitable for continuous production. On the above described basis, the authors have made preliminary study on the production of sintered ceramics by the addition of several oxides to HLLW. The simulated waste and additive oxides were pressed in a mold to make the preforms of 50 mm diameter and 10 to 15 mm thick. The preforms were then normally sintered at temperature from 1000 to 1400 deg C for 2 to 4 hours. The characterization of the sintered solids revealed the following facts. (1) X-ray diffraction analysis showed that the expected crystals were formed by normal-sintering as well as by hot-pressing. (2) The bulk density of the ceramics by normal-sintering was around 90 to 95% of the assumed theoretical values. (3) The leach-rate of the solids was affected by the bulk density. (4) Other properties of the solids, such as thermal expansion or thermal conductivity, are dominantly determined by those of main crystals in the solids. Sintering process is generally simple and productive as far as normal sintering is concerned. However, hot-pressing is an intermittent and time consuming process. From this fact, the authors intended to adopt the normal sintering process for the ceramic solidification of high level liquid wastes. (Wakatsuki, Y.)

Nuclear waste disposal: alternatives to solidification in glass proposed

International Nuclear Information System (INIS)

Kerr, R.A.

1979-01-01

More than a quarter-million cubic meters of liquid radioactive wastes are now being held at government installations awaiting final disposal. During the past 20 years, the disposal plan of choice has been to incorporate the 40 to 50 radioactive elements dissolved in liquid wastes into blocks of glass, seal the glass in metal canisters, and insert the canisters into deep, geologically stable salt beds. Over the last few years, some geologists and materials scientists have become concerned that perhaps not enough is known yet about the interaction of waste, container, and salt (or any rock) to have a reasonable assurance that the hazardous wastes will be contained successfully. The biggest advantage of glass at present is the demonstrated practicality of producing large, highly radioactive blocks of it. The frontrunner as a successor to glass is ceramics, which are nonmetallic crystalline materials formed at high temperature, such as chinaware or natural minerals. An apparent advantage of ceramics is that they already have an ordered atomic structure, whose properties can be tailored to a particular waste element and to conditions of a specific disposal site. A ceramic tailored for waste disposal called supercalcine-ceramic has been developed. It was emphasized that the best minerals for waste solidification may be those that have proved most stable under natural conditions over geologic time. Disadvantage to ceramics are radiation damage and transmutation. However, it is now obvious that some ceramics are more stable than glass under certain conditions. Metal-encapsulated ceramic, called cermet, is being developed as a waste form. Cermets are considerably more resistant at 100 0 C than a borosilicate waste glass. Researchers are now testing prospective waste forms under the most extreme conditions that might prevail in a waste disposal site

Study of stabilization/solidification processes (of solid porous wastes) based on hydraulic or bituminous binders; Etude des procedes de stabilisation/solidification (des dechets solides poreux) a base de liants hydrauliques ou de liants bitumineux

Energy Technology Data Exchange (ETDEWEB)

Sing-Teniere, Ch.

1998-02-01

The first part of this thesis presents the regulatory framework and the technical context linked with the study of stabilized/solidified wastes and with the evaluation of stabilization/solidification processes. A presentation of the two type of ultimate wastes under study (a used catalyst and an activated charcoal) and an analysis of the processes is given. The second part is devoted to the experimental characterization of both types of porous wastes. The third part deals with the processing of such wastes using an hydraulic binder. The study stresses on both on the stabilization/solidification efficiency of the process and on the conditions of its implementation. The same work is made for a process that uses a bituminous binder. Some choice criteria for the selection of the better process are deduced from the examination of the overall data collected. The waste characterization methodology is applied six times: two times for the raw wastes, two times for the same wastes processed with an hydraulic binder, and two times for the same wastes processed with a bituminous binder. (J.S.)

AECL's mixed waste management program

International Nuclear Information System (INIS)

Peori, R.; Hulley, V.

2006-01-01

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

Overview of mixed waste issues

International Nuclear Information System (INIS)

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

1986-01-01

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

Solidification Tests Conducted on Transuranic Mixed Oil Waste (TRUM) at the Rocky Flats Environmental Technology Site (RFETS)

International Nuclear Information System (INIS)

Brunkow, W. G.; Campbell, D.; Geimer, R.; Gilbreath, C.; Rivera, M.

2002-01-01

Rocky Flats Environmental Technology Site (RFETS) near Golden, Colorado is the first major nuclear weapons site within the DOE complex that has been declared a full closure site. RFETS has been given the challenge of closing the site by 2006. Key to meeting this challenge is the removal of all waste from the site followed by site restoration. Crucial to meeting this challenge is Kaiser-Hill's (RFETS Operating Contractor) ability to dispose of significant quantities of ''orphan'' wastes. Orphan wastes are those with no current disposition for treatment or disposal. Once such waste stream, generically referred to as Transuranic oils, poses a significant threat to meeting the closure schedule. Historically, this waste stream, which consist of a variety of oil contaminated with a range of organic solvents were treated by simply mixing with Environstone. This treatment method rendered a solidified waste form, but unfortunately not a TRUPACT-II transportable waste. So for the last ten years, RFETS has been accumulating these TRU oils while searching for a non-controversial treatment option

Plastic solidification system at Hamaoka Nuclear Power Station

International Nuclear Information System (INIS)

Okajima, Hiroyuki; Iokibe, Hiroyuki; Tsukiyama, Shigeru; Suzuki, Michio; Yamaguchi, Masato

1987-01-01

In Unit 1 and 2 of the Hamaoka Nuclear Power Station, radioactive waste was previously solidified in cement. By this method, the quantity of waste thus treated is relatively small, resulting in large number of the solidified drums. In order to solve this problem, the solidification facility using a thermosetting resin was employed, which is in operation since January 1986 for Unit 1, 2 and 3. As compared with the cement solidification, the solidified volume of concentrated liquid is about 1/12 and of spent-resin slurry is about 1/4 in plastic solidification. The following are described: course leading to the employment, the plastic solidification facility, features of the facility, operation results so far with the facility, etc. (Mori, K.)

Managing a mixed waste program

International Nuclear Information System (INIS)

Koch, J.D.

1994-01-01

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

Spray solidification of nuclear waste

International Nuclear Information System (INIS)

Bonner, W.F.; Blair, H.T.; Romero, L.S.

1976-08-01

The spray calciner is a relatively simple machine. Operation is simple and is easily automated. Startup and shutdown can be performed in less than an hour. A wide variety of waste compositions and concentrations can be calcined under easily maintainable conditions. Spray calcination of high-level and mixed high- and intermediate-level liquid wastes has been demonstrated. Waste concentrations of from near infinite dilution to less than 225 liters per tonne of fuel are calcinable. Wastes have been calcined containing over 2M sodium. Feed concentration, composition, and flowrate can vary rapidly by over a factor of two without requiring operator action. Wastes containing mainly sodium cations can be spray calcined by addition of finely divided silica to the feedstock. A remotely replaceable atomizing nozzle has been developed for use in plant-scale equipment. Calciner capacity of over 75 l/h has been demonstrated in pilot-scale equipment. Sintered stainless steel filters are effective in deentraining over 99.9 percent of the solids that result from calcining the feedstock. The volume of recycle required from the effluent treatment system is very small. Vibrator action maintains the calcine holdup in the calciner at less than 1 kg. Successful remote operation and maintenance of a heated-wall spray calciner have been demonstrated while processing high-level waste. Radionuclide volatilization was acceptably low

Volume reduction and solidification of liquid and solid low-level radioactive waste

International Nuclear Information System (INIS)

May, J.R.

1979-01-01

This paper presents a brief background of the development of a method of radioactive waste volume reduction using a unique fluidized bed calciner/incinerator. The volume reduction system is capable of processing a variety of liquid chemical wastes, spent ion exchange resin beads, filter treatment sludges, contaminated lubricating oils, and miscellaneous combustible solids such as paper, rags, protective clothing, wood, etc. All of these wastes are processed in one chemical reaction vessel. Detailed process data is presented that shows the system is capable of reducing the total volume of disposable radioactive waste generated by light water reactors by a factor of 10. Equally important to reducing the volume of power reactor radwaste is the final form of the stored or disposable radwaste. This paper also presents process data related to a new radwaste solidification system, presently being developed, that is particularly suited for immobilizing the granular solids and ashes resulting from volume reduction by calcination and/or incineration

WasteChem Corporation's Volume Reduction and Solidification (VRS) system for low-level radwaste treatment: Final report

International Nuclear Information System (INIS)

1988-01-01

Since 1965, low and medium level radwastes from nuclear power stations, reprocessing plants and nuclear research centers have been stabilized using the Volume Reduction and Solidification (VRS) system. The VRS system uses an extruder/evaporator to evaporate the liquids from waste influents, while simultaneously incorporating the remaining radioactive solids in an asphalt binder. In the period 1965 to 1986 a minimum of 700,000 cubic feet of wastes have been processed with the VRS system. This report provides current operating data from various systems including the volume reduction factors achieved, and the progress of start-ups in the US. The report also provides previously unpublished experience with mixed wastes including uranium raffinate and nitrate-bearing sludges from surface impoundments. VRS systems in the US are currently operating at the Palisades and Hope Creek nuclear stations. These systems produce a variety of waste types including boric acid, bead resin, sodium sulfate and powdered resins. There are three start-ups of VRS systems scheduled in the US in 1987. These systems are at Fermi 2, Seabrook, and Nine Mile Point 2. Overseas, the startup of new systems continues with three VRS process lines coming on-line at the LaHague Reprocessing Center in France in 1986 and a start-up scheduled for 1987 at the Laguna Verde plant in Mexico. The US systems are operating continuously and with little required maintenance. Data on maintenance and the operator exposure are provided in this report. 6 refs., 11 figs., 13 tabs

Mixed wasted integrated program: Logic diagram

International Nuclear Information System (INIS)

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

1994-01-01

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

Mixed wasted integrated program: Logic diagram

Energy Technology Data Exchange (ETDEWEB)

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

1994-11-30

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

Management of metal-bearing industrial solid waste by stabilization/solidification process

Energy Technology Data Exchange (ETDEWEB)

Sunitha, C.; Palanivelu, K. [Anna University, Chennai (India). Centre for Environmental Studies

2005-07-01

Metal-bearing sludge from an electroplating industry was immobilised by the solidification stabilisation treatment method. Reduction of the leachability of metals from the waste was studied in different combinations of waste and additives - cement, lime and fly ash. The study revealed that the optimum proportion for cement: metal hydroxide sludge: fly ash as 1:2:2 is the best. The encapsulation efficiency calculated for the metals such as Cu, Cr, Ni, Pb, and Zn was above 92%. The unconfined compressive strength (UCS) for the developed block was found to be 11.5 kg/cm{sup 2} after curing. The toxicity characteristic leach test (TCLP) test reveals that the heavy metal content in the leachate was well below the maximum permissible limit of WHO drinking water standard. 10 refs., 6 tabs.

Method of solidifying and disposing radioactive waste plastic

International Nuclear Information System (INIS)

Matsuura, Hiroyuki; Yasumura, Keijiro

1981-01-01

Purpose: To solidify radioactive waste as it is with plastic by forming a W/O (Water-in-Oil) emulsion with the radioactive waste and a plastic solidifier, and treating it with a polymerization starting agent, an accelerator, and the like. Method: A predetermined amount of alkaline substance such as sodium hydroxide, triethanol, or the like is added quantitatively to radioactive waste and it is mixed by an agitator. A predetermined amount of solidifier such as unsaturated polyester or the like is added to the mixture and it is further mixed by the agitator to form a stable W/O emulsion. Subsequently, predetermined amounts of polymerization starting agent such as methyl ethyl ketone peroxide and polymerization accelerator such as cobalt naphthenate or the like are added thereto, the mixture is mixed, and is then allowed to stand for at room temperature for the plastic solidification thereof. No reaction occurs after the solidification. (Sekiya, K.)

Proceedings of the 1991 Joint International Waste Management Conference

International Nuclear Information System (INIS)

1991-01-01

This proceedings contains articles of 1991 joint international waste management conference. It was held on October 21-23, 1991 in Seoul, Korea. The main subject titles are as follows: national waste management programs, waste management in developing countries, incineration - development and experience, site characterization and performance assessment, waste disposal, decontamination and decommissioning, waste solidification and waste form, radioactive waste processing, mixed waste and others (Yi, J. H.)

Regulatory aspects of mixed waste

International Nuclear Information System (INIS)

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

1990-01-01

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

Solidification/stabilization of technetium in cement-based grouts

International Nuclear Information System (INIS)

Gilliam, T.M.; Bostick, W.D.; Spence, R.D.; Shoemaker, J.L.

1990-01-01

Mixed low-level radioactive and chemically hazardous process treatment wastes from the Portsmouth Gaseous Diffusion Plant are stabilized by solidification in cement-based grouts. Conventional portland cement and fly ash grouts have been shown to be effective for retention of hydrolyzable metals (e.g., lead, cadmium, uranium and nickel) but are marginally acceptable for retention of radioactive Tc-99, which is present in the waste as the highly mobile pertechnate anion. Addition of ground blast furnace slag to the grout is shown to reduce the leachability of technetium by several orders of magnitude. The selective effect of slag is believed to be due to its ability to reduce Tc(VII) to the less soluble Tc(IV) species. 12 refs., 4 tabs

Mixed Waste Focus Area - Waste form initiative

International Nuclear Information System (INIS)

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

1998-01-01

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

The influence of radioactive waste solidification methods and storage conditions on the migration and dispersion of radionuclides in the terrestrial environment

International Nuclear Information System (INIS)

Golinski, M.; Ferenc, M.; Tomczak, W.; Cholerzynski, A.

1979-01-01

In the first part of the paper a survey of literature data on the methods and apparatus used for solidification of low- and intermediate-level radioactive wastes is done and the methods for the determination of the solidification product properties are discussed. The second part of the paper contains the experimental leachability data for 60 Co, 90 Sr and 137 Cs from simulated radioactive waste solidification products obtained with the help of bitumen, cement and urea formaldehyde resin. The leachability of radionuclides from the bituminization products decreases in the following order: 137 Cs 90 Sr 60 Co while that form concrete and the urea formaldehyde resin blocks as follows: 137 Cs 60 Co 90 Sr. A very good resistance of bitumen blocks against changeable atmospheric conditions and saline has been observed. The results obtained show that bitumen is the best binder while urea formaldehyde resin is the worst. (author)

Mixed waste, preparing for 1996

International Nuclear Information System (INIS)

Duke, D.L.

1995-01-01

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

Mercury removal from solid mixed waste

International Nuclear Information System (INIS)

Gates, D.D.; Morrissey, M.; Chava, K.K.; Chao, K.

1994-01-01

The removal of mercury from mixed wastes is an essential step in eliminating the temporary storage of large inventories of mixed waste throughout the Department of Energy (DOE) complex. Currently thermal treatment has been identified as a baseline technology and is being developed as part of the DOE Mixed Waste Integrated Program (MWIP). Since thermal treatment will not be applicable to all mercury containing mixed waste and the removal of mercury prior to thermal treatment may be desirable, laboratory studies have been initiated at Oak Ridge National Laboratory (ORNL) to develop alternative remediation technologies capable of removing mercury from certain mixed waste. This paper describes laboratory investigations of the KI/I 2 leaching processes to determine the applicability of this process to mercury containing solid mixed waste

Evaluation of concrete as a matrix for solidification of Savannah River Plant waste

International Nuclear Information System (INIS)

Stone, J.A.

1978-01-01

Some of the favorable and unfavorable characteristics of concrete as a matrix for solidification of SRP waste, as found in this study, are listed. Compressive strength and leachability of waste forms containing 90 Sr and alpha emitters are very good. The waste forms have reasonable long-term thermal stability up to 400 0 C, although water is evolved above 100 0 C. Long-term radiation stability of the solid, as measured by strength and leachability, is excellent. For the unfavorable characteristics, methods are available to overcome any problems these properties might cause. 137 Cs leachability can be reduced by additives such as zeolite. Steam generation can be reduced by an initial degassing step; however, radiolytic gassing may require further study. Set times can be retarded with additives. 10 figs

Proceedings of the international seminar on chemistry and process engineering for high-level liquid waste solidification

International Nuclear Information System (INIS)

Odoj, R.; Merz, E.

1981-06-01

The proceedings record a very distinct phase of the chemistry and process engineering for high-level liquid waste solidification in the past years. The main purpose is to provide solutions which guarantee sufficient safe and economically acceptable measure causing no adverse consequence to man and his environment. (DG)

SPECIFIC FEATURES OF OLIGOMERIC PRODUCT SOLIDIFICATION FROM POLYURETHANE WASTES AND THEIR PRACTICAL APPLICATION

OpenAIRE

V. Belyatsky; Yu. Kryvogus

2012-01-01

The paper considers a possibility to use secondary polyurethane obtained by  thermal depolymerization of wastes on the basis of cross-linked polyurethane (polyurethane adduct) and isocyanate. An effect of density dependence of the obtained polyurethane samples on nature and quantity of solvent has been revealed and it is significantly observed while using low-boiling solvents. The influence of adduct/solidification agent ratio on mechanical hardness of the obtained samples has been studied in...

Solidification/stabilisation of liquid oil waste in metakaolin-based geopolymer

Energy Technology Data Exchange (ETDEWEB)

Cantarel, V.; Nouaille, F.; Rooses, A.; Lambertin, D., E-mail: david.lambertin@cea.fr; Poulesquen, A.; Frizon, F.

2015-09-15

Highlights: • Formulation with 20 vol.% of oil in a geopolymer have been successful tested. • Oil waste is encapsulated as oil droplets in metakaolin-based geopolymer. • Oil/geopolymer composite present good mechanical performance. • Carbon lixiviation of oil/geopolymer composite is very low. - Abstract: The solidification/stabilisation of liquid oil waste in metakaolin based geopolymer was studied in the present work. The process consists of obtaining a stabilised emulsion of oil in a water-glass solution and then adding metakaolin to engage the setting of a geopolymer block with an oil emulsion stabilised in the material. Geopolymer/oil composites have been made with various oil fraction (7, 14 and 20 vol.%). The rigidity and the good mechanical properties have been demonstrated with compressive strength tests. Leaching tests evidenced the release of oil from the composite material is very limited whereas the constitutive components of the geopolymer (Na, Si and OH{sup −}) are involved into diffusion process.

Status of mixed-waste regulation

International Nuclear Information System (INIS)

Bahadur, S.

1988-01-01

Mixed waste is waste containing radionuclides regulated by the US Nuclear Regulatory Commission (NRC) under the Atomic Energy Act, as well as hazardous waste materials regulated by the Environmental Protection Agency (EPA) under the Resource Conservation and Recovery Act (RCRA). This has led to a situation of dual regulation in which both NRC and EPA regulate the same waste under requirements that at times appear conflicting. The NRC has been working with the EPA to resolve the issues associated with the dual regulation of mixed waste. Discussions between the two agencies indicate that dual regulation of mixed wastes appears technically achievable, although the procedures may be complex and burdensome to the regulated community. The staffs of both agencies have been coordinating their efforts to minimize the burden of dual regulation on state agencies and the industry. Three major issues were identified as sources of potential regulatory conflict: (a) definition and identification of mixed waste, (b) siting guidelines for disposal facilities, and (c) design concepts for disposal units

The Hazardous Waste/Mixed Waste Disposal Facility

International Nuclear Information System (INIS)

Bailey, L.L.

1991-01-01

The Hazardous Waste/Mixed Waste Disposal Facility (HW/MWDF) will provide permanent Resource Conservation and Recovery Act (RCRA) permitted storage, treatment, and disposal for hazardous and mixed waste generated at the Department of Energy's (DOE) Savannah River Site (SRS) that cannot be disposed of in existing or planned SRS facilities. Final design is complete for Phase I of the project, the Disposal Vaults. The Vaults will provide RCRA permitted, above-grade disposal capacity for treated hazardous and mixed waste generated at the SRS. The RCRA Part B Permit application was submitted upon approval of the Permit application, the first Disposal Vault is scheduled to be operational in mid 1994. The technical baseline has been established for Phase II, the Treatment Building, and preliminary design work has been performed. The Treatment Building will provide RCRA permitted treatment processes to handle a variety of hazardous and mixed waste generated at SRS in preparation for disposal. The processes will treat wastes for disposal in accordance with the Environmental Protection Agency's (EPA's) Land Disposal Restrictions (LDR). A RCRA Part B Permit application has not yet been submitted to SCDHEC for this phase of the project. The Treatment Building is currently scheduled to be operational in late 1996

Mixed Waste Management Options: 1995 Update. National Low-Level Waste Management Program

Energy Technology Data Exchange (ETDEWEB)

Kirner, N.; Kelly, J.; Faison, G.; Johnson, D. [Foster Wheeler Environmental Corp. (United States)

1995-05-01

In the original mixed Waste Management Options (DOE/LLW-134) issued in December 1991, the question was posed, ``Can mixed waste be managed out of existence?`` That study found that most, but not all, of the Nation`s mixed waste can theoretically be managed out of existence. Four years later, the Nation is still faced with a lack of disposal options for commercially generated mixed waste. However, since publication of the original Mixed Waste Management Options report in 1991, limited disposal capacity and new technologies to treat mixed waste have become available. A more detailed estimate of the Nation`s mixed waste also became available when the US Environmental Protection Agency (EPA) and the US Nuclear Regulatory Commission (NRC) published their comprehensive assessment, titled National Profile on Commercially Generated Low-Level Radioactive Mixed Waste (National Profile). These advancements in our knowledge about mixed waste inventories and generation, coupled with greater treatment and disposal options, lead to a more applied question posed for this updated report: ``Which mixed waste has no treatment option?`` Beyond estimating the volume of mixed waste requiring jointly regulated disposal, this report also provides a general background on the Atomic Energy Act (AEA) and the Resource Conservation and Recovery Act (RCRA). It also presents a methodical approach for generators to use when deciding how to manage their mixed waste. The volume of mixed waste that may require land disposal in a jointly regulated facility each year was estimated through the application of this methodology.

Mixed Waste Management Options: 1995 Update. National Low-Level Waste Management Program

International Nuclear Information System (INIS)

Kirner, N.; Kelly, J.; Faison, G.; Johnson, D.

1995-05-01

In the original mixed Waste Management Options (DOE/LLW-134) issued in December 1991, the question was posed, ''Can mixed waste be managed out of existence?'' That study found that most, but not all, of the Nation's mixed waste can theoretically be managed out of existence. Four years later, the Nation is still faced with a lack of disposal options for commercially generated mixed waste. However, since publication of the original Mixed Waste Management Options report in 1991, limited disposal capacity and new technologies to treat mixed waste have become available. A more detailed estimate of the Nation's mixed waste also became available when the US Environmental Protection Agency (EPA) and the US Nuclear Regulatory Commission (NRC) published their comprehensive assessment, titled National Profile on Commercially Generated Low-Level Radioactive Mixed Waste (National Profile). These advancements in our knowledge about mixed waste inventories and generation, coupled with greater treatment and disposal options, lead to a more applied question posed for this updated report: ''Which mixed waste has no treatment option?'' Beyond estimating the volume of mixed waste requiring jointly regulated disposal, this report also provides a general background on the Atomic Energy Act (AEA) and the Resource Conservation and Recovery Act (RCRA). It also presents a methodical approach for generators to use when deciding how to manage their mixed waste. The volume of mixed waste that may require land disposal in a jointly regulated facility each year was estimated through the application of this methodology

ENGINEERING BULLETIN: SOLIDIFICATION/STABILIZATION OF ORGANICS AND INORGANICS

Science.gov (United States)

Solidification refers to techniques that encapsulate hazardous waste into a solid material of high structural integrity. Encapsulation involves either fine waste particles (microencapsulation) or a large block or container of wastes (macroencapsulation). Stabilization refe...

Hanford facility dangerous waste permit application, 325 hazardous waste treatment units. Revision 1

International Nuclear Information System (INIS)

1997-07-01

This report contains the Hanford Facility Dangerous Waste Permit Application for the 325 Hazardous Waste Treatment Units (325 HWTUs) which consist of the Shielded Analytical Laboratory, the 325 Building, and the 325 Collection/Loadout Station Tank. The 325 HWTUs receive, store, and treat dangerous waste generated by Hanford Facility programs. Routine dangerous and/or mixed waste treatment that will be conducted in the 325 HWTUs will include pH adjustment, ion exchange, carbon absorption, oxidation, reduction, waste concentration by evaporation, precipitation, filtration, solvent extraction, solids washing, phase separation, catalytic destruction, and solidification/stabilization

Hanford facility dangerous waste permit application, 325 hazardous waste treatment units. Revision 1

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-07-01

This report contains the Hanford Facility Dangerous Waste Permit Application for the 325 Hazardous Waste Treatment Units (325 HWTUs) which consist of the Shielded Analytical Laboratory, the 325 Building, and the 325 Collection/Loadout Station Tank. The 325 HWTUs receive, store, and treat dangerous waste generated by Hanford Facility programs. Routine dangerous and/or mixed waste treatment that will be conducted in the 325 HWTUs will include pH adjustment, ion exchange, carbon absorption, oxidation, reduction, waste concentration by evaporation, precipitation, filtration, solvent extraction, solids washing, phase separation, catalytic destruction, and solidification/stabilization.

Mixed and Low-Level Treatment Facility Project. Appendix B, Waste stream engineering files, Part 1, Mixed waste streams

Energy Technology Data Exchange (ETDEWEB)

1992-04-01

This appendix contains the mixed and low-level waste engineering design files (EDFS) documenting each low-level and mixed waste stream investigated during preengineering studies for Mixed and Low-Level Waste Treatment Facility Project. The EDFs provide background information on mixed and low-level waste generated at the Idaho National Engineering Laboratory. They identify, characterize, and provide treatment strategies for the waste streams. Mixed waste is waste containing both radioactive and hazardous components as defined by the Atomic Energy Act and the Resource Conservation and Recovery Act, respectively. Low-level waste is waste that contains radioactivity and is not classified as high-level waste, transuranic waste, spent nuclear fuel, or 11e(2) byproduct material as defined by DOE 5820.2A. Test specimens of fissionable material irradiated for research and development only, and not for the production of power or plutonium, may be classified as low-level waste, provided the concentration of transuranic is less than 100 nCi/g. This appendix is a tool that clarifies presentation format for the EDFS. The EDFs contain waste stream characterization data and potential treatment strategies that will facilitate system tradeoff studies and conceptual design development. A total of 43 mixed waste and 55 low-level waste EDFs are provided.

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.

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

Mixed waste study, Lawrence Livermore National Laboratory Hazardous Waste Management facilities

International Nuclear Information System (INIS)

1990-11-01

This document addresses the generation and storage of mixed waste at Lawrence Livermore National Laboratory (LLNL) from 1984 to 1990. Additionally, an estimate of remaining storage capacity based on the current inventory of low-level mixed waste and an approximation of current generation rates is provided. Section 2 of this study presents a narrative description of Environmental Protection Agency (EPA) and Department of Energy (DOE) requirements as they apply to mixed waste in storage at LLNL's Hazardous Waste Management (HWM) facilities. Based on information collected from the HWM non-TRU radioactive waste database, Section 3 presents a data consolidation -- by year of storage, location, LLNL generator, EPA code, and DHS code -- of the quantities of low-level mixed waste in storage. Related figures provide the distribution of mixed waste according to each of these variables. A historical review follows in Section 4. The trends in type and quantity of mixed waste managed by HWM during the past five years are delineated and graphically illustrated. Section 5 provides an estimate of remaining low-level mixed waste storage capacity at HWM. The estimate of remaining mixed waste storage capacity is based on operational storage capacity of HWM facilities and the volume of all waste currently in storage. An estimate of the time remaining to reach maximum storage capacity is based on waste generation rates inferred from the HWM database and recent HWM documents. 14 refs., 18 figs., 9 tabs

Solidification control in continuous casting of steel

Indian Academy of Sciences (India)

Unknown

Solidification in continuous casting (CC) technology is initiated in a water- ..... to fully austenitic solidification, and FP between 0 and 1 indicates mixed mode. ... the temperature interval (LIT – TSA) corresponding to fs = 0⋅9 → 1, is in reality the.

Pilot plant experience on high-level waste solidification and design of the engineering prototype VERA

Energy Technology Data Exchange (ETDEWEB)

Guber, W; Diefenbacher, W; Hild, W; Krause, H; Schneider, E; Schubert, G

1972-11-01

In the present paper the solidification process for highly active waste solutions as developed in the Karlsruhe Nuclear Research Center is presented. Its principal steps are: denitration, calcination in a spray calciner operated with superheated steam, melting of the calcine with appropriate additives to borosilicate glass in an induction-heated melting furnace. The operational experiences gained so far in the inactive 1:1 pilot plant are reported. Furthermore, a description is given of the projected multi-purpose experimental facility VERA 2 which is provided for processing the highly active waste solutions from the first German reprocessing plant WAK.

Addressing mixed waste in plutonium processing

International Nuclear Information System (INIS)

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

1991-01-01

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

Mixed Waste Landfill Integrated Demonstration

International Nuclear Information System (INIS)

1994-02-01

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

Assessing mixed waste treatment technologies

International Nuclear Information System (INIS)

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

1994-01-01

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

Determining how much mixed waste will require disposal

International Nuclear Information System (INIS)

Kirner, N.P.

1990-01-01

Estimating needed mixed-waste disposal capacity to 1995 and beyond is an essential element in the safe management of low-level radioactive waste disposal capacity. Information on the types and quantities of mixed waste generated is needed by industry to allow development of treatment facilities and by states and others responsible for disposal and storage of this type of low-level radioactive waste. The design of a mixed waste disposal facility hinges on a detailed assessment of the types and quantities of mixed waste that will ultimately require land disposal. Although traditional liquid scintillation counting fluids using toluene and xylene are clearly recognized as mixed waste, characterization of other types of mixed waste has, however, been difficult. Liquid scintillation counting fluids comprise most of the mixed waste generated and this type of mixed waste is generally incinerated under the supplemental fuel provisions of the Resource Conservation and Recovery Act (RCRA) Because there are no Currently operating mixed waste land disposal facilities, it is impossible to make projections of waste requiring land disposal based on a continuation of current waste disposal practices. Evidence indicates the volume of mixed waste requiring land disposal is not large, since generators are apparently storing these wastes. Surveys conducted to date confirm that relatively small volumes of commercially generated mixed waste volume have relied heavily oil generators' knowledge of their wastes. Evidence exists that many generators are confused by the differences between the Atomic Energy Act and the Resource Conservation and Recovery Act (RCRA) on the issue of when a material becomes a waste. In spite of uncertainties, estimates of waste volumes requiring disposal can be made. This paper proposes an eight-step process for such estimates

Solidification of metal chloride waste from pyrochemical process via dechlorination-chlorination reaction system

Energy Technology Data Exchange (ETDEWEB)

Park, H.S.; Cho, I.H.; Lee, K.R.; Choi, J.H.; Eun, H.C.; Kim, I.T.; Park, G.I. [Korea Atomic Energy Research Inst., Deajeon (Korea, Republic of)

2014-07-01

The metal chloride wastes generated from the pyro-chemical process to recover uranium and TRUs has been considered as a problematic waste due to the high volatility and low compatibility with conventional silicate glass. Our research group has suggested the dechlorination approach for the solidification of this kind of waste by using a synthetic composite, SAP (SiO{sub 2}-Al{sub 2}O{sub 3}-P{sub 2}O{sub 5}). During the dechlorination, metal elements are chemically interacted with the inorganic composite, SAP, while chlorine is vaporized as gaseous chlorine. Metal elements in the salt were immobilized into phosphate and silicate glass which are uniformly distributed in tens of nm scale. During the dechlorination, gaseous chlorine is captured by Li{sub 2}O-Li{sub 2}O{sub 2} composite that can be converted into metal chloride (LiCl). About 98wt% of oxide composite was converted into LiCl that can be used as an electrolyte in the electrochemical process. The method suggested in this study can provide a chance to minimize the waste volume for the final disposal of salt wastes from a pyro-chemical process. (author)

Intense volume reduction of mixed and low-level waste, solidification in sulphur polymer concrete, and excellent disposal at minimum cost

International Nuclear Information System (INIS)

Darnell, G.R.

1990-01-01

Progressive changes in regulations governing the disposal of the nation's radioactive and hazardous wastes demand the development of more advanced treatment and disposal systems. The U.S. Department of Energy's Radioactive Waste Technology Support Program (formerly the Defense Low-Level Waste Management Program) was given the task of demonstrating the degree of excellence that could be achieved at reasonable cost using existing technology. The resulting concept is a Waste Treatment and Disposal Complex that will fully treat contact-handled mixed and low-level radioactive waste to a disposable product that is totally liquid-free and approximately 98% inorganic. An excellent volume reduction factor is achieved through sorting, sizing, incineration, vitrification, and final grouting. Inorganic waste items larger than 1/4 in. will be placed in inexpensive, uniform-sized, smooth-sided, thin-walled steel boxes. The smaller particles will be mixed with sulfur polymer concrete and pumped into the boxes, filling most voids. The appendage-free boxes measuring 1 by 1 by 1 m will be stacked tightly in an abovegrade, earth-mounded, concrete disposal vault where a temporary roof will protect them from rain and snow. A concrete roof poured directly on top of the dense, essentially voidless waste stack will be topped by an engineered, water-shedding earthen cover. Total cost for design, construction, testing, 30 years of treatment and disposal, administration, decontamination and decommissioning, site closure, and postclosure monitoring and maintenance will cost less per cubic foot than is currently expended for subsurface disposal. A radiological performance assessment shows this concept will exceed the nation's existing disposal systems and governmental performance objectives for the protection of the general public by a factor of 30,000

Effects Disposal Condition and Ground Water to Leaching Rate of Radionuclides from Solidification Products

International Nuclear Information System (INIS)

Herlan Martono; Wati

2008-01-01

Effects disposal condition and ground water to leaching rate of radionuclides from solidification products have been studied. The aims of leaching test at laboratory to get the best composition of solidified products for continuous process or handling. The leaching rate of radionuclides from the many kinds of matrix from smallest to bigger are glass, thermosetting plastic, urea formaldehyde, asphalt, and cement. Glass for solidification of high level waste, thermosetting plastic and urea formaldehyde for solidification of low and intermediate waste, asphalt and cement for solidification of low and intermediate level waste. In shallow land burial, ground water rate is fast, debit is high, and high permeability, so the probability contact between solidification products and ground water is occur. The pH of ground water increasing leaching rate, but cation in the ground water retard leaching rate. Effects temperature radiation and radiolysis to solidification products is not occur. In the deep repository, ground water rate is slow, debit is small, and low permeability, so the probability contact between solidification products and ground water is very small. There are effect cooling time and distance between pits to rock temperature. Alfa radiation effects can be occur, but there is no contact between solidification products and ground water, so that there is not radiolysis. (author)

Robotics for mixed waste operations, demonstration description

International Nuclear Information System (INIS)

Ward, C.R.

1993-01-01

The Department of Energy (DOE) Office of Technology Development (OTD) is developing technology to aid in the cleanup of DOE sites. Included in the OTD program are the Robotics Technology Development Program and the Mixed Waste Integrated Program. These two programs are working together to provide technology for the cleanup of mixed waste, which is waste that has both radioactive and hazardous constituents. There are over 240,000 cubic meters of mixed low level waste accumulated at DOE sites and the cleanup is expected to generate about 900,000 cubic meters of mixed low level waste over the next five years. This waste must be monitored during storage and then treated and disposed of in a cost effective manner acceptable to regulators and the states involved. The Robotics Technology Development Program is developing robotics technology to make these tasks safer, better, faster and cheaper through the Mixed Waste Operations team. This technology will also apply to treatment of transuranic waste. The demonstration at the Savannah River Site on November 2-4, 1993, showed the progress of this technology by DOE, universities and industry over the previous year. Robotics technology for the handling, characterization and treatment of mixed waste as well robotics technology for monitoring of stored waste was demonstrated. It was shown that robotics technology can make future waste storage and waste treatment facilities better, faster, safer and cheaper

Hanford Central Waste Complex: Radioactive mixed waste storage facility dangerous waste permit application

International Nuclear Information System (INIS)

1991-10-01

The Hanford Site is owned by the US Government and operated by the US Department of Energy Field Office, Richland. The Hanford Site manages and produces dangerous waste and mixed waste (containing both radioactive and dangerous components). The dangerous waste is regulated in accordance with the Resource Conversation and Recovery Act of 1976 and the State of Washington Hazardous Waste Management Act of 1976. The radioactive component of mixed waste is interpreted by the US Department of Energy to be regulated under the Atomic Energy Act of 1954; the nonradioactive dangerous component of mixed waste is interpreted to be regulated under the Resource Conservation and Recovery Act of 1976 and Washington Administrative Code 173--303. Westinghouse Hanford Company is a major contractor to the US Department of Energy Field Office, Richland and serves as co-operator of the Hanford Central Waste Complex. The Hanford Central Waste Complex is an existing and planned series of treatment, storage, and/or disposal units that will centralize the management of solid waste operations at a single location on the Hanford facility. The Hanford Central Waste Complex units include the Radioactive Mixed Waste Storage Facility, the unit addressed by this permit application, and the Waste Receiving and Processing Facility. The Waste Receiving and Processing Facility is covered in a separate permit application submittal

Preliminary study on immobilization of buffing dust by solidification method in ceramic brick

Science.gov (United States)

Yuliansyah, Ahmad Tawfiequrrahman; Prasetya, Agus; Putra, Arif Eka; Satriawan, Humam Budi

2017-11-01

Leather-based industries generate a substantial amount of hazardous solid and liquid wastes in their process. One of the solid wastes is buffing dust, which is fine particulates containing fat, tanning, dyes and chromium. From 1 ton of leather processed, approximately 2-6 kg of buffing dust is generated. Chromium in the buffing dust is carcinogenic, so a proper handling is highly required. Solidification is a method commonly used to immobilize toxic material. Hence, the material is trapped in a matrix made of binding agents to minimize its mobility. However, a very small amount of the materials is sometimes released to the environment during storage. This study investigates leaching process of chromium from immobilized buffing dust in ceramic brick. Buffing dust, which contains chromium, is solidified by mixing it with clay at certain compositions and fired in a muffle furnace to produce a ceramic brick. Performance of the solidification process is evaluated by measuring the leaching of chromium in the leaching test. The results show that the solidification has significantly reduced the potential release of chromium to the environment. Higher of the firing temperature, less chromium is leached from ceramic brick. The chromium concentration of leachate water from 800°C brick is 0.376 ppm, while those from 850 and 900°C brick are 0.212 and 0.179 ppm respectively.

Solidification of radioactive aqueous solution

Energy Technology Data Exchange (ETDEWEB)

Aikawa, Hideaki; Kato, Kiyoshi; Wadachi, Yoshiki

1970-09-07

A process for solidifying a radioactive waste solution is provided, using as a solidifying agent a mixture of calcined gypsum and burnt vermiculite. The quantity ratio of the mixture is preferred to be 1:1 by volume. The quantity of impregnation is 1/2 of the volume of the total quantity of the solidifying agent. In embodiments, 10 liters of plutonium waste solution was mixed with a mixture of 1:1 calcined gypsum and burnt vermiculite contained in a 20-liter cylindrical steel container lined with asphalt. The plutonium waste solution from the laboratory was neutralized with a caustic soda aqueous solution to prevent explosion due to the nitration of organic compounds. The neutralization is not always necessary. A market available dental gypsum was calcined at 400 to 500/sup 0/C and a vermiculite from Illinois was burnt at 1,100/sup 0/C to prepare the agents. The time required for the impregnation with 10 liters of plutonium solution was four minutes. After impregnation, the temperature rose to 40/sup 0/C within 30 minutes to one hour. Next, it was cooled to room temperature by standing for 3-4 hours. Solidification time was about 1 hour. The Japan Atomic Energy Research Insitute had treated and disposed about 1,000 tons of plutonium waste by this process as of August 19, 1970.

Optimisation of industrial wastes reuse as construction materials.

Science.gov (United States)

Collivignarelli, C; Sorlini, S

2001-12-01

This study concerns the reuse of two inorganic wastes, foundry residues and fly ashes from municipal solid waste incineration, as "recycled aggregate" in concrete production. This kind of reuse was optimised by waste treatment with the following steps: waste washing with water; waste stabilisation-solidification treatment with inorganic reagents; final grinding of the stabilised waste after curing for about 10-20 days. Both the treated wastes were reused in concrete production with different mix-designs. Concrete specimens were characterised by means of conventional physical-mechanical tests (compression, elasticity modulus, shrinkage) and different leaching tests. Experimental results showed that a good structural and environmental quality of "recycled concrete" is due both to a correct waste treatment and to a correct mix-design for concrete mixture.

In-situ stabilization of mixed waste contaminated soil

International Nuclear Information System (INIS)

Siegrist, R.L.; Cline, S.R.; Gilliam, T.M.; Conner, J.R.

1993-01-01

A full-scale field demonstration was conducted to evaluate in for stabilizing an inactive RCRA land treatment site at a DOE facility in Ohio. Subsurface silt and clay deposits were contaminated principally with up to 500 mg/kg of trichloroethylene and other halocarbons, but also trace to low levels of Pb, Cr, 235 U, and 99 Tc. In situ solidification was studied in three, 3.1 m diameter by 4.6 m deep columns. During mixing, a cement-based grout was injected and any missions from the mixed region were captured in a shroud and treated by filtration and carbon adsorption. During in situ processing, operation and performance parameters were measured, and soil cores were obtained from a solidified column 15 months later. Despite previous site-specific treatability experience, there were difficulties in selecting a grout with the requisite treatment agents amenable to subsurface injection and at a volume adequate for distribution throughout the mixed region while minimizing volume expansion. observations during the demonstration revealed that in situ solidification was rapidly accomplished (e.g., >90 m 3 /d) with limited emissions of volatile organics (i.e., -6 cm/s vs. 10 -8 cm/s). Leaching tests performed on the treated samples revealed non-detectable to acceptably low concentrations of all target contaminants

High Solids Consolidated Incinerator Facility (CIF) Wastes Stabilization with Ceramicrete and Super Cement

International Nuclear Information System (INIS)

Walker, B.W.

1999-01-01

High Solids ash and scrubber solution waste streams were generated at the incinerator facility at SRS by burning radioactive diatomaceous filter rolls which contained small amounts of uranium, and listed solvents (F and U). This report details solidification activities using selected Mixed Waste Focus Area (MWFA) technologies with the High Solids waste streams

Solidification of TRU wastes in a ceramic matrix

International Nuclear Information System (INIS)

Loida, A.; Schubert, G.

1991-01-01

Aluminumsilicate based ceramic materials have been evaluated as an alternative waste form for the incorporation of TRU wastes. These waste forms are free of water and - cannot generate hydrogen radiolyticly, - they show good compatibility between the compounds of the waste and the matrix, - they are resistent against aqueous solutions, heat and radiation. R and D-work has been performed to demonstrate the suitability of this waste form for the immobilization of TRU-wastes. Four kinds of original TRU-waste streams and a mixture of all of them have been immobilized by ceramization, using glove box and remote operation technique as well. Clay minerals, (kaolinite, bentonite) and reactive corundum were selected as ceramic raw materials (KAB 78) in an appropriate ratio yielding 78 wt% Al 2 O 3 and 22 wt%SiO 2 . The main process steps are (i) pretreatment of the liquid waste (concentration, denitration, neutralization, solid- liquid separation), (ii) mixing with ceramic raw materials and forming, (iii) heat treatment with T max. of 1300 0 C for 15 minutes. The waste load of the ceramic matrix has been increased gradually from 20 to 50, in some cases to 60 wt.%

DOE regulatory reform initiative vitrified mixed waste

International Nuclear Information System (INIS)

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

1997-01-01

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

Method and apparatus for glass solidification porcessing for radioactive liquid waste

International Nuclear Information System (INIS)

Torada, Shin-ichiro; Masaki, Toshio; Sakai, Akira.

1989-01-01

Glass material supplied to a glass melting furnace is made in the form of a glass container. Then, radioactive liquid wastes are directly injected into the glass vessel and the glass vessel injected with the radioactive liquid wastes is charged into the glass melting furnace. The glass material and the radioactive liquid wastes are supplied simultaneously to the glass melting furnace. Then, corresponding to the amount of the glass material used for the glass vessel, the amount of the radioactive liquid wastes injected to the inside thereof is controlled to thereby set the mixing ratio between the glass material and the radioactive liquid wastes. Further, by controlling the number of the glass vessels injected with the radioactive liquid wastes to be charged into the glass melting furnace, the amount of supplying the radioactive liquid wastes and the glass material is controlled. This can easily maintain constant the amount of the glass material and the radioacative liquid wastes supplied to the glass melting furnace and the mixing ratio thereof. (T.M.)

Solidification of highly active liquid waste

International Nuclear Information System (INIS)

Morris, J.B.

1985-03-01

This document contains the annual progress reports on the following subjects: Joule ceramic melter; microwave vitrification; glass technology; identification, evaluation and review of potential alternative solidification processes; rotary kiln calcination; alternative glass feedstocks; volatile ruthenium trapping by solid adsorbents; irrigated baffle column dust scrubber. (author)

Waste-form development

International Nuclear Information System (INIS)

Neilson, R.M. Jr.; Colombo, P.

1982-01-01

Contemporary solidification agents are being investigated relative to their applications to major fuel cycle and non-fuel cycle low-level waste (LLW) streams. Work is being conducted to determine the range of conditions under which these solidification agents can be applied to specific LLW streams. These studies are directed primarily towards defining operating parameters for both improved solidification of problem wastes and solidification of new LLW streams generated from advanced volume reduction technologies. Work is being conducted to measure relevant waste form properties. These data will be compiled and evaluated to demonstrate compliance with waste form performance and shallow land burial acceptance criteria and transportation requirements

Evaluation of concrete as a matrix for solidification of Savannah River Plant waste

International Nuclear Information System (INIS)

Stone, J.A.

1977-06-01

The properties of concrete as a matrix for solidification of Savannah River Plant (SRP) high-level radioactive wastes were studied. In an experimental, laboratory-scale program, concrete specimens were prepared and evaluated with both simulated and actual SRP waste sludges. Properties of concrete were found adequate for fixation of SRP wastes. Procedures were developed for preparation of simulated sludges and concrete-sludge castings. Effects of cement type, simulated sludge type, sludge loading, and water content on concrete formulations were tested in a factorial experiment. Compressive strength, leachability of strontium and plutonium, thermal stability, and radiation stability were measured for each formulation. From these studies, high-alumina cement and a portland-pozzolanic cement were selected for additional tests. Incorporation of cesium-loaded zeolite into cement-sludge mixtures had no adverse effects on mechanical or chemical properties of waste forms. Effects of heating concrete-sludge castings were investigated; thermal conductivity and DTA-TGA-EGA data are reported. Formulations of actual SRP waste sludges in concrete were prepared and tested for compressive strength; for leachability of 90 Sr, 137 Cs, and alpha emitters; and for long-term thermal stability. The radioactive sludges were generally similar in behavior to simulated sludges in concrete. 37 tables, 34 figures

Experiences with treatment of mixed waste

International Nuclear Information System (INIS)

Dziewinski, J.; Marczak, S.; Smith, W.H.; Nuttall, E.

1996-01-01

During its many years of research activities involving toxic chemicals and radioactive materials, Los Alamos National Laboratory (Los Alamos) has generated considerable amounts of waste. Much of this waste includes chemically hazardous components and radioisotopes. Los Alamos chose to use an electrochemical process for the treatment of many mixed waste components. The electro-chemical process, which the authors are developing, can treat a great variety of waste using one type of equipment built at a moderate expense. Such a process can extract heavy metals, destroy cyanides, dissolve contamination from surfaces, oxidize toxic organic compounds, separate salts into acids and bases, and reduce the nitrates. All this can be accomplished using the equipment and one crew of trained operating personnel. Results of a treatability study of chosen mixed wastes from Los Alamos Mixed Waste Inventory are presented. Using electrochemical methods cyanide and heavy metals bearing wastes were treated to below disposal limits

Experiences with treatment of mixed waste

Energy Technology Data Exchange (ETDEWEB)

Dziewinski, J.; Marczak, S.; Smith, W.H. [Los Alamos National Lab., NM (United States); Nuttall, E. [Univ. of New Mexico, Albuquerque, NM (United States). Chemical and Nuclear Engineering Dept.

1996-04-10

During its many years of research activities involving toxic chemicals and radioactive materials, Los Alamos National Laboratory (Los Alamos) has generated considerable amounts of waste. Much of this waste includes chemically hazardous components and radioisotopes. Los Alamos chose to use an electrochemical process for the treatment of many mixed waste components. The electro-chemical process, which the authors are developing, can treat a great variety of waste using one type of equipment built at a moderate expense. Such a process can extract heavy metals, destroy cyanides, dissolve contamination from surfaces, oxidize toxic organic compounds, separate salts into acids and bases, and reduce the nitrates. All this can be accomplished using the equipment and one crew of trained operating personnel. Results of a treatability study of chosen mixed wastes from Los Alamos Mixed Waste Inventory are presented. Using electrochemical methods cyanide and heavy metals bearing wastes were treated to below disposal limits.

Permitting mixed waste treatment, storage and disposal facilities: A mixed bag

International Nuclear Information System (INIS)

Ranek, N.L.; Coalgate, J.L.

1995-01-01

The Federal Facility Compliance Act of 1992 (FFCAct) requires the U.S. Department of Energy (DOE) to make a comprehensive national inventory of its mixed wastes (i.e., wastes that contain both a hazardous component that meets the Resource Conservation and Recovery Act (RCRA) definition of hazardous waste and a radioactive component consisting of source, special nuclear, or byproduct material regulated under the Atomic Energy Act (AEA)), and of its mixed waste treatment technologies and facilities. It also requires each DOE facility that stores or generates mixed waste to develop a treatment plan that includes, in part, a schedule for constructing units to treat those wastes that can be treated using existing technologies. Inherent in constructing treatment units for mixed wastes is, of course, permitting. This paper identifies Federal regulatory program requirements that are likely to apply to new DOE mixed waste treatment units. The paper concentrates on showing how RCRA permitting requirements interrelate with the permitting or licensing requirements of such other laws as the Atomic Energy Act, the Clean Water Act, and the Clean Air Act. Documentation needed to support permit applications under these laws are compared with RCRA permit application documentation. National Environmental Policy Act (NEPA) documentation requirements are also addressed, and throughout the paper, suggestions are made for managing the permitting process

Strategy for managing mixed waste at a plant site

International Nuclear Information System (INIS)

Fentiman, A.

1991-01-01

No waste disposal site is currently accepting mixed waste, but facilities across the country continue to generate it. The waste manager at each site is faced with two problems: how to manage the mixed waste already on-site and how to minimize the amount of new waste generated. A strategy has been developed to address each problem. A key element of the strategy is a building-by-building survey of the site. The survey provides information on how and where mixed waste is generated and stored. This paper describes a method for planning and conducting a site-wide mixed-waste survey. It then outlines approaches to managing existing mixed waste and to minimizing mixed-waste generation using information from the survey

Summary of BNL studies regarding commercial mixed waste

International Nuclear Information System (INIS)

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

1986-09-01

Based on BNL's study it was concluded that there are low-level radioactive wastes (LLWs) which contain chemically hazardous components. Scintillation liquids may be considered an EPA listed hazardous waste and are, therefore, potential mixed wastes. Since November 1985, no operating LLW disposal site will accept these wastes for disposal. Unless such wastes contain de minimis quantities of radionuclides, they cannot be disposed of at an EPA permitted site. Currently generators of liquid scintillation wastes can ship de minimis wastes to be burned at commercial facilities. Oil wastes may also eventually be an EPA listed waste and thus will have to be considered a potential radioactive mixed waste unless NRC establishes de minimis levels of radionuclides below which oils can be managed as hazardous wastes. Regarding wastes containing lead metal there is some question as to the extent of the hazard posed by lead disposed in a LLW burial trench. Chromium-containing wastes would have to be tested to determine whether they are potential mixed wastes. There may be other wastes that are mixed wastes; the responsibility for determining this rests with the waste generator. While management options for handling potential mixed wastes are available, there is limited regulatory guidance for generators. BNL has identified and evaluated a variety of treatment options for the management of potential radioactive mixed wastes. The findings of that study showed that application of a management option with the purpose of addressing EPA concerns can, at the same time, address stabilization and volume reduction concerns of NRC. 6 refs., 1 tab

National Institutes of Health: Mixed waste stream analysis

International Nuclear Information System (INIS)

Kirner, N.P.; Faison, G.P.; Johnson, D.R.

1994-08-01

The Low-Level Radioactive Waste Policy Amendments Act of 1985 requires that the US Department of Energy (DOE) provide technical assistance to host States, compact regions, and unaffiliated States to fulfill their responsibilities under the Act. The National Low-Level Waste Management Program (NLLWMP) operated for DOE by EG ampersand G Idaho, Inc. provides technical assistance in the development of new commercial low-level radioactive waste disposal capacity. The NLLWMP has been requested by the Appalachian Compact to help the biomedical community become better acquainted with its mixed waste streams, to help minimize the mixed waste streams generated by the biomedical community, and to provide applicable treatment technologies to those particular mixed waste streams. Mixed waste is waste that satisfies the definition of low-level radioactive waste (LLW) in the Low-Level Radioactive Waste Policy Act of 1980 (LLRWPA) and contains hazardous waste that either (a) is listed as a hazardous waste in Subpart D of 40 CFR 261, or (b) causes the LLW to exhibit any of the hazardous waste characteristics identified in 40 CFR 261. The purpose of this report is to clearly define and characterize the mixed waste streams generated by the biomedical community so that an identification can be made of the waste streams that can and cannot be minimized and treated by current options. An understanding of the processes and complexities of generation of mixed waste in the biomedical community may encourage more treatment and storage options to become available

Mixed Low-Level Radioactive Waste (MLLW) Primer

International Nuclear Information System (INIS)

Schwinkendorf, W.E.

1999-01-01

This document presents a general overview of mixed low-level waste, including the regulatory definitions and drivers, the manner in which the various kinds of mixed waste are regulated, and a discussion of the waste treatment options

Mixed Low-Level Radioactive Waste (MLLW) Primer

Energy Technology Data Exchange (ETDEWEB)

W. E. Schwinkendorf

1999-04-01

This document presents a general overview of mixed low-level waste, including the regulatory definitions and drivers, the manner in which the various kinds of mixed waste are regulated, and a discussion of the waste treatment options.

Sulfur polymer cement for macroencapsulation of mixed waste debris

International Nuclear Information System (INIS)

Mattus, C.H.

1998-01-01

In FY 1997, the US DOE Mixed Waste Focus Area (MWFA) sponsored a demonstration of the macroencapsulation of mixed waste debris using sulfur polymer cement (SPC). Two mixed wastes were tested--a D006 waste comprised of sheets of cadmium and a D008/D009 waste comprised of lead pipes and joints contaminated with mercury. The demonstration was successful in rendering these wastes compliant with Land Disposal Restrictions (LDR), thereby eliminating one Mixed Waste Inventory Report (MWIR) waste stream from the national inventory

Green remediation and recycling of contaminated sediment by waste-incorporated stabilization/solidification.

Science.gov (United States)

Wang, Lei; Tsang, Daniel C W; Poon, Chi-Sun

2015-03-01

Navigational/environmental dredging of contaminated sediment conventionally requires contained marine disposal and continuous monitoring. This study proposed a green remediation approach to treat and recycle the contaminated sediment by means of stabilization/solidification enhanced by the addition of selected solid wastes. With an increasing amount of contaminated sediment (20-70%), the 28-d compressive strength of sediment blocks decreased from greater than 10MPa to slightly above 1MPa. For augmenting the cement hydration, coal fly ash was more effective than lime and ground seashells, especially at low sediment content. The microscopic and spectroscopic analyses showed varying amounts of hydration products (primarily calcium hydroxide and calcium silicate hydrate) in the presence of coal fly ash, signifying the influence of pozzolanic reaction. To facilitate the waste utilization, cullet from beverage glass bottles and bottom ashes from coal combustion and waste incineration were found suitable to substitute coarse aggregate at 33% replacement ratio, beyond which the compressive strength decreased accordingly. The mercury intrusion porosimetry analysis indicated that the increase in the total pore area and average pore diameter were linearly correlated with the decrease of compressive strength due to waste replacement. All the sediment blocks complied with the acceptance criteria for reuse in terms of metal leachability. These results suggest that, with an appropriate mixture design, contaminated sediment and waste materials are useful resources for producing non-load-bearing masonry units or fill materials for construction uses. Copyright © 2014 Elsevier Ltd. All rights reserved.

Mixed waste characterization reference document

International Nuclear Information System (INIS)

1997-09-01

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

Solidification of radioactive liquid wastes. A comparison of treatment options for spent resins and concentrates

International Nuclear Information System (INIS)

Roth, A.; Willmann, F.; Ebata, M.; Wendt, S.

2008-01-01

Ion exchange is one of the most common and effective treatment methods for liquid radioactive waste. However, spent ion exchange resins are considered to be problematic waste that in many cases require special approaches and pre-conditioning during its immobilization to meet the acceptance criteria for disposal. Because of the function that they fulfill, spent ion exchange resins often contain high concentrations of radioactivity and pose special handling and treatment problems. Another very common method of liquid radioactive waste treatment and water cleaning is the evaporation or diaphragm filtration. Both treatment options offer a high volume reduction of the total volume of liquids treated but generate concentrates which contain high concentrations of radioactivity. Both mentioned waste streams, spent resins as well as concentrates, resulting from first step liquid radioactive waste treatment systems have to be conditioned in a suitable manner to achieve stable waste products for final disposal. The most common method of treatment of such waste streams is the solidification in a solid matrix with additional inactive material like cement, polymer etc. In the past good results have been achieved and the high concentration of radioactivity can be reduced by adding the inactive material. On the other hand, under the environment of limited space for interim storage and the absence of a final repository site, the built-up of additional volume has to be considered as very critical. Moreover, corrosive effects on cemented drums during long-term interim storage at the surface have raised doubts about the long-term stability of such waste products. In order to avoid such disadvantages solidification methods have been improved in order to get a well-defined product with a better load factor of wastes in the matrix. In a complete different approach, other technologies solidify the liquid radioactive wastes without adding of any inactive material by means of drying

Commercial mixed waste treatment and disposal

International Nuclear Information System (INIS)

Vance, J.K.

1994-01-01

At the South Clive, Utah, site, Envirocare of Utah, Inc., (Envirocare), currently operates a commercial low-activity, low-level radioactive waste facility, a mixed waste RCRA Part B storage and disposal facility, and an 11e.(2) disposal facility. Envirocare is also in the process of constructing a Mixed Waste Treatment Facility. As the nation's first and only commercial treatment and disposal facility for such waste, the information presented in this segment will provide insight into their current and prospective operations

Mixed Waste Focus Area program management plan

International Nuclear Information System (INIS)

Beitel, G.A.

1996-10-01

This plan describes the program management principles and functions to be implemented in the Mixed Waste Focus Area (MWFA). The mission of the MWFA is to provide acceptable technologies that enable implementation of mixed waste treatment systems developed in partnership with end-users, stakeholders, tribal governments and regulators. The MWFA will develop, demonstrate and deliver implementable technologies for treatment of mixed waste within the DOE Complex. Treatment refers to all post waste-generation activities including sampling and analysis, characterization, storage, processing, packaging, transportation and disposal

Rock solidification method

International Nuclear Information System (INIS)

Nakaya, Iwao; Murakami, Tadashi; Miyake, Takafumi; Funakoshi, Toshio; Inagaki, Yuzo; Hashimoto, Yasuhide.

1985-01-01

Purpose: To convert radioactive wastes into the final state for storage (artificial rocks) in a short period of time. Method: Radioactive burnable wastes such as spent papers, cloths and oils and activated carbons are burnt into ashes in a burning furnace, while radioactive liquid wastes such as liquid wastes of boric acid, exhausted cleaning water and decontaminating liquid wastes are powderized in a drying furnace or calcining furnace. These powders are joined with silicates as such as white clay, silica and glass powder and a liquid alkali such as NaOH or Ca(OH) 2 and transferred to a solidifying vessel. Then, the vessel is set to a hydrothermal reactor, heated and pressurized, then taken out about 20 min after and tightly sealed. In this way, radioactive wastes are converted through the hydrothermal reactions into aqueous rock stable for a long period of time to obtain solidification products insoluble to water and with an extremely low leaching rate. (Ikeda, J.)

Radioactive mixed waste disposal

International Nuclear Information System (INIS)

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

1993-02-01

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

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

Hanford's Radioactive Mixed Waste Disposal Facility

International Nuclear Information System (INIS)

McKenney, D.E.

1995-01-01

The Radioactive Mixed Waste Disposal Facility, is located in the Hanford Site Low-Level Burial Grounds and is designated as Trench 31 in the 218-W-5 Burial Ground. Trench 31 is a Resource Conservation and Recovery Act compliant landfill and will receive wastes generated from both remediation and waste management activities. On December 30, 1994, Westinghouse Hanford Company declared readiness to operate Trench 31, which is the Hanford Site's (and the Department of Energy complex's) first facility for disposal of low-level radioactive mixed wastes

Waste form development

International Nuclear Information System (INIS)

Neilson, R.M. Jr.; Colombo, P.

1982-01-01

In this program, contemporary solidification agents are being investigated relative to their applications to major fuel cycle and non-fuel cycle low-level waste (LLW) streams. Work is being conducted to determine the range of conditions under which these solidification agents can be applied to specific LLW streams. These studies are directed primarily towards defining operating parameters for both improved solidification of problem wastes and solidification of new LLW streams generated from advanced volume reduction technologies. Work is being conducted to measure relevant waste form properties. These data will be compiled and evaluated to demonstrate compliance with waste form performance and shallow land burial acceptance criteria and transportation requirements (both as they exist and as they are modified with time). 6 tables

Assessment of LANL transuranic mixed waste management documentation

International Nuclear Information System (INIS)

Davis, K.D.; Hoevemeyer, S.S.; McCance, C.H.; Jennrich, E.A.; Lund, D.M.

1991-04-01

The objective of this report is to present findings from the evaluation of the Los Alamos National Laboratory (LANL) TRU Mixed Waste Acceptance Criteria to determine its compliance with applicable DOE requirements. The driving requirements for s TRU Mixed Waste Acceptance Criteria are essentially those contained in the ''TRU Waste Acceptance Criteria for the Waste Isolation Pilot Plant'' or WIPP WAC (DOE Report WIPP-DOE-069), 40 CFR 261-270, and DOE Order 5820.2A (Radioactive Waste Management), specifically Chapter II which is entitled ''Management of Transuranic Waste''. The primary purpose of the LANL WAC is the establishment of those criteria that must be met by generators of TRU mixed waste before such waste can be accepted by the Waste Management Group. An annotated outline of a genetic TRU mixed waste acceptance criteria document was prepared from those requirements contained in the WIPP WAC, 40 CFR 261-270, and 5820.2A, and is based solely upon those requirements

Small hazardous waste generators in developing countries: use of stabilization/solidification process as an economic tool for metal wastewater treatment and appropriate sludge disposal.

Science.gov (United States)

Silva, Marcos A R; Mater, Luciana; Souza-Sierra, Maria M; Corrêa, Albertina X R; Sperb, Rafael; Radetski, Claudemir M

2007-08-25

The aim of this study was to propose a profitable destination for an industrial sludge that can cover the wastewater treatment costs of small waste generators. Optimized stabilization/solidification technology was used to treat hazardous waste from an electroplating industry that is currently released untreated to the environment. The stabilized/solidified (S/S) waste product was used as a raw material to build concrete blocks, to be sold as pavement blocks or used in roadbeds and/or parking lots. The quality of the blocks containing a mixture of cement, lime, clay and waste was evaluated by means of leaching and solubility tests according to the current Brazilian waste regulations. Results showed very low metal leachability and solubility of the block constituents, indicating a low environmental impact. Concerning economic benefits from the S/S process and reuse of the resultant product, the cost of untreated heavy metal-containing sludge disposal to landfill is usually on the order of US$ 150-200 per tonne of waste, while 1tonne of concrete roadbed blocks (with 25% of S/S waste constitution) has a value of around US$ 100. The results of this work showed that the cement, clay and lime-based process of stabilization/solidification of hazardous waste sludge is sufficiently effective and economically viable to stimulate the treatment of wastewater from small industrial waste generators.

Mixed Waste Salt Encapsulation Using Polysiloxane - Final Report

International Nuclear Information System (INIS)

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

1997-01-01

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

Low temperature setting iron phosphate ceramics as a stabilization and solidification agent for incinerator ash contaminated with transuranic and RCRA metals

International Nuclear Information System (INIS)

Medvedev, P.G.; Hansen, M.; Wood, E.L.; Frank, S.M.; Sidwell, R.W.; Giglio, J.J.; Johnson, S.G.; Macheret, J.

1997-01-01

Incineration of combustible Mixed Transuranic Waste yields an ash residue that contains oxides of Resource Conservation and Recovery Act (RCRA) and transuranic metals. In order to dispose of this ash safely, it has to be solidified and stabilized to satisfy appropriate requirements for repository disposal. This paper describes a new method for solidification of incinerator ash, using room temperature setting iron phosphate ceramics, and includes fabrication procedures for these waste forms as well as results of the MCC-1 static leach test, XRD analysis, scanning electron microscopy studies and density measurements of the solidified waste form produced

Conversion of three mixed-waste streams

International Nuclear Information System (INIS)

Harmer, D.E.; Porter, D.L.; Conley, C.W.

1990-01-01

At the present time, commercial mixed waste (containing both radioactive and hazardous components) is not handled by any disposal site in this country. Thus, a generator of such material is faced with the prospect of separating or altering the nature of the waste components. A chemical or physical separation may be possible. However, if separation fails there remains the opportunity of chemically transforming the hazardous ingredients to non-hazardous substances, allowing disposal at an existing radioactive burial site. Finally, chemical or physical stabilization can be used as a tool to achieve an acceptable waste form lacking the characteristics of mixed waste. A practical application of these principles has been made in the case of certain mixed waste streams at Aerojet Ordnance Tennessee. Three different streams were involved: (1) lead and lead oxide contaminated with uranium, (2) mixed chloride salts including barium chloride, contaminated with uranium, and (3) bricks impregnated with the barium salt mixture. This paper summarizes the approach of this mixed-waste problem, the laboratory solutions found, and the intended field remediations to be followed. Mixture (1), above, was successfully converted to a vitreous insoluble form. Mixture (2) was separated into radioactive and non-radioactive streams, and the hazardous characteristics of the latter altered chemically. Mixture (3) was treated to an extraction process, after which the extractant could be treated by the methods of Mixture (2). Field application of these methods is scheduled in the near future

EPA/DOE joint efforts on mixed waste treatment

International Nuclear Information System (INIS)

Lee, C.C.; Huffman, G.L.; Nalesnik, R.P.

1995-01-01

Under the requirements of the Federal Facility Compliance Act (FFCA), the Department of Energy (DOE) is directed to develop treatment plans for their stockpile of wastes generated at their various sites. As a result, DOE is facing the monumental problem associated with the treatment and ultimate disposal of their mixed (radioactive and hazardous) waste. Meanwhile, the Environmental Protection Agency (EPA) issued a final open-quotes Hazardous Waste Combustion Strategyclose quotes in November 1994. Under the Combustion Strategy, EPA permit writers have been given the authority to use the Omnibus Provision of the Resource Conservation and Recovery Act (RCRA) to impose more stringent emission limits for waste combustors prior to the development of new regulations. EPA and DOE established a multi-year Interagency Agreement (IAG) in 1991. The main objective of the IAG (and of the second IAG that was added in 1993) is to conduct a research program on thermal technologies for treating mixed waste and to establish permit procedures for these technologies particularly under the new requirements of the above-mentioned EPA Combustion Strategy. The objective of this Paper is to summarize the results of the EPA/DOE joint efforts on mixed waste treatment since the establishment of the original Interagency Agreement. Specifically, this Paper will discuss six activities that have been underway; namely: (1) National Technical Workgroup (NTW) on Mixed Waste Treatment, (2) State-of-the-Art Assessment of APC (Air Pollution Control) and Monitoring Technologies for the Rocky Flats Fluidized Bed Unit, (3) Initial Study of Permit open-quotes Roadmapclose quotes Development for Mixed Waste Treatment, (4) Risk Assessment Approach for a Mixed Waste Thermal Treatment Facility, (5) Development and Application of Technology Selection Criteria for Mixed Waste Thermal Treatment, and (6) Performance Testing of Mixed Waste Incineration: In-Situ Chlorine Capture in a Fluidized Bed Unit

Method for processing powdery radioactive wastes

International Nuclear Information System (INIS)

Yasumura, Keijiro; Matsuura, Hiroyuki; Tomita, Toshihide; Nakayama, Yasuyuki.

1978-01-01

Purpose: To solidify radioactive wastes with ease and safety at a high reaction speed but with no boiling by impregnating the radioactive wastes with chlorostyrene. Method: Beads-like dried ion exchange resin, powdery ion exchange resin, filter sludges, concentrated dried waste liquor or the like are mixed or impregnated with a chlorostyrene monomer dissolving therein a polymerization initiator such as methyl ethyl ketone peroxide and benzoyl peroxide. Mixed or impregnated products are polymerized to solid after a predetermined of time through curing reaction to produce solidified radioactive wastes. Since inflammable materials are used, this process has a high safety. About 70% wastes can be incorporated. The solidified products have a strength as high as 300 - 400 kg/cm 3 and are suitable to ocean disposal. The products have a greater radioactive resistance than other plastic solidification products. (Seki, T.)

Volume reduction/solidification of liquid radioactive waste using bitumen at Ontario Hydro's Bruce Nuclear Generating Station 'A'

International Nuclear Information System (INIS)

Day, J.E.; Baker, R.L.

1995-01-01

Ontario Hydro at the Bruce Nuclear Generating Station 'A' has undertaken a program to render the station's liquid radioactive waste suitable for discharge to Lake Huron by removing sufficient radiological and chemical contaminants to satisfy regulatory requirements for emissions. The system will remove radionuclide and chemical contaminants from five different plant waste streams. The contaminants will be immobilized and stored at on-site radioactive waste storage facilities and the purified streams will be discharged. The discharge targets established by Ontario Hydro are set well below the limits established by the Ontario Ministry of Environment (MOE) and are based on the Best Available Technology Economically Achievable Approach (B.A.T.E.A.). ADTECHS Corporation has been selected by Ontario Hydro to provide volume reduction/solidification technology for one of the five waste streams. The system will dry and immobilize the contaminants from a liquid waste stream in emulsified asphalt using thin film evaporation technology

Grout and glass performance in support of stabilization/solidification of ORNL tank sludges

International Nuclear Information System (INIS)

Spence, R.D.; Mattus, C.H.; Mattus, A.J.

1998-09-01

Wastewater at Oak Ridge National Laboratory (ORNL) is collected, evaporated, and stored in the Melton Valley Storage Tanks (MVST) and Bethel Valley Evaporator Storage Tanks (BVEST) pending treatment for disposal. In addition, some sludges and supernatants also requiring treatment remain in two inactive tank systems: the gunite and associated tanks (GAAT) and the old hydrofracture (OHF) tank. The waste consists of two phases: sludge and supernatant. The sludges contain a high amount of radioactivity, and some are classified as TRU sludges. Some Resource Conservation and Recovery Act (RCRA) metal concentrations are high enough to be defined as RCRA hazardous; therefore, these sludges are presumed to be mixed TRU waste. Grouting and vitrification are currently two likely stabilization/solidification alternatives for mixed wastes. Grouting has been used to stabilize/solidify hazardous and low-level radioactive waste for decades. Vitrification has been developed as a high-level radioactive alternative for decades and has been under development recently as an alternative disposal technology for mixed waste. The objective of this project is to define an envelope, or operating window, for grout and glass formulations for ORNL tank sludges. Formulations will be defined for the average composition of each of the major tank farms (BVEST/MVST, GAAT, and OHF) and for an overall average composition of all tank farms. This objective is to be accomplished using surrogates of the tank sludges with hot testing of actual tank sludges to check the efficacy of the surrogates

Assessment of LANL solid low-level mixed waste documentation

International Nuclear Information System (INIS)

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

1991-04-01

DOE Order 5820.2A requires that a system performance assessment be conducted to assure efficient and compliant management of all radioactive waste. The objective of this report is to determine the present status of the Radioactive Waste Operations Section and the Chemical Waste Operations Section capabilities regarding preparation and maintenance of appropriate criteria, plans, and procedures. Additionally, a comparison is made which identifies areas where these documents are not presently in existence or being fully implemented. The documents being assessed in this report are: Solid Low-Level Mixed Waste Acceptance Criteria, Solid Low-Level Mixed Waste Characterization Plan, Solid Low-Level Mixed waste Certification Plan, Solid Low-Level Mixed Waste Acceptance Procedures, Solid Low-Level Mixed Waste characterization Procedures, Solid Low-Level Mixed Waste Certification Procedures, Solid Low-Level Mixed Waste Training Procedures, and Solid Low-Level Mixed Waste Recordkeeping Requirements. This report compares the current status of preparation and implementation, by the Radioactive Waste Operations Section and the Chemical Waste Operations Section, of these documents to the requirements of DOE 5820.2A,. 40 CFR 260 to 270, and to recommended practice. Chapters 2 through 9 of the report presents the results of the comparison in tabular form for each of the documents being assessed, followed by narrative discussion of all areas which are perceived to be unsatisfactory or out of compliance with respect to the availability and content of the documents. The final subpart of each of the following chapters provides recommendations where documentation practices may be improved to achieve compliance or to follow the recommended practice

Overcoming mixed waste management obstacles - A company wide approach

International Nuclear Information System (INIS)

Buckley, R.N.

1996-01-01

The dual regulation of mixed waste by the Nuclear Regulatory Commission and the Environmental Protection Agency has significantly complicated the treatment, storage and disposal of this waste. Because of the limited treatment and disposal options available, facilities generating mixed waste are also being forced to acquire storage permits to meet requirements associated with the Resource Conservation and Recovery Act. Due to the burdens imposed by the regulatory climate, Entergy Operations has undertaken a proactive approach to managing its mixed waste. Their approach is company wide and simplistic in nature. Utilizing the peer groups to develop strategies and a company wide procedure for guidance on mixed waste activities, they have focused on areas where they have the most control and can achieve the greatest benefits from their efforts. A key aspect of the program includes training and employee awareness regarding mixed waste minimization practices. In addition, Entergy Operations is optimizing the implementation of regulatory provisions that facilitate more flexible management practices for mixed waste. This presentation focuses on the team approach to developing mixed waste managements programs and the utilization of innovative thinking and planning to minimize the regulatory burdens. It will also describe management practices and philosophies that have provided more flexibility in implementing a safe and effective company wide mixed waste management program

Status of high level and alpha bearing waste management in PNC

International Nuclear Information System (INIS)

Uematsu, Kunihiko

1982-04-01

For completing the nuclear fuel cycle in Japan, Power Reactor and Nuclear Fuel Development Corporation (PNC) has a role to promote the management of high level and alpha bearing wastes. For high level waste management, it is planned in Japan to initiate the operation of a vitrification pilot plant by 1987 for the development of the solidification process, and to make it possible to initiate trial disposal by 2015 for the development of geological disposal technology. In PNC, monolithic borosilicate glass was selected as the final form of solidification. Alpha bearing wastes have been produced in the mixed oxide fuel fabrication facility and the reprocessing plant in PNC; and the amount should increase considerably in the future in Japan. About these two areas of waste management, the policy and the research/development programs are described. (J.P.N.)

Fixation process for radioactive waste

International Nuclear Information System (INIS)

Theysohn, F.

1977-01-01

An improvement on the method of solidification of radioactive liquid waste in bitumen with the aid of extruders is described. So far, it has been difficult to remove large amounts of water. The waste sludge, as proposed here, is pre-dried in the extruder and then mixed with the bitumen. The extruder is inclined upward in the transport direction, and its barrel extruders have through holes parallel to the direction of transport in the raised sides of the passages, so that water runs back. Also the waste steam nozzles are arranged before the bitumen inlet. (UWI) [de

Evaluation of Secondary Streams in Mixed Waste Treatment

International Nuclear Information System (INIS)

Haywood, Fred F.; Goldsmith, William A.; Allen, Douglas F.; Mezga, Lance J.

1995-12-01

The United States Department of Energy (DOE) and its predecessors have generated waste containing radioactive and hazardous chemical components (mixed wastes) for over 50 years. Facilities and processes generating these wastes as well as the regulations governing their management have changed. Now, DOE has 49 sites where mixed waste streams exist. The Federal Facility Compliance Act of 1992 (1) required DOE to prepare and obtain regulatory approval of plans for treating these mixed waste streams. Each of the involved DOE sites submitted its respective plan to regulators in April 1995 (2). Most of the individual plans were approved by the respective regulatory agencies in October 1995. The implementation of these plans has begun accordance with compliance instruments (orders) issued by the cognizant regulatory authority. Most of these orders include milestones that are fixed, firm and enforceable as defined in each compliance order. In many cases, mixed waste treatment that was already being carried out and survived the alternative selection process is being used now to treat selected mixed waste streams. For other waste streams at sites throughout the DOE complex treatment methods and schedules are subject to negotiation as the realties of ever decreasing budgets begin to drive the available options. Secondary wastes generated by individual waste treatment systems are also mixed wastes that require treatment in the appropriate treatment system. These secondary wastes may be solid or liquid waste (or both). For example debris washing will generate wastewater requiring treatment; wastewater treatment, in turn, will generate sludge or other residuals requiring treatment; liquid effluents must meet applicable limits of discharge permits. At large DOE sites, secondary waste streams will be a major influence in optimizing design for primary treatment. Understanding these impacts is important not only foe system design, but also for assurances that radiation releases and

Characterization of radioactive mixed wastes: The industrial perspective

International Nuclear Information System (INIS)

Leasure, C.S.

1992-01-01

Physical and chemical characterization of Radioactive Mixed Wastes (RMW) is necessary for determination of appropriate treatment options and to satisfy environmental regulations. Radioactive mixed waste can be classified as two main categories; contact-handled (low level) RMW and remote-handled RMW. Ibis discussion will focus mainly on characterization of contact handled RMW. The characterization of wastes usually follows one of two pathways: (1) characterization to determine necessary parameters for treatment or (2) characterization to determine if the material is a hazardous waste. Sometimes, however, wastes can be declared as hazardous waste without testing and then treated as hazardous waste. Characterization of radioactive mixed wastes pose some unique issues, however, that will require special solutions. Below, five issues affecting sampling and analysis of RMW will be discussed

Enthalpies of a binary alloy during solidification

Science.gov (United States)

Poirier, D. R.; Nandapurkar, P.

1988-01-01

The purpose of the paper is to present a method of calculating the enthalpy of a dendritic alloy during solidification. The enthalpies of the dendritic solid and interdendritic liquid of alloys of the Pb-Sn system are evaluated, but the method could be applied to other binaries, as well. The enthalpies are consistent with a recent evaluation of the thermodynamics of Pb-Sn alloys and with the redistribution of solute in the same during dendritic solidification. Because of the heat of mixing in Pb-Sn alloys, the interdendritic liquid of hypoeutectic alloys (Pb-rich) of less than 50 wt pct Sn has enthalpies that increase as temperature decreases during solidification.

Mixed Waste Integrated Program emerging technology development

International Nuclear Information System (INIS)

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

1994-01-01

The US Department of Energy (DOE) is responsible for the management and treatment of its mixed low-level wastes (MLLW). MLLW are regulated under both the Resource Conservation and Recovery Act and various DOE orders. Over the next 5 years, DOE will manage over 1.2 m 3 of MLLW and mixed transuranic (MTRU) wastes. In order to successfully manage and treat these mixed wastes, DOE must adapt and develop characterization, treatment, and disposal technologies which will meet performance criteria, regulatory approvals, and public acceptance. Although technology to treat MLLW is not currently available without modification, DOE is committed to developing such treatment technologies and demonstrating them at the field scale by FY 1997. The Office of Research and Development's Mixed Waste Integrated Program (MWIP) within the DOE Office of Environmental Management (EM), OfFice of Technology Development, is responsible for the development and demonstration of such technologies for MLLW and MTRU wastes. MWIP advocates and sponsors expedited technology development and demonstrations for the treatment of MLLW

Study on the High Volume Reduction of Radioactive Wastes

Energy Technology Data Exchange (ETDEWEB)

Kim, Ki Hong; Sik, Kang Il; Seok, Hong Dae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Ho, Jeon Gil [RADIN Co. Ltd., Daejeon (Korea, Republic of)

2013-10-15

The solidification of radioactive wastes by the mixing method always increases their volume due to the limitation of incorporation ratio (waste/solidification agent). But if the powdered wastes can be compacted as the high density pellets and also the pellets can be filled up in a waste drum as much as possible while solidifying them with a very sticky solidification agent including a void formed in the filling step of pellets, it might be more desirable to reduce the waste volume as compared with the mixing method. So in this study, we designed and manufactured a high volume reduction machine which has the special size and shape of a pellet pocket, which the pellets can be extracted from easily and filled up in a large amount in drum, a pressurizing device to press 2 rolls, and the uniform feeding device of powder to the roll tyre. Some operational parameters which affect the formation of pellets from a powder were investigated, and then the volume reduction of a powder was evaluated. The briquetting machine, popular in general industry, was modified to apply for the volume reduction of the powered radioactive wastes (dried concentrate, sludge, spent ion-exchange resin, ash, depleted uranium powder, and etc.). In this developed high volume reduction machine, the capacity was 25 ∼ 62.5 kg/h at the optimum conditions, and the estimated volume reduction was about 2.95 (2.74/0.93) on the basis of between a powder (bulk density = 0.93 g/cm{sup 3}) and the pellet (2.74 g/cm{sup 3}). But on the basis of 200L drum, the calculated volume reduction was about 1.34 in consideration of a void volume originated in the filling step of the pellets.

Hazardous and mixed waste transportation program

International Nuclear Information System (INIS)

Hohnstreiter, G.F.; Glass, R.E.; McAllaster, M.E.; Nigrey, P.J.; Trennel, A.J.; Yoshimura, H.R.

1993-01-01

Sandia National Laboratories (SNL) has developed a program to address the packaging needs associated with the transport of hazardous and mixed waste during the United States' Department of Energy (DOE) remediation efforts. The program addresses the technology needs associated with the transport of materials which have components that are radioactive and chemically hazardous. The mixed waste transportation activities focus on on-site specific applications of technology to the transport of hazardous and mixed wastes. These activities were identified at a series of DOE-sponsored workshops. These activities will be composed of the following: (1) packaging concepts, (2) chemical compatibility studies, and (3) systems studies. This paper will address activities in each of these areas. (J.P.N.)

Hazardous and Mixed Waste Transportation Program

International Nuclear Information System (INIS)

Hohnstreiter, G.F.; Glass, R.E.; McAllaster, M.E.; Nigrey, P.J.; Trennel, A.J.; Yoshimura, H.R.

1991-01-01

Sandia National Laboratories (SNL) has developed a program to address the packaging needs associated with the transport of hazardous and mixed waste during the United States' Department of Energy (DOE) remediation efforts. The program addresses the technology needs associated with the transport of materials which have components that are radioactive and chemically hazardous. The mixed waste transportation activities focus on on-site specific applications of technology to the transport of hazardous and mixed wastes. These activities were identified at a series of DOE-sponsored workshops. These activities will be composed of the following: (1) packaging concepts, (2) chemical compatibility studies, and (3) systems studies. This paper will address activities in each of these areas

The Mixed Waste Focus Area: Status and accomplishments

International Nuclear Information System (INIS)

Conner, J.E.

1997-01-01

The Mixed Waste Focus Area began operations in February of 1995. Its mission is to provide acceptable technologies that enable implementation of mixed waste treatment systems developed in partnership with end-users, stakeholders, tribal governments, and regulators. The MWFA will develop, demonstrate, and deliver implementable technologies for treatment of mixed waste within the DOE complex. Treatment refers to all post waste-generation activities including sampling and analysis, characterization, storage, processing, packaging, transportation, and disposal. The MWFA's mission arises from the Resources Conservation and Recovery Act (RCRA) as amended by the Federal Facility Compliance Act. Each DOE site facility that generates or stores mixed waste prepared a plan, the Site Treatment Plan, for developing treatment capacities and treating that waste. Agreements for each site were concluded with state regulators, resulting in Consent Orders providing enforceable milestones for achieving treatment of the waste. The paper discusses the implementation of the program, its status, accomplishments and goals for FY1996, and plans for 1997

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

International Nuclear Information System (INIS)

Mayberry, J.L.; Huebner, T.L.; Ross, W.; Nakaoka, R.; Schumacher, R.; Cunnane, J.; Singh, D.; Darnell, R.; Greenhalgh, W.

1993-08-01

This report presents information on low-level mixed waste forms.The descriptions of the low-level mixed waste (LLMW) streams that are considered by the Mixed Waste Integrated Program (MWIP) are given in Appendix A. This information was taken from descriptions generated by the Mixed Waste Treatment Program (MWTP). Appendix B provides a list of characteristic properties initially considered by the Final Waste Form (FWF) Working Group (WG). A description of facilities available to test the various FWFs discussed in Volume I of DOE/MWIP-3 are given in Appendix C. Appendix D provides a summary of numerous articles that were reviewed on testing of FWFS. Information that was collected by the tests on the characteristic properties considered in this report are documented in Appendix D. The articles reviewed are not a comprehensive list, but are provided to give an indication of the data that are available

Mixed waste disposal facility at the Nevada Test Site

International Nuclear Information System (INIS)

Dickman, P.T.; Kendall, E.W.

1987-01-01

In 1984, a law suit brought against DOE resulted in the requirement that DOE be subject to regulation by the state and US Environmental Protection Agency (EPA) for all hazardous wastes, including mixed wastes. Therefore, all DOE facilities generating, storing, treating, or disposing of mixed wastes will be regulated under the Resource Conservation and Recovery Act (RCTA). In FY 1985, DOE Headquarters requested DOE low-level waste (LLW) sites to apply for a RCRA Part B Permit to operate radioactive mixed waste facilities. An application for the Nevada Test Site (NTS) was prepared and submitted to the EPA, Region IX in November 1985 for review and approval. At that time, the state of Nevada had not yet received authorization for hazardous wastes nor had they applied for regulatory authority for mixed wastes. In October 1986, DOE Nevada Operations Office was informed by the Rocky Flats Plant that some past waste shipments to NTS contained trace quantities of hazardous substances. Under Colorado law, these wastes are defined as mixed. A DOE Headquarters task force was convened by the Under Secretary to investigate the situation. The task force concluded that DOE has a high priority need to develop a permitted mixed waste site and that DOE Nevada Operations Office should develop a fast track project to obtain this site and all necessary permits. The status and issues to be resolved on the permit for a mixed waste site are discussed

Solidification of radioactive liquid wastes, Treatment options for spent resins and concentrates - 16405

International Nuclear Information System (INIS)

Roth, Andreas

2009-01-01

Ion exchange is one of the most common and effective treatment methods for liquid radioactive waste. However, spent ion exchange resins are considered to be problematic waste that in many cases require special approaches and pre-conditioning during its immobilization to meet the acceptance criteria for disposal. Because of the function that they fulfill, spent ion exchange resins often contain high concentrations of radioactivity and pose special handling and treatment problems. Another very common method of liquid radioactive waste treatment and water cleaning is the evaporation or diaphragm filtration. Both treatment options offer a high volume reduction of the total volume of liquids treated but generate concentrates which contain high concentrations of radioactivity. Both mentioned waste streams, spent resins as well as concentrates, resulting from first step liquid radioactive waste treatment systems have to be conditioned in a suitable manner to achieve stable waste products for final disposal. Spent resin and concentrate treatment often appear as a specific task in decommissioning projects, because in the past those waste streams typically had been stored in tanks for the lifetime of the plant and needs to be retrieved, conditioned and packed prior to dismantling activities. Additionally a large amount of contaminated liquids will be generated by utilizing decontamination processes and needs to be processed further on. Such treatment options need to achieve waste products acceptable for final disposal, because due to the closure of the site no interim storage can be envisaged. The most common method of treatment of such waste streams is the solidification in a solid matrix with additional inactive material like cement, polymer etc. In the past good results have been achieved and the high concentration of radioactivity can be reduced by adding the inactive material. On the other hand, under the environment of limited space for interim storage and the absence

Mixed waste focus area alternative technologies workshop

International Nuclear Information System (INIS)

Borduin, L.C.; Palmer, B.A.; Pendergrass, J.A.

1995-01-01

This report documents the Mixed Waste Focus Area (MWFA)-sponsored Alternative Technology Workshop held in Salt Lake City, Utah, from January 24--27, 1995. The primary workshop goal was identifying potential applications for emerging technologies within the Options Analysis Team (OAT) ''wise'' configuration. Consistent with the scope of the OAT analysis, the review was limited to the Mixed Low-Level Waste (MLLW) fraction of DOE's mixed waste inventory. The Los Alamos team prepared workshop materials (databases and compilations) to be used as bases for participant review and recommendations. These materials derived from the Mixed Waste Inventory Report (MWIR) data base (May 1994), the Draft Site Treatment Plan (DSTP) data base, and the OAT treatment facility configuration of December 7, 1994. In reviewing workshop results, the reader should note several caveats regarding data limitations. Link-up of the MWIR and DSTP data bases, while representing the most comprehensive array of mixed waste information available at the time of the workshop, requires additional data to completely characterize all waste streams. A number of changes in waste identification (new and redefined streams) occurred during the interval from compilation of the data base to compilation of the DSTP data base with the end result that precise identification of radiological and contaminant characteristics was not possible for these streams. To a degree, these shortcomings compromise the workshop results; however, the preponderance of waste data was linked adequately, and therefore, these analyses should provide useful insight into potential applications of alternative technologies to DOE MLLW treatment facilities

Solidification paths in modified Inconel 625 weld overlay material

DEFF Research Database (Denmark)

Chandrasekaran, Karthik; Tiedje, Niels Skat; Hald, John

2009-01-01

Inconel 625 is commonly used for overlay welding to protect the base metal against high temperature corrosion. The efficiency of corrosion protection depends on effective mixing of the overlay weld with the base metal and the subsequent segregation of alloy elements during solidification....... Metallographic analysis of solidified samples of Inconel 625 with addition of selected elements is compared with thermodynamic modelling of segregation during solidification. The influence of changes in the melt chemistry on the formation of intermetallic phases during solidification is shown. In particular...

Hanford Site annual dangerous waste report: Volume 2, Generator dangerous waste report, radioactive mixed waste

International Nuclear Information System (INIS)

1994-01-01

This report contains information on radioactive mixed wastes at the Hanford Site. Information consists of shipment date, physical state, chemical nature, waste description, waste number, waste designation, weight, and waste designation

DOE acceptance of commercial mixed waste -- Studies are under way

Energy Technology Data Exchange (ETDEWEB)

Plummer, T.L. [Dept. of Energy, Washington, DC (United States). Technical Support Program; Owens, C.M. [Idaho National Engineering Lab., Idaho Falls, ID (United States). National Low-Level Waste Management Program

1993-03-01

The topic of the Department of Energy acceptance of commercial mixed waste at DOE facilities has been proposed by host States and compact regions that are developing low-level radioactive waste disposal facilities. States support the idea of DOE accepting commercial mixed waste because (a) very little commercial mixed waste is generated compared to generation by DOE facilities (Department of Energy--26,300 cubic meters annually vs. commercial--3400 cubic meters annually); (b) estimated costs for commercial disposal are estimated to be $15,000 to $40,000 per cubic foot; (c) once treatment capability becomes available, 70% of the current levels of commercial mixed waste will be eliminated, (d) some State laws prohibit the development of mixed waste disposal facilities in their States; (e) DOE is developing a nationwide strategy that will include treatment and disposal capacity for its own mixed waste and the incremental burden on the DOE facilities would be minuscule, and (6) no States are developing mixed waste disposal facilities. DOE senior management has repeatedly expressed willingness to consider investigating the feasibility of DOE accepting commercial mixed waste. In January 1991, Leo Duffy of the Department of energy met with members of the Low-Level Radioactive Waste Forum, which led to an agreement to explore such an arrangement. He stated that this seems like a cost-effective way to solve commercial mixed waste management problems.

National Institutes of Health: Mixed waste minimization and treatment

International Nuclear Information System (INIS)

1995-08-01

The Appalachian States Low-Level Radioactive Waste Commission requested the US Department of Energy's National Low-Level Waste Management Program (NLLWMP) to assist the biomedical community in becoming more knowledgeable about its mixed waste streams, to help minimize the mixed waste stream generated by the biomedical community, and to identify applicable treatment technologies for these mixed waste streams. As the first step in the waste minimization process, liquid low-level radioactive mixed waste (LLMW) streams generated at the National Institutes of Health (NIH) were characterized and combined into similar process categories. This report identifies possible waste minimization and treatment approaches for the LLMW generated by the biomedical community identified in DOE/LLW-208. In development of the report, on site meetings were conducted with NIH personnel responsible for generating each category of waste identified as lacking disposal options. Based on the meetings and general waste minimization guidelines, potential waste minimization options were identified

National Institutes of Health: Mixed waste minimization and treatment

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-08-01

The Appalachian States Low-Level Radioactive Waste Commission requested the US Department of Energy`s National Low-Level Waste Management Program (NLLWMP) to assist the biomedical community in becoming more knowledgeable about its mixed waste streams, to help minimize the mixed waste stream generated by the biomedical community, and to identify applicable treatment technologies for these mixed waste streams. As the first step in the waste minimization process, liquid low-level radioactive mixed waste (LLMW) streams generated at the National Institutes of Health (NIH) were characterized and combined into similar process categories. This report identifies possible waste minimization and treatment approaches for the LLMW generated by the biomedical community identified in DOE/LLW-208. In development of the report, on site meetings were conducted with NIH personnel responsible for generating each category of waste identified as lacking disposal options. Based on the meetings and general waste minimization guidelines, potential waste minimization options were identified.

Macroencapsulated and elemental lead mixed waste sites report

International Nuclear Information System (INIS)

Kalia, A.; Jacobson, R.

1996-09-01

The purpose of this study was to compile a list of the Macroencapsulated (MACRO) and Elemental Lead (EL) Mixed Wastes sites that will be treated and require disposal at the Nevada Test Site within the next five to ten years. The five sites selected were: Hanford Site, Richland, Washington; Idaho National Engineering Laboratory (INEL), Idaho Falls, Idaho; Oak Ridge National Laboratory (ORNL), Oak Ridge, Tennessee; Rocky Flats Environmental Technology (RF), Golden, Colorado; and Savannah River (SRS), Charleston, South Carolina. A summary of total lead mixed waste forms at the five selected DOE sites is described in Table E-1. This table provides a summary of total waste and grand total of the current inventory and five-year projected generation of lead mixed waste for each site. This report provides conclusions and recommendations for further investigations. The major conclusions are: (1) the quantity of lead mixed current inventory waste is 500.1 m 3 located at the INEL, and (2) the five sites contain several other waste types contaminated with mercury, organics, heavy metal solids, and mixed sludges

Mixed and Low-Level Waste Treatment Facility project

International Nuclear Information System (INIS)

1992-04-01

Mixed and low-level wastes generated at the Idaho National Engineering Laboratory (INEL) are required to be managed according to applicable State and Federal regulations, and Department of Energy Orders that provide for the protection of human health and the environment. The Mixed and Low-Level Waste Treatment Facility Project was chartered in 1991, by the Department of Energy to provide treatment capability for these mixed and low-level waste streams. The first project task consisted of conducting engineering studies to identify the waste streams, their potential treatment strategies, and the requirements that would be imposed on the waste streams and the facilities used to process them. The engineering studies, initiated in July 1991, identified 37 mixed waste streams, and 55 low-level waste streams. This report documents the waste stream information and potential treatment strategies, as well as the regulatory requirements for the Department of Energy-owned treatment facility option. The total report comprises three volumes and two appendices. This report consists of Volume 1, which explains the overall program mission, the guiding assumptions for the engineering studies, and summarizes the waste stream and regulatory information, and Volume 2, the Waste Stream Technical Summary which, encompasses the studies conducted to identify the INEL's waste streams and their potential treatment strategies

Method and apparatus for solidifying radioactive waste

International Nuclear Information System (INIS)

Kadota, Hiroko; Kikuchi, Makoto; Tsuchiya, Hiroyuki; Tamada, Shin.

1989-01-01

The present invention concerns a method of solidifying radioactive wastes that generate heat with water curing solidifying material and the object there of is suppress the effect of heat generation of the wastes given on the solidification material. That is, it is a feature of the invention to inject water content contained in the water curable solidification material in the form of ice into the wastes. Thus, since the water content in the water curable solidification material is ice, the solidification products can be obtained by way of the following three steps: (1) ice is dissolved into water, (2) solid content of the solidification material is dissolved into water, and(3) curing reaction of the solidification material is started. Acccordingly, since the heat generated from the wastes contributes as heat of reaction when ice is dissolved into water till the solidification material has been completely filled, promotion for the curing reaction causing problems so far can be suppressed to enable easy filling. Then, after the completion of the filling of the solidification material, the heat of the wastes has an effect of promoting the second and the third steps described above to accelerate the curing reaction. (K.M.)

Mixed waste management in Washington and the Northwest Compact Region

International Nuclear Information System (INIS)

Carlin, E.M.

1988-01-01

The state of Washington's concerns about the management of mixed waste have evolved over the past year. One concern that receives increasing attention is the Northwest Compact Region's need to plan for disposal of its own mixed waste. An informal survey of the region's potential mixed waste generators has indicated that mixed waste volumes are low. However, the opening of a disposal facility may result in increased waste volumes. A preliminary proposal for such a facility has been reviewed by the federal and state agencies that dually regulate mixed waste. Initial conclusions reached by the regulators are presented

Alternative solidification techniques for radioactive ion exchange resins and liquid concentrates

International Nuclear Information System (INIS)

Thegerstroem, C.

1980-01-01

Methods, that are used or are under development for solidification of radioactive ion exchange resins or liquid concentrates, utilize normally cement, bitumen or some polymere as matrix material. This report contains a review and a description of these solidification processes and their products, especially of relatively new techniques that are under development in different countries. It is possible that solidification in thermosetting resins will be more used in the future, especially when product quality requirements are high (for instance when solidifying medium level resins) or when special waste categories has to be solidified. However it is not probable that thermosetting resins will be extensively used in a broad application as matrix material. In that case the methods are to complicated and expensive compared to, for instance, solidification in concrete. Systems for incorporation in polyesteremulsions (Dow-process) have a potential as they are quite simple and can accept a large variation of liquid wastes. Some methods in an early stage of development (for instance Inert Carrier Radwaste Process) will have to be tested in active application before they can be further evaluated. (author)

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

MIxed Waste Integrated Program (MWIP): Technology summary

International Nuclear Information System (INIS)

1994-02-01

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

Molten salt destruction process for mixed wastes

International Nuclear Information System (INIS)

Upadhye, R.S.; Wilder, J.G.; Karlsen, C.E.

1993-04-01

We are developing an advanced two-stage process for the treatment of mixed wastes, which contain both hazardous and radioactive components. The wastes, together with an oxidant gas, such as air, are injected into a bed of molten salt comprising a mixture of sodium-, potassium-, and lithium-carbonates, with a melting point of about 580 degree C. The organic constituents of the mixed waste are destroyed through the combined effect of pyrolysis and oxidation. Heteroatoms. such as chlorine, in the mixed waste form stable salts, such as sodium chloride, and are retained in the melt. The radioactive actinides in the mixed waste are also retained in the melt because of the combined action of wetting and partial dissolution. The original process, consists of a one-stage unit, operated at 900--1000 degree C. The advanced two-stage process has two stages, one for pyrolysis and one for oxidation. The pyrolysis stage is designed to operate at 700 degree C. The oxidation stage can be operated at a higher temperature, if necessary

Mixed Waste Integrated Program emerging technology development

Energy Technology Data Exchange (ETDEWEB)

Berry, J.B. [Oak Ridge National Lab., TN (United States); Hart, P.W. [USDOE, Washington, DC (United States)

1994-06-01

The US Department of Energy (DOE) is responsible for the management and treatment of its mixed low-level wastes (MLLW). MLLW are regulated under both the Resource Conservation and Recovery Act and various DOE orders. Over the next 5 years, DOE will manage over 1.2 m{sup 3} of MLLW and mixed transuranic (MTRU) wastes. In order to successfully manage and treat these mixed wastes, DOE must adapt and develop characterization, treatment, and disposal technologies which will meet performance criteria, regulatory approvals, and public acceptance. Although technology to treat MLLW is not currently available without modification, DOE is committed to developing such treatment technologies and demonstrating them at the field scale by FY 1997. The Office of Research and Development`s Mixed Waste Integrated Program (MWIP) within the DOE Office of Environmental Management (EM), OfFice of Technology Development, is responsible for the development and demonstration of such technologies for MLLW and MTRU wastes. MWIP advocates and sponsors expedited technology development and demonstrations for the treatment of MLLW.

Hazardous and mixed waste management at UMTRA sites

International Nuclear Information System (INIS)

Hampill, H.G.

1988-01-01

During the early stages of the Uranium Mill Tailings Remedial Action Project, there were some serious questions regarding the ownership of and consequently the responsibility for disposal of hazardous wastes at UMTRA sites. In addition to State and Indian Tribe waste disposal regulations, UMTRA must also conform to guidelines established by the NRC, OSHA, EPA, and DOT. Because of the differing regulatory thrusts of these agencies, UMTRA has to be vigilant in order to ensure that the disposal of each parcel of waste material is in compliance with all regulations. Mixed-waste disposal presents a particularly difficult problem. No single agency is willing to lay claim to the regulation of mixed-wastes, and no conventional waste disposal facility is willing to accept it. Consequently, the disposal of each lot of mixed-waste at UMTRA sites must be handled on a case by case basis. A recently published position paper which spells out UMTRA policy on waste materials indicates that wastes found at UMTRA sites are either residual radioactive wastes, or mixed-wastes, or for the disposal of hazardous waste is determined by the time the original material arrived. If it arrived prior to the termination of the AEC uranium supply contract, its disposal is the responsibility of UMTRA. If it arrived after the end of the contract, the responsibility for disposal lies with the former operator

Bioprocessing of low-level radioactive and mixed hazard wastes

International Nuclear Information System (INIS)

Stoner, D.L.

1990-01-01

Biologically-based treatment technologies are currently being developed at the Idaho National Engineering Laboratory (INEL) to aid in volume reduction and/or reclassification of low-level radioactive and mixed hazardous wastes prior to processing for disposal. The approaches taken to treat low-level radioactive and mixed wastes will reflect the physical (e.g., liquid, solid, slurry) and chemical (inorganic and/or organic) nature of the waste material being processed. Bioprocessing utilizes the diverse metabolic and biochemical characteristics of microorganisms. The application of bioadsorption and bioflocculation to reduce the volume of low-level radioactive waste are strategies comparable to the use of ion-exchange resins and coagulants that are currently used in waste reduction processes. Mixed hazardous waste would require organic as well as radionuclide treatment processes. Biodegradation of organic wastes or bioemulsification could be used in conjunction with radioisotope bioadsorption methods to treat mixed hazardous radioactive wastes. The degradation of the organic constituents of mixed wastes can be considered an alternative to incineration, while the use of bioemulsification may simply be used as a means to separate inorganic and organics to enable reclassification of wastes. The proposed technology base for the biological treatment of low-level radioactive and mixed hazardous waste has been established. Biodegradation of a variety of organic compounds that are typically found in mixed hazardous wastes has been demonstrated, degradative pathways determined and the nutritional requirements of the microorganisms are understood. Accumulation, adsorption and concentration of heavy and transition metal species and transuranics by microorganisms is widely recognized. Work at the INEL focuses on the application of demonstrated microbial transformations to process development

Hanford Site annual dangerous waste report: Volume 4, Waste Management Facility report, Radioactive mixed waste

International Nuclear Information System (INIS)

1994-01-01

This report contains information on radioactive mixed wastes at the Hanford Site. Information consists of shipment date, physical state, chemical nature, waste description, handling method and containment vessel, waste number, waste designation and amount of waste

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)

Savannah River Plant Separations Department mixed waste program

International Nuclear Information System (INIS)

Wierzbicki, W.M.

1988-01-01

The Department of Energy's (DOE) Savannah River Plant (SRP) generates radioactive and mixed waste as a result of the manufacture of nuclear material for the national defense program. The radioactive portion of the mixed waste and all nonhazardous radioactive wastes would continue to be regulated by DOE under the Atomic Energy Act. The Separations Department is the largest generator of solid radioactive waste at the Savannah River Plant. Over the last three years, the Separations Department has developed and implemented a program to characterize candidate mixed-waste streams. The program consisted of facility personnel interviews, a waste-generation characterization program and waste testing to determine whether a particular waste form was hazardous. The Separations Department changed waste-handling practices and procedures to meet the requirements of the generator standards. For each Separation Department Facility, staging areas were established, inventory and reporting requirements were developed, operating procedures were revised to ensure proper waste handling, and personnel were provided hazardous waste training. To emphasize the importance of the new requirements, a newsletter was developed and issued to all Separations supervisory personnel

Secondary Waste Form Development and Optimization—Cast Stone

Energy Technology Data Exchange (ETDEWEB)

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

2011-07-14

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

Liquid secondary waste: Waste form formulation and qualification

Energy Technology Data Exchange (ETDEWEB)

Cozzi, A. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Dixon, K. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hill, K. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Nichols, R. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-07-31

The Hanford Site Effluent Treatment Facility (ETF) currently treats aqueous waste streams generated during site cleanup activities. When the Hanford Tank Waste Treatment and Immobilization Plant (WTP) begins operations, including Direct Feed Low Activity Waste (DFLAW) vitrification, a liquid secondary waste (LSW) stream from the WTP will need to be treated. The volume of effluent for treatment at the ETF will increase significantly. The powdered salt waste form produced by the ETF will be replaced by a stabilized solidified waste form for disposal in Hanford’s Integrated Disposal Facility (IDF). Washington River Protection Solutions is implementing a Secondary Liquid Waste Immobilization Technology Development Plan to address the technology needs for a waste form and solidification process to treat the increased volume of waste planned for disposal at the IDF. Waste form testing to support this plan is composed of work in the near term to provide data as input to a performance assessment (PA) for Hanford’s IDF. In 2015, three Hanford Liquid Secondary Waste simulants were developed based on existing and projected waste streams. Using these waste simulants, fourteen mixes of Hanford Liquid Secondary Waste were prepared and tested varying the waste simulant, the water-to-dry materials ratio, and the dry materials blend composition.1 In FY16, testing was performed using a simulant of the EMF process condensate blended with the caustic scrubber—from the Low Activity Waste (LAW) melter—, processed through the ETF. The initial EMF-16 simulant will be based on modeling efforts performed to determine the mass balance of the ETF for the DFLAW.2 The compressive strength of all of the mixes exceeded the target of 3.4 MPa (500 psi) to meet the requirements identified as potential IDF Waste Acceptance Criteria in Table 1 of the Secondary Liquid Waste Immobilization Technology Development Plan.3 The hydraulic properties of the waste forms tested (hydraulic conductivity

Mixed Waste Treatment Project: Computer simulations of integrated flowsheets

International Nuclear Information System (INIS)

Dietsche, L.J.

1993-12-01

The disposal of mixed waste, that is waste containing both hazardous and radioactive components, is a challenging waste management problem of particular concern to DOE sites throughout the United States. Traditional technologies used for the destruction of hazardous wastes need to be re-evaluated for their ability to handle mixed wastes, and in some cases new technologies need to be developed. The Mixed Waste Treatment Project (MWTP) was set up by DOE's Waste Operations Program (EM30) to provide guidance on mixed waste treatment options. One of MWTP's charters is to develop flowsheets for prototype integrated mixed waste treatment facilities which can serve as models for sites developing their own treatment strategies. Evaluation of these flowsheets is being facilitated through the use of computer modelling. The objective of the flowsheet simulations is to provide mass and energy balances, product compositions, and equipment sizing (leading to cost) information. The modelled flowsheets need to be easily modified to examine how alternative technologies and varying feed streams effect the overall integrated process. One such commercially available simulation program is ASPEN PLUS. This report contains details of the Aspen Plus program

Is radioactive mixed waste packaging and transportation really a problem

International Nuclear Information System (INIS)

McCall, D.L.; Calihan, T.W. III.

1992-01-01

Recently, there has been significant concern expressed in the nuclear community over the packaging and transportation of radioactive mixed waste under US Department of Transportation regulation. This concern has grown more intense over the last 5 to 10 years. Generators and regulators have realized that much of the waste shipped as ''low-level radioactive waste'' was in fact ''radioactive mixed waste'' and that these wastes pose unique transportation and disposal problems. Radioactive mixed wastes must, therefore, be correctly identified and classed for shipment. If must also be packaged, marked, labeled, and otherwise prepared to ensure safe transportation and meet applicable storage and disposal requirements, when established. This paper discusses regulations applicable to the packaging and transportation of radioactive mixed waste and identifies effective methods that waste shippers can adopt to meet the current transportation requirements. This paper will include a characterization and description of the waste, authorized packaging, and hazard communication requirements during transportation. Case studies will be sued to assist generators in understanding mixed waste shipment requirements and clarify the requirements necessary to establish a waste shipment program. Although management and disposal of radioactive mixed waste is clearly a critical issue, packaging and transportation of these waste materials is well defined in existing US Department of Transportation hazardous material regulations

Minimization of mixed waste in explosive testing operations

International Nuclear Information System (INIS)

Gonzalez, M.A.; Sator, F.E.; Simmons, L.F.

1993-02-01

In the 1970s and 1980s, efforts to manage mixed waste and reduce pollution focused largely on post-process measures. In the late 1980s, the approach to waste management and pollution control changed, focusing on minimization and prevention rather than abatement, treatment, and disposal. The new approach, and the formulated guidance from the US Department of Energy, was to take all necessary measures to minimize waste and prevent the release of pollutants to the environment. Two measures emphasized in particular were source reduction (reducing the volume and toxicity of the waste source) and recycling. In 1988, a waste minimization and pollution prevention program was initiated at Site 300, where the Lawrence Livermore National Laboratory (LLNL) conducts explosives testing. LLNL's Defense Systems/Nuclear Design (DS/ND) Program has adopted a variety of conservation techniques to minimize waste generation and cut disposal costs associated with ongoing operations. The techniques include minimizing the generation of depleted uranium and lead mixed waste through inventory control and material substitution measures and through developing a management system to recycle surplus explosives. The changes implemented have reduced annual mixed waste volumes by more than 95% and reduced overall radioactive waste generation (low-level and mixed) by more than 75%. The measures employed were cost-effective and easily implemented

PD and I works to improve the waste solidification performance of Angra 1

International Nuclear Information System (INIS)

Tello, Cledola C.O. de; Gomes, Nelson J.P.O.

2009-01-01

Angra 1 Power Plant is the first Brazilian NPP, and started its commercial operation in 1985. The wastes from its operation are spent ion exchange resins, expended filter cartridges, boric acid evaporator concentrates; and solid wastes such as gloves, cleaning tools, plastics and protective clothing. Liquid waste streams are treated by evaporation to reduce the volume, and the concentrate is solidified in cement. The original cementation unit was designed to meet the US product acceptance criteria of the early seventies, when environmental and safety requirements were less restrictive than now. This system had a very low efficiency, and the waste product quality was very poor. A new solidification plant was bought, and R and D project was carried out to study and develop a suitable cementation process and the PCP - process control program - to guarantee the licensing of the plant and maximize its efficiency using Brazilian materials. To accomplish these objectives, a huge amount of experiments were carried on. Tests were performed to determine the main properties of the process and the product, e.g. the workability, set time, compressive strength (stress resistance) and leachability. After reaching the established parameters, the mixture system was optimized. The paddle and container designs were modified and also the motor specifications, so that the process efficiency could be improved. The performance of this new system is increased allowing solidifying from 56 till 69% in volume of waste, much better in comparison of the old one, whose maximum value was about 20%. (author)

Volume reduction/solidification of liquid radioactive waste using bitumen at Ontario hydro's Bruce nuclear generating station open-quotes Aclose quotes

International Nuclear Information System (INIS)

Day, J.E.; Baker, R.L.

1994-01-01

Ontario Hydro at the Bruce Nuclear Generating Station open-quotes Aclose quotes has undertaken a program to render the station's liquid radioactive waste suitable for discharge to Lake Huron by removing sufficient radiological and chemical contaminants from five different plant waste streams. The contaminants will be immobilized and stored at on-site radioactive waste storage facilities and the purified streams will be discharged. The discharge targets established by Ontario Hydro are set well below the limits established by the Ontario Ministry of Environment (MOE) and are based on the Best Available Technology Economically Achievable Approach (B.A.T.E.A.). ADTECHS Corporation has been selected by Ontario Hydro to provide volume reduction/solidification technology for one of the five waste streams. The system will dry and immobilize the contaminants from a liquid waste stream in emulsified asphalt using thin film evaporation technology

Comparison of costs for solidification of high-level radioactive waste solutions: glass monoliths vs metal matrices

International Nuclear Information System (INIS)

Jardine, L.J.; Carlton, R.E.; Steindler, M.J.

1981-05-01

A comparative economic analysis was made of four solidification processes for liquid high-level radioactive waste. Two processes produced borosilicate glass monoliths and two others produced metal matrix composites of lead and borosilicate glass beads and lead and supercalcine pellets. Within the uncertainties of the cost (1979 dollars) estimates, the cost of the four processes was about the same, with the major cost component being the cost of the primary building structure. Equipment costs and operating and maintenance costs formed only a small portion of the building structure costs for all processes

Waste form development/test

International Nuclear Information System (INIS)

Kalb, P.D.; Colombo, P.

1983-01-01

The main objective of this study is to investigate new solidification agents relative to their potential application to wastes generated by advanced high volume reduction technologies, e.g., incinerator ash, dry solids, and ion exchange resins. Candidate materials selected for the solidification of these wastes include a modified sulfur cement and low-density polyethylene, neither of which are currently employed commerically for the solidification of low-level waste (LLW). As both the modified sulfur cement and the polyethylene are thermoplastic materials, a heated screw type extruder is utilized in the production of waste form samples for testing and evaluation. In this regard, work is being conducted to determine the range of conditions under which these solidification agents can be satisfactorily applied to the specific LLW streams and to provide information relevant to operating parameters and process control

Scope and approach to management of mixed wastes: introduction to the session

International Nuclear Information System (INIS)

Ausmus, B.S.

1986-01-01

Mixed wastes are those that are termed both radioactive and chemically hazardous based on regulatory criteria in the United States. Historically, mixed wastes that could be classified as radioactive wastes were treated, stored, and disposed under statutes governing radioactive wastes. In recent years, it has become apparent that: (a) hazardous wastes are generated in nuclear facilities; (b) many wastes are both radioactive and chemically hazardous; and (c) the management of chemically hazardous wastes and mixed wastes requires reexamination of current waste treatment/disposal methods and development/implementation of modified methods. The purpose of this session is to discuss specific aspects of the mixed waste management problems and to provide a forum for discussion of the technical and institutional barriers to problem solutions. The paper addresses several mixed waste problems and current approaches to their solutions, including: (1) mixed waste management in fuel cycle facilities; (2) mixed waste management in a US Dept. of Energy production facility; and (3) mixed wastes impacts on 10CFR61 compliance. Technical and institutional approaches to mixed waste management are explored in three areas: (1) alternatives for treatment prior to shallow land disposal; (2) potential benefits of recovery of strategic/critical materials from mixed wastes; and (3) shallow land disposal system compatibilities/problems

Development of radioactive waste treatment technique

International Nuclear Information System (INIS)

Kikuchi, Makoto; Amamiya, Shigeru; Yusa, Hideo.

1984-01-01

The techniques of radioactive waste treatment are generally reviewed, placing emphasis on volume reduction and solidification techniques. After a brief description on the general process of radioactive waste treatment, some special technologies being developed by Hitachi Ltd. are explained. From the viewpoints of the volume reduction, long term management and final disposal of wastes, the pelletization of dried waste and the solidification with inorganic substances are considered. One of the features of the pelletization system is to treat various kinds of wastes such as concentrated liquid wastes and used resins by the same system. The flow diagram of the system and its special features are shown. The volume reduction achieved by this system as compared to the conventional method is about 1/7. The first commercial plant for the treatment of concentrated liquid waste is scheduled to begin operation in June, 1984. As for the solidification technique for waste disposal, the use of cement glass is considered. The solidification system being developed is shortly described. (Aoki, K.)

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

Radioactive and mixed waste management plan for the Lawrence Berkeley Laboratory Hazardous Waste Handling Facility

International Nuclear Information System (INIS)

1995-01-01

This Radioactive and Mixed Waste Management Plan for the Hazardous Waste Handling Facility at Lawrence Berkeley Laboratory is written to meet the requirements for an annual report of radioactive and mixed waste management activities outlined in DOE Order 5820.2A. Radioactive and mixed waste management activities during FY 1994 listed here include principal regulatory and environmental issues and the degree to which planned activities were accomplished

Finite-element solidification modelling of metals and binary alloys

International Nuclear Information System (INIS)

Mathew, P.M.

1986-12-01

In the Canadian Nuclear Fuel Waste Management Program, cast metals and alloys are being evaluated for their ability to support a metallic fuel waste container shell under disposal vault conditions and to determine their performance as an additional barrier to radionuclide release. These materials would be cast to fill residual free space inside the container and allowed to solidify without major voids. To model their solidification characteristics following casting, a finite-element model, FAXMOD-3, was adopted. Input parameters were modified to account for the latent heat of fusion of the metals and alloys considered. This report describes the development of the solidification model and its theoretical verification. To model the solidification of pure metals and alloys that melt at a distinct temperature, the latent heat of fusion was incorporated as a double-ramp function in the specific heat-temperature relationship, within an interval of +- 1 K around the solidification temperature. Comparison of calculated results for lead, tin and lead-tin eutectic melts, unidirectionally cooled with and without superheat, showed good agreement with an alternative technique called the integral profile method. To model the solidification of alloys that melt over a temperature interval, the fraction of solid in the solid-liquid region, as calculated from the Scheil equation, was used to determine the fraction of latent heat to be liberated over a temperature interval within the solid-liquid zone. Comparison of calculated results for unidirectionally cooled aluminum-4 wt.% copper melt, with and without superheat, showed good agreement with alternative finite-difference techniques

Bioprocessing of a stored mixed liquid waste

Energy Technology Data Exchange (ETDEWEB)

Wolfram, J.H.; Rogers, R.D. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Finney, R. [Mound Applied Technologies, Miamisburg, OH (United States)] [and others

1995-12-31

This paper describes the development and results of a demonstration for a continuous bioprocess for mixed waste treatment. A key element of the process is an unique microbial strain which tolerates high levels of aromatic solvents and surfactants. This microorganism is the biocatalysis of the continuous flow system designed for the processing of stored liquid scintillation wastes. During the past year a process demonstration has been conducted on commercial formulation of liquid scintillation cocktails (LSC). Based on data obtained from this demonstration, the Ohio EPA granted the Mound Applied Technologies Lab a treatability permit allowing the limited processing of actual mixed waste. Since August 1994, the system has been successfully processing stored, {open_quotes}hot{close_quotes} LSC waste. The initial LSC waste fed into the system contained 11% pseudocumene and detectable quantities of plutonium. Another treated waste stream contained pseudocumene and tritium. Data from this initial work shows that the hazardous organic solvent, and pseudocumene have been removed due to processing, leaving the aqueous low level radioactive waste. Results to date have shown that living cells are not affected by the dissolved plutonium and that 95% of the plutonium was sorbed to the biomass. This paper discusses the bioprocess, rates of processing, effluent, and the implications of bioprocessing for mixed waste management.

Treatment of Petroleum Sludge By Using Solidification/Stabilization (S/S) Method : Effect of Hydration Days to Heavy Metals Leaching and Strength

Science.gov (United States)

Murshid, N.; Kamil, N. A. F. M.; Kadir, A. A.

2018-04-01

Petroleum sludge is one of the major solid wastes generated in the petroleum industry. Generally, there are numbers of heavy metals in petroleum sludge and one treatment that is gaining prominence to treat a variety of mixed organic and inorganic waste is solidification/stabilization (S/S) method. The treatment protects human health and the environment by immobilizing contaminants within the treated material and prevents migration of the contaminants. In this study, solidification/stabilization (S/S) method has been used to treat the petroleum sludge. The comparison of hydration days, namely, 7th and 28th days in these cement-based waste materials were studied by using Synthetic Precipitate Leaching Procedure (SPLP). The results were compared to the United States Environmental Protection Agency (USEPA) standards. The results for leaching test concluded that less percentage OPC gave maximum concentration of heavy metals leaching due to deficient in Calcium Oxide (CaO), which is can caused weak solidification in the mixture. Physical and mechanical properties conducted such as compressive strength and density test. From the results, it shows addition up to of 30percentage PS give results which comply with minimum landfill dispose limit. The results shows correlation between strength and density are strong regression coefficient of 82.7%. In conclusion, S/S method can be alternative disposal method for PS in the same time complies with standard for minimum landfill disposal limit. The results for leaching test concluded the less OPC percentage gave maximum concentration of heavy metals leaching.

Mixed waste: An alternative solution. The utility perspective

International Nuclear Information System (INIS)

Seizert, R.D.

1988-01-01

The issue of mixed waste is one of significant interest to the utility industry. The interest is focused on the current regulatory scheme of dual regulation. A fundamental concern of the commercial nuclear utilities resulting from dual regulation is that there are currently no facilities in the US to dispose of mixed low-level radioactive and hazardous waste. The lack of available sites renders mixed waste an orphan, requiring generators of such material to store the waste on-site. This in turn causes commercial nuclear power plants to be subjected to the full gamut of Environmental Protection Agency (EPA) Resource Conservation and Recovery Act (RCRA) regulation in addition to the existing Nuclear Regulatory Commission (NRC) regulations. Superimposing dual regulatory schemes will have impacts which extend far beyond the mere management of mixed waste. Certainly the burdens, complexities and costs of complying with the overlapping regulatory schemes will not have a commensurate increase in protection from the real risks being addressed. For these reasons, the commercial nuclear utility industry is working toward an alternative solution which will protect the public health and the environment from all hazards of mixed waste and will minimize the impacts on both the regulators and the regulated community

Recommendations for continuous emissions monitoring of mixed waste incinerators

International Nuclear Information System (INIS)

Quigley, G.P.

1992-01-01

Considerable quantities of incinerable mixed waste are being stored in and generated by the DOE complex. Mixed waste is defined as containing a hazardous component and a radioactive component. At the present time, there is only one incinerator in the complex which has the proper TSCA and RCRA permits to handle mixed waste. This report describes monitoring techniques needed for the incinerator

Tracking mixed waste from environmental restoration through waste management for the Federal Facility Compliance Act

International Nuclear Information System (INIS)

Isbell, D.; Tolbert-Smith, M.; MacDonell, M.; Peterson, J.

1994-01-01

The Federal Facility Compliance Act required the US Department of Energy (DOE) to prepare an inventory report that presents comprehensive information on mixed wastes. Additional documents, such as site treatment plans, were also required of facilities with mixed waste. For a number of reasons, not all DOE mixed waste sites are able to provide detailed characterization and planning data at this time. Thus, an effort is currently under way to develop a reporting format that will permit mixed waste information across the DOE complex to be tracked as it becomes available

Identification of permit and waste acceptance criteria provisions requiring modification for acceptance of commercial mixed waste

International Nuclear Information System (INIS)

1994-03-01

In October 1990, representatives of States and compact regions requested that the US Department of Energy (DOE) explore an agreement with host States and compact regions under which DOE would accept commercial mixed low-level radioactive waste (LLW) at DOE's own treatment and disposal facilities. A program for DOE management of commercial mixed waste is made potentially more attractive in light of the low commercial mixed waste volumes, high regulatory burdens, public opposition to new disposal sites, and relatively high cost of constructing commercial disposal facilities. Several studies were identified as essential in determining the feasibility of DOE accepting commercial mixed waste for disposal. The purpose of this report is to identify any current or proposed waste acceptance criteria (WAC) or Resource Conservation and Recovery Act (RCRA) provisions that would have to be modified for commercial mixed waste acceptance at specified DOE facilities. Following the introduction, Section 2 of this report (a) provides a background summary of existing and proposed mixed waste disposal facilities at each DOE site, and (b) summarizes the status of any RCRA Part B permit and WAC provisions relating to the disposal of mixed waste, including provisions relating to acceptance of offsite waste. Section 3 provides overall conclusions regarding the current status and permit modifications that must be implemented in order to grant DOE sites authority under their permits to accept commercial mixed waste for disposal. Section 4 contains a list of references

Mixed waste focus area technical baseline report. Volume 2

International Nuclear Information System (INIS)

1997-04-01

As part of its overall program, the MWFA uses a national mixed waste data set to develop approaches for treating mixed waste that cannot be treated using existing capabilities at DOE or commercial facilities. The current data set was originally compiled under the auspices of the 1995 Mixed Waste Inventory Report. The data set has been updated over the past two years based on Site Treatment Plan revisions and clarifications provided by individual sites. The current data set is maintained by the MWFA staff and is known as MWFA97. In 1996, the MWFA developed waste groupings, process flow diagrams, and treatment train diagrams to systematically model the treatment of all mixed waste in the DOE complex. The purpose of the modeling process was to identify treatment gaps and corresponding technology development needs for the DOE complex. Each diagram provides the general steps needed to treat a specific type of waste. The NWFA categorized each MWFA97 waste stream by waste group, treatment train, and process flow. Appendices B through F provide the complete listing of waste streams by waste group, treatment train, and process flow. The MWFA97 waste strewn information provided in the appendices is defined in Table A-1

Method of solidifying radioactive laundry wastes

International Nuclear Information System (INIS)

Yasumura, Keijiro

1984-01-01

Purpose: To enable to solidify radioactive laundry wastes containing non-ionic liquid detergents less solidifiable by plastic solidification process in liquid laundry wastes for cloths or the likes discharged from a nuclear power plant. Method: Radioactive laundry wastes are solidified by using plastic solidifying agent comprising, as a main ingredient, unsaturated polyester resins and methylmethacrylate monomers. The plastic solidifying agents usable herein include, for example, unsaturated polyester resins prepared by condensating maleic anhydride and phthalic anhydride with propylene glycol and incorporated with methylmethacrylate monomers. The mixing ratio of the methylmethacrylate monomers is preferably 30 % by weight based on the unsaturated polyester resins. (Aizawa, K.)

Phase III (full scale) agitated mixing test plan

International Nuclear Information System (INIS)

Ruff, D.T.

1994-01-01

Waste Receiving and Processing Facility Module 2A (WRAP 2A) is the proposed second module of the WRAP facility. This facility will provide the required treatment for contact Handled (CH) Low Level (LL) Mixed Waste (MW) to allow its permanent disposal. Solidification of a portion of this waste using a cement based grout has been selected in order to reduce the toxicity and mobility of the waste in the disposal site. Mixing of the waste with the cement paste and material handling constraints/requirements associated with the mixed material is, therefore, a key process in the overall treatment strategy. This test plan addresses Phase 3, Full Scale Testing. The objectives of these tests are to determine if there are scale-up issues associated with the mixing results obtained in Phase 1 and 2 mixing tests, verify the workability of mixtures resulting from previous formulation development efforts (Waste Immobilization Development [WID]), and provide a baseline for WRAP 2A mixing equipment design. To this end, the following objectives are of particular interest: determine geometric influence of mixing blade at full scale (i.e., size, type, and location: height/offset); determine if similar results in terms of mixing effectiveness and product quality are achievable at this scale; determine if vibration is as effective at this larger scale in fluidizing the mixture and aiding in cleaning the vessel; determine if baffles or sweeping blades are needed to aid in mixing at the larger size and for cleaning the vessel; and determine quality of the poured monolithic product and investigate exotherm and filling influences at this larger size

Supported liquid membrane based removal of lead(II) and cadmium(II) from mixed feed: Conversion to solid waste by precipitation

Energy Technology Data Exchange (ETDEWEB)

Bhatluri, Kamal Kumar; Manna, Mriganka Sekhar; Ghoshal, Aloke Kumar; Saha, Prabirkumar, E-mail: p.saha@iitg.ac.in

2015-12-15

Highlights: • Simultaneous removal of two heavy metals lead and cadmium. • Conversion of liquid waste to solid precipitation. • Precipitation facilitates the metals transportation through LM. • Solidification of liquid waste minimizes the final removal of waste. - Abstract: Simultaneous removal of two heavy metals, lead(II) and cadmium(II), from mixed feed using supported liquid membrane (SLM) based technique is investigated in this work. The carrier-solvent combination of “sodium salt of Di-2-ethylhexylphosphoric acid (D2EHPA) (4% w/w) in environmentally benign coconut oil” was immobilized into the pores of solid polymeric polyvinylidene fluoride (PVDF) support. Sodium carbonate (Na{sub 2}CO{sub 3}) was used as the stripping agent. Carbonate salts of lead(II) and cadmium(II) were formed in the stripping side interface and they were insoluble in water leading to precipitation inside the stripping solution. The transportation of solute is positively affected due to the precipitation. Lead(II) removal was found to be preferential due to its favorable electronic configuration. The conversion of the liquid waste to the solid one was added advantage for the final removal of hazardous heavy metals.

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

The mixed waste landfill integrated demonstration

International Nuclear Information System (INIS)

Burford, T.D.; Williams, C.V.

1994-01-01

The Mixed Waste Landfill Integrated Demonstration (MWLID) focuses on ''in-situ'' characterization, monitoring, remediation, and containment of landfills in arid environments that contain hazardous and mixed waste. The MWLID mission is to assess, demonstrate, and transfer technologies and systems that lead to faster, better, cheaper, and safer cleanup. Most important, the demonstrated technologies will be evaluated against the baseline of conventional technologies and systems. The comparison will include the cost, efficiency, risk, and feasibility of using these innovative technologies at other sites

Mixed and Low-Level Waste Treatment Facility Project

International Nuclear Information System (INIS)

1992-04-01

Mixed and low-level wastes generated at the Idaho National Engineering Laboratory (INEL) are required to be managed according to applicable State and Federal regulations, and Department of Energy Orders that provide for the protection of human health and the environment. The Mixed and Low-Level Waste Treatment Facility Project was chartered in 1991, by the Department of Energy to provide treatment capability for these mixed and low-level waste streams. The first project task consisted of conducting engineering studies to identify the waste streams, their potential treatment strategies, and the requirements that would be imposed on the waste streams and the facilities used to process them. This report documents those studies so the project can continue with an evaluation of programmatic options, system tradeoff studies, and the conceptual design phase of the project. This report, appendix B, comprises the engineering design files for this project study. The engineering design files document each waste steam, its characteristics, and identified treatment strategies

Presidential Rapid Commercialization Initiative for mixed waste solvent extraction

International Nuclear Information System (INIS)

Honigford, L.; Dilday, D.; Cook, D.

1997-01-01

Recently, the Fernald Environmental Management Project (FEMP) has made some major steps in mixed waste treatment which have taken it closer to meeting final remediation goals. However, one major hurdle remains for the FEMP mixed waste treatment program, and that hurdle is tri-mixed waste. Tri-mixed is a term coined to describe low-level waste containing RCRA hazardous constituents along with polychlorinated biphenyls (PCB). The prescribed method for disposal of PCBs is incineration. In mixed waste treatment plans developed by the FEMP with public input, the FEMP committed to pursue non-thermal treatment methods and avoid the use of incineration. Through the SITE Program, the FEMP identified a non-thermal treatment technology which uses solvents to extract PCBs. The technology belongs to a small company called Terra-Kleen Response Group, Inc. A question arose as to how can this new and innovative technology be implemented by a small company at a Department of Energy (DOE) facility. The answer came in the form of the Rapid Commercialization Initiative (RCI) and the Mixed Waste Focus Area (MWFA). RCI is a program sponsored by the Department of commerce (DOC), DOE, Department of Defense (DOD), US EPA and various state agencies to aid companies to market new and innovative technologies

The solidification of low level radioactive organic fluids with Envirostone Gypsum Cement

International Nuclear Information System (INIS)

Rosenstiel, T.L.; Lange, R.G.

1984-01-01

The primary method for the management of low level radioactive waste (LLW) has been and continues to be the isolation of the waste in a solid mass. Of the four typical LLW streams, organic fluids pose the most significant waste isolation problem. The organic fluids comprised of lubrication oils, hydraulic fluids, sludges, scintillation fluids, etc., result from the operation and maintenance of nuclear power generating stations, research activities, tooling operations, and diagnostic analyses. The United States Gypsum Company developed the patented Envirostone Gypsum Cement system for the solidification of all types of low level radioactive wastes to facilitate handling and transportation to regulated LLW disposal sites. For the solidification of organic fluids, Envirostone Gypsum Cement is used in conjunction with Envirostone Emulsifier, selected for its ability to emulsify a broad range of organic fluids in aqueous solutions. In the solidification process it is theorized that as the crystalline matrix of the gypsum forms, the micelles of the emulsifier behave as a chemical bridge which draws the organic fluid into the crystalline structure via the hydration water. Initial testing of physical properties of solidified waste forms, including leachability, per the requirements and the procedures specified for 10 CFR Part 61 as outlined in the Branch Technical Position Report from the United States Nuclear Regulatory Commission were in progress as of the writing of this paper. Upon completion of this testing a Topical Report will be submitted to the USNRC for review and approval. The presentation reviews field experience in the use of Envirostone Gypsum Cement for the solidification of low level radioactive organic fluids from nuclear power generating stations and makes an economic comparison between Envirostone Gypsum Cement and portland cement systems

System for disposing of radioactive waste

International Nuclear Information System (INIS)

Gablin, K.A.; Hansen, L.J.

1979-01-01

A system is described for disposing of radioactive waste material from nuclear reactors by solidifying the liquid components to produce an encapsulated mass adapted for disposal by burial. The method contemplates mixing of radioactive waste materials, with or without contained solids, with a setting agent capable of solidifying the waste liquids into a free standing hardened mass, placing the resulting liquid mixture in a container with a proportionate amount of a curing agent to effect solidification under controlled conditions, and thereafter burying the container and contained solidified mixture. The setting agent is a water-extendable polymer consisting of a suspension of partially polymerized particles of urea formaldehyde in water, and the curing agent is sodium bisulfate. Methods are disclosed for dewatering slurry-like mixtures of liquid and particulate radioactive waste materials, such as spent ion exchange resin beads, and for effecting desired distribution of non-liquid radioactive materials in the central area of the container prior to solidification, so that the surrounding mass of lower specific radioactivity acts as a partial shield against higher radioactivity of the non-liquid radioactive materials. The methods also provide for addition of non-radioactive filler materials to dilute the mixture and lower the overall radioactivity of the hardened mixture to desired Lowest Specific Activity counts. An inhibiting agent is added to the liquid mixture to adjust the solidification time, and provision is made for adding additional amounts of setting agent and curing agent to take up any free water and further encapsulate the hardened material within the container. 30 claims

System for disposing of radioactive waste

International Nuclear Information System (INIS)

Gablin, K.A.; Hansen, L.J.

1977-01-01

A system is described for disposing of radioactive waste material from nuclear reactors by solidifying the liquid components to produce an encapsulated mass adapted for disposal by burial. The method contemplates mixing of radioactive waste materials, with or without contained solids, with a setting agent capable of solidifying the waste liquids into a free standing hardened mass, placing the resulting liquid mixture in a container with a proportionate amount of a curing agent to effect solidification under controlled conditions, and thereafter burying the container and contained solidified mixture. The setting agent is a water-extendable polymer consisting of a suspension of partially polymerized particles of urea formaldehyde in water, and the curing agent is sodium bisulfate. Methods are disclosed for dewatering slurry-like mixtures of liquid and particulate radioactive waste materials, such as spent ion exchange resin beads, and for effecting desired distribution of non-liquid radioactive materials in the central area of the container prior to solidification, so that the surrounding mass of lower specific radioactivity acts as a partial shield against higher radioactivity of the non-liquid radioactive materials. The methods also provide for addition of non-radioactive filler materials to dilute the mixture and lower the overall radioactivity of the hardened mixture to desired Lowest Specific Activity counts. An inhibiting agent is added to the liquid mixture to adjust the solidification time, and provision is made for adding additional amounts of setting agent and curing agent to take up any free water and further encapsulate the hardened material within the container

Managing mixed wastes: technical issues

International Nuclear Information System (INIS)

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

1986-01-01

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

Mixed waste disposal facilities at the Savannah River Site

International Nuclear Information System (INIS)

Wells, M.N.; Bailey, L.L.

1991-01-01

The Savannah River Site (SRS) is a key installation of the US Department of Energy (DOE). The site is managed by DOE's Savannah River Field Office and operated under contract by the Westinghouse Savannah River Company (WSRC). The Site's waste management policies reflect a continuing commitment to the environment. Waste minimization, recycling, use of effective pre-disposal treatments, and repository monitoring are high priorities at the site. One primary objective is to safely treat and dispose of process wastes from operations at the site. To meet this objective, several new projects are currently being developed, including the M-Area Waste Disposal Project (Y-Area) which will treat and dispose of mixed liquid wastes, and the Hazardous Waste/Mixed Waste Disposal Facility (HW/MWDF), which will store, treat, and dispose of solid mixed and hazardous wastes. This document provides a description of this facility and its mission

The Hybrid Treatment Process for mixed radioactive and hazardous waste treatment

International Nuclear Information System (INIS)

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

1992-06-01

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

The setting time of a clay-slag geopolymer matrix: the influence of blast-furnace-slag addition and the mixing method

Czech Academy of Sciences Publication Activity Database

Perná, Ivana; Hanzlíček, Tomáš

112, Part 1, JAN 20 (2016), s. 1150-1155 ISSN 0959-6526 Institutional support: RVO:67985891 Keywords : blast-furnace slag * geopolymer * setting time * mixing method * solidification * recycling Subject RIV: DM - Solid Waste and Recycling Impact factor: 5.715, year: 2016

Electrochemical treatment of mixed and hazardous waste

International Nuclear Information System (INIS)

Dziewinski, J.; Marczak, S.; Smith, W.; Nuttall, E.

1995-01-01

Los Alamos National Laboratory (LANL) and The University of New Mexico are jointly developing an electrochemical process for treating hazardous and radioactive wastes. The wastes treatable by the process include toxic metal solutions, cyanide solutions, and various organic wastes that may contain chlorinated organic compounds. The main component of the process is a stack of electrolytic cells with peripheral equipment such as a rectifier, feed system, tanks with feed and treated solutions, and a gas-venting system. During the treatment, toxic metals are deposited on the cathode, cyanides are oxidized on the anode, and organic compounds are anodically oxidized by direct or mediated electrooxidation, depending on their type. Bench scale experimental studies have confirmed the feasibility of applying electrochemical systems to processing of a great variety of hazardous and mixed wastes. The operating parameters have been defined for different waste compositions using surrogate wastes. Mixed wastes are currently treated at bench scale as part of the treatability study

Decontamination impacts on solidification and waste disposal

International Nuclear Information System (INIS)

Kempf, C.R.; Soo, P.

1988-01-01

Research to determine chemical and physical conditions which could lead to thermal excursions, gas generation, and/or general degradation of decontamination-reagent-loaded resins has shown that IRN-78, IONAC A-365, and IRN-77 organic ion exchange resin moisture contents vary significantly depending on the counter ion ''loading.'' The extent/vigor of the reaction is very highly dependent on the degree of dewatering of the resins and on the method of solution addition. The heat generation may be due, in part, to the heat of neutralization. In studies of the long-term compatibility effects of decontamination waste resins in contact with waste package container materials in the presence of decontamination reagents, radiolysis products and gamma irradiation, it has been found that the corrosion of carbon steel and austenitic stainless steel in mixed bed resins is enhanced by gamma irradiation. However, cracking in high density polyethylene is essentially eliminated because of the rapid removal of oxygen from the environment by gamma-induced oxidation of the large resin mass. 13 refs., 10 figs., 3 tabs

Mixed Waste Integrated Program -- Problem-oriented technology development

International Nuclear Information System (INIS)

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

1994-01-01

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

Chemodynamics of EDTA in a simulated mixed waste: the Hanford Site's complex concentrate waste

International Nuclear Information System (INIS)

Toste, A.P.; Ohnuki, Toshihiko

1999-01-01

Enormous stockpiles of mixed wastes at the USDOE's Hanford Site, the original US plutonium production facility, await permanent disposal. One mixed waste derived from reprocessing spent fuel was found to contain numerous nuclear related organics including chelating agents like EDTA and complexing agents, which have been used as decontamination agents, etc. Their presence in actual mixed wastes indicates that the organic content of nuclear wastes is dynamic and complicate waste management efforts. The subjects of this report is the chemo-degradation of EDTA degradation in a simulant Hanford's complex concentrate waste. The simulant was prepared by adding EDTA to an inorganic matrix, which was formulated based on past analyses of the actual waste. Aliquots of the EDTA simulant were withdrawn at different time points, derivatized via methylation and analyzed by gas chromatography and Gc/MS to monitor the disappearance of EDTA and the appearance of its' degradation products. This report also compares the results of EDTA's chemo-degradation to the g-radiolysis of EDTA in the simulant, the subject of a recently published article. Finally based on the results of these two studies, an assesment of the potential impact of EDTA degradation on the management of mixed wastes is offered. (J.P.N.)

Retrofit of existing Dow solidification system at quad cities nuclear station

International Nuclear Information System (INIS)

Dekitis, L.; Jarvis, N.; Petri, R.; Testa, J.

1983-01-01

Over the past year ATCOR has been involved in the design, testing and supply of an In-container Mixing System Retrofit for Commonwealth Edison's Quad Cities Nuclear Station Solidification System. The system supplied by DOW, itself a retrofit of a urea formaldehyde system, was based upon use of 50 cubic foot containers (liners). ATCOR's retrofit increased liner capacity to 170 cubic feet and allowed in-cask solidification of highly radioactive material. This paper discusses the reasons for the decision to replace equipment within the originally furnished system and the development of the ATCOR plan to proof-test this equipment prior to delivery at the site. Results of this pre-testing, and a comparison between pre-tested conditions and the actual in-plan start-up tests are presented. Development of improved instrumentation and mechanical modifications which enhance the reliability of the ATCOR/DOW process In-container Mixing System was provided as a part of this project. Test results are presented on instruments, controls and the unique method of mechanical attachment of the Mixing Head to the solidification container

Cement radwaste solidification studies third annual report

International Nuclear Information System (INIS)

Brown, D.J.; James, J.M.; Lee, D.J.; Smith, D.L.; Walker, A.T.

1982-03-01

This report summarises cement radwaste studies carried out at AEE Winfrith during 1981 on the encapsulation of medium and low active waste in cement. During the year more emphasis has been placed on the work which is directly related to the solidification of SGHWR active sludge. Information has been obtained on the properties of 220 dm 3 drums of cemented waste. The use of cement grouts for the encapsulation of solid items has also been investigated during 1981. (U.K.)

Method of solidifying radioactive liquid wastes

International Nuclear Information System (INIS)

Uetake, Naoto; Kawamura, Fumio; Kikuchi, Makoto; Fukazawa, Tetsuo.

1983-01-01

Purpose: To enable to confine the volatiling ingredients such as cesium in liquid wastes safely in glass solidification products while suppressing the volatilization thereof. Method: Acid salt of tetravalent metal such as titanium phosphate has an intense selective adsorption property to cesium. So liquid wastes stored in a high level liquid wastes tank is mixed with titanium phosphate gels stored in an adsorbent tank, then supplied to a mixer and mixed with a sodium silicate solution stored in a sodium silicate storage tank and boric acid stored in an additive tank, into gel-like state. The gel-like material thus formed is supplied to a drier. After being dried at a temperature of 200sup(o)C - 300sup(o)C, the material is melted under heating at a temperature of 1000sup(o)C - 1100sup(o)C, and then cooled to solidify. (Horiuchi, T.)

Mixed and Low-Level Waste Treatment Facility project

International Nuclear Information System (INIS)

1992-04-01

Mixed and low-level wastes generated at the Idaho National Engineering Laboratory (INEL) are required to be managed according to applicable State and Federal regulations, and Department of Energy Orders that provide for the protection of human health and the environment. The Mixed and Low-Level Waste Treatment Facility Project was chartered in 1991, by the Department of Energy to provide treatment capability for these mixed and low-level waste streams. The first project task consisted of conducting engineering studies to identify the waste streams, their potential treatment strategies, and the requirements that would be imposed on the waste streams and the facilities used to process them. This report, Appendix A, Environmental ampersand Regulatory Planning ampersand Documentation, identifies the regulatory requirements that would be imposed on the operation or construction of a facility designed to process the INEL's waste streams. These requirements are contained in five reports that discuss the following topics: (1) an environmental compliance plan and schedule, (2) National Environmental Policy Act requirements, (3) preliminary siting requirements, (4) regulatory justification for the project, and (5) health and safety criteria

Waste-Form Development Program. Annual progress report, October 1981-September 1982

International Nuclear Information System (INIS)

Neilson, R.M. Jr.; Colombo, P.

1982-09-01

Low-level wastes (LLW) at nuclear facilities have traditionally been solidified using portland cement (with and without additives). Urea-formaldehyde has been used for LLW solidification while bitumen (asphalt) and thermosetting polymers will be applied to domestic wastes in the near future. Operational difficulties have been observed with each of these solidification agents. Such difficulties include incompatibility with waste constitutents inhibiting solidification, premature setting, free standing water and fires. Some specific waste types have proven difficult to solidify with one or more of the contemporary agents. Similar problems are also anticipated for the solidification of new wastes, which are generated using advanced volume reduction technologies, and with the application of additional agents which may be introduced in the near future for the solidification of LLW. In the Waste Form Development program, contemporary solidification agents are being investigated relative to their potential applications to major fuel cycle and non-fuel cycle LLW streams. The range of conditions under which these solidification agents can be satisfactorily applied to specific LLW streams is being determined. These studies are primarily directed towards defining operating parameters for both improved solidification of problem wastes such as ion exchange resins, organic liquids and oils for which prevailing processes, as currently employed, appear to be inadequate, and solidification of new LLW streams including high solids content evaporator concentrates, dry solids, and incinerator ash generated from advanced volume reduction technologies. Solidified waste forms are tested and evaluated to demonstrate compliance with waste form performance and shallow land burial (SLB) acceptance criteria and transportation requirements (both as they currently exist and as they are anticipated to be modified with time)

Hanford land disposal restrictions plan for mixed wastes

International Nuclear Information System (INIS)

1990-10-01

Since the early 1940s, the Hanford Site has been involved in the production and purification of nuclear defense materials. These production activities have resulted in the generation of large quantities of liquid and solid radioactive mixed waste. This waste is subject to regulation under authority of both the Resource Conservation and Recovery Act of 1976 (RCRA) and the Atomic Energy Act. The State of Washington Department of Ecology (Ecology), the US Environmental Protection Agency (EPA), and the US Department of Energy (DOE) have entered into an agreement, the Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) to bring Hanford Site Operations into compliance with dangerous waste regulations. The Tri-Party Agreement was amended to require development of the Hanford Land Disposal Restrictions Plan for Mixed Wastes (this plan) to comply with land disposal restrictions requirements for radioactive mixed waste. The Tri-Party Agreement requires, and the this plan provides, the following sections: Waste Characterization Plan, Storage Report, Treatment Report, Treatment Plan, Waste Minimization Plan, a schedule, depicting the events necessary to achieve full compliance with land disposal restriction requirements, and a process for establishing interim milestones. 34 refs., 28 figs., 35 tabs

Hanford land disposal restrictions plan for mixed wastes

Energy Technology Data Exchange (ETDEWEB)

1990-10-01

Since the early 1940s, the Hanford Site has been involved in the production and purification of nuclear defense materials. These production activities have resulted in the generation of large quantities of liquid and solid radioactive mixed waste. This waste is subject to regulation under authority of both the Resource Conservation and Recovery Act of 1976 (RCRA) and the Atomic Energy Act. The State of Washington Department of Ecology (Ecology), the US Environmental Protection Agency (EPA), and the US Department of Energy (DOE) have entered into an agreement, the Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) to bring Hanford Site Operations into compliance with dangerous waste regulations. The Tri-Party Agreement was amended to require development of the Hanford Land Disposal Restrictions Plan for Mixed Wastes (this plan) to comply with land disposal restrictions requirements for radioactive mixed waste. The Tri-Party Agreement requires, and the this plan provides, the following sections: Waste Characterization Plan, Storage Report, Treatment Report, Treatment Plan, Waste Minimization Plan, a schedule, depicting the events necessary to achieve full compliance with land disposal restriction requirements, and a process for establishing interim milestones. 34 refs., 28 figs., 35 tabs.

Feasibility study on the solidification of liquid low-level radioactive mixed waste in the inactive tank system at Oak Ridge National Laboratory, Oak Ridge, Tennessee

International Nuclear Information System (INIS)