High level waste forms: glass marbles and thermal spray coatings

International Nuclear Information System (INIS)

Treat, R.L.; Oma, K.H.; Slate, S.C.

1982-01-01

A process that converts high-level waste to glass marbles and then coats the marbles has been developed at Pacific Northwest Laboratory (PNL) under sponsorship of the US Department of Energy. The process consists of a joule-heated glass melter, a marble-making device based on a patent issued to Corning Glass Works, and a coating system that includes a plasma spray coater and a marble tumbler. The process was developed under the Alternative Waste Forms Program which strived to improve upon monolithic glass for immobilizing high-level wastes. Coated glass marbles were found to be more leach-resistant, and the marbles, before coating were found to be very homogeneous, highly impact resistant, and conductive to encapsulation in a metal matric for improved heat transfer and containment. Marbles are also ideally suited for quality assurance and recycling. However, the marble process is more complex, and marbles require a larger number of canisters for waste containment and have a higher surface area than do glass monoliths

Comparison of the corrosion behaviors of the glass-bonded sodalite ceramic waste form and reference HLW glasses

International Nuclear Information System (INIS)

Ebert, W. L.; Lewis, M. A.

1999-01-01

A glass-bonded sodalite ceramic waste form is being developed for the long-term immobilization of salt wastes that are generated during spent nuclear fuel conditioning activities. A durable waste form is prepared by hot isostatic pressing (HIP) a mixture of salt-loaded zeolite powders and glass frit. A mechanistic description of the corrosion processes is being developed to support qualification of the CWF for disposal. The initial set of characterization tests included two standard tests that have been used extensively to study the corrosion behavior of high level waste (HLW) glasses: the Material Characterization Center-1 (MCC-1) Test and the Product Consistency Test (PCT). Direct comparison of the results of tests with the reference CWF and HLW glasses indicate that the corrosion behaviors of the CWF and HLW glasses are very similar

Characterization and durability testing of a glass-bonded ceramic waste form

International Nuclear Information System (INIS)

Johnson, S. G.

1998-01-01

Argonne National Laboratory is developing a glass bonded ceramic waste form for encapsulating the fission products and transuranics from the conditioning of metallic reactor fuel. This waste form is currently being scaled to the multi-kilogram size for encapsulation of actual high level waste. This paper will present characterization and durability testing of the ceramic waste form. An emphasis on results from application of glass durability tests such as the Product Consistency Test and characterization methods such as X-ray diffraction and scanning electron microscopy. The information presented is based on a suite of tests utilized for assessing product quality during scale-up and parametric testing

Low sintering temperature glass waste forms for sequestering radioactive iodine

Science.gov (United States)

Nenoff, Tina M.; Krumhansl, James L.; Garino, Terry J.; Ockwig, Nathan W.

2012-09-11

Materials and methods of making low-sintering-temperature glass waste forms that sequester radioactive iodine in a strong and durable structure. First, the iodine is captured by an adsorbant, which forms an iodine-loaded material, e.g., AgI, AgI-zeolite, AgI-mordenite, Ag-silica aerogel, ZnI.sub.2, CuI, or Bi.sub.5O.sub.7I. Next, particles of the iodine-loaded material are mixed with powdered frits of low-sintering-temperature glasses (comprising various oxides of Si, B, Bi, Pb, and Zn), and then sintered at a relatively low temperature, ranging from 425.degree. C. to 550.degree. C. The sintering converts the mixed powders into a solid block of a glassy waste form, having low iodine leaching rates. The vitrified glassy waste form can contain as much as 60 wt % AgI. A preferred glass, having a sintering temperature of 500.degree. C. (below the silver iodide sublimation temperature of 500.degree. C.) was identified that contains oxides of boron, bismuth, and zinc, while containing essentially no lead or silicon.

Glass composite waste forms for iodine confined in bismuth-embedded SBA-15

Energy Technology Data Exchange (ETDEWEB)

Yang, Jae Hwan [Nuclear Fuel Cycle Process Development Division, Korea Atomic Energy Research Institute, 989-111 Daeduk-daero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of); Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon, 305-701 (Korea, Republic of); Park, Hwan Seo; Ahn, Do-Hee [Nuclear Fuel Cycle Process Development Division, Korea Atomic Energy Research Institute, 989-111 Daeduk-daero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of); Yim, Man-Sung, E-mail: msyim@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon, 305-701 (Korea, Republic of)

2016-11-15

The aim of this study was to stabilize bismuth-embedded SBA-15 that captured iodine gas by fabrication of monolithic waste forms. The iodine containing waste was mixed with Bi{sub 2}O{sub 3} (a stabilizing additive) and low-temperature sintering glass followed by pelletizing and the sintering process to produce glass composite materials. Iodine volatility during the sintering process was significantly affected by the ratio of Bi{sub 2}O{sub 3} and the glass composition. It was confirmed that BiI{sub 3}, the main iodine phase within bismuth-embedded SBA-15, was effectively transformed to the mixed phases of Bi{sub 5}O{sub 7}I and BiOI. The initial leaching rates of iodine from the glass composite waste forms ranged 10{sup −3}–10{sup −2} g/m{sup 2} day, showing the stability of the iodine phases encapsulated by the glassy networks. It was also observed that common groundwater anions (e.g., chloride, carbonate, sulfite, and fluoride) elevated the iodine leaching rate by anion exchange reactions. The present results suggest that the glass composite waste form of bismuth-embedded SBA-15 could be a candidate material for stable storage of {sup 129}I. - Highlights: • Glass composite waste forms were developed to stabilize iodine confined in Bi-embedded SBA-15. • BiI{sub 3} within Bi-embedded SBA-15 was transformed to BiOI and Bi{sub 5}O{sub 7}I during sintering process. • Iodine volatility was significantly affected by glass composition and Bi{sub 2}O{sub 3} additive. • Iodine leaching rates were 10{sup −3}–10{sup −2} g/m{sup 2} day due to the stable iodine phases encapsulated by glassy networks. • Glass composite waste form of Bi-embedded SBA-15 is expected to be a candidate material for stable storage of {sup 129}I.

Glassy slag: A complementary waste form to homogeneous glass for the implementation of MAWS in treating DOE low level/mixed wastes

International Nuclear Information System (INIS)

Feng, X.; Ordaz, G.; Krumrine, P.

1994-01-01

Glassy slag waste forms are being developed to complement glass waste forms in implementing the Minimum Additive Waste Stabilization (MAWS) Program for supporting DOE's environmental restoration efforts. These glassy slags are composed of various metal oxide crystalline phases embedded in an alumino-silicate glass phase. The slags are appropriate final waste forms for waste streams that contain large amounts of scrap metals and elements with low solubilities in glass, and that have low-flux contents. Homogeneous glass waste forms are appropriate for wastes with sufficient fluxes and low metal contents. Therefore, utilization of both glass and glassy slag waste forms will make vitrification technology applicable to the treatment of a much larger range of radioactive and mixed wastes. The MAWS approach was a plied to glassy slags by blending multiple waste streams to produce the final waste form, minimizing overall waste form volume and reducing costs. The crystalline oxide phases formed in the glassy slags can be specially formulated so that they are very durable and contain hazardous and radioactive elements in their lattice structures. The Structural Bond Strength (SBS) Model was used to predict the chemical durability of the product from the slag composition so that optimized slag compositions could be obtain with a limited number of crucible melts and testing

Glass composite waste forms for iodine confined in bismuth-embedded SBA-15

Science.gov (United States)

Yang, Jae Hwan; Park, Hwan Seo; Ahn, Do-Hee; Yim, Man-Sung

2016-11-01

The aim of this study was to stabilize bismuth-embedded SBA-15 that captured iodine gas by fabrication of monolithic waste forms. The iodine containing waste was mixed with Bi2O3 (a stabilizing additive) and low-temperature sintering glass followed by pelletizing and the sintering process to produce glass composite materials. Iodine volatility during the sintering process was significantly affected by the ratio of Bi2O3 and the glass composition. It was confirmed that BiI3, the main iodine phase within bismuth-embedded SBA-15, was effectively transformed to the mixed phases of Bi5O7I and BiOI. The initial leaching rates of iodine from the glass composite waste forms ranged 10-3-10-2 g/m2 day, showing the stability of the iodine phases encapsulated by the glassy networks. It was also observed that common groundwater anions (e.g., chloride, carbonate, sulfite, and fluoride) elevated the iodine leaching rate by anion exchange reactions. The present results suggest that the glass composite waste form of bismuth-embedded SBA-15 could be a candidate material for stable storage of 129I.

Consolidated waste forms: glass marbles and ceramic pellets

International Nuclear Information System (INIS)

Treat, R.L.; Rusin, J.M.

1982-05-01

Glass marbles and ceramic pellets have been developed at Pacific Northwest Laboratory as part of the multibarrier concept for immobilizing high-level radioactive waste. These consolidated waste forms served as substrates for the application of various inert coatings and as ideal-sized particles for encapsulation in protective matrices. Marble and pellet formulations were based on existing defense wastes at Savannah River Plant and proposed commercial wastes. To produce marbles, glass is poured from a melter in a continuous stream into a marble-making device. Marbles were produced at PNL on a vibratory marble machine at rates as high as 60 kg/h. Other marble-making concepts were also investigated. The marble process, including a lead-encapsulation step, was judged as one of the more feasible processes for immobilizing high-level wastes. To produce ceramic pellets, a series of processing steps are required, which include: spray calcining - to dry liquid wastes to a powder; disc pelletizing - to convert waste powders to spherical pellets; sintering - to densify pellets and cause desired crystal formation. These processing steps are quite complex, and thereby render the ceramic pellet process as one of the least feasible processes for immobilizing high-level wastes

Studies of glass waste form performance at Japan Atomic Energy Research Institute

International Nuclear Information System (INIS)

Banba, Tsunetaka; Kamizono, Hiroshi; Nakayama, Shinichi; Tashiro, Shingo

1989-08-01

The recent studies of glass waste form performance at Japan Atomic Energy Research Institute can be classified into the following three categories; (1) Study on the volatilization of radionuclides from the waste glass, which is necessary to estimate the safety in relation to operation of a storage facility. (2) Study on the radiation (alpha-radiation) effects which have relation to the long-term stability of the waste glass. (3) Study on the leaching behavior of actinides under the repository conditions, which is necessary to predict the long-term release rate of radionuclides from the waste glass. In the present report, the recent results corresponding to the above categories are described. (author)

Determination of the Structure of Vitrified Hydroceramic/CBC Waste Form Glasses Manufactured from DOE Reprocessing Waste

International Nuclear Information System (INIS)

Scheetz, B.E.; White, W. B.; Chesleigh, M.; Portanova, A.; Olanrewaju, J.

2005-01-01

The selection of a glass-making option for the solidification of nuclear waste has dominated DOE waste form programs since the early 1980's. Both West Valley and Savannah River are routinely manufacturing glass logs from the high level waste inventory in tank sludges. However, for some wastes, direct conversion to glass is clearly not the optimum strategy for immobilization. INEEL, for example, has approximately 4400 m 3 of calcined high level waste with an activity that produces approximately 45 watts/m 3 , a rather low concentration of radioactive constituents. For these wastes, there is value in seeking alternatives to glass. An alternative approach has been developed and the efficacy of the process demonstrated that offers a significant savings in both human health and safety exposures and also a lower cost relative to the vitrification option. The alternative approach utilizes the intrinsic chemical reactivity of the highly alkaline waste with the addition of aluminosilicate admixtures in the appropriate proportions to form zeolites. The process is one in which a chemically bonded ceramic is produced. The driving force for reaction is derived from the chemical system itself at very modest temperatures and yet forms predominantly crystalline phases. Because the chemically bonded ceramic requires an aqueous medium to serve as a vehicle for the chemical reaction, the proposed zeolite-containing waste form can more adequately be described as a hydroceramic. The hydrated crystalline materials are then subject to hot isostatic pressing (HIP) which partially melts the material to form a glass ceramic. The scientific advantages of the hydroceramic/CBC approach are: (1) Low temperature processing; (2) High waste loading and thus only modest volumetric bulking from the addition of admixtures; (3) Ability to immobilize sodium; (4) Ability to handle low levels of nitrate (2-3% NO 3 - ); (5) The flexibility of a vitrifiable waste; and (6) A process that is based on an

Performance of borosilicate glass, Synroc and spent fuel as nuclear waste forms

International Nuclear Information System (INIS)

Lutze, W.; Grambow, B.; Ewing, R.C.

1990-01-01

Presently, there are three prominent waste forms under consideration for the disposal of high-level waste: Borosilicate glass and Synroc for high-level radioactive waste from fuel reprocessing and spent fuel as the waste form for non-reprocessed fuel. Using the present experimental data base, one may compare the performance of these three waste forms under repository relevant conditions. In low water flow regimes and at temperatures less than 100 degree C, the fractional release rates of all three waste forms are low, on the order of 10-7/d or less and may decrease with time. Under these conditions the three waste forms behave similarly. At elevated temperatures or in high flow regimes, the durability of borosilicate glass will be much less than that of Synroc, and thus, for certain disposal schemes (e.g., deep burial) Synroc is preferable. All predictions of the long-term behavior are based on the extrapolation of short term experimental data, we point out that appropriate and useful natural analogues are available for each of these waste forms and should be used in the performance assessment of each waste form's long-term behavior. 14 refs

Glass-bonded iodosodalite waste form for immobilization of 129I

Science.gov (United States)

Chong, Saehwa; Peterson, Jacob A.; Riley, Brian J.; Tabada, Diana; Wall, Donald; Corkhill, Claire L.; McCloy, John S.

2018-06-01

Immobilization of radioiodine is an important requirement for current and future nuclear fuel cycles. Iodosodalite [Na8(AlSiO4)6I2] was synthesized hydrothermally from metakaolin, NaI, and NaOH. Dried unwashed sodalite powders were used to synthesize glass-bonded iodosodalite waste forms (glass composite materials) by heating pressed pellets at 650, 750, or 850 °C with two types of sodium borosilicate glass binders. These heat-treated specimens were characterized with X-ray diffraction, Fourier-transform infrared spectroscopy, scanning electron microscopy, energy dispersive spectroscopy, thermal analysis, porosity and density measurements, neutron activation analysis, and inductively-coupled plasma mass spectrometry. For the best waste form produced (pellets mixed with 10 mass% of glass binder and heat-treated at 750 °C), the maximum possible elemental iodine loading was 19.8 mass%, but only ∼8-9 mass% waste loading of iodine was retained in the waste form after thermal processing. Other pellets with higher iodine retention either contained higher porosity or were incompletely sintered. ASTM C1308 and C1285 (product consistency test, PCT) experiments were performed to understand chemical durability under diffusive and static conditions. The C1308 test resulted in significantly higher normalized loss compared to the C1285 test, most likely because of the strong effect of neutral pH solution renewal and prevention of ion saturation in solution. Both experiments indicated that release rates of Na and Si were higher than for Al and I, probably due to a poorly durable Na-Si-O phase from the glass bonding matrix or from initial sodalite synthesis; however the C1308 test result indicated that congruent dissolution of iodosodalite occurred. The average release rates of iodine obtained from C1308 were 0.17 and 1.29 g m-2 d-1 for 80 or 8 m-1, respectively, and the C1285 analysis gave a value of 2 × 10-5 g m-2 d-1, which is comparable to or better than the durability of

Development of iodine waste forms using low-temperature sintering glass

International Nuclear Information System (INIS)

Krumhansl, James Lee; Nenoff, Tina Maria; Garino, Terry J.; Rademacher, David

2010-01-01

This presentation will describe our recent work on the use of low temperature-sintering glass powders mixed with either AgI or AgI-zeolite to produce a stable waste form. Radioactive iodine ( 129 I, half-life of 1.6 x 10 7 years) is generated in the nuclear fuel cycle and is of particular concern due to its extremely long half-life and its effects on human health. As part of the DOE/NE Advanced Fuel Cycle Initiative (AFCI), the separation of 129 I from spent fuel during fuel reprocessing is being studied. In the spent fuel reprocessing scheme under consideration, the iodine is released in gaseous form and collected using Ag-loaded zeolites, to form AgI. Although AgI has extremely low solubility in water, it has a relatively high vapor pressure at moderate temperatures (>550 C), thus limiting the thermal processing. Because of this, immobilization using borosilicate glass is not feasible. Therefore, a bismuth oxide-based glasses are being studied due to the low solubility of bismuth oxide in aqueous solution at pH > 7. These waste forms were processed at 500 C, where AgI volatility is low but the glass powder is able to first densify by viscous sintering and then crystallize. Since the glass is not melted, a more chemically stable glass can be used. The AgI-glass mixture was found to have high iodine leach resistance in these initial studies.

Relating structural parameters to leachability in a glass-bonded ceramic waste form

International Nuclear Information System (INIS)

Frank, S. M.; Johnson, S. G.; Moschetti, T. L.

1998-01-01

Lattice parameters for a crystalline material can be obtained by several methods, notably by analyzing x-ray powder diffraction patterns. By utilizing a computer program to fit a pattern, one can follow the evolution or subtle changes in a structure of a crystalline species in different environments. This work involves such a study for an essential component of the ceramic waste form that is under development at Argonne National Laboratory. Zeolite 4A and zeolite 5A are used to produce two different types of waste forms: a glass-bonded sodalite and a glass-bonded zeolite, respectively. Changes in structure during production of the waste forms are discussed. Specific salt-loadings in the sodalite waste form are related to relative peak intensities of certain reflections in the XRD patterns. Structural parameters for the final waste forms will also be given and related to leachability under standard conditions

Polyphase ceramic and glass-ceramic forms for immobilizing ICPP high-level nuclear waste

International Nuclear Information System (INIS)

Harker, A.B.; Flintoff, J.F.

1984-01-01

Polyphase ceramic and glass-ceramic forms have been consolidated from simulated Idaho Chemical Processing Plant wastes by hot isostatic pressing calcined waste and chemical additives by 1000 0 C or less. The ceramic forms can contain over 70 wt% waste with densities ranging from 3.5 to 3.85 g/cm 3 , depending upon the formulation. Major phases are CaF 2 , CaZrTi 207 , CaTiO 3 , monoclinic ZrO 2 , and amorphous intergranular material. The relative fraction of the phases is a function of the chemical additives (TiO 2 , CaO, and SiO 2 ) and consolidation temperature. Zirconolite, the major actinide host, makes the ceramic forms extremely leach resistant for the actinide simulant U 238 . The amorphous phase controls the leach performance for Sr and Cs which is improved by the addition of SiO 2 . Glass-ceramic forms were also consolidated by HIP at waste loadings of 30 to 70 wt% with densities of 2.73 to 3.1 g/cm 3 using Exxon 127 borosilicate glass frit. The glass-ceramic forms contain crystalline CaF 2 , Al 203 , and ZrSi 04 (zircon) in a glass matrix. Natural mineral zircon is a stable host for 4+ valent actinides. 17 references, 3 figures, 5 tables

A Glass-Ceramic Waste Forms for the Immobilization of Rare Earth Oxides from the Pyroprocessing Waste salt

International Nuclear Information System (INIS)

Ahn, Byung-Gil; Park, Hwan-Seo; Kim, Hwan-Young; Kim, In-Tae

2008-01-01

The fission product of rare earth (RE) oxide wastes are generates during the pyroprocess . Borosilicate glass or some ceramic materials such as monazite, apatite or sodium zirconium phosphate (NZP) have been a prospective host matrix through lots of experimental results. Silicate glasses have long been the preferred waste form for the immobilization of HLW. In immobilization of the RE oxides, the developed process on an industrial scale involves their incorporation into a glass matrix, by melting under 1200 ∼ 1300 .deg. C. Instead of the melting process, glass powder sintering is lower temperature (∼ 900 .deg. C) required for the process which implies less demanding conditions for the equipment and a less evaporation of volatile radionuclides. This study reports the behaviors, direct vitrification of RE oxides with glass frit, glass powder sintering of REceramic with glass frit, formation of RE-apatite (or REmonazite) ceramic according to reaction temperature, and the leach resistance of the solidified waste forms

Glass forms for immobilization of Hanford wastes

International Nuclear Information System (INIS)

Schulz, W.W.; Dressen, A.L.; Hobbick, C.W.; Babad, H.

1975-03-01

Approximately 140 million liters of solid salt cake (mainly NaNO 3 ), produced by evaporation of aged alkaline high-level liquid wastes, will be stored in underground tanks when the present Hanford Waste Management Program is completed in the early 1980's. At this time also, large volumes of various other solid radioactive wastes (sludges, excavated Pu-contaminated soil, and doubly encapsulated 137 CsCl and 90 SrF 2 ) will be stored on the Hanford Reservation. All these solid wastes can be converted to immobile silicate and aluminosilicate glasses of low water leachability by melting them at 1100 0 to 1400 0 C with appropriate amounts of basalt (or sand) and other glass-formers such as B 2 O 3 or CaO. Reviewed in this paper are formulations and other melt conditions used successfully in batch tests to make glasses from actual and synthetic wastes; leachability and other properties of these glasses show them to be satisfactory vehicles for immobilization of the Hanford wastes. (U.S.)

Corrosion behaviors of a glass-bonded sodalite ceramic waste form and its constituents

International Nuclear Information System (INIS)

Lewis, M. A.; Ebert, W. L.; Morss, L.

1999-01-01

A ceramic waste form (CWF) of glass bonded sodalite is being developed as a waste form for the long-term immobilization of fission products and transuranic elements from the U.S. Department of Energy's activities on spent nuclear fuel conditioning. A durable waste form was prepared by hot isostatic pressing (HIP) a mixture of salt-loaded zeolite powders and glass frit. During HIP the zeolite is converted to sodalite, and the resultant CWF is been completed for durations of up to 182 days. Four dissolution modes were identified: dissolution of free salt, dissolution of the aluminosilicate matrix of sodalite and the accompanying dissolution of occluded salt, dissolution of the boroaluminosilicate matrix of the glass, and ion exchange. Synergies inherent to the CWF were identified by comparing the results of the tests with pure glass and sodalite with those of the composite CWF

Characteristics of borosilicate waste glass form for high-level radioactive waste

Energy Technology Data Exchange (ETDEWEB)

Kim, Seung Soo; Chun, Kwan Sik; Choi, Jong Won; Kang, Chul Hyung

2001-03-01

Basic data, required for the design and the performance assessment of a repository of HLW, suchas the chemical composition and the characteristics of the borosilicate waste glass have been identified according to the burn-ups of spent PWR fuels. The diemnsion of waste canister is 430mm in diameter and 1135mm in length, and the canister should hold less than 2kwatts of heat from their decay of radionuclides contained in the HLW. Based on the reprocessing of 5 years-cooled spent fuel, one canister could hold about 11.5wt.% and 10.8wt.% of oxidized HLW corresponding to their burn-ups of 45,000MWD/MTU and 55,000MWD/MTU, respectively. These waste forms have been recommanded as the reference waste forms of HLW. The characteristics of these wastes as a function of decay time been evaluated. However, after a specific waste form and a specific site for the disposal would be selected, the characteristics of the waste should be reevaluated under the consideration of solidification period, loaded waste, storage condition and duration, site circumstances for the repository system and its performance assessment.

AN ALTERNATIVE HOST MATRIX BASED ON IRON PHOSPHATE GLASSES FOR THE VITRIFICATION OF SPECIALIZED WASTE FORMS

International Nuclear Information System (INIS)

Day, Delbert D.

2000-01-01

As mentioned above, the overall goal of this research project was to collect the scientific information essential to develop iron phosphate glass based nuclear wasteforms. The specific objectives of the project were: (1) Investigate the structure of binary iron phosphate glasses and it's dependence on the composition and melting atmosphere: Understand atomic arrangements and nature of the bonding. Establish structure-property relationships. Determine the compositions and melting conditions which optimize the critical properties of the base glass. (2) Understand the structure of iron phosphate wasteforms and it's dependence on the composition and melting atmosphere: Investigate how the waste elements are bonded and coordinated within the glass structure. Establish structure-property relationships for the waste glasses. Determine the compositions and melting atmosphere for which the critical properties of the waste forms would be optimum. (3) Determine the role(s) played by the valence states of iron ions and it's dependence on the composition and melting atmosphere: Understand the different roles of iron(II) and iron(III) ions in determining the critical properties of the base glass and the waste forms. Investigate how the iron valence and its significance depend on the composition and melting atmosphere. (4) Investigate glass forming and crystallization processes of the iron phosphate glasses and their waste forms: Understand the dependence of the glass forming and crystallization characteristics on overall glass composition and valence states of iron ions. Identify the products of devitrification and investigate the critical properties of these crystalline compounds which may adversely affect the chemical and physical properties of the waste forms

Durability, mechanical, and thermal properties of experimental glass-ceramic forms for immobilizing ICPP high level waste

International Nuclear Information System (INIS)

Vinjamuri, K.

1990-01-01

The high-level liquid waste generated at the Idaho Chemical Processing Plant (ICPP) is routinely solidified into granular calcined high-level waste (HLW) and stored onsite. Research is being conducted at the ICPP on methods of immobilizing the HLW, including developing a durable glass-ceramic form which has the potential to significantly reduce the final waste volume by up to 60% compared to a glass form. Simulated, pilot plant, non-radioactive, calcines similar to the composition of the calcined HLW and glass forming additives are used to produce experimental glass-ceramic forms. The objective of the research reported in this paper is to study the impact of ground calcine particle size on durability and mechanical and thermal properties of experimental glass-ceramic forms

Glass containing radioactive nuclear waste

International Nuclear Information System (INIS)

Boatner, L.A.; Sales, B.C.

1985-01-01

Lead-iron phosphate glasses containing a high level of Fe 2 O 3 for use as a storage medium for high-level-radioactive nuclear waste. By combining lead-iron phosphate glass with various types of simulated high-level nuclear waste, a highly corrosion resistant, homogeneous, easily processed glass can be formed. For corroding solutions at 90 C, with solution pH values in the range between 5 and 9, the corrosion rate of the lead-iron phosphate nuclear waste glass is at least 10 2 to 10 3 times lower than the corrosion rate of a comparable borosilicate nuclear waste glass. The presence of Fe 2 O 3 in forming the lead-iron phosphate glass is critical. Lead-iron phosphate nuclear waste glass can be prepared at temperatures as low as 800 C, since they exhibit very low melt viscosities in the 800 to 1050 C temperature range. These waste-loaded glasses do not readily devitrify at temperatures as high as 550 C and are not adversely affected by large doses of gamma radiation in H 2 O at 135 C. The lead-iron phosphate waste glasses can be prepared with minimal modification of the technology developed for processing borosilicate glass nuclear waste forms. (author)

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

International Nuclear Information System (INIS)

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

1996-10-01

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

Crystallization behavior of nuclear waste forms

International Nuclear Information System (INIS)

Rusin, J.M.; Lokken, R.O.; May, R.P.; Wald, J.W.

1981-09-01

Several waste form options have been or are being developed for the immobilization of high-level wastes. The final selection of a waste form must take into consideration both waste form product as well as process factors. Crystallization behavior has an important role in nuclear waste form technology. For glass or vitreous waste forms, crystallization is generally controlled to a minimum by appropriate glass formulation and heat treatment schedules. With glass ceramic waste forms, crystallization is essential to convert glass products to highly crystalline waste forms with a minimum residual glass content. In the case of ceramic waste forms, additives and controlled sintering schedules are used to contain the radionuclides in specific tailored crystalline phases

Development and testing of matrices for the encapsulation of glass and ceramic nuclear waste forms

International Nuclear Information System (INIS)

Wald, J.W.; Brite, D.W.; Gurwell, W.E.; Buckwalter, C.Q.; Bunnell, L.R.; Gray, W.J.; Blair, H.T.; Rusin, J.M.

1982-02-01

This report details the results of research on the matrix encapsulation of high level wastes at PML over the past few years. The demonstrations and tests described were designed to illustrate how the waste materials are effected when encapsulated in an inert matrix. Candidate materials evaluated for potential use as matrices for encapslation of pelletized ceramics or glass marbles were categorized into four groups: metals, glasses, ceramics, and graphite. Two processing techniques, casting and hot pressing, were investigated as the most promising methods of formation or densification of the matrices. The major results reported deal with the development aspects. However, chemical durability tests (leach tests) of the matrix materials themselves and matrix-waste form composites are also reported. Matrix waste forms can provide a low porosity, waste-free barrier resulting in increased leach protection, higher impact strength and improved thermal conductivity compared to unencapsulated glass or ceramic waste materials. Glass marbles encapsulated in a lead matrix offer the most significant improvement in waste form stability of all combinations evaluated. This form represents a readily demonstrable process that provides high thermal conductivity, mechanical shock resistance, radiation shielding and increased chemical durability through both a chemical passivation mechanism and as a physical barrier. Other durable matrix waste forms evaluated, applicable primarily to ceramic pellets, involved hot-pressed titanium or TiO 2 materials. In the processing of these forms, near 100% dense matrices were obtained. The matrix materials had excellent compatibility with the waste materials and superior potential chemical durability. Cracking of the hot-pressed ceramic matrix forms, in general, prevented the realization of their optimum properties

Secondary phases formed during nuclear waste glass-water interactions: Thermodynamic and derived properties

International Nuclear Information System (INIS)

McKenzie, W.F.

1992-08-01

The thermodynamic properties of secondary phases observed to form during nuclear waste glass-water interactions are of particular interest as it is with the application of these properties together with the thermodynamic properties of other solid phases, fluid phases, and aqueous species that one may predict the environmental consequences of introducing radionuclides contained in the glass into groundwater at a high-level nuclear waste repository. The validation of these predicted consequences can be obtained from laboratory experiments and field observations at natural analogue sites. The purpose of this report is to update and expand the previous compilation (McKenzie, 1991) of thermodynamic data retrieved from the literature and/or estimated for secondary phases observed to form (and candidate phases from observed chemical compositions) during nuclear waste glass-water interactions. In addition, this report includes provisionally recommended thermodynamic data of secondary phases

Glasses and nuclear waste vitrification

International Nuclear Information System (INIS)

Ojovan, Michael I.

2012-01-01

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

Solution exchange corrosion testing with the glass-zeolite ceramic waste form in demineralized water at 900C

International Nuclear Information System (INIS)

Simpson, L. J.

1998-01-01

A ceramic waste form of glass-bonded zeolite is being developed for the long-term disposition of fission products and transuranic elements in wastes from the U.S. Department of Energy's spent nuclear fuel conditioning activities. Solution exchange corrosion tests were performed on the ceramic waste form and its potential base constituents of glass, zeolite 5A, and sodalite as part of an effort to qualify the ceramic waste form for acceptance into the Civilian Radioactive Waste Management System. Solution exchange tests were performed at 90 C by replacing 80 to 90% of the leachate with fresh demineralized water after set time intervals. The results from these tests provide information about corrosion mechanisms and the ability of the ceramic waste form and its constituent materials to retain waste components. The results from solution exchange tests indicate that radionuclides will be preferentially retained in the zeolites without the glass matrix and in the ceramic waste form, with respect to cations like Li, K, and Na. Release results have been compared for simulated waste from candidate ceramic waste forms with zeolite 5A and its constituent materials to determine the corrosion behavior of each component

Comparative assessment of TRU waste forms and processes. Volume I. Waste form and process evaluations

International Nuclear Information System (INIS)

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

1982-09-01

This study provides an assesses seven waste forms and eight processes for immobilizing transuranic (TRU) wastes. The waste forms considered are cast cement, cold-pressed cement, FUETAP (formed under elevated temperature and pressure) cement, borosilicate glass, aluminosilicate glass, basalt glass-ceramic, and cold-pressed and sintered silicate ceramic. The waste-immobilization processes considered are in-can glass melting, joule-heated glass melting, glass marble forming, cement casting, cement cold-pressing, FUETAP cement processing, ceramic cold-pressing and sintering, basalt glass-ceramic processing. Properties considered included gas generation, chemical durability, mechanical strength, thermal stability, and radiation stability. The ceramic products demonstrated the best properties, except for plutonium release during leaching. The glass and ceramic products had similar properties. The cement products generally had poorer properties than the other forms, except for plutonium release during leaching. Calculations of the Pu release indicated that the waste forms met the proposed NRC release rate limit of 1 part in 10 5 per year in most test conditions. The cast-cement process had the lowest processing cost, followed closely by the cold-pressed and FUETAP cement processes. Joule-heated glass melting had the lower cost of the glass processes. In-can melting in a high-quality canister had the highest cost, and cold-pressed and sintered ceramic the second highest. Labor and canister costs for in-can melting were identified. The major contributor to costs of disposing of TRU wastes in a defense waste repository is waste processing costs. Repository costs could become the dominant cost for disposing of TRU wastes in a commercial repository. It is recommended that cast and FUETAP cement and borosilicate glass waste-form systems be considered. 13 figures, 16 tables

Comparative waste forms study

International Nuclear Information System (INIS)

Wald, J.W.; Lokken, R.O.; Shade, J.W.; Rusin, J.M.

1980-12-01

A number of alternative process and waste form options exist for the immobilization of nuclear wastes. Although data exists on the characterization of these alternative waste forms, a straightforward comparison of product properties is difficult, due to the lack of standardized testing procedures. The characterization study described in this report involved the application of the same volatility, mechanical strength and leach tests to ten alternative waste forms, to assess product durability. Bulk property, phase analysis and microstructural examination of the simulated products, whose waste loading varied from 5% to 100% was also conducted. The specific waste forms investigated were as follows: Cold Pressed and Sintered PW-9 Calcine; Hot Pressed PW-9 Calcine; Hot Isostatic Pressed PW-9 Calcine; Cold Pressed and Sintered SPC-5B Supercalcine; Hot Isostatic pressed SPC-5B Supercalcine; Sintered PW-9 and 50% Glass Frit; Glass 76-68; Celsian Glass Ceramic; Type II Portland Cement and 10% PW-9 Calcine; and Type II Portland Cement and 10% SPC-5B Supercalcine. Bulk property data were used to calculate and compare the relative quantities of waste form volume produced at a spent fuel processing rate of 5 metric ton uranium/day. This quantity ranged from 3173 L/day (5280 Kg/day) for 10% SPC-5B supercalcine in cement to 83 L/day (294 Kg/day) for 100% calcine. Mechanical strength, volatility, and leach resistance tests provide data related to waste form durability. Glass, glass-ceramic and supercalcine ranked high in waste form durability where as the 100% PW-9 calcine ranked low. All other materials ranked between these two groupings

Fracture toughness measurements on a glass bonded sodalite high-level waste form

International Nuclear Information System (INIS)

DiSanto, T.; Goff, K. M.; Johnson, S. G.; O'Holleran, T. P.

1999-01-01

The electrometallurgical treatment of metallic spent nuclear fuel produces two high-level waste streams; cladding hulls and chloride salt. Argonne National Laboratory is developing a glass bonded sodalite waste form to immobilize the salt waste stream. The waste form consists of 75 Vol.% crystalline sodalite (containing the salt) with 25 Vol.% of an ''intergranular'' glassy phase. Microindentation fracture toughness measurements were performed on representative samples of this material using a Vickers indenter. Palmqvist cracking was confirmed by post-indentation polishing of a test sample. Young's modulus was measured by an acoustic technique. Fracture toughness, microhardness, and Young's modulus values are reported, along with results from scanning electron microscopy studies

Evaluation of final waste forms and recommendations for baseline alternatives to group and glass

Energy Technology Data Exchange (ETDEWEB)

Bleier, A.

1997-09-01

An assessment of final waste forms was made as part of the Federal Facilities Compliance Agreement/Development, Demonstration, Testing, and Evaluation (FFCA/DDT&E) Program because supplemental waste-form technologies are needed for the hazardous, radioactive, and mixed wastes of concern to the Department of Energy and the problematic wastes on the Oak Ridge Reservation. The principal objective was to identify a primary waste-form candidate as an alternative to grout (cement) and glass. The effort principally comprised a literature search, the goal of which was to establish a knowledge base regarding four areas: (1) the waste-form technologies based on grout and glass, (2) candidate alternatives, (3) the wastes that need to be immobilized, and (4) the technical and regulatory constraints on the waste-from technologies. This report serves, in part, to meet this goal. Six families of materials emerged as relevant; inorganic, organic, vitrified, devitrified, ceramic, and metallic matrices. Multiple members of each family were assessed, emphasizing the materials-oriented factors and accounting for the fact that the two most prevalent types of wastes for the FFCA/DDT&E Program are aqueous liquids and inorganic sludges and solids. Presently, no individual matrix is sufficiently developed to permit its immediate implementation as a baseline alternative. Three thermoplastic materials, sulfur-polymer cement (inorganic), bitumen (organic), and polyethylene (organic), are the most technologically developed candidates. Each warrants further study, emphasizing the engineering and economic factors, but each also has limitations that regulate it to a status of short-term alternative. The crystallinity and flexible processing of sulfur provide sulfur-polymer cement with the highest potential for short-term success via encapsulation. Long-term immobilization demands chemical stabilization, which the thermoplastic matrices do not offer. Among the properties of the remaining

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

International Nuclear Information System (INIS)

Crum, Jarrod; Maio, Vince; McCloy, John; Scott, Clark; Riley, Brian; Benefiel, Brad; Vienna, John; Archibald, Kip; Rodriguez, Carmen; Rutledge, Veronica; Zhu, Zihua; Ryan, Joe; Olszta, Matthew

2014-01-01

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

Infrared and Raman investigation of rare-earth phosphate glasses for potential use as radioactive waste forms

International Nuclear Information System (INIS)

Morgan, S.H.

1989-01-01

This project was designed to investigate the properties of the rare-earth phosphate glass systems CeO 2 -P 2 O 5 and Pr 2 O 3 -P 2 O 5 for potential use as radioactive waste glasses. The glass-forming region and optimum processing parameters of these glass systems were investigated. The structure of the host glasses and glassed loaded with simulated waste elements was investigated using Raman and infrared spectroscopy. Because of the radical differences in the spectra of the molybdenum-loaded glasses, the structure of the MoO 3 -P 2 O 5 glass system was also investigated. 29 refs., 8 figs., 2 tabs

Production of a High-Level Waste Glass from Hanford Waste Samples

International Nuclear Information System (INIS)

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

1998-09-01

The HLW glass was produced from a HLW sludge slurry (Envelope D Waste), eluate waste streams containing high levels of Cs-137 and Tc-99, solids containing both Sr-90 and transuranics (TRU), and glass-forming chemicals. The eluates and Sr-90/TRU solids were obtained from ion-exchange and precipitation pretreatments, respectively, of other Hanford supernate samples (Envelopes A, B and C Waste). The glass was vitrified by mixing the different waste streams with glass-forming chemicals in platinum/gold crucibles and heating the mixture to 1150 degree C. Resulting glass analyses indicated that the HLW glass waste form composition was close to the target composition. The targeted waste loading of Envelope D sludge solids in the HLW glass was 30.7 wt percent, exclusive of Na and Si oxides. Condensate samples from the off-gas condenser and off-gas dry-ice trap indicated that very little of the radionuclides were volatilized during vitrification. Microstructure analysis of the HLW glass using Scanning Electron Microscopy (SEM) and Energy Dispersive X-Ray Analysis (EDAX) showed what appeared to be iron spinel in the HLW glass. Further X-Ray Diffraction (XRD) analysis confirmed the presence of nickel spinel trevorite (NiFe2O4). These crystals did not degrade the leaching characteristics of the glass. The HLW glass waste form passed leach tests that included a standard 90 degree C Product Consistency Test (PCT) and a modified version of the United States Environmental Protection Agency Toxicity Characteristic Leaching Procedure (TCLP)

Effect of different glass and zeolite A compositions on the leach resistance of ceramic waste forms

International Nuclear Information System (INIS)

Lewis, M.A.; Hash, M.; Glandorf, D.

1996-01-01

A ceramic waste form is being developed for waste generated during electrometallurgical treatment of spent nuclear fuel. The waste is generated when fission products are removed from the electrolyte, LiCl-KCl eutectic. The waste form is a composite fabricated by hot isostatic pressing a mixture of glass frit and zeolite occluded with fission products and salt. Normalized release rate is less than 1 g/m 2 d for all elements in MCC-1 leach test run for 28 days in deionized water at 90 C. This leach resistance is comparable to that of early Savannah River glasses. We are investigating how leach resistance is affected by changes in cationic form of zeolite and in glass composition. Composites were made with 3 forms of zeolite A and 6 glasses. We used 3-day ASTM C1220-92 (formerly MCC-1) leach tests to screen samples for development purposes only. The leach test results show that the glass composites of zeolites 5A and 4A retain fission products equally well. Loss of Cs is small (0.1-0.5 wt%), while the loss of divalent and trivalent fission products is one or more orders of magnitude smaller. Composites of 5A retain chloride ion better in these short-term screens than 4A and 3A. The more leach resistant composites were made with durable glasses rich in silica and poor in alkaline earth oxides. XRD show that a salt phase was absent in the leach resistant composites of 5A and the better glasses but was present in the other composites with poorer leach performance. Thus, absence of salt phase corresponds to improved leach resistance. Interactions between zeolite and glass depend on composition of both

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

International Nuclear Information System (INIS)

Ellison, A.J.G.; Mazer, J.J.; Ebert, W.L.

1994-11-01

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

Production of sodalite waste forms by addition of glass

International Nuclear Information System (INIS)

Pereira, C.

1995-01-01

Spent nuclear fuel can be treated in a molten salt electrorefiner for conversion into metal and mineral waste forms for geologic disposal. Sodalite is one of the mineral waste forms under study. Fission products in the molten salt are ion-exchanged into zeolite A, which is converted to sodalite and consolidated. Sodalite can be formed directly from mixtures of salt and zeolite A at temperatures above 975 K; however, nepheline is usually produced as a secondary phase. Addition of small amounts of glass frit to the mixture reduced nepheline formation significantly. Loss of fission products was not observed for reaction below 1000 K. Hot-pressing of the sodalite powders yielded dense pellets (∼2.3 g/cm 3 ) without any loss of fission product species. Normalized release rates were below 1 g/m 2 ·day for pre-washed samples in 28-day leach tests based on standard MCC-1 tests but increased with the presence of free salt on the sodalite

Apatite and sodalite based glass-bonded waste forms for immobilization of 129I and mixed halide radioactive wastes

Energy Technology Data Exchange (ETDEWEB)

Goel, Ashutosh [Rutgers Univ., New Brunswick, NJ (United States); McCloy, John S. [Washington State Univ., Pullman, WA (United States); Riley, Brian J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Matyas, Josef [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2017-12-30

The goal of the project was to utilize the knowledge accumulated by the team, in working with minerals for chloride wastes and biological apatites, toward the development of advanced waste forms for immobilizing 129I and mixed-halide wastes. Based on our knowledge, experience, and thorough literature review, we had selected two minerals with different crystal structures and potential for high chemical durability, sodalite and CaP/PbV-apatite, to form the basis of this project. The focus of the proposed effort was towards: (i) low temperature synthesis of proposed minerals (iodine containing sodalite and apatite) leading to the development of monolithic waste forms, (ii) development of a fundamental understanding of the atomic-scale to meso-scale mechanisms of radionuclide incorporation in them, and (iii) understanding of the mechanism of their chemical corrosion, alteration mechanism, and rates. The proposed work was divided into four broad sections. deliverables. 1. Synthesis of materials 2. Materials structural and thermal characterization 3. Design of glass compositions and synthesis glass-bonded minerals, and 4. Chemical durability testing of materials.

Synroc tailored waste forms for actinide immobilization

Energy Technology Data Exchange (ETDEWEB)

Gregg, Daniel J.; Vance, Eric R. [Australian Nuclear Science and Technology Organisation, Kirrawee (Australia). ANSTOsynroc, Inst. of Materials Engineering

2017-07-01

Since the end of the 1970s, Synroc at the Australian Nuclear Science and Technology Organisation (ANSTO) has evolved from a focus on titanate ceramics directed at PUREX waste to a platform waste treatment technology to fabricate tailored glass-ceramic and ceramic waste forms for different types of actinide, high- and intermediate level wastes. The particular emphasis for Synroc is on wastes which are problematic for glass matrices or existing vitrification process technologies. In particular, nuclear wastes containing actinides, notably plutonium, pose a unique set of requirements for a waste form, which Synroc ceramic and glass-ceramic waste forms can be tailored to meet. Key aspects to waste form design include maximising the waste loading, producing a chemically durable product, maintaining flexibility to accommodate waste variations, a proliferation resistance to prevent theft and diversion, and appropriate process technology to produce waste forms that meet requirements for actinide waste streams. Synroc waste forms incorporate the actinides within mineral phases, producing products which are much more durable in water than baseline borosilicate glasses. Further, Synroc waste forms can incorporate neutron absorbers and {sup 238}U which provide criticality control both during processing and whilst within the repository. Synroc waste forms offer proliferation resistance advantages over baseline borosilicate glasses as it is much more difficult to retrieve the actinide and they can reduce the radiation dose to workers compared to borosilicate glasses. Major research and development into Synroc at ANSTO over the past 40 years has included the development of waste forms for excess weapons plutonium immobilization in collaboration with the US and for impure plutonium residues in collaboration with the UK, as examples. With a waste loading of 40-50 wt.%, Synroc would also be considered a strong candidate as an engineered waste form for used nuclear fuel and highly

Evaluation of final waste forms and recommendations for baseline alternatives to grout and glass

International Nuclear Information System (INIS)

Bleier, A.

1997-09-01

An assessment of final waste forms was made as part of the Federal Facilities Compliance Agreement/Development, Demonstration, Testing, and Evaluation (FFCA/DDT ampersand E) Program because supplemental waste-form technologies are needed for the hazardous, radioactive, and mixed wastes of concern to the Department of Energy and the problematic wastes on the Oak Ridge Reservation. The principal objective was to identify a primary waste-form candidate as an alternative to grout (cement) and glass. The effort principally comprised a literature search, the goal of which was to establish a knowledge base regarding four areas: (1) the waste-form technologies based on grout and glass, (2) candidate alternatives, (3) the wastes that need to be immobilized, and (4) the technical and regulatory constraints on the waste-from technologies. This report serves, in part, to meet this goal. Six families of materials emerged as relevant; inorganic, organic, vitrified, devitrified, ceramic, and metallic matrices. Multiple members of each family were assessed, emphasizing the materials-oriented factors and accounting for the fact that the two most prevalent types of wastes for the FFCA/DDT ampersand E Program are aqueous liquids and inorganic sludges and solids. Presently, no individual matrix is sufficiently developed to permit its immediate implementation as a baseline alternative. Three thermoplastic materials, sulfur-polymer cement (inorganic), bitumen (organic), and polyethylene (organic), are the most technologically developed candidates. Each warrants further study, emphasizing the engineering and economic factors, but each also has limitations that regulate it to a status of short-term alternative. The crystallinity and flexible processing of sulfur provide sulfur-polymer cement with the highest potential for short-term success via encapsulation. Long-term immobilization demands chemical stabilization, which the thermoplastic matrices do not offer. Among the properties of the

Processability analysis of candidate waste forms

International Nuclear Information System (INIS)

Gould, T.H. Jr.; Dunson, J.B. Jr.; Eisenberg, A.M.; Haight, H.G. Jr.; Mello, V.E.; Schuyler, R.L. III.

1982-01-01

A quantitative merit evaluation, or processability analysis, was performed to assess the relative difficulty of remote processing of Savannah River Plant high-level wastes for seven alternative waste form candidates. The reference borosilicate glass process was rated as the simplest, followed by FUETAP concrete, glass marbles in a lead matrix, high-silica glass, crystalline ceramics (SYNROC-D and tailored ceramics), and coated ceramic particles. Cost estimates for the borosilicate glass, high-silica glass, and ceramic waste form processing facilities are also reported

Designing Advanced Ceramic Waste Forms for Electrochemical Processing Salt Waste

International Nuclear Information System (INIS)

Ebert, W. L.; Snyder, C. T.; Frank, Steven; Riley, Brian

2016-01-01

This report describes the scientific basis underlying the approach being followed to design and develop ''advanced'' glass-bonded sodalite ceramic waste form (ACWF) materials that can (1) accommodate higher salt waste loadings than the waste form developed in the 1990s for EBR-II waste salt and (2) provide greater flexibility for immobilizing extreme waste salt compositions. This is accomplished by using a binder glass having a much higher Na_2O content than glass compositions used previously to provide enough Na+ to react with all of the Cl- in the waste salt and generate the maximum amount of sodalite. The phase compositions and degradation behaviors of prototype ACWF products that were made using five new binder glass formulations and with 11-14 mass% representative LiCl/KCl-based salt waste were evaluated and compared with results of similar tests run with CWF products made using the original binder glass with 8 mass% of the same salt to demonstrate the approach and select a composition for further studies. About twice the amount of sodalite was generated in all ACWF materials and the microstructures and degradation behaviors confirmed our understanding of the reactions occurring during waste form production and the efficacy of the approach. However, the porosities of the resulting ACWF materials were higher than is desired. These results indicate the capacity of these ACWF waste forms to accommodate LiCl/KCl-based salt wastes becomes limited by porosity due to the low glass-to-sodalite volume ratio. Three of the new binder glass compositions were acceptable and there is no benefit to further increasing the Na content as initially planned. Instead, further studies are needed to develop and evaluate alternative production methods to decrease the porosity, such as by increasing the amount of binder glass in the formulation or by processing waste forms in a hot isostatic press. Increasing the amount of binder glass to eliminate porosity will decrease the waste

Designing Advanced Ceramic Waste Forms for Electrochemical Processing Salt Waste

Energy Technology Data Exchange (ETDEWEB)

Ebert, W. L. [Argonne National Lab. (ANL), Argonne, IL (United States); Snyder, C. T. [Argonne National Lab. (ANL), Argonne, IL (United States); Frank, Steven [Argonne National Lab. (ANL), Argonne, IL (United States); Riley, Brian [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-03-01

This report describes the scientific basis underlying the approach being followed to design and develop “advanced” glass-bonded sodalite ceramic waste form (ACWF) materials that can (1) accommodate higher salt waste loadings than the waste form developed in the 1990s for EBR-II waste salt and (2) provide greater flexibility for immobilizing extreme waste salt compositions. This is accomplished by using a binder glass having a much higher Na₂O content than glass compositions used previously to provide enough Na+ to react with all of the Cl– in the waste salt and generate the maximum amount of sodalite. The phase compositions and degradation behaviors of prototype ACWF products that were made using five new binder glass formulations and with 11-14 mass% representative LiCl/KCl-based salt waste were evaluated and compared with results of similar tests run with CWF products made using the original binder glass with 8 mass% of the same salt to demonstrate the approach and select a composition for further studies. About twice the amount of sodalite was generated in all ACWF materials and the microstructures and degradation behaviors confirmed our understanding of the reactions occurring during waste form production and the efficacy of the approach. However, the porosities of the resulting ACWF materials were higher than is desired. These results indicate the capacity of these ACWF waste forms to accommodate LiCl/KCl-based salt wastes becomes limited by porosity due to the low glass-to-sodalite volume ratio. Three of the new binder glass compositions were acceptable and there is no benefit to further increasing the Na content as initially planned. Instead, further studies are needed to develop and evaluate alternative production methods to decrease the porosity, such as by increasing the amount of binder glass in the formulation or by processing waste forms in a hot isostatic press. Increasing the amount of binder glass to eliminate porosity will decrease

Iron phosphate glass containing simulated fast reactor waste: Characterization and comparison with pristine iron phosphate glass

International Nuclear Information System (INIS)

Joseph, Kitheri; Asuvathraman, R.; Venkata Krishnan, R.; Ravindran, T.R.; Govindaraj, R.; Govindan Kutty, K.V.; Vasudeva Rao, P.R.

2014-01-01

Detailed characterization was carried out on an iron phosphate glass waste form containing 20 wt.% of a simulated nuclear waste. High temperature viscosity measurement was carried out by the rotating spindle method. The Fe 3+ /Fe ratio and structure of this waste loaded iron phosphate glass was investigated using Mössbauer and Raman spectroscopy respectively. Specific heat measurement was carried out in the temperature range of 300–700 K using differential scanning calorimeter. Isoconversional kinetic analysis was employed to understand the crystallization behavior of the waste loaded iron phosphate glass. The glass forming ability and glass stability of the waste loaded glass were also evaluated. All the measured properties of the waste loaded glass were compared with the characteristics of pristine iron phosphate glass

Review of high-level waste form properties

International Nuclear Information System (INIS)

Rusin, J.M.

1980-12-01

This report is a review of waste form options for the immobilization of high-level-liquid wastes from the nuclear fuel cycle. This review covers the status of international research and development on waste forms as of May 1979. Although the emphasis in this report is on waste form properties, process parameters are discussed where they may affect final waste form properties. A summary table is provided listing properties of various nuclear waste form options. It is concluded that proposed waste forms have properties falling within a relatively narrow range. In regard to crystalline versus glass waste forms, the conclusion is that either glass of crystalline materials can be shown to have some advantage when a single property is considered; however, at this date no single waste form offers optimum properties over the entire range of characteristics investigated. A long-term effort has been applied to the development of glass and calcine waste forms. Several additional waste forms have enough promise to warrant continued research and development to bring their state of development up to that of glass and calcine. Synthetic minerals, the multibarrier approach with coated particles in a metal matrix, and high pressure-high temperature ceramics offer potential advantages and need further study. Although this report discusses waste form properties, the total waste management system should be considered in the final selection of a waste form option. Canister design, canister materials, overpacks, engineered barriers, and repository characteristics, as well as the waste form, affect the overall performance of a waste management system. These parameters were not considered in this comparison

Glass Property Data and Models for Estimating High-Level Waste Glass Volume

Energy Technology Data Exchange (ETDEWEB)

Vienna, John D.; Fluegel, Alexander; Kim, Dong-Sang; Hrma, Pavel R.

2009-10-05

This report describes recent efforts to develop glass property models that can be used to help estimate the volume of high-level waste (HLW) glass that will result from vitrification of Hanford tank waste. The compositions of acceptable and processable HLW glasses need to be optimized to minimize the waste-form volume and, hence, to save cost. A database of properties and associated compositions for simulated waste glasses was collected for developing property-composition models. This database, although not comprehensive, represents a large fraction of data on waste-glass compositions and properties that were available at the time of this report. Glass property-composition models were fit to subsets of the database for several key glass properties. These models apply to a significantly broader composition space than those previously publised. These models should be considered for interim use in calculating properties of Hanford waste glasses.

Glass Property Data and Models for Estimating High-Level Waste Glass Volume

International Nuclear Information System (INIS)

Vienna, John D.; Fluegel, Alexander; Kim, Dong-Sang; Hrma, Pavel R.

2009-01-01

This report describes recent efforts to develop glass property models that can be used to help estimate the volume of high-level waste (HLW) glass that will result from vitrification of Hanford tank waste. The compositions of acceptable and processable HLW glasses need to be optimized to minimize the waste-form volume and, hence, to save cost. A database of properties and associated compositions for simulated waste glasses was collected for developing property-composition models. This database, although not comprehensive, represents a large fraction of data on waste-glass compositions and properties that were available at the time of this report. Glass property-composition models were fit to subsets of the database for several key glass properties. These models apply to a significantly broader composition space than those previously publised. These models should be considered for interim use in calculating properties of Hanford waste glasses.

Glassy slags as novel waste forms for remediating mixed wastes with high metal contents

International Nuclear Information System (INIS)

Feng, X.; Wronkiewicz, D.J.; Bates, J.K.; Brown, N.R.; Buck, E.C.; Gong, M.; Ebert, W.L.

1994-01-01

Argonne National Laboratory (ANL) is developing a glassy slag final waste form for the remediation of low-level radioactive and mixed wastes with high metal contents. This waste form is composed of various crystalline and metal oxide phases embedded in a silicate glass phase. This work indicates that glassy slag shows promise as final waste form because (1) it has similar or better chemical durability than high-level nuclear waste (HLW) glasses, (2) it can incorporate large amounts of metal wastes, (3) it can incorporate waste streams having low contents of flux components (boron and alkalis), (4) it has less stringent processing requirements (e.g., viscosity and electric conductivity) than glass waste forms, (5) its production can require little or no purchased additives, which can result in greater reduction in waste volume and overall treatment costs. By using glassy slag waste forms, minimum additive waste stabilization approach can be applied to a much wider range of waste streams than those amenable only to glass waste forms

Candidate Low-Temperature Glass Waste Forms for Technetium-99 Recovered from Hanford Effluent Management Facility Evaporator Concentrate

Energy Technology Data Exchange (ETDEWEB)

Ding, Mei [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tang, Ming [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rim, Jung Ho [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Chamberlin, Rebecca M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-07-24

Alternative treatment and disposition options may exist for technetium-99 (99Tc) in secondary liquid waste from the Hanford Direct-Feed Low-Activity Waste (DFLAW) process. One approach includes development of an alternate glass waste form that is suitable for on-site disposition of technetium, including salts and other species recovered by ion exchange or precipitation from the EMF evaporator concentrate. By recovering the Tc content from the stream, and not recycling the treated concentrate, the DFLAW process can potentially be operated in a more efficient manner that lowers the cost to the Department of Energy. This report provides a survey of candidate glass formulations and glass-making processes that can potentially incorporate technetium at temperatures <700 °C to avoid volatilization. Three candidate technetium feed streams are considered: (1) dilute sodium pertechnetate loaded on a non-elutable ion exchange resin; (2) dilute sodium-bearing aqueous eluent from ion exchange recovery of pertechnetate, or (3) technetium(IV) oxide precipitate containing Sn and Cr solids in an aqueous slurry. From the technical literature, promising candidate glasses are identified based on their processing temperatures and chemical durability data. The suitability and technical risk of three low-temperature glass processing routes (vitrification, encapsulation by sintering into a glass composite material, and sol-gel chemical condensation) for the three waste streams was assessed, based on available low-temperature glass data. For a subset of candidate glasses, their long-term thermodynamic behavior with exposure to water and oxygen was modeled using Geochemist’s Workbench, with and without addition of reducing stannous ion. For further evaluation and development, encapsulation of precipitated TcO2/Sn/Cr in a glass composite material based on lead-free sealing glasses is recommended as a high priority. Vitrification of pertechnetate in aqueous anion exchange eluent solution

Immobilization of radioactive waste in glass matrices

International Nuclear Information System (INIS)

Wicks, G.G.

1978-01-01

A promising process for long-term management of high-level radioactive waste is to immobilize the waste in a borosilicate glass matrix. Among the most important criteria characterizing the integrity of the large-scale glass-waste forms are that they possess good chemical stability (including low leachability), thermal stability, mechanical integrity, and high radiation stability. Fulfillment of these criteria ensures the maximum margin of safety of glass-waste products, following solidification, handling, transportation, and long-term storage

Database for waste glass composition and properties

International Nuclear Information System (INIS)

Peters, R.D.; Chapman, C.C.; Mendel, J.E.; Williams, C.G.

1993-09-01

A database of waste glass composition and properties, called PNL Waste Glass Database, has been developed. The source of data is published literature and files from projects funded by the US Department of Energy. The glass data have been organized into categories and corresponding data files have been prepared. These categories are glass chemical composition, thermal properties, leaching data, waste composition, glass radionuclide composition and crystallinity data. The data files are compatible with commercial database software. Glass compositions are linked to properties across the various files using a unique glass code. Programs have been written in database software language to permit searches and retrievals of data. The database provides easy access to the vast quantities of glass compositions and properties that have been studied. It will be a tool for researchers and others investigating vitrification and glass waste forms

Waste acceptance product specifications for vitrified high-level waste forms

International Nuclear Information System (INIS)

Applewhite-Ramsey, A.; Sproull, J.F.

1993-01-01

The Nuclear Waste Policy Act of 1982 mandated that all high-level waste (HLW) be sent to a federal geologic repository for permanent disposal. DOE published the Environmental Assessment in 1982 which identified borosilicate glass as the chosen HLW form. 1 In 1985 the Department of Energy instituted a Waste Acceptance Process to assure that DWPF glass waste forms would be acceptable to such a repository. This assurance was important since production of waste forms will precede repository construction and licensing. As part of this Waste Acceptance Process, the DOE Office of Civilian Radioactive Waste Management (RW) formed the Waste Acceptance Committee (WAC). The WAC included representatives from the candidate repository sites, the waste producing sites and DOE. The WAC was responsible for developing the Waste Acceptance Preliminary Specifications (WAPS) which defined the requirements the waste forms must meet to be compatible with the candidate repository geologies

Cerium, uranium, and plutonium behavior in glass-bonded sodalite, a ceramic nuclear waste form

International Nuclear Information System (INIS)

Lewis, M. A.; Lexa, D.; Morss, L. R.; Richmann, M. K.

1999-01-01

Glass-bonded sodalite is being developed as a ceramic waste form (CWF) to immobilize radioactive fission products, actinides, and salt residues from electrometallurgical treatment of spent nuclear reactor fuel. The CWF consists of about 75 mass % sodalite, 25 mass % glass, and small amounts of other phases. This paper presents some results and interpretation of physical measurements to characterize the CWF structure, and dissolution tests to measure the release of matrix components and radionuclides from the waste form. Tests have been carried out with specimens of the CWF that contain rare earths at concentrations similar to those expected in the waste form. Parallel tests have been carried out on specimens that have uranium or plutonium as well as the rare earths at concentrations similar to those expected in the waste forms; in these specimens UCl 3 forms UO 2 and PuCl 3 forms PuO 2 . The normalized releases of rare earths in dissolution tests were found to be much lower than those of matrix elements (B, Si, Al, Na). When there is no uranium in the CWF, the release of cerium is two to ten times lower than the release of the other rare earths. The low release of cerium may be due to its tetravalent state in uranium-free CWF. However, when there is uranium in the CWF, the release of cerium is similar to that of the other rare earths. This trivalent behavior of cerium is attributed to charge transfer or covalent interactions among cerium, uranium, and oxygen in (U,Ce)O 2

Estimation of centerline temperature of the waste form for the rare earth waste generated from pyrochemical process

Energy Technology Data Exchange (ETDEWEB)

Choi, Jung-Hoon, E-mail: mrchoijh@kaeri.re.kr; Eun, Hee-Chul; Lee, Tae-Kyo; Lee, Ki-Rak; Han, Seung-Youb; Jeon, Min-Ku; Park, Hwan-Seo; Ahn, Do-Hee

2017-01-15

Estimation of centerline temperature of nuclear glass waste form for each waste stream is very essential in the period of storage because the centerline temperature being over its glass transition temperature results in the increase of leaching rate of radioactive nuclides due to the devitrification of glass waste form. Here, to verify the effects of waste form diameter and transuranic element content in the rare earth waste on the centerline temperature of the waste form, the surrogate rare earth glass waste generated from pyrochemical process was immobilized with SiO{sub 2}−Al{sub 2}O{sub 3}−B{sub 2}O{sub 3} glass frit system, and thermal properties of the rare earth glass waste form were determined by thermomechanical analysis and thermal conductivity analysis. The estimation of centerline temperature was carried out using the experimental thermal data and steady-state conduction equation in a long and solid cylinder type waste form. It was revealed that thermal stability of waste form in case of 0.3 m diameter was not affected by the TRU content even in the case of 80% TRU recovery ratio in the electrowinning process, meaning that the waste form of 0.3 m diameter is thermally stable due to the low centerline temperature relative to its glass transition temperature of the rare earth glass waste form.

Estimates of radionuclide release from glass waste forms in a tuff repository and the effects on regulatory compliance

International Nuclear Information System (INIS)

Aines, R.D.

1986-04-01

This paper discusses preliminary estimates of the release of radionuclides from waste packages containing glass-based waste forms under the expected conditions at Yucca Mountain. These estimates can be used to evaluate the contribution of waste package performance toward meeting repository regulatory restrictions on radionuclide release. Glass waste will be held in double stainless steel canisters. After failure of the container sometime after the 300 to 1000 year containment period, the open headspace in these cans will provide the only area where standing water can accumulate and react with the glass. A maximum release rate of 0.177 g/m 2 x year or 1.3 grams per year was obtained. Normalized loss of 1.3 grams per year corresponds to 0.08 parts in 100,000 per year of the 1660 kg reference weight of DWPF glass

Conceptual waste package interim product specifications and data requirements for disposal of borosilicate glass defense high-level waste forms in salt geologic repositories

International Nuclear Information System (INIS)

1983-06-01

The conceptual waste package interim product specifications and data requirements presented are applicable specifically to the normal borosilicate glass product of the Defense Waste Processing Facility (DWPF). They provide preliminary numerical values for the defense high-level waste form parameters and properties identified in the waste form performance specification for geologic isolation in salt repositories. Subject areas treated include containment and isolation, operational period safety, criticality control, waste form/production canister identification, and waste package performance testing requirements. This document was generated for use in the development of conceptual waste package designs in salt. It will be revised as additional data, analyses, and regulatory requirements become available

Thermal Predictions of the Cooling of Waste Glass Canisters

Energy Technology Data Exchange (ETDEWEB)

Donna Post Guillen

2014-11-01

Radioactive liquid waste from five decades of weapons production is slated for vitrification at the Hanford site. The waste will be mixed with glass forming additives and heated to a high temperature, then poured into canisters within a pour cave where the glass will cool and solidify into a stable waste form for disposal. Computer simulations were performed to predict the heat rejected from the canisters and the temperatures within the glass during cooling. Four different waste glass compositions with different thermophysical properties were evaluated. Canister centerline temperatures and the total amount of heat transfer from the canisters to the surrounding air are reported.

The production of advanced glass ceramic HLW forms using cold crucible induction melter

International Nuclear Information System (INIS)

Rutledge, V.J.; Maio, V.

2013-01-01

Cold Crucible Induction Melters (CCIM) will favorably change how High-Level radioactive Waste (from nuclear fuel recovery) is treated in a near future. Unlike the existing Joule-Heated Melters (JHM) currently in operation for the glass-based immobilization of High-Level Waste (HLW), CCIM offers unique material features that will increase melt temperatures, increase throughput, increase mixing, increase loading in the waste form, lower melter foot prints, eliminate melter corrosion and lower costs. These features not only enhance the technology for producing HLW forms, but also provide advantageous attributes to the waste form by allowing more durable alternatives to glass. It is concluded that glass ceramic waste forms that are tailored to immobilize fission products of HLW can be can be made from the HLW processed with the CCIM. The advantageous higher temperatures reached with the CCIM and unachievable with JHM allows the lanthanides, alkali, alkaline earths, and molybdenum to dissolve into a molten glass. Upon controlled cooling they go into targeted crystalline phases to form a glass ceramic waste form with higher waste loadings than achievable with borosilicate glass waste forms. Natural cooling proves to be too fast for the formation of all targeted crystalline phases

Conceptual waste package interim product specifications and data requirements for disposal of glass commercial high-level waste forms in salt geologic repositories

International Nuclear Information System (INIS)

1983-10-01

The conceptual waste package interim product specifications and data requirements presented are applicable to the reference glass composition described in PNL-3838 and carbon steel canister described in ONWI-438. They provide preliminary numerical values for the commercial high-level waste form parameters and properties identified in the waste form performance specification for geologic isolation in salt repositories. Subject areas treated include containment and isolation, operational period safety, criticality control, waste form/production canister identification, and waste package performance testing requirements. This document was generated for use in the development of conceptual waste package designs in salt. It will be revised as additional data, analyses and regulatory requirements become available. 13 references, 1 figure

High-level radioactive waste glass and storage canister design

International Nuclear Information System (INIS)

Slate, S.C.; Ross, W.A.

1979-01-01

Management of high-level radioactive wastes is a primary concern in nuclear operations today. The main objective in managing these wastes is to convert them into a solid, durable form which is then isolated from man. A description is given of the design and evaluation of this waste form. The waste form has two main components: the solidified waste and the storage canister. The solid waste form discussed in this study is glass. Waste glasses have been designed to be inert to water attack, physically rugged, low in volatility, and stable over time. Two glass-making processes are under development at PNL. The storage canister is being designed to provide high-integrity containment for solidified wastes from processing to terminal storage. An outline is given of the steps in canister design: material selection, stress and thermal analyses, quality verification, and postfill processing. Examples are given of results obtained from actual nonradioactive demonstration tests. 14 refs

Decontamination processes for waste glass canisters

International Nuclear Information System (INIS)

Rankin, W.N.

1982-01-01

A Defense Waste Processing Facility (DWPF) is currently being designed to convert Savannah River Plant liquid, high-level radioactive waste into a solid form, such as borosilicate glass. To prevent the spread of radioactivity, the outside of the canisters of waste glass must have very low levels of smearable radioactive contamination before they are removed from the DWPF. Several techniques were considered for canister decontamination: high-pressure water spray, electropolishing, chemical dissolution, and abrasive blasting. An abrasive blasting technique using a glass frit slurry has been selected for use in the DWPF. No additional equipment is needed to process waste generated from decontamination. Frit used as the abrasive will be mixed with the waste and fed to the glass melter. In contrast, chemical and electrochemical techniques require more space in the DWPF, and produce large amounts of contaminated by-products, which are difficult to immobilize by vitrification

Alternative-waste-form evaluation for Savannah River Plant high-level waste

International Nuclear Information System (INIS)

Gould, T.H. Jr.; Crandall, J.L.

1982-01-01

Results of the waste form evaluation are summarized as: risks of human exposure are comparable and extremely small for either borosilicate glass or Synroc ceramic. Waste form properties are more than adequate for either form. The waste form decision can therefore be made on the basis of practicality and cost effectiveness. Synroc offers lower costs for transportation and emplacement. The borosilicate glass form offers the lowest total disposal cost, much simpler and less costly production, an established and proven process, lower future development costs, and an earlier startup of the DWPF

Preparation and leaching of radioactive INEL waste forms

International Nuclear Information System (INIS)

Schuman, R.P.; Welch, J.M.; Staples, B.A.

1982-01-01

The purpose of this study is to prepare and leach test ceramic and glass waste form specimens produced from actual transuranic waste sludges and high-level waste calcines, respectively. Description of wastes, specimen fabrication, leaching procedure, analysis of leachates and results are discussed. The conclusion is that radioactive waste stored at INEL can be readily incorporated in fused ceramic and glass forms. Initial leach testing results indicate that these forms show great promise for safe long-term containment of radioactive wastes

Review of high-level waste form properties. [146 bibliographies

Energy Technology Data Exchange (ETDEWEB)

Rusin, J.M.

1980-12-01

This report is a review of waste form options for the immobilization of high-level-liquid wastes from the nuclear fuel cycle. This review covers the status of international research and development on waste forms as of May 1979. Although the emphasis in this report is on waste form properties, process parameters are discussed where they may affect final waste form properties. A summary table is provided listing properties of various nuclear waste form options. It is concluded that proposed waste forms have properties falling within a relatively narrow range. In regard to crystalline versus glass waste forms, the conclusion is that either glass of crystalline materials can be shown to have some advantage when a single property is considered; however, at this date no single waste form offers optimum properties over the entire range of characteristics investigated. A long-term effort has been applied to the development of glass and calcine waste forms. Several additional waste forms have enough promise to warrant continued research and development to bring their state of development up to that of glass and calcine. Synthetic minerals, the multibarrier approach with coated particles in a metal matrix, and high pressure-high temperature ceramics offer potential advantages and need further study. Although this report discusses waste form properties, the total waste management system should be considered in the final selection of a waste form option. Canister design, canister materials, overpacks, engineered barriers, and repository characteristics, as well as the waste form, affect the overall performance of a waste management system. These parameters were not considered in this comparison.

Methods of vitrifying waste with low melting high lithia glass compositions

Science.gov (United States)

Jantzen, Carol M.; Pickett, John B.; Cicero-Herman, Connie A.; Marra, James C.

2001-01-01

The invention relates to methods of vitrifying waste and for lowering the melting point of glass forming systems by including lithia formers in the glass forming composition in significant amounts, typically from about 0.16 wt % to about 11 wt %, based on the total glass forming oxides. The lithia is typically included as a replacement for alkali oxide glass formers that would normally be present in a particular glass forming system. Replacement can occur on a mole percent or weight percent basis, and typically results in a composition wherein lithia forms about 10 wt % to about 100 wt % of the alkali oxide glass formers present in the composition. The present invention also relates to the high lithia glass compositions formed by these methods. The invention is useful for stabilization of numerous types of waste materials, including aqueous waste streams, sludge solids, mixtures of aqueous supernate and sludge solids, combinations of spent filter aids from waste water treatment and waste sludges, supernate alone, incinerator ash, incinerator offgas blowdown, or combinations thereof, geological mine tailings and sludges, asbestos, inorganic filter media, cement waste forms in need of remediation, spent or partially spent ion exchange resins or zeolites, contaminated soils, lead paint, etc. The decrease in melting point achieved by the present invention desirably prevents volatilization of hazardous or radioactive species during vitrification.

Secondary Waste Cast Stone Waste Form Qualification Testing Plan

Energy Technology Data Exchange (ETDEWEB)

Westsik, Joseph H.; Serne, R. Jeffrey

2012-09-26

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

Retention of Halogens in Waste Glass

Energy Technology Data Exchange (ETDEWEB)

Hrma, Pavel R.

2010-05-01

In spite of their potential roles as melting rate accelerators and foam breakers, halogens are generally viewed as troublesome components for glass processing. Of five halogens, F, Cl, Br, I, and At, all but At may occur in nuclear waste. A nuclear waste feed may contain up to 10 g of F, 4 g of Cl, and ≤100 mg of Br and I per kg of glass. The main concern is halogen volatility, producing hazardous fumes and particulates, and the radioactive iodine 129 isotope of 1.7x10^7-year half life. Because F and Cl are soluble in oxide glasses and tend to precipitate on cooling, they can be retained in the waste glass in the form of dissolved constituents or as dispersed crystalline inclusions. This report compiles known halogen-retention data in both high-level waste (HLW) and low-activity waste (LAW) glasses. Because of its radioactivity, the main focus is on I. Available data on F and Cl were compiled for comparison. Though Br is present in nuclear wastes, it is usually ignored; no data on Br retention were found.

The effects of crystallization and residual glass on the chemical durability of iron phosphate waste forms containing 40 wt% of a high MoO3 Collins-CLT waste

Science.gov (United States)

Hsu, Jen-Hsien; Bai, Jincheng; Kim, Cheol-Woon; Brow, Richard K.; Szabo, Joe; Zervos, Adam

2018-03-01

The effects of cooling rate on the chemical durability of iron phosphate waste forms containing up to 40 wt% of a high MoO3 Collins-CLT waste simulant were determined at 90 °C using the product consistency test (PCT). The waste form, designated 40wt%-5, meets appropriate Department of Energy (DOE) standards when rapidly quenched from the melt (as-cast) and after slow cooling following the CCC (canister centerline cooling)-protocol, although the quenched glass is more durable. The analysis of samples from the vapor hydration test (VHT) and the aqueous corrosion test (differential recession test) reveals that rare earth orthophosphate (monazite) and Zr-pyrophosphate crystals that form on cooling are more durable than the residual glass in the 40wt%-5 waste form. The residual glass in the CCC-treated samples has a greater average phosphate chain length and a lower Fe/P ratio, and those contribute to its faster corrosion kinetics.

Waste glass weathering

International Nuclear Information System (INIS)

Bates, J.K.; Buck, E.C.

1994-01-01

The weathering of glass is reviewed by examining processes that affect the reaction of commercial, historical, natural, and nuclear waste glass under conditions of contact with humid air and slowly dripping water, which may lead to immersion in nearly static solution. Radionuclide release data from weathered glass under conditions that may exist in an unsaturated environment are presented and compared to release under standard leaching conditions. While the comparison between the release under weathering and leaching conditions is not exact, due to variability of reaction in humid air, evidence is presented of radionuclide release under a variety of conditions. These results suggest that both the amount and form of radionuclide release can be affected by the weathering of glass

Effect of aluminum and silicon reactants and process parameters on glass-ceramic waste form characteristics for immobilization of high-level fluorinel-sodium calcined waste

International Nuclear Information System (INIS)

Vinjamuri, K.

1993-06-01

In this report, the effects of aluminum and silicon reactants, process soak time and the initial calcine particle size on glass-ceramic waste form characteristics for immobilization of the high-level fluorinel-sodium calcined waste stored at the Idaho Chemical Processing Plant (ICPP) are investigated. The waste form characteristics include density, total and normalized elemental leach rates, and microstructure. Glass-ceramic waste forms were prepared by hot isostatically pressing (HIPing) a pre-compacted mixture of pilot plant fluorinel-sodium calcine, Al, and Si metal powders at 1050 degrees C, 20,000 psi for 4 hours. One of the formulations with 2 wt % Al was HIPed for 4, 8, 16 and 24 hours at the same temperature and pressure. The calcine particle size range include as calcined particle size smaller than 600 μm (finer than -30 mesh, or 215 μm Mass Median Diameter, MMD) and 180 μm (finer than 80 mesh, or 49 μm MMD)

Physical and chemical characterization of borosilicate glasses containing Hanford high-level wastes

International Nuclear Information System (INIS)

Kupfer, M.J.; Palmer, R.A.

1980-10-01

Scouting studies are being performed to develop and evaluate silicate glass forms for immobilization of Hanford high-level wastes. Detailed knowledge of the physical and chemical properties of these glasses is required to assess their suitability for long-term storage or disposal. Some key properties to be considered in selecting a glass waste form include leach resistance, resistance to radiation, microstructure (includes devitrification behavior or crystallinity), homogeneity, viscosity, electrical resistivity, mechanical ruggedness, thermal expansion, thermal conductivity, density, softening point, annealing point, strain point, glass transformation temperature, and refractive index. Other properties that are important during processing of the glass include volatilization of glass and waste components, and corrosivity of the glass on melter components. Experimental procedures used to characterize silicate waste glass forms and typical properties of selected glass compositions containing simulated Hanford sludge and residual liquid wastes are presented. A discussion of the significance and use of each measured property is also presented

Combined Waste Form Cost Trade Study

International Nuclear Information System (INIS)

Gombert, Dirk; Piet, Steve; Trickel, Timothy; Carter, Joe; Vienna, John; Ebert, Bill; Matthern, Gretchen

2008-01-01

A new generation of aqueous nuclear fuel reprocessing, now in development under the auspices of the DOE Office of Nuclear Energy (NE), separates fuel into several fractions, thereby partitioning the wastes into groups of common chemistry. This technology advance enables development of waste management strategies that were not conceivable with simple PUREX reprocessing. Conventional wisdom suggests minimizing high level waste (HLW) volume is desirable, but logical extrapolation of this concept suggests that at some point the cost of reducing volume further will reach a point of diminishing return and may cease to be cost-effective. This report summarizes an evaluation considering three groupings of wastes in terms of cost-benefit for the reprocessing system. Internationally, the typical waste form for HLW from the PUREX process is borosilicate glass containing waste elements as oxides. Unfortunately several fission products (primarily Mo and the noble metals Ru, Rh, Pd) have limited solubility in glass, yielding relatively low waste loading, producing more glass, and greater disposal costs. Advanced separations allow matching the waste form to waste stream chemistry, allowing the disposal system to achieve more optimum waste loading with improved performance. Metals can be segregated from oxides and each can be stabilized in forms to minimize the HLW volume for repository disposal. Thus, a more efficient waste management system making the most effective use of advanced waste forms and disposal design for each waste is enabled by advanced separations and how the waste streams are combined. This trade-study was designed to juxtapose a combined waste form baseline waste treatment scheme with two options and to evaluate the cost-benefit using available data from the conceptual design studies supported by DOE-NE

Preliminary assessment of nine waste-form products/processes for immobilizing transuranic wastes

International Nuclear Information System (INIS)

Crisler, L.R.

1980-09-01

Nine waste-form processes for reduction of the present and projected Transuranic (TRU) waste inventory to an immobilized product have been evaluated. Product formulations, selected properties, preparation methods, technology status, problem areas needing resolution and location of current research development being pursued in the United States are discussed for each process. No definitive utility ranking is attempted due to the early stage of product/process development for TRU waste containing products and the uncertainties in the state of current knowledge of TRU waste feed compositional and quantitative makeup. Of the nine waste form products/processes included in this discussion, bitumen and cements (encapsulation agents) demonstrate the degree of flexibility necessary to immobilize the wide composition range present in the TRU waste inventory. A demonstrated process called Slagging Pyrolysis Incineration converts a varied compositional feed (municipal wastes) to a ''basalt'' like product. This process/product appears to have potential for TRU waste immobilization. The remaining waste forms (borosilicate glass, high-silica glass, glass ceramics, ''SYNROC B'' and cermets) have potential for immobilizing a smaller fraction of the TRU waste inventory than the above discussed waste forms

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

Energy Technology Data Exchange (ETDEWEB)

Veronica J Rutledge; Vince Maio

2013-10-01

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

Investigation of lead-iron-phosphate glass for SRP waste

International Nuclear Information System (INIS)

Jantzen, C.M.

1986-10-01

The search for a host solid for the immobilization of nuclear waste has focused on various vitreous waste forms. Recently, lead-iron-phosphate (LIP) glasses have been proposed for solidification of all types of HLLW. Investigation of this glass for vitrification of SRP waste demonstrated that the phosphate glass is incompatible with the current borosilicate glass technology. The durability of LIP glasses in deionized water was comparable to current borosilicate waste glass formulations, and the LIP glass has a low melt temperature. However, many of the defense waste constituents have low solubility in the phosphate melt, producing an inhomogeneous product. Also, the LIP melt is highly corrosive which prevents the use of current melter materials, in particular Inconel 690, and thus requires more exotic materials of construction such as platinum

Baseline Glass Development for Combined Fission Products Waste Streams

International Nuclear Information System (INIS)

Crum, Jarrod V.; Billings, Amanda Y.; Lang, Jesse B.; Marra, James C.; Rodriguez, Carmen P.; Ryan, Joseph V.; Vienna, John D.

2009-01-01

Borosilicate glass was selected as the baseline technology for immobilization of the Cs/Sr/Ba/Rb (Cs), lanthanide (Ln) and transition metal fission product (TM) waste steams as part of a cost benefit analysis study.(1) Vitrification of the combined waste streams have several advantages, minimization of the number of waste forms, a proven technology, and similarity to waste forms currently accepted for repository disposal. A joint study was undertaken by Pacific Northwest National Laboratory (PNNL) and Savannah River National Laboratory (SRNL) to develop acceptable glasses for the combined Cs + Ln + TM waste streams (Option 1) and Cs + Ln combined waste streams (Option 2) generated by the AFCI UREX+ set of processes. This study is aimed to develop baseline glasses for both combined waste stream options and identify key waste components and their impact on waste loading. The elemental compositions of the four-corners study were used along with the available separations data to determine the effect of burnup, decay, and separations variability on estimated waste stream compositions.(2-5) Two different components/scenarios were identified that could limit waste loading of the combined Cs + LN + TM waste streams, where as the combined Cs + LN waste stream has no single component that is perceived to limit waste loading. Combined Cs + LN waste stream in a glass waste form will most likely be limited by heat due to the high activity of Cs and Sr isotopes.

Sampling and analysis strategies to support waste form qualification

International Nuclear Information System (INIS)

Westsik, J.H. Jr.; Pulsipher, B.A.; Eggett, D.L.; Kuhn, W.L.

1989-04-01

As part of the waste acceptance process, waste form producers will be required to (1) demonstrate that their glass waste form will meet minimum specifications, (2) show that the process can be controlled to consistently produce an acceptable waste form, and (3) provide documentation that the waste form produced meets specifications. Key to the success of these endeavors is adequate sampling and chemical and radiochemical analyses of the waste streams from the waste tanks through the process to the final glass product. This paper suggests sampling and analysis strategies for meeting specific statistical objectives of (1) detection of compositions outside specification limits, (2) prediction of final glass product composition, and (3) estimation of composition in process vessels for both reporting and guiding succeeding process steps. 2 refs., 1 fig., 3 tabs

Effects of beta/gamma radiation on nuclear waste glasses

Energy Technology Data Exchange (ETDEWEB)

Weber, W.J. [Pacific Northwest National Lab., Richland, WA (United States)

1997-07-01

A key challenge in the disposal of high-level nuclear waste (HLW) in glass waste forms is the development of models of long-term performance based on sound scientific understanding of relevant phenomena. Beta decay of fission products is one source of radiation that can impact the performance of HLW glasses through the interactions of the emitted {beta}-particles and g-rays with the atoms in the glass by ionization processes. Fused silica, alkali silicate glasses, alkali borosilicate glasses, and nuclear waste glasses are all susceptible to radiation effects from ionization. In simple glasses, defects (e.g., non-bridging oxygen and interstitial molecular oxygen) are observed experimentally. In more complex glasses, including nuclear waste glasses, similar defects are expected, and changes in microstructure, such as the formation of bubbles, have been reported. The current state of knowledge regarding the effects of {beta}/{gamma} radiation on the properties and microstructure of nuclear waste glasses are reviewed. (author)

Effects of beta/gamma radiation on nuclear waste glasses

International Nuclear Information System (INIS)

Weber, W.J.

1997-01-01

A key challenge in the disposal of high-level nuclear waste (HLW) in glass waste forms is the development of models of long-term performance based on sound scientific understanding of relevant phenomena. Beta decay of fission products is one source of radiation that can impact the performance of HLW glasses through the interactions of the emitted β-particles and g-rays with the atoms in the glass by ionization processes. Fused silica, alkali silicate glasses, alkali borosilicate glasses, and nuclear waste glasses are all susceptible to radiation effects from ionization. In simple glasses, defects (e.g., non-bridging oxygen and interstitial molecular oxygen) are observed experimentally. In more complex glasses, including nuclear waste glasses, similar defects are expected, and changes in microstructure, such as the formation of bubbles, have been reported. The current state of knowledge regarding the effects of β/γ radiation on the properties and microstructure of nuclear waste glasses are reviewed. (author)

Borosilicate glass as a matrix for the immobilization of Savannah River Plant waste

International Nuclear Information System (INIS)

Plodinec, M.J.; Wicks, G.G.; Bibler, N.E.

1982-01-01

The reference waste form for immobilization of Savannah River Plant (SRP) waste is borosilicate glass. In the reference process, waste is mixed with glass-forming chemicals and melted in a Joule-heated ceramic melter at 1150 0 C. Waste glass made with actual or simulated waste on a small scale and glass made with simulated waste on a large scale confirm that the current reference process and glass-former composition are able to accommodate all SRP waste compositions and can produce a glass with: high waste loading; low leach rates; good thermal stability; high resistance to radiation effects; and good impact resistance. Borosilicate glass has been studied as a matrix for the immobilization of SRP waste since 1974. This paper reviews the results of extensive characterization and performance testing of the glass product. These results show that borosilicate glass is a very suitable matrix for the immobilization of SRP waste. 18 references, 3 figures, 10 tables

Radiation transport in high-level waste form

International Nuclear Information System (INIS)

Arakali, V.S.; Barnes, S.M.

1992-01-01

The waste form selected for vitrifying high-level nuclear waste stored in underground tanks at West Valley, NY is borosilicate glass. The maximum radiation level at the surface of a canister filled with the high-level waste form is prescribed by repository design criteria for handling and disposition of the vitrified waste. This paper presents an evaluation of the radiation transport characteristics for the vitreous waste form expected to be produced at West Valley and the resulting neutron and gamma dose rates. The maximum gamma and neutron dose rates are estimated to be less than 7500 R/h and 10 mRem/h respectively at the surface of a West Valley canister filled with borosilicate waste glass

Development, evaluation, and selection of candidate high-level waste forms

International Nuclear Information System (INIS)

Bernadzikowski, T.A.; Allender, J.S.; Gordon, D.E.; Gould, T.H. Jr.

1982-01-01

The seven candidate waste forms, evaluated as potential media for the immobilization and gelogic disposal of high-level nuclear wastes were borosilicate glass, SYNROC, tailored ceramic, high-silica glass, FUETAP concrete, coated sol-gel particles, and glass marbles in a lead matrix. The evaluation, completed on August 1, 1981, combined preliminary waste form evaluations conducted at Department of Energy (DOE) defense waste-sites and at independent laboratories, peer review assessments, a product performance evaluation, and a processability analysis. Based on the combined results of these four inputs, two of the seven forms, borosilicate glass and a titanate-based ceramic, SYNROC, were selected as the reference and alternative forms, respectively, for continued development and evaluation in the National HLW Program. The borosilicate glass and ceramic forms were further compared during FY-1982 on the basis of risk assessments, cost comparisons, properties comparisons, and conformance with proposed regulatory and repository criteria. Both the glass and ceramic forms are viable candidates for use at DOE defense HLW sites; they are also candidates for immobilization of commercial reprocessing wastes. This paper describes the waste form screening process, discusses each of the four major inputs considered in the selection of the two forms in 1981, and presents a brief summary of the comparisons of the two forms during 1982 and the selection process to determine the final form for SRP defense HLW

Glass as a waste form for the immobilization of plutonium

International Nuclear Information System (INIS)

Bates, J.K.; Ellison, A.J.G.; Emery, J.W.; Hoh, J.C.

1995-01-01

Several alternatives for disposal of surplus plutonium are being considered. One method is incorporating Pu into glass and in this paper we discuss the development and corrosion behavior of an alkali-tin-silicate glass and update results in testing Pu doped Defense Waste Processing Facility (DWPF) reference glasses. The alkali-tin-silicate glass was engineered to accommodate a high Pu loading and to be durable under conditions likely to accelerate glass reaction. The glass dissolves about 7 wt% Pu together with the neutron absorber Gd, and under test conditions expected to accelerate the glass reaction with water, is resistant to corrosion. The Pu and the Gd are released from the glass at nearly the same rate in static corrosion tests in water, and are not segregated into surface alteration phases when the glass is reacted in water vapor. Similar results for the behavior of Pu and Gd are found for the DWPF reference glasses, although the long-term rate of reaction for the reference glasses is more rapid than for the alkali-tin-silicate glass

Package materials, waste form

International Nuclear Information System (INIS)

Anon.

1980-01-01

The schedules for waste package development for the various host rocks were presented. The waste form subtask activities were reviewed, with the papers focusing on high-level waste, transuranic waste, and spent fuel. The following ten papers were presented: (1) Waste Package Development Approach; (2) Borosilicate Glass as a Matrix for Savannah River Plant Waste; (3) Development of Alternative High-Level Waste Forms; (4) Overview of the Transuranic Waste Management Program; (5) Assessment of the Impacts of Spent Fuel Disassembly - Alternatives on the Nuclear Waste Isolation System; (6) Reactions of Spent Fuel and Reprocessing Waste Forms with Water in the Presence of Basalt; (7) Spent Fuel Stabilizer Screening Studies; (8) Chemical Interactions of Shale Rock, Prototype Waste Forms, and Prototype Canister Metals in a Simulated Wet Repository Environment; (9) Impact of Fission Gas and Volatiles on Spent Fuel During Geologic Disposal; and (10) Spent Fuel Assembly Decay Heat Measurement and Analysis

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

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.)

High-level waste-form-product performance evaluation

International Nuclear Information System (INIS)

Bernadzikowski, T.A.; Allender, J.S.; Stone, J.A.; Gordon, D.E.; Gould, T.H. Jr.; Westberry, C.F. III.

1982-01-01

Seven candidate waste forms were evaluated for immobilization and geologic disposal of high-level radioactive wastes. The waste forms were compared on the basis of leach resistance, mechanical stability, and waste loading. All forms performed well at leaching temperatures of 40, 90, and 150 0 C. Ceramic forms ranked highest, followed by glasses, a metal matrix form, and concrete. 11 tables

Processing of high-temperature simulated waste glass in a continuous ceramic melter

International Nuclear Information System (INIS)

Barnes, S.M.; Brouns, R.A.; Hanson, M.S.

1980-01-01

Recent operations have demonstrated that high-melting-point glasses and glass-ceramics can be successfully processed in joule-heated, ceramic-lined melters with minor modifications to the existing technology. Over 500 kg of simulated waste glasses have been processed at temperatures up to 1410 0 C. The processability of the two high-temperature waste forms tested is similar to existing borosilicate waste glasses. High-temperature waste glass formulations produced in the bench-scale melter exhibit quality comparing favorably to standard waste glass formulations

DEFENSE HIGH LEVEL WASTE GLASS DEGRADATION

International Nuclear Information System (INIS)

Ebert, W.

2001-01-01

The purpose of this Analysis/Model Report (AMR) is to document the analyses that were done to develop models for radionuclide release from high-level waste (HLW) glass dissolution that can be integrated into performance assessment (PA) calculations conducted to support site recommendation and license application for the Yucca Mountain site. This report was developed in accordance with the ''Technical Work Plan for Waste Form Degradation Process Model Report for SR'' (CRWMS M andO 2000a). It specifically addresses the item, ''Defense High Level Waste Glass Degradation'', of the product technical work plan. The AP-3.15Q Attachment 1 screening criteria determines the importance for its intended use of the HLW glass model derived herein to be in the category ''Other Factors for the Postclosure Safety Case-Waste Form Performance'', and thus indicates that this factor does not contribute significantly to the postclosure safety strategy. Because the release of radionuclides from the glass will depend on the prior dissolution of the glass, the dissolution rate of the glass imposes an upper bound on the radionuclide release rate. The approach taken to provide a bound for the radionuclide release is to develop models that can be used to calculate the dissolution rate of waste glass when contacted by water in the disposal site. The release rate of a particular radionuclide can then be calculated by multiplying the glass dissolution rate by the mass fraction of that radionuclide in the glass and by the surface area of glass contacted by water. The scope includes consideration of the three modes by which water may contact waste glass in the disposal system: contact by humid air, dripping water, and immersion. The models for glass dissolution under these contact modes are all based on the rate expression for aqueous dissolution of borosilicate glasses. The mechanism and rate expression for aqueous dissolution are adequately understood; the analyses in this AMR were conducted to

Corrosion of inconel in high-temperature borosilicate glass melts containing simulant nuclear waste

Science.gov (United States)

Mao, Xianhe; Yuan, Xiaoning; Brigden, Clive T.; Tao, Jun; Hyatt, Neil C.; Miekina, Michal

2017-10-01

The corrosion behaviors of Inconel 601 in the borosilicate glass (MW glass) containing 25 wt.% of simulant Magnox waste, and in ZnO, Mn2O3 and Fe2O3 modified Mg/Ca borosilicate glasses (MZMF and CZMF glasses) containing 15 wt.% of simulant POCO waste, were evaluated by dimensional changes, the formation of internal defects and changes in alloy composition near corrosion surfaces. In all three kinds of glass melts, Cr at the inconel surface forms a protective Cr2O3 scale between the metal surface and the glass, and alumina precipitates penetrate from the metal surface or formed in-situ. The corrosion depths of inconel 601 in MW waste glass melt are greater than those in the other two glass melts. In MW glass, the Cr2O3 layer between inconel and glass is fragmented because of the reaction between MgO and Cr2O3, which forms the crystal phase MgCr2O4. In MZMF and CZMF waste glasses the layers are continuous and a thin (Zn, Fe, Ni, B)-containing layer forms on the surface of the chromium oxide layer and prevents Cr2O3 from reacting with MgO or other constituents. MgCr2O4 was observed in the XRD analysis of the bulk MW waste glass after the corrosion test, and ZrSiO4 in the MZMF waste glass, and ZrSiO4 and CaMoO4 in the CZMF waste glass.

Lead-iron phosphate glass: a stable storage medium for high-level nuclear waste

International Nuclear Information System (INIS)

Sales, B.C.; Boatner, L.A.

1984-01-01

Results are presented which show that lead-iron phosphate glasses are a promising new waste form for the safe immobilization of both high-level defense and high-level commercial radioactive waste. Relative to the borosilicate nuclear waste glasses that are currently the ''reference'' waste form for the long-term disposal of nuclear waste, lead-iron phosphate glasses have several distinct advantages: (1) an aqueous corrosion rate that is about 1000 times lower, (2) a processing temperature that is 100 0 to 250 0 C lower and, (3) a much lower melt viscosity in the temperature range from 800 0 to 1000 0 C. Most significantly, the lead-iron phosphate waste form can be processed using a technology similar to that developed for borosilicate nuclear waste glasses

Application of PCT to the EBR II ceramic waste form

International Nuclear Information System (INIS)

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

2002-01-01

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

Crystallization in high-level waste glass: A review of glass theory and noteworthy literature

Energy Technology Data Exchange (ETDEWEB)

Christian, J. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2015-08-18

There is a fundamental need to continue research aimed at understanding nepheline and spinel crystal formation in high-level waste (HLW) glass. Specifically, the formation of nepheline solids (K/NaAlSiO₄) during slow cooling of HLW glass can reduce the chemical durability of the glass, which can cause a decrease in the overall durability of the glass waste form. The accumulation of spinel solids ((Fe, Ni, Mn, Zn)(Fe, Cr)₂O₄), while not detrimental to glass durability, can cause an array of processing problems inside HLW glass melters. In this review, the fundamental differences between glass and solid-crystals are explained using kinetic, thermodynamic, and viscosity arguments, and several highlights of glass-crystallization research, as it pertains to high-level waste vitrification, are described. In terms of mitigating spinel in the melter and both spinel and nepheline formation in the canister, the complexity of HLW glass and the intricate interplay between thermal, chemical, and kinetic factors further complicates this understanding. However, new experiments seeking to elucidate the contributing factors of crystal nucleation and growth in waste glass, and the compilation of data from older experiments, may go a long way towards helping to achieve higher waste loadings while developing more efficient processing strategies. Higher waste loadings and more efficient processing strategies will reduce the overall HLW Hanford Tank Waste Treatment and Immobilization Plant (WTP) vitrification facilities mission life.

Chemical compatibility of DWPF canistered waste forms

International Nuclear Information System (INIS)

Harbour, J.R.

1993-01-01

The Waste Acceptance Preliminary Specifications (WAPS) require that the contents of the canistered waste form are compatible with one another and the stainless steel canister. The canistered waste form is a closed system comprised of a stainless steel vessel containing waste glass, air, and condensate. This system will experience a radiation field and an elevated temperature due to radionuclide decay. This report discusses possible chemical reactions, radiation interactions, and corrosive reactions within this system both under normal storage conditions and after exposure to temperatures up to the normal glass transition temperature, which for DWPF waste glass will be between 440 and 460 degrees C. Specific conclusions regarding reactions and corrosion are provided. This document is based on the assumption that the period of interim storage prior to packaging at the federal repository may be as long as 50 years

Crystallization in high-level waste glass: A review of glass theory and noteworthy literature

Energy Technology Data Exchange (ETDEWEB)

Christian, J. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2015-08-01

There is a fundamental need to continue research aimed at understanding nepheline and spinel crystal formation in high-level waste (HLW) glass. Specifically, the formation of nepheline solids (K/NaAlSiO₄) during slow cooling of HLW glass can reduce the chemical durability of the glass, which can cause a decrease in the overall durability of the glass waste form. The accumulation of spinel solids ((Fe, Ni, Mn, Zn)(Fe,Cr)₂O₄), while not detrimental to glass durability, can cause an array of processing problems inside of HLW glass melters. In this review, the fundamental differences between glass and solid-crystals are explained using kinetic, thermodynamic, and viscosity arguments, and several highlights of glass-crystallization research, as it pertains to high-level waste vitrification, are described. In terms of mitigating spinel in the melter and both spinel and nepheline formation in the canister, the complexity of HLW glass and the intricate interplay between thermal, chemical, and kinetic factors further complicates this understanding. However, new experiments seeking to elucidate the contributing factors of crystal nucleation and growth in waste glass, and the compilation of data from older experiments, may go a long way towards helping to achieve higher waste loadings while developing more efficient processing strategies.

Development Of Glass Matrices For HLW Radioactive Wastes

International Nuclear Information System (INIS)

Jantzen, C.

2010-01-01

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

Glasses and ceramics for immobilisation of radioactive wastes for disposal

International Nuclear Information System (INIS)

Johnson, K.D.B.; Marples, J.A.C.

1979-05-01

The U.K. Research Programme on Radioactive Waste Management includes the development of processes for the conversion of high level liquid reprocessing wastes from thermal and fast reactors to borosilicate glasses. The properties of these glasses and their behaviour under storage and disposal conditions have been examined. Methods for immobilising activity from other wastes by conversion to glass or ceramic forms is described. The U.K. philosophy of final solutions to waste management and disposal is presented. (author)

Characterization of the Defense Waste Processing Facility (DWPF) Environmental Assessment (EA) glass Standard Reference Material. Revision 1

Energy Technology Data Exchange (ETDEWEB)

Jantzen, C.M.; Bibler, N.E.; Beam, D.C.; Crawford, C.L.; Pickett, M.A.

1993-06-01

Liquid high-level nuclear waste at the Savannah River Site (SRS) will be immobilized by vitrification in borosilicate glass. The glass will be produced and poured into stainless steel canisters in the Defense Waste Processing Facility (DWPF). Other waste form producers, such as West Valley Nuclear Services (WVNS) and the Hanford Waste Vitrification Project (HWVP), will also immobilize high-level radioactive waste in borosilicate glass. The canistered waste will be stored temporarily at each facility for eventual permanent disposal in a geologic repository. The Department of Energy has defined a set of requirements for the canistered waste forms, the Waste Acceptance Product Specifications (WAPS). The current Waste Acceptance Primary Specification (WAPS) 1.3, the product consistency specification, requires the waste form producers to demonstrate control of the consistency of the final waste form using a crushed glass durability test, the Product Consistency Test (PCI). In order to be acceptable, a waste glass must be more durable during PCT analysis than the waste glass identified in the DWPF Environmental Assessment (EA). In order to supply all the waste form producers with the same standard benchmark glass, 1000 pounds of the EA glass was fabricated. The chemical analyses and characterization of the benchmark EA glass are reported. This material is now available to act as a durability and/or redox Standard Reference Material (SRM) for all waste form producers.

Ceramic nuclear waste forms. II. A ceramic-waste composite prepared by hot pressing. Progress report and preprint

International Nuclear Information System (INIS)

McCarthy, G.J.

1975-01-01

A feasibility study was conducted to determine whether nuclear waste calcine and a crystalline ceramic matrix can be fabricated by hot pressing into a composite waste form with suitable leaching resistance and thermal stability. It was found that a hard, dense composite could be formed using the typical commercial waste formulation PW-4b and a matrix of α-quartz with a small amount of a lead borosilicate glass added as a consolidation aide. Its density, waste loading, and leaching resistance are comparable to the glasses currently being considered for fixation of nuclear wastes. The hot pressed composite offers a closer approach to thermodynamic stability and improved thermal stability (in monolithic form) compared to glass waste forms. Recommendations for further optimization of the hot pressed waste form are given. (U.S.)

Glass science tutorial: Lecture No. 7, Waste glass technology for Hanford

International Nuclear Information System (INIS)

Kruger, A.A.

1995-07-01

This paper presents the details of the waste glass tutorial session that was held to promote knowledge of waste glass technology and how this can be used at the Hanford Reservation. Topics discussed include: glass properties; statistical approach to glass development; processing properties of nuclear waste glass; glass composition and the effects of composition on durability; model comparisons of free energy of hydration; LLW glass structure; glass crystallization; amorphous phase separation; corrosion of refractories and electrodes in waste glass melters; and glass formulation for maximum waste loading

SETTLING OF SPINEL IN A HIGH-LEVEL WASTE GLASS MELTER

International Nuclear Information System (INIS)

Pavel Hrma; Pert Schill; Lubomir Nemec

2002-01-01

High-level nuclear waste is being vitrified, i.e., converted to a durable glass that can be stored in a safe repository for hundreds of thousands of years. Waste vitrification is accomplished in reactors called melters to which the waste is charged together with glass-forming additives. The mixture is electrically heated to a temperature as high as 1150 decrees C to create a melt that becomes glass on cooling

Decontamination of Savannah River Plant waste glass canisters

International Nuclear Information System (INIS)

Rankin, W.N.

1982-01-01

A Defense Waste Processing Facility (DWPF) is currently being designed to convert Savannah River Plant (SRP) liquid, high-level radioactive waste into a solid form, such as borosilicate glass. The outside of the canisters of waste glass must have very low levels of smearable radioactive contamination before they are removed from the DWPF to prevent the spread of radioactivity. Several techniques were considered for canister decontamination: high-pressure water spray, electropolishing, chemical dissolution, and abrasive blasting. An abrasive blasting technique using a glass frit slurry has been selected for use in the DWPF. No additional equipment is needed to process waste generated from decontamination. Frit used as the abrasive will be mixed with the waste and fed to the glass melter. In contrast, chemical and electrochemical techniques require more space in the DWPF, and produce large amounts of contaminated byproducts which are difficult to immobilize by vitrification

Transuranic waste management program waste form development

International Nuclear Information System (INIS)

Bennett, W.S.; Crisler, L.R.

1981-01-01

To ensure that all technology necessary for long term management of transuranic (TRU) wastes is available, the Department of Energy has established the Transuranic Waste Management Program. A principal focus of the program is development of waste forms that can accommodate the very diverse TRU waste inventory and meet geologic isolation criteria. The TRU Program is following two approaches. First, decontamination processes are being developed to allow removal of sufficient surface contamination to permit management of some of the waste as low level waste. The other approach is to develop processes which will allow immobilization by encapsulation of the solids or incorporate head end processes which will make the solids compatible with more typical waste form processes. The assessment of available data indicates that dewatered concretes, synthetic basalts, and borosilicate glass waste forms appear to be viable candidates for immobilization of large fractions of the TRU waste inventory in a geologic repository

DEVELOPMENT OF GLASS MATRICES FOR HLW RADIOACTIVE WASTES

Energy Technology Data Exchange (ETDEWEB)

Jantzen, C.

2010-03-18

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

Development and characterization of basalt-glass ceramics for the immobilization of transuranic wastes

International Nuclear Information System (INIS)

Lokken, R.O.; Chick, L.A.; Thomas, L.E.

1982-09-01

Basalt-based waste forms were developed for the immobilization of transuranic (TRU) contaminated wastes. The specific waste studied is a 3:1 blend of process sludge and incinerator ash. Various amounts of TRU blended waste were melted with Pomona basalt powder. The vitreous products were subjected to a variety of heat treatment conditions to form glass ceramics. The total crystallinity of the glass ceramic, ranging from 20 to 45 wt %, was moderately dependent on composition and heat treatment conditions. Three parent glasses and four glass ceramics with varied composition and heat treatment were produced for detailed phase characterization and leaching. Both parent glasses and glass ceramics were mainly composed of a continuous, glassy matrix phase. This glass matrix entered into solution during leaching in both types of materials. The Fe-Ti rich dispersed glass phase was not significantly degraded by leaching. The glass ceramics, however, exhibited four to ten times less elemental releases during leaching than the parent glasses. The glass ceramic matrix probably contains higher Fe and Na and lower Ca and Mg relative to the parent glass matrix. The crystallization of augite in the glass ceramics is believed to contribute to the improved leach rates. Leach rates of the basalt glass ceramic are compared to those of other TRU nuclear waste forms containing 239 Pu

Preparation of plutonium waste forms with ICPP calcined high-level waste

Energy Technology Data Exchange (ETDEWEB)

Staples, B.A.; Knecht, D.A. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States); O`Holleran, T.P. [Argonne National Lab.-West, Idaho Falls, ID (United States)] [and others

1997-05-01

Glass and glass-ceramic forms developed for the immobilization of calcined high-level wastes generated by Idaho Chemical Processing Plant (ICPP) fuel reprocessing activities have been investigated for ability to immobilize plutonium and to simultaneously incorporate calcined waste as an anti-proliferation barrier. Within the forms investigated, crystallization of host phases result in an increased loading of plutonium as well as its incorporation into potentially more durable phases than the glass. The host phases were initially formed and characterized with cerium (Ce{sup +4}) as a surrogate for plutonium (Pu{sup +4}) and samarium as a neutron absorber for criticality control. Verification of the surrogate testing results were then performed replacing cerium with plutonium. All testing was performed with surrogate calcined high-level waste. The results of these tests indicated that a potentially useful host phase, based on zirconia, can be formed either by devitrification or solid state reaction in the glass studied. This phase incorporates plutonium as well as samarium and the calcined waste becomes part of the matrix. Its ease of formation makes it potentially useful in excess plutonium dispositioning. Other durable host phases for plutonium and samarium, including zirconolite and zircon have been formed from zirconia or alumina calcine through cold press-sintering techniques and hot isostatic pressing. Host phase formation experiments conducted through vitrification or by cold press-sintering techniques are described and the results discussed. Recommendations are given for future work that extends the results of this study.

Preparation of plutonium waste forms with ICPP calcined high-level waste

International Nuclear Information System (INIS)

Staples, B.A.; Knecht, D.A.; O'Holleran, T.P.

1997-05-01

Glass and glass-ceramic forms developed for the immobilization of calcined high-level wastes generated by Idaho Chemical Processing Plant (ICPP) fuel reprocessing activities have been investigated for ability to immobilize plutonium and to simultaneously incorporate calcined waste as an anti-proliferation barrier. Within the forms investigated, crystallization of host phases result in an increased loading of plutonium as well as its incorporation into potentially more durable phases than the glass. The host phases were initially formed and characterized with cerium (Ce +4 ) as a surrogate for plutonium (Pu +4 ) and samarium as a neutron absorber for criticality control. Verification of the surrogate testing results were then performed replacing cerium with plutonium. All testing was performed with surrogate calcined high-level waste. The results of these tests indicated that a potentially useful host phase, based on zirconia, can be formed either by devitrification or solid state reaction in the glass studied. This phase incorporates plutonium as well as samarium and the calcined waste becomes part of the matrix. Its ease of formation makes it potentially useful in excess plutonium dispositioning. Other durable host phases for plutonium and samarium, including zirconolite and zircon have been formed from zirconia or alumina calcine through cold press-sintering techniques and hot isostatic pressing. Host phase formation experiments conducted through vitrification or by cold press-sintering techniques are described and the results discussed. Recommendations are given for future work that extends the results of this study

Effects of radionuclide decay on waste glass behavior: A critical review

International Nuclear Information System (INIS)

Wronkiewicz, D.J.

1993-12-01

This paper is an extension of a chapter in an earlier report [1] that provides an updated review on the status of radiation damage problems in nuclear waste glasses. This report will focus on radiation effects on vitrified borosilicate nuclear waste glasses under conditions expected in the proposed Yucca mountain repository. Radiation effects on high-level waste glasses and their surrounding repository environment are important considerations for radionuclide immobilization because of the potential to alter the glass stability and thereby influence the radionuclide retentive properties of this waste form. The influence of radionuclide decay on vitrified nuclear waste may be manifested by several changes, including volume, stored energy, structure, microstructure, mechanical properties, and phase separation. Radiation may also affect the composition of aqueous fluids and atmospheric gases in relatively close proximity to the waste form. What is important to the radionuclide retentive properties of the repository is how these radiation effects collectively or individually influence the durability and radionuclide release from the glass in the event of liquid water contact

Evaluation and selection of candidate high-level waste forms

International Nuclear Information System (INIS)

1982-03-01

Seven candidate waste forms being developed under the direction of the Department of Energy's National High-Level Waste (HLW) Technology Program, were evaluated as potential media for the immobilization and geologic disposal of high-level nuclear wastes. The evaluation combined preliminary waste form evaluations conducted at DOE defense waste-sites and independent laboratories, peer review assessments, a product performance evaluation, and a processability analysis. Based on the combined results of these four inputs, two of the seven forms, borosilicate glass and a titanate based ceramic, SYNROC, were selected as the reference and alternative forms for continued development and evaluation in the National HLW Program. Both the glass and ceramic forms are viable candidates for use at each of the DOE defense waste-sites; they are also potential candidates for immobilization of commercial reprocessing wastes. This report describes the waste form screening process, and discusses each of the four major inputs considered in the selection of the two forms

Waste form development for a DC arc furnace

Energy Technology Data Exchange (ETDEWEB)

Feng, X.; Bloomer, P.E.; Chantaraprachoom, N.; Gong, M.; Lamar, D.A.

1996-09-01

A laboratory crucible study was conducted to develop waste forms to treat nonradioactive simulated {sup 238}Pu heterogeneous debris waste from Savannah River, metal waste from the Idaho National Engineering Laboratory (INEL), and nominal waste also from INEL using DC arc melting. The preliminary results showed that the different waste form compositions had vastly different responses for each processing effect. The reducing condition of DC arc melting had no significant effects on the durability of some waste forms while it decreased the waste form durability from 300 to 700% for other waste forms, which resulted in the failure of some TCLP tests. The right formulations of waste can benefit from devitrification and showed an increase in durability by 40%. Some formulations showed no devitrification effects while others decreased durability by 200%. Increased waste loading also affected waste form behavior, decreasing durability for one waste, increasing durability by 240% for another, and showing no effect for the third waste. All of these responses to the processing and composition variations were dictated by the fundamental glass chemistry and can be adjusted to achieve maximal waste loading, acceptable durability, and desired processing characteristics if each waste formulation is designed for the result according to the glass chemistry.

Chemical and mechanical performance properties for various final waste forms -- PSPI scoping study

International Nuclear Information System (INIS)

Farnsworth, R.K.; Larsen, E.D.; Sears, J.W.; Eddy, T.L.; Anderson, G.L.

1996-09-01

The US DOE is obtaining data on the performance properties of the various final waste forms that may be chosen as primary treatment products for the alpha-contaminated low-level and transuranic waste at the INEL's Transuranic Storage Area. This report collects and compares selected properties that are key indicators of mechanical and chemical durability for Portland cement concrete, concrete formed under elevated temperature and pressure, sulfur polymer cement, borosilicate glass, and various forms of alumino-silicate glass, including in situ vitrification glass and various compositions of iron-enriched basalt (IEB) and iron-enriched basalt IV (IEB4). Compressive strength and impact resistance properties were used as performance indicators in comparative evaluation of the mechanical durability of each waste form, while various leachability data were used in comparative evaluation of each waste form's chemical durability. The vitrified waste forms were generally more durable than the non-vitrified waste forms, with the iron-enriched alumino-silicate glasses and glass/ceramics exhibiting the most favorable chemical and mechanical durabilities. It appears that the addition of zirconia and titania to IEB (forming IEB4) increases the leach resistance of the lanthanides. The large compositional ranges for IEB and IEB4 more easily accommodate the compositions of the waste stored at the INEL than does the composition of borosilicate glass. It appears, however, that the large potential variation in IEB and IEB4 compositions resulting from differing waste feed compositions can impact waste form durability. Further work is needed to determine the range of waste stream feed compositions and rates of waste form cooling that will result in acceptable and optimized IEB or IEB4 waste form performance. 43 refs

Impact test for solid waste forms

International Nuclear Information System (INIS)

Wallace, R.M.; Kelley, J.A.

1976-03-01

Samples of concretes and glasses being considered for incorporation of radioactive waste sludge were subjected to impact tests to determine the relationship between the energy of the impact and the resulting increase in surface area of the damaged sample. Test results indicate that the increased surface area per unit of energy input for glass waste forms is less by a factor of about three than that for concretes containing 40 wt percent simulated sludge (average values of 9.6 cm 2 /Joule and 24.7 cm 2 /Joule for glass and concrete, respectively)

Control of radioactive waste-glass melters

International Nuclear Information System (INIS)

Bickford, D.F.; Smith, P.K.; Hrma, P.; Bowan, B.W.

1987-01-01

Radioactive waste-glass melters require physical control limits and redox control of glass to assure continuous operation, and maximize production rates. Typical waste-glass melter operating conditions, and waste-glass chemical reaction paths are discussed. Glass composition, batching and melter temperature control are used to avoid the information of phases which are disruptive to melting or reduce melter life. The necessity and probable limitations of control for electric melters with complex waste feed compositions are discussed. Preliminary control limits, their bases, and alternative control methods are described for use in the Defense Waste Processing Facility (DWPF) at the US Department of Energy's Savannah River Plant (SRP), and at the West Valley Demonstration Project (WVDP). Slurries of simulated high level radioactive waste and ground glass frit or glass formers have been isothermally reacted and analyzed to identify the sequence of the major chemical reactions in waste vitrification, and their effect on waste-glass production rates. Relatively high melting rates of waste batches containing mixtures of reducing agents (formic acid, sucrose) and nitrates are attributable to exothermic reactions which occur at critical stages in the vitrification process. The effect of foaming on waste glass production rates is analyzed, and limits defined for existing waste-glass melters, based upon measurable thermophysical properties. Through balancing the high nitrate wastes of the WVDP with reducing agents, the high glass melting rates and sustained melting without foaming required for successful WVDP operations have been demonstrated. 65 refs., 4 figs., 15 tabs

Multibarrier waste forms. Part III: Process considerations

International Nuclear Information System (INIS)

Lokken, R.O.

1979-10-01

The multibarrier concept for the solidification and storage of radioactive waste utilizes up to three barriers to isolate radionuclides from the environment: a solidified waste inner core, an impervious coating, and a metal matrix. The coating and metal matrix give the composite waste form enhanced inertness with improvements in thermal stability, mechanical strength, and leach resistance. Preliminary process flow rates and material costs were evaluated for four multibarrier waste forms with the process complexity increasing thusly: glass marbles, uncoated supercalcine, glass-coated supercalcine, and PyC/Al 2 O 3 -coated supercalcine. This report discusses the process variables and their effect on optimization of product quality, processing simplicity, and material cost. 11 figures, 2 tables

Leach rate characterization of solid radioactive waste forms

International Nuclear Information System (INIS)

Flynn, K.F.; Barletta, R.E.; Jardine, L.J.; Steindler, M.J.

1978-01-01

Leach rates were measured using distilled water on four types of waste forms: spray calcined waste mixed with silica and borosilicate glass and sintered, the same pulverized, the same in a lead matrix, and waste glass containing U. Twenty isotopes ranging from 22 Na to 239 Np were measured using activation analysis. Leach rates were also measured for a variety of matrix materials (Zircaloy, Al, Pb, glass, Pb 3 RE 6 (SiO 4 ) 6 ), using one isotope each. 2 tables

Development of solid radionuclide waste forms in the United States

International Nuclear Information System (INIS)

Crandall, J.L.

1979-01-01

New ways of reworking the wastes require a new classification in terms of the final waste forms. This paper surveys the candidate forms: encapsulation binders, in-place solidification waste forms, glass and ceramic waste forms, mineral waste forms, matrix waste forms, gaseous waste forms (fixation), and canisters and engineered barriers. Participants in the US-high-level waste form development program are listed. Requirements and selection of waste forms are also discussed. 26 references

Fracturing of simulated high-level waste glass in canisters

International Nuclear Information System (INIS)

Peters, R.D.; Slate, S.C.

1981-09-01

Waste-glass castings generated from engineering-scale developmental processes at the Pacific Northwest Laboratory are generally found to have significant levels of cracks. The causes and extent of fracturing in full-scale canisters of waste glass as a result of cooling and accidental impact are discussed. Although the effects of cracking on waste-form performance in a repository are not well understood, cracks in waste forms can potentially increase leaching surface area. If cracks are minimized or absent in the waste-glass canisters, the potential for radionuclide release from the canister package can be reduced. Additional work on the effects of cracks on leaching of glass is needed. In addition to investigating the extent of fracturing of glass in waste-glass canisters, methods to reduce cracking by controlling cooling conditions were explored. Overall, the study shows that the extent of glass cracking in full-scale, passively-cooled, continuous melting-produced canisters is strongly dependent on the cooling rate. This observation agrees with results of previously reported Pacific Northwest Laboratory experiments on bench-scale annealed canisters. Thus, the cause of cracking is principally bulk thermal stresses. Fracture damage resulting from shearing at the glass/metal interface also contributes to cracking, more so in stainless steel canisters than in carbon steel canisters. This effect can be reduced or eliminated with a graphite coating applied to the inside of the canister. Thermal fracturing can be controlled by using a fixed amount of insulation for filling and cooling of canisters. In order to maintain production rates, a small amount of additional facility space is needed to accomodate slow-cooling canisters. Alternatively, faster cooling can be achieved using the multi-staged approach. Additional development is needed before this approach can be used on full-scale (60-cm) canisters

Alternative solidified forms for nuclear wastes

International Nuclear Information System (INIS)

McElroy, J.L.; Ross, W.A.

1976-01-01

Radioactive wastes will occur in various parts of the nuclear fuel cycle. These wastes have been classified in this paper as high-level waste, intermediate and low-level waste, cladding hulls, and residues. Solidification methods for each type of waste are discussed in a multiple barrier context of primary waste form, applicable coatings or films, matrix encapsulation, canister, engineered structures, and geological storage. The four major primary forms which have been most highly developed are glass for HLW, cement for ILW, organics for LLW, and metals for hulls

GLASS COMPOSITION-TCLP RESPONSE MODEL FOR WASTE GLASSES

International Nuclear Information System (INIS)

Kim, Dong-Sang; Vienna, John D.

2004-01-01

A first-order property model for normalized Toxicity Characteristic Leaching Procedure (TCLP) release as a function of glass composition was developed using data collected from various studies. The normalized boron release is used to estimate the release of toxic elements based on the observation that the boron release represents the conservative release for those constituents of interest. The current TCLP model has two targeted application areas: (1) delisting of waste-glass product as radioactive (not mixed) waste and (2) designating the glass wastes generated from waste-glass research activities as hazardous or non-hazardous. This paper describes the data collection and model development for TCLP releases and discusses the issues related to the application of the model

Standard test methods for determining chemical durability of nuclear, hazardous, and mixed waste glasses and multiphase glass ceramics: The product consistency test (PCT)

CERN Document Server

American Society for Testing and Materials. Philadelphia

2002-01-01

1.1 These product consistency test methods A and B evaluate the chemical durability of homogeneous glasses, phase separated glasses, devitrified glasses, glass ceramics, and/or multiphase glass ceramic waste forms hereafter collectively referred to as “glass waste forms” by measuring the concentrations of the chemical species released to a test solution. 1.1.1 Test Method A is a seven-day chemical durability test performed at 90 ± 2°C in a leachant of ASTM-Type I water. The test method is static and conducted in stainless steel vessels. Test Method A can specifically be used to evaluate whether the chemical durability and elemental release characteristics of nuclear, hazardous, and mixed glass waste forms have been consistently controlled during production. This test method is applicable to radioactive and simulated glass waste forms as defined above. 1.1.2 Test Method B is a durability test that allows testing at various test durations, test temperatures, mesh size, mass of sample, leachant volume, a...

Alternatives for high-level waste forms, containers, and container processing systems

International Nuclear Information System (INIS)

Crawford, T.W.

1995-01-01

This study evaluates alternatives for high-level waste forms, containers, container processing systems, and onsite interim storage. Glass waste forms considered are cullet, marbles, gems, and monolithic glass. Small and large containers configured with several combinations of overpack confinement and shield casks are evaluated for these waste forms. Onsite interim storage concepts including canister storage building, bore holes, and storage pad were configured with various glass forms and canister alternatives. All favorable options include the monolithic glass production process as the waste form. Of the favorable options the unshielded 4- and 7-canister overpack options have the greatest technical assurance associated with their design concepts due to their process packaging and storage methods. These canisters are 0.68 m and 0.54 m in diameter respectively and 4.57 m tall. Life-cycle costs are not a discriminating factor in most cases, varying typically less than 15 percent

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

International Nuclear Information System (INIS)

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

1992-01-01

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

Comparative risk assessments for the production and interim storage of glass and ceramic waste forms: defense waste processing facility

International Nuclear Information System (INIS)

Huang, J.C.; Wright, W.V.

1982-04-01

The Defense Waste Processing Facility (DWPF) for immobilizing nuclear high level waste (HLW) is scheduled to be built at the Savannah River Plant (SRP). High level waste is produced when SRP reactor components are subjected to chemical separation operations. Two candidates for immobilizing this HLW are borosilicate glass and crystalline ceramic, either being contained in weld-sealed stainless steel canisters. A number of technical analyses are being conducted to support a selection between these two waste forms. The present document compares the risks associated with the manufacture and interim storage of these two forms in the DWPF. Process information used in the risk analysis was taken primarily from a DWPF processibility analysis. The DWPF environmental analysis provided much of the necessary environmental information. To perform the comparative risk assessments, consequences of the postulated accidents are calculated in terms of: (1) the maximum dose to an off-site individual; and (2) the dose to off-site population within 80 kilometers of the DWPF, both taken in terms of the 50-year inhalation dose commitment. The consequences are then multiplied by the estimated accident probabilities to obtain the risks. The analyses indicate that the maximum exposure risk to an individual resulting from the accidents postulated for both the production and interim storage of either waste form represents only an insignificant fraction of the natural background radiation of about 90 mrem per year per person in the local area. They also show that there is no disaster potential to the off-site population. Therefore, the risks from abnormal events in the production and the interim storage of the DWPF waste forms should not be considered as a dominant factor in the selection of the final waste form

Standard test method for determining liquidus temperature of immobilized waste glasses and simulated waste glasses

CERN Document Server

American Society for Testing and Materials. Philadelphia

2011-01-01

1.1 These practices cover procedures for determining the liquidus temperature (TL) of nuclear waste, mixed nuclear waste, simulated nuclear waste, or hazardous waste glass in the temperature range from 600°C to 1600°C. This method differs from Practice C829 in that it employs additional methods to determine TL. TL is useful in waste glass plant operation, glass formulation, and melter design to determine the minimum temperature that must be maintained in a waste glass melt to make sure that crystallization does not occur or is below a particular constraint, for example, 1 volume % crystallinity or T1%. As of now, many institutions studying waste and simulated waste vitrification are not in agreement regarding this constraint (1). 1.2 Three methods are included, differing in (1) the type of equipment available to the analyst (that is, type of furnace and characterization equipment), (2) the quantity of glass available to the analyst, (3) the precision and accuracy desired for the measurement, and (4) candi...

Development and evaluation of candidate high-level waste forms

International Nuclear Information System (INIS)

Bernadzikowski, T.A.

1981-01-01

Some seventeen candidate waste forms have been investigated under US Department of Energy programs as potential media for the immobilization and geologic disposal of the high-level radioactive wastes (HLW) resulting from chemical processing of nuclear reactor fuels and targets. Two of these HLW forms were selected at the end of fiscal year (FY) 1981 for intensive development if FY 1982 to 1983. Borosilicate glass was continued as the reference form. A crystalline ceramic waste form, SYNROC, was selected for further product formulation and process development as the alternative to borosilicate glass. This paper describes the bases on which this decision was made

Stability of High-Level Waste Forms

Energy Technology Data Exchange (ETDEWEB)

Besmann, Theodore M.; Vienna, John D.

2005-09-30

The objective of the proposed effort is to use a new approach to develop solution models of complex waste glass systems and spent fuel that are predictive with regard to composition, phase separation, and volatility. The effort will also yield thermodynamic values for waste components that are fundamentally required for corrosion models used to predict the leaching/corrosion behavior for waste glass and spent fuel material. This basic information and understanding of chemical behavior can subsequently be used directly in computational models of leaching and transport in geologic media, in designing and engineering waste forms and barrier systems, and in prediction of chemical interactions.

Mechanical properties of nuclear waste glasses

International Nuclear Information System (INIS)

Connelly, A.J.; Hand, R.J.; Bingham, P.A.; Hyatt, N.C.

2011-01-01

The mechanical properties of nuclear waste glasses are important as they will determine the degree of cracking that may occur either on cooling or following a handling accident. Recent interest in the vitrification of intermediate level radioactive waste (ILW) as well as high level radioactive waste (HLW) has led to the development of new waste glass compositions that have not previously been characterised. Therefore the mechanical properties, including Young's modulus, Poisson's ratio, hardness, indentation fracture toughness and brittleness of a series of glasses designed to safely incorporate wet ILW have been investigated. The results are presented and compared with the equivalent properties of an inactive simulant of the current UK HLW glass and other nuclear waste glasses from the literature. The higher density glasses tend to have slightly lower hardness and indentation fracture toughness values and slightly higher brittleness values, however, it is shown that the variations in mechanical properties between these different glasses are limited, are well within the range of published values for nuclear waste glasses, and that the surveyed data for all radioactive waste glasses fall within relatively narrow range.

Materials interactions relating to long-term geologic disposal of nuclear waste glass

International Nuclear Information System (INIS)

Bibler, N.E.; Jantzen, C.M.

1987-01-01

In the geologic disposal of nuclear waste glass, the glass will eventually interact with groundwater in the repository system. Interactions can also occur between the glass and other waste package materials that are present. These include the steel canister that holds the glass, the metal overpack over the canister, backfill materials that may be used, and the repository host rock. This review paper systematizes the additional interactions that materials in the waste package will impose on the borosilicate glass waste form-groundwater interactions. The repository geologies reviewed are tuff, salt, basalt, and granite. The interactions emphasized are those appropriate to conditions expected after repository closure, e.g. oxic vs anoxic conditions. Whenever possible, the effect of radiation from the waste form on the interactions is examined. The interactions are evaluated based on their effect on the release and speciation of various elements including radionuclides from the glass. It is noted when further tests of repository interactions are needed before long-term predictions can be made. 63 references, 1 table

UK program: glasses and ceramics for immobilization of radioactive wastes for disposal

International Nuclear Information System (INIS)

Johnson, K.D.B.

1979-01-01

The UK Research Program on Radioactive Waste Management includes the development of processes for the conversion of high-level-liquid-reprocessing wastes from thermal and fast reactors to borosilicate glasses. The properties of these glasses and their behavior under storage and disposal conditions have been examined. Methods for immobilizing activity from other wastes by conversion to glass or ceramic forms are described. The UK philosophy of final solutions to waste management and disposal is presented

Task plan: Temperatures in DWPF Glass Waste Storage Building

International Nuclear Information System (INIS)

Hardy, B.J.

1993-01-01

The Bechtel National, Inc. Detailed Design Instructions for Structural Design (DDI-02) requires that concrete components of the GWSB not exceed 150 degrees F for structural elements and 200 degrees F locally over a 24 hour period. In addition, the Waste Acceptance Product Specifications (WAPS) sets the maximum post cooldown temperature of the glass waste-form at 400 degrees C. Various scenarios can be postulated which result in elevated glass and concrete temperatures in the GWSB. Therefore, it is important to determine the concrete and glass temperatures during both normal and off-normal conditions. This document details specific tasks required to develop a technically defensible and verifiable methodology for determining maximum temperatures for the waste-forms and the GWSB concrete structures. All models used in this analysis will satisfy Quality Assurance requirements and be defensible to review and oversight committees

Alternate nuclear waste forms and interactions in geologic media

International Nuclear Information System (INIS)

Boatner, L.A.; Battle, G.C. Jr.

1981-04-01

The primary purposes of the conference on Alternate Nuclear Waste Forms and Interactions in Geologic Media were: First, to provide an opportunity for a review of the status of the research on some of the candidate alternative waste forms; second, to provide an opportunity for comparing the characteristics of alternate waste forms to those of glasses; and third, to stimulate increased interactions between those research groups that were engaged in a more basic approach to characterizing waste forms and those who were concerned with more applied aspects such as the processing of these materials. The motivating philosophy behind this third purpose of the conference was based on the idea that by operating from the soundest possible fundamental base for any of the candidate waste forms, hopefully any future unpleasant surprise - such as that alluded to earlier in the case of glass waste forms - could be avoided. Separate abstracts have been prepared for individual papers for inclusion in the Energy Data Base

Lead iron phosphate glass as a containment medium for disposal of high-level nuclear waste

International Nuclear Information System (INIS)

Boatner, L.A.; Sales, B.C.

1989-01-01

This patent describes lead-iron phosphate glasses containing a high level of Fe 2 O 3 for use as a storage medium for high-level radioactive nuclear waste. By combining lead-iron phosphate glass with various types of simulated high-level nuclear waste, a highly corrosion resistant, homogeneous, easily processed glass can be formed. For corroding solutions at 90 0 C, with solution pH values in the range between 5 and 9, the corrosion rate of the lead-iron phosphate nuclear waste glass is at least 10 2 to 10 3 times lower than the corrosion rate of a comparable borosilicate nuclear waste glass. The presence of Fe 2 O 3 in forming the lead-iron phosphate glass is critical. The lead-iron phosphate waste glasses can be prepared with minimal modification of the technology developed for processing borosilicate glass nuclear wasteforms

Applications of High Energy Ion Beam Techniques in Environmental Science: Investigation Associated with Glass and Ceramic Waste Forms

Energy Technology Data Exchange (ETDEWEB)

Thevuthasan, Suntharampillai; Shutthanandan, V; Zhang, Yanwen

2006-02-01

High energy ion beam capabilities including Rutherford backscattering spectrometry (RBS) and nuclear reaction analysis (NRA) have been very effectively used in environmental science to investigate the ion exchange mechanisms in glass waste forms and the effects of irradiation in glass and ceramic waste forms in the past. In this study, RBS and NRA along with SIMNRA simulations were used to monitor the Na depletion and D and 18O uptake in alumina silicate glasses, respectively, after the glass coupons were exposed to aqueous solution. These results show that the formation of a reaction layer and an establishment of a region where diffusion limited ion exchange occur in these glasses during exposure to silica-saturated solutions. Different regions including reaction and diffusion regions were identified on the basis of the depth distributions of these elements. In the case of ceramics, damage accumulation was studied as a function of ion dose at different irradiation temperatures. A sigmoidal dependence of relative disorder on the ion dose was observed. The defect dechanneling factors were calculated for two irradiated regions in SrTiO? using the critical angles determined from the angular yield curves. The dependence of defect dechanneling parameter on the incident energy was investigated and it was observed that the generated defects are mostly interstitial atoms and amorphous clusters. Thermal recovery experiments were performed to study the damage recovery processes up to a maximum temperature of 870 K.

An alternative host matrix based on iron phosphate glasses for the vitrification of specialized nuclear waste forms. Annual progress report, September 15, 1996 - September 14, 1997

International Nuclear Information System (INIS)

Day, D.E.; Marasinghe, K.; Ray, C.S.

1997-01-01

'Objectives of this project are to: (1) investigate the glass composition and processing conditions that yield optimum properties for iron phosphate glasses for vitrifying radioactive waste, (2) determine the atomic structure of iron phosphate glasses and the structure-property relationships, (3) determine how the physical and structural properties of iron phosphate glasses are affected by the addition of simulated high level nuclear waste components, and (4) investigate the process and products of devitrification of iron phosphate waste forms. The glass forming ability of about 125 iron phosphate melts has been investigated in different oxidizing to reducing atmospheres using various iron oxide raw materials such as Fe 2 O 3 , FeO, Fe 3 O 4 , and FeC 2 O 4 2H 2 O. The chemical durability, redox equilibria between Fe(II) and Fe(III), crystallization behavior and structural features for these glasses and their crystalline forms have been investigated using a variety of techniques including Mossbauer spectroscopy, X-ray absorption spectroscopy (XAS), X-ray photoelectron spectroscopy (XPS), Extended x-ray absorption fine structure (EXAFS) and X-ray absorption near edge structure (XANES) analysis, differential thermal and thermogravimetric analysis (DTA/TGA), and X-ray and neutron diffraction.'

Waste forms for plutonium disposition

International Nuclear Information System (INIS)

Johnson, S.G.; O'Holleran, T.P.; Frank, S.M.; Meyer, M.K.; Hanson, M.; Staples, B.A.; Knecht, D.A.; Kong, P.C.

1997-01-01

The field of plutonium disposition is varied and of much importance, since the Department of Energy has decided on the hybrid option for disposing of the weapons materials. This consists of either placing the Pu into mixed oxide fuel for reactors or placing the material into a stable waste form such as glass. The waste form used for Pu disposition should exhibit certain qualities: (1) provide for a suitable deterrent to guard against proliferation; (2) be of minimal volume, i.e., maximize the loading; and (3) be reasonably durable under repository-like conditions. This paper will discuss several Pu waste forms that display promising characteristics

Hydrogen speciation in hydrated layers on nuclear waste glass

International Nuclear Information System (INIS)

Aines, R.D.; Weed, H.C.; Bates, J.K.

1987-01-01

The hydration of an outer layer on nuclear waste glasses is known to occur during leaching, but the actual speciation of hydrogen (as water or hydroxyl groups) in these layers has not been determined. As part of the Nevada Nuclear Waste Storage Investigations Project, we have used infrared spectroscopy to determine hydrogen speciations in three nuclear waste glass compositions (SRL-131 and 165, and PNL 76-68), which were leached at 90 0 C (all glasses) or hydrated in a vapor-saturated atmosphere at 202 0 C (SRL-131 only). Hydroxyl groups were found in the surface layers of all the glasses. Molecular water was found in the surface of SRL-131 and PNL 76-68 glasses that had been leached for several months in deionized water, and in the vapor-hydrated sample. The water/hydroxyl ratio increases with increasing reaction time; molecular water makes up most of the hydrogen in the thick reaction layers on vapor-phase hydrated glass while only hydroxyl occurs in the least reacted samples. Using the known molar absorptivities of water and hydroxyl in silica-rich glass the vapor-phase layer contained 4.8 moles/liter of molecular water, and 0.6 moles water in the form hydroxyl. A 15 μm layer on SRL-131 glass formed by leaching at 90 0 C contained a total of 4.9 moles/liter of water, 2/3 of which was as hydroxyl. The unreacted bulk glass contains about 0.018 moles/liter water, all as hydroxyl. The amount of hydrogen added to the SRL-131 glass was about 70% of the original Na + Li content, not the 300% that would result from alkali=hydronium ion interdiffusion. If all the hydrogen is then assumed to be added as the result of alkali-H + interdiffusion, the molecular water observed may have formed from condensation of the original hydroxyl groups

Alternative waste form development - low-temperature pyrolytic carbon coatings

International Nuclear Information System (INIS)

Oma, K.H.; Rusin, J.M.; Kidd, R.W.; Browning, M.F.

1981-01-01

Although several chemical vapor deposition (CVD) - coated waste forms have been successfully produced, some major disadvantages associated with the high-temperature fluidized-bed CVD coating process exist. To overcome these disadvantages, the Pacific Northwest Laboratory has initiated the development of a pyrolytic carbon CVD coating system to coat large waste-form particles at temperatures ranging from 400 to 500/degree/C. This relatively simple system has been used to coat kilogram quantities of simulated waste-glass marbles. Further development of this system could result in a viable process to coat bulk quantities of both glass and ceramic waste forms. This paper discusses various aspects of the development work, including coating techniques, parametric study, and coater equipment. 10 refs

Low-risk alternative waste forms for problematic high-level and long-lived nuclear wastes

International Nuclear Information System (INIS)

Stewart, M.W.A.; Begg, B.D.; Moricca, S.; Day, R.A.

2006-01-01

Full text: The highest cost component the nuclear waste clean up challenge centres on high-level waste (HLW) and consequently the greatest opportunity for cost and schedule savings lies with optimising the approach to HLW cleanup. The waste form is the key component of the immobilisation process. To achieve maximum cost savings and optimum performance the selection of the waste form should be driven by the characteristics of the specific nuclear waste to be immobilised, rather than adopting a single baseline approach. This is particularly true for problematic nuclear wastes that are often not amenable to a single baseline approach. The use of tailored, high-performance, alternative waste forms that include ceramics and glass-ceramics, coupled with mature process technologies offer significant performance improvements and efficiency savings for a nuclear waste cleanup program. It is the waste form that determines how well the waste is locked up (chemical durability), and the number of repository disposal canisters required (waste loading efficiency). The use of alternative waste forms for problematic wastes also lowers the overall risk by providing high performance HLW treatment alternatives. The benefits tailored alternative waste forms bring to the HLW cleanup program will be briefly reviewed with reference to work carried out on the following: The HLW calcines at the Idaho National Laboratory; SYNROC ANSTO has developed a process utilising a glass-ceramic combined with mature hot-isostatic pressing (HIP) technology and has demonstrated this at a waste loading of 80 % and at a 30 kg HIP scale. The use of this technology has recently been estimated to result in a 70 % reduction in waste canisters, compared to the baseline borosilicate glass technology; Actinide-rich waste streams, particularly the work being done by SYNROC ANSTO with Nexia Solutions on the Plutonium-residues wastes at Sellafield in the UK, which if implemented is forecast to result in substantial

Molecular environmental science using synchrotron radiation: Chemistry and physics of waste form materials

International Nuclear Information System (INIS)

Lindle, Dennis W.; Shuh, David K.

2005-01-01

Production of defense-related nuclear materials has generated large volumes of complex chemical wastes containing a mixture of radionuclides. The disposition of these wastes requires conversion of the liquid and solid-phase components into durable, solid forms suitable for long-term immobilization [1]. Specially formulated glass compositions, many of which have been derived from glass developed for commercial purposes, and ceramics such as pyrochlores and apatites, will be the main recipients for these wastes. The performance characteristics of waste-form glasses and ceramics are largely determined by the loading capacity for the waste constituents (radioactive and non-radioactive) and the resultant chemical and radiation resistance of the waste-form package to leaching (durability). There are unique opportunities for the use of near-edge soft-x-ray absorption fine structure (NEXAFS) spectroscopy to investigate speciation of low-Z elements forming the backbone of waste-form glasses and ceramics. Although nuclear magnetic resonance (NMR) is the primary technique employed to obtain speciation information from low-Z elements in waste forms, NMR is incompatible with the metallic impurities contained in real waste and is thus limited to studies of idealized model systems. In contrast, NEXAFS can yield element-specific speciation information from glass constituents without sensitivity to paramagnetic species. Development and use of NEXAFS for eventual studies of real waste glasses has significant implications, especially for the low-Z elements comprising glass matrices [5-7]. The NEXAFS measurements were performed at Beamline 6.3.1, an entrance-slitless bend-magnet beamline operating from 200 eV to 2000 eV with a Hettrick-Underwood varied-line-space (VLS) grating monochromator, of the Advanced Light Source (ALS) at LBNL. Complete characterization and optimization of this beamline was conducted to enable high-performance measurements

Molecular environmental science using synchrotron radiation:Chemistry and physics of waste form materials

Energy Technology Data Exchange (ETDEWEB)

Lindle, Dennis W.; Shuh, David K.

2005-02-28

Production of defense-related nuclear materials has generated large volumes of complex chemical wastes containing a mixture of radionuclides. The disposition of these wastes requires conversion of the liquid and solid-phase components into durable, solid forms suitable for long-term immobilization [1]. Specially formulated glass compositions, many of which have been derived from glass developed for commercial purposes, and ceramics such as pyrochlores and apatites, will be the main recipients for these wastes. The performance characteristics of waste-form glasses and ceramics are largely determined by the loading capacity for the waste constituents (radioactive and non-radioactive) and the resultant chemical and radiation resistance of the waste-form package to leaching (durability). There are unique opportunities for the use of near-edge soft-x-ray absorption fine structure (NEXAFS) spectroscopy to investigate speciation of low-Z elements forming the backbone of waste-form glasses and ceramics. Although nuclear magnetic resonance (NMR) is the primary technique employed to obtain speciation information from low-Z elements in waste forms, NMR is incompatible with the metallic impurities contained in real waste and is thus limited to studies of idealized model systems. In contrast, NEXAFS can yield element-specific speciation information from glass constituents without sensitivity to paramagnetic species. Development and use of NEXAFS for eventual studies of real waste glasses has significant implications, especially for the low-Z elements comprising glass matrices [5-7]. The NEXAFS measurements were performed at Beamline 6.3.1, an entrance-slitless bend-magnet beamline operating from 200 eV to 2000 eV with a Hettrick-Underwood varied-line-space (VLS) grating monochromator, of the Advanced Light Source (ALS) at LBNL. Complete characterization and optimization of this beamline was conducted to enable high-performance measurements.

Waste glass melting stages

International Nuclear Information System (INIS)

Anderson, L.D.; Dennis, T.; Elliott, M.L.; Hrma, P.

1993-04-01

Three different simulated nuclear waste glass feeds, consisting of dried waste and glass frit, were heat treated for 1 hour in a gradient furnace at temperatures ranging from approximately 600 degrees C--1000 degrees C. Simulated melter feeds from the Hanford Waste Vitrification Plant (HWVP), the Defense Waste Processing Facility (DWPF), and Kernforschungszentrum Karlsruhe (KfK) in Germany were used. The samples were thin-sectioned and examined by optical microscopy to investigate the stages of the conversion from feed to glass. Various phenomena were seen, such as frit softening, bubble formation, foaming, bubble motion and removal, convective mixing, and homogenization. Behavior of different feeds was similar, although the degree of gas generation and melt homogenization varied

Equilibrium Temperature Profiles within Fission Product Waste Forms

Energy Technology Data Exchange (ETDEWEB)

Kaminski, Michael D. [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-10-01

We studied waste form strategies for advanced fuel cycle schemes. Several options were considered for three waste streams with the following fission products: cesium and strontium, transition metals, and lanthanides. These three waste streams may be combined or disposed separately. The decay of several isotopes will generate heat that must be accommodated by the waste form, and this heat will affect the waste loadings. To help make an informed decision on the best option, we present computational data on the equilibrium temperature of glass waste forms containing a combination of these three streams.

MIIT: International in-situ testing of nuclear-waste glasses: Performance of SRS simulated waste glass after five years of burial at the Waste Isolation Pilot Plant (WIPP)

International Nuclear Information System (INIS)

Wicks, G.G.; Lodding, A.R.; Macedo, P.B.; Clark, D.E.

1991-01-01

In July of 1986, the first in-situ test involving burial of simulated high-level waste (HLW) forms conducted in the United States was started. This program, called the Materials Interface Interactions Test or MIIT, comprises the largest, most cooperative field-testing venture in the international waste management community. In July of 1991, the experimental portion of the 5-year MIIT study was completed on schedule. During this time interval, many in-situ measurements were performed, thousands of brine analyses conducted, and hundreds of waste glass and package components exhumed and evaluated after 6 mo., 1 yr., 2 yr. and 5 yr. burial periods. Although analyses are still in progress, the performance of SRS waste glass based on all data currently available has been seen to be excellent thus far. Initial analyses and assessment of Savannah River (SR) waste glass after burial in WIPP at 90 degrees C for 5 years are presented in this document

Nuclear waste glass corrosion mechanisms

International Nuclear Information System (INIS)

Jantzen, C.M.

1987-04-01

Dissolution of nuclear waste glass occurs by corrosion mechanisms similar to those of other solids, e.g., metallurgical and mineralogic systems. Metallurgical phenomena such as active corrosion, passivation and immunity have been observed to be a function of the glass composition and the solution pH. Hydration thermodynamics was used to quantify the role of glass composition and its effect on the solution pH during dissolution. A wide compositional range of natural, lunar, medieval, and nuclear waste glasses, as well as some glass-ceramics were investigated. The factors observed to affect dissolution in deionized water are pertinent to the dissolution of glass in natural environments such as the groundwaters anticipated to interact with nuclear waste glass in a geologic repository. The effects of imposed pH and oxidation potential (Eh) conditions existing in natural environments on glass dissolution is described in the context of Pourbaix diagrams, pH potential diagrams, for glass

Glass-ceramics with multibarrier structure obtained from industrial waste

Energy Technology Data Exchange (ETDEWEB)

Berzina, L.; Cimdins, R.; Rozenstrauha, I. [Riga Tech. Univ. (Latvia). Fac. of Chem. Technol.; Bossert, J. [Technisches Inst.: Materialwissenschaft, Friedrich-Schiller-Univ., Jena (Germany); Kravtchenko, I. [Inst. for Problems of Material Science, Kiev (Ukraine)

1997-12-31

Recycling problem for various kind of waste is solved by processing the waste to ecological depositable products with multibarrier structure. In order to form a multibarrier structure the ecologically incompatible substances may be diluted and chemically bound until their recycling products gain a structure like natural mineral or glass (I. barrier). After that, remineralized materials are converted into a new product by melting or powder technology using an ecological compatible type of waste as a matrix phase (II. barrier). Waste which are treated this way could be applied to produce ceramic building materials and goods such as floor tiles, stone pavement and casting products. Industrial waste from the metallurgical factory in Latvia ``Liepajas metalurgs`` are metallurgical slag, filter dust, etching waste and sewage used in technologies. The main constituents of chemical compositions of these waste are: Fe, Ca, Si, Mg, Al, Mn etc. In some types of waste a small amount of ecologically risky elements such as Cr, Ni, Zr, Sn and Pb can occur. The combination of metallurgical waste with peat ashes from Riga thermal power station, oil shale ashes or glass waste under controlled sintering procedure gives bulk materials with surface or/and bulkcrystallization. The structure of glass-ceramics built this way may prevent the migration of ecologically risky elements into environment due to corrosion or friction. Physical-chemical properties and thermal behaviour (DTA, dilatometry, melting) of waste define the range of sintering for production of glass-ceramics (powder technology) and decorative glass-ceramic materials (melting and powder technology). (orig.) 5 refs.

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

International Nuclear Information System (INIS)

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

1980-12-01

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

Secondary Waste Form Down-Selection Data Package—Fluidized Bed Steam Reforming Waste Form

Energy Technology Data Exchange (ETDEWEB)

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

2011-09-12

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

Waste glass melting stages

International Nuclear Information System (INIS)

Anderson, L.D.; Dennis, T.; Elliott, M.L.; Hrma, P.

1994-01-01

Three simulated nuclear waste glass feeds, consisting of dried waste and glass frit, were heat treated for 1 hour in a gradient furnace at temperatures ranging from approximately 600 degrees C to 1000 degrees C. Simulated melter feeds from the Hanford Waste Vitrification Plant (HWVP), the Defense Waste Processing Facility (DWPF), and Kernforschungszentru Karlsruhe (KfK) in Germany were used. The samples were thin sectioned and examined by optical microscopy to investigate the stages of the conversion from feed to glass. Various phenomena were seen, such as frit softening, bubble formation, foaming, bubble motion and removal, convective mixing, and homogenization. The behavior of different feeds was similar, although the degree of gas generation and melt homogenization varied. 2 refs., 8 tabs

Strontium chloroapatite based glass-ceramics composites for nuclear waste immobilisation

International Nuclear Information System (INIS)

Jena, Hrudananda; Maji, Binoy Kumar; Asuvathraman, R.; Govindan Kutty, K.V.

2013-01-01

Apatites are naturally occurring minerals with a general formula of M 10 (PO 4 ) 6 X 2 , (M= Ca, Sr, Ba, X= OH, Cl, F) with a hexagonal crystal structure (S.G :P6 3 /m) and can accommodate alkaline earth and various other aliovalent cations and anions into its crystal structure. Apatites are also known to have high resistance to leaching of the constituent elements under geological conditions. It may not often be possible to immobilize the whole spectrum of the radioactive waste in a single phase M 10 (PO 4 ) 6 Cl 2 , then a combination of M-chloroapatite encapsulated in borosilicate glass (BSG) can immobilize most of the radwaste elements in the composite glass-ceramic matrix (glass bonded chloroapatite), thus utilizing the immobilizing efficiency of both the ceramic phase and glass. In the present study, the synthesis, characterization and thermo-physical property measurements of the Sr-chloroapatite (SrApCI) and some glass-bonded composites based on it have been investigated. The Sr-chloroapatite glass-ceramics were prepared by solid state reactions among stoichiometric concentrations of apatite forming reagents, 20 wt. % borosilicate glass (BSG), and known concentrations (10, 13 and 16 wt. %) of a simulated waste in chloride form. The products were characterized by XRD to confirm the formation of Sr 10 (PO 4 ) 6 Cl 2 and glass bonded-chloroapatite composites. The surface morphology and qualitative chemical composition of the powders were examined by SEM and EDX. Thermal expansion and glass transition temperature of the matrices were measured by dilatometry. Glass transition temperature of the glass-bonded composites was also examined by differential scanning calorimetry and differential thermal analysis. The 10-16 wt.% waste loaded matrices showed similar thermal expansion as that of SrApCI, indicating the thermal stability of the matrix to chloride waste immobilization. The glass transition temperature of the waste loaded matrices decreases on increasing the

Effect of different glasses in glass bonded zeolite

International Nuclear Information System (INIS)

Lewis, M.A.; Ackerman, J.P.; Verma, S.

1995-01-01

A mineral waste form has been developed for chloride waste salt generated during the pyrochemical treatment of spent nuclear fuel. The waste form consists of salt-occluded zeolite powders bound within a glass matrix. The zeolite contains the salt and immobilizes the fission products. The zeolite powders are hot pressed to form a mechanically stable, durable glass bonded zeolite. Further development of glass bonded zeolite as a waste form requires an understanding of the interaction between the glass and the zeolite. Properties of the glass that enhance binding and durability of the glass bonded zeolite need to be identified. Three types of glass, boroaluminosilicate, soda-lime silicate, and high silica glasses, have a range of properties and are now being investigated. Each glass was hot pressed by itself and with an equal amount of zeolite. MCC-1 leach tests were run on both. Soda-lime silicate and high silica glasses did not give a durable glass bonded zeolite. Boroaluminosilicate glasses rich in alkaline earths did bind the zeolite and gave a durable glass bonded zeolite. Scanning electron micrographs suggest that the boroaluminosilicate glasses wetted the zeolite powders better than the other glasses. Development of the glass bonded zeolite as a waste form for chloride waste salt is continuing

High-level waste glass compendium; what it tells us concerning the durability of borosilicate waste glass

International Nuclear Information System (INIS)

Cunnane, J.C.; Allison, J.

1993-01-01

Facilities for vitrification of high-level nuclear waste in the United States are scheduled for startup in the next few years. It is, therefore, appropriate to examine the current scientific basis for understanding the corrosion of high-level waste borosilicate glass for the range of service conditions to which the glass products from these facilities may be exposed. To this end, a document has been prepared which compiles worldwide information on borosilicate waste glass corrosion. Based on the content of this document, the acceptability of canistered waste glass for geological disposal is addressed. Waste glass corrosion in a geologic repository may be due to groundwater and/or water vapor contact. The important processes that determine the glass corrosion kinetics under these conditions are discussed based on experimental evidence from laboratory testing. Testing data together with understanding of the long-term corrosion kinetics are used to estimate radionuclide release rates. These rates are discussed in terms of regulatory performance standards

WTP Waste Feed Qualification: Glass Fabrication Unit Operation Testing Report

Energy Technology Data Exchange (ETDEWEB)

Stone, M. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL). Hanford Missions Programs; Newell, J. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL). Process Technology Programs; Johnson, F. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL). Engineering Process Development; Edwards, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL). Engineering Process Development

2016-07-14

The waste feed qualification program is being developed to protect the Hanford Tank Waste Treatment and Immobilization Plant (WTP) design, safety basis, and technical basis by assuring waste acceptance requirements are met for each staged waste feed campaign prior to transfer from the Tank Operations Contractor to the feed receipt vessels inside the Pretreatment Facility. The Waste Feed Qualification Program Plan describes the three components of waste feed qualification: 1. Demonstrate compliance with the waste acceptance criteria 2. Determine waste processability 3. Test unit operations at laboratory scale. The glass fabrication unit operation is the final step in the process demonstration portion of the waste feed qualification process. This unit operation generally consists of combining each of the waste feed streams (high-level waste (HLW) and low-activity waste (LAW)) with Glass Forming Chemicals (GFCs), fabricating glass coupons, performing chemical composition analysis before and after glass fabrication, measuring hydrogen generation rate either before or after glass former addition, measuring rheological properties before and after glass former addition, and visual observation of the resulting glass coupons. Critical aspects of this unit operation are mixing and sampling of the waste and melter feeds to ensure representative samples are obtained as well as ensuring the fabrication process for the glass coupon is adequate. Testing was performed using a range of simulants (LAW and HLW simulants), and these simulants were mixed with high and low bounding amounts of GFCs to evaluate the mixing, sampling, and glass preparation steps in shielded cells using laboratory techniques. The tests were performed with off-the-shelf equipment at the Savannah River National Laboratory (SRNL) that is similar to equipment used in the SRNL work during qualification of waste feed for the Defense Waste Processing Facility (DWPF) and other waste treatment facilities at the

Natural analogues of nuclear waste glass corrosion

International Nuclear Information System (INIS)

Abrajano, T.A. Jr.; Ebert, W.L.; Luo, J.S.

1999-01-01

This report reviews and summarizes studies performed to characterize the products and processes involved in the corrosion of natural glasses. Studies are also reviewed and evaluated on how well the corrosion of natural glasses in natural environments serves as an analogue for the corrosion of high-level radioactive waste glasses in an engineered geologic disposal system. A wide range of natural and experimental corrosion studies has been performed on three major groups of natural glasses: tektite, obsidian, and basalt. Studies of the corrosion of natural glass attempt to characterize both the nature of alteration products and the reaction kinetics. Information available on natural glass was then compared to corresponding information on the corrosion of nuclear waste glasses, specifically to resolve two key questions: (1) whether one or more natural glasses behave similarly to nuclear waste glasses in laboratory tests, and (2) how these similarities can be used to support projections of the long-term corrosion of nuclear waste glasses. The corrosion behavior of basaltic glasses was most similar to that of nuclear waste glasses, but the corrosion of tektite and obsidian glasses involves certain processes that also occur during the corrosion of nuclear waste glasses. The reactions and processes that control basalt glass dissolution are similar to those that are important in nuclear waste glass dissolution. The key reaction of the overall corrosion mechanism is network hydrolysis, which eventually breaks down the glass network structure that remains after the initial ion-exchange and diffusion processes. This review also highlights some unresolved issues related to the application of an analogue approach to predicting long-term behavior of nuclear waste glass corrosion, such as discrepancies between experimental and field-based estimates of kinetic parameters for basaltic glasses

Natural analogues of nuclear waste glass corrosion.

Energy Technology Data Exchange (ETDEWEB)

Abrajano, T.A. Jr.; Ebert, W.L.; Luo, J.S.

1999-01-06

This report reviews and summarizes studies performed to characterize the products and processes involved in the corrosion of natural glasses. Studies are also reviewed and evaluated on how well the corrosion of natural glasses in natural environments serves as an analogue for the corrosion of high-level radioactive waste glasses in an engineered geologic disposal system. A wide range of natural and experimental corrosion studies has been performed on three major groups of natural glasses: tektite, obsidian, and basalt. Studies of the corrosion of natural glass attempt to characterize both the nature of alteration products and the reaction kinetics. Information available on natural glass was then compared to corresponding information on the corrosion of nuclear waste glasses, specifically to resolve two key questions: (1) whether one or more natural glasses behave similarly to nuclear waste glasses in laboratory tests, and (2) how these similarities can be used to support projections of the long-term corrosion of nuclear waste glasses. The corrosion behavior of basaltic glasses was most similar to that of nuclear waste glasses, but the corrosion of tektite and obsidian glasses involves certain processes that also occur during the corrosion of nuclear waste glasses. The reactions and processes that control basalt glass dissolution are similar to those that are important in nuclear waste glass dissolution. The key reaction of the overall corrosion mechanism is network hydrolysis, which eventually breaks down the glass network structure that remains after the initial ion-exchange and diffusion processes. This review also highlights some unresolved issues related to the application of an analogue approach to predicting long-term behavior of nuclear waste glass corrosion, such as discrepancies between experimental and field-based estimates of kinetic parameters for basaltic glasses.

Description of a ceramic waste form and canister for Savannah River Plant high-level waste

International Nuclear Information System (INIS)

Butler, J.L.; Allender, J.S.; Gould, T.H. Jr.

1982-04-01

A canistered ceramic waste form for possible immobilization of Savannah River Plant (SRP) high-level radioactive wastes is described. Characteristics reported for the form include waste loading, chemical composition, heat content, isotope inventory, mechanical and thermal properties, and leach rates. A conceptual design of a potential production process for making this canistered form are also described. The ceramic form was selected in November 1981 as the primary alternative to the reference waste form, borosilicate glass, for making a final waste form decision for SRP waste by FY-1983. 11 tables

Characteristics of high-level radioactive waste forms for their disposal

International Nuclear Information System (INIS)

Kim, Seung Soo; Chun, Kwan Sik; Kang, Chul Hyung

2000-12-01

In order to develop a deep geological repository for a high-level radioactive waste coming from reprocessing of spent nuclear fuels discharged from our domestic nuclear power plants, the the required characteristics of waste form are dependent upon a solidifying medium and the amount of waste loading in the medium. And so, by the comparative analysis of the characteristics of various waste forms developed up to the present, a suitable medium is recommended.The overall characteristics of the latter is much better than those of the former, but the change of the properties due to an amorphysation by radiation exposure and its thermal expansion has not been clearly identified yet. And its process has not been commercialized. However, the overall properties of the borosilicate glass waste forms are acceptable for their disposal, their production cost is reasonable and their processes have already been commercialized. And plenty informations of their characteristics and operational experiences have been accumulated. Consequently, it is recommended that a suitable medium solidifying the HLW is a borosilicate glass and its composition for the identification of a reference waste form would be based on the glass frit of R7T7

Mixed low-level waste form evaluation

International Nuclear Information System (INIS)

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

1997-01-01

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

Settling of Spinel in A High-Level Waste Glass Melter

International Nuclear Information System (INIS)

Pavel Hrma; Pert Schill; Lubomir Nemec

2002-01-01

High-level nuclear waste is being vitrified, i.e., converted to a durable glass that can be stored in a safe repository for hundreds of thousands of years. Waste vitrification is accomplished in reactors call melters to which the waste is charged together with glass-forming additives. The mixture is electrically heated to a temperature as high as 1150 degree C (or even higher in advanced melters) to create a melt that becomes glass on cooling. This process is slow and expensive. Moreover, the melters that are currently in use or are going to be used in the U.S. are sensitive to clogging and thus cannot process melt in which solid particles are suspended. These particles settle and gradually accumulate on the melter bottom. Such particles, most often small crystals of spinel ( a mineral containing iron, nickel, chromium, and other minor oxides), inevitably occurred in the melt when the content of the waste in the glass (called waste loading) increases above a certain limit. To avoid the presence of solid particles in the melter, the waste loading is kept rather low, in average 15% lower than in glass formulated for more robust melters

Direct conversion of plutonium metal, scrap, residue, and transuranic waste to glass

International Nuclear Information System (INIS)

Forsberg, C.W.; Beahm, E.C.; Parker, G.W.; Malling, J.F.; Rudolph, J.

1995-01-01

A method for the direct conversion of metals, ceramics, organics, and amorphous solids to borosilicate glass has been invented. The process is called the Glass Material Oxidation and Dissolution System (GMODS). Traditional glass-making processes can convert only oxide materials to glass. However, many wastes contain complex mixtures of metals, ceramics, organics, and amorphous solids. Conversion of such mixtures to oxides followed by their conversion to glass is often impractical. GMODS may create a practical method to convert such mixtures to glass. Plutonium-containing materials (PCMS) exist in many forms, including metals, ceramics, organics, amorphous solids, and mixtures thereof. These PCMs vary from plutonium metal to filters made of metal, organic binders, and glass fibers. For storage and/or disposal of PCMS, it is desirable to convert PCMs to borosilicate glass. Borosilicate glass is the preferred repository waste form for high-level waste (HLW) because of its properties. PCMs converted to a transuranic borosilicate homogeneous glass would easily pass all waste acceptance and storage criteria. Conversion of PCMs to a glass would also simplify safeguards by conversion of heterogeneous PCMs to homogeneous glass. Thermodynamic calculations and proof-of-principle experiments on the GMODS process with cerium (plutonium surrogate), uranium, stainless steel, aluminum, Zircaloy-2, and carbon were successfully conducted. Initial analysis has identified potential flowsheets and equipment. Major unknowns remain, but the preliminary data suggests that GMODS may be a major new treatment option for PCMs

High-silica glass matrix process for high-level waste solidification

International Nuclear Information System (INIS)

Simmons, J.H.; Macedo, P.B.

1981-01-01

In the search for an optimum glass matrix composition, we have determined that chemical durability and thermal stability are maximized, and that stress development is minimized for glass compositions containing large concentrations of glass-forming oxides, of which silica is the major component (80 mol%). These properties and characteristics were recently demonstrated to belong to very old geological glasses known as tektites (ages of 750,000 to 34 million years.) The barrier to simulating tektite compositions for the waste glasses was the high melting temperature (1600 to 1800 0 C) needed for these glasses. Such temperatures greatly complicate furnace design and maintenance and lead to an intolerable vaporization of many of the radioisotopes into the off-gas system. Research conducted at our laboratory led to the development of a porous high-silica waste glass material with approximately 80% SiO 2 by mole and 30% waste loading by weight. The process can handle a wide variety of compositions, and yields long, elliptical, monolithic samples, which consist of a loaded high-silica core completely enveloped in a high-silica glass tube, which has collapsed upon the core and sealed it from the outside. The outer glass layer is totally free of waste isotopes and provides an integral multibarrier protection system

Spinel dissolution via addition of glass forming chemicals. Results of preliminary experiments

Energy Technology Data Exchange (ETDEWEB)

Fox, K. M. [Savannah River Site (SRS), Aiken, SC (United States); Johnson, F. C. [Savannah River Site (SRS), Aiken, SC (United States)

2015-11-01

Increased loading of high level waste in glass can lead to crystallization within the glass. Some crystalline species, such as spinel, have no practical impact on the chemical durability of the glass, and therefore may be acceptable from both a processing and a product performance standpoint. In order to operate a melter with a controlled amount of crystallization, options must be developed for remediating an unacceptable accumulation of crystals. This report describes preliminary experiments designed to evaluate the ability to dissolve spinel crystals in simulated waste glass melts via the addition of glass forming chemicals (GFCs).

The role of ceramics, cement and glass in the immobilization of radioactive wastes

International Nuclear Information System (INIS)

Glasser, F.P.

1985-01-01

A brief account is given of the constitution and origin of nuclear waste. The immobilization of wastes is discussed: borosilicate glasses are considered as possible matrices; ceramic forms are dealt with in more detail. The principles of the use of ceramics are explained, with examples of different ceramic structures; cements are mentioned as being suitable for wet, medium- to low-active wastes. The effects of radiation on cement, ceramic and glass waste forms are indicated. The account concludes with 'summary and future progress'. (U.K.)

Immobilization of radioactive wastes in glasses and ceramics

International Nuclear Information System (INIS)

Zanotto, E.D.

1983-01-01

A large amount of radioactive liquid wastes arises from the reprocessing of spent nuclear fuels to recover uranium and plutonium. Immobilization of such wastes in solid form and disposal of the solidified wastes in safe places, to prevent contamination of the human environment, are topics of considerable interest for both the scientific community and the public in general. The great majority of materials candidate for the encapsulation of radioactive wastes are inorganic non-metalic, such as glasses, glass-ceramics, special cements, calcined ceramics and few more. Among these materials, certain glasses have received special attention, and are being studied for over twenty years. It is estimated that about US$2 billion have already been spent in these studies. The disposal (long term storage) of these solid wastes may be possible in deep geological formations, salt mines, the ocean bed, by evacuation to the outer space, etc. A brief review on the several options avaiable for encapsulation and disposal of high level radioactive liquid wastes is presented, along with the relative merits and disadvantages of the candidate materials for encapsulation. A few suggestions for the solution of the Brazilian problem are advanced. (Author) [pt

Molecular Environmental Science Using Synchrotron Radiation: Chemistry and Physics of Waste Form Materials

Energy Technology Data Exchange (ETDEWEB)

Lindle, Dennis W.

2011-04-21

Production of defense-related nuclear materials has generated large volumes of complex chemical wastes containing a mixture of radionuclides. The disposition of these wastes requires conversion of the liquid and solid-phase components into durable, solid forms suitable for long-term immobilization. Specially formulated glass compositions and ceramics such as pyrochlores and apatites are the main candidates for these wastes. An important consideration linked to the durability of waste-form materials is the local structure around the waste components. Equally important is the local structure of constituents of the glass and ceramic host matrix. Knowledge of the structure in the waste-form host matrices is essential, prior to and subsequent to waste incorporation, to evaluate and develop improved waste-form compositions based on scientific considerations. This project used the soft-x-ray synchrotron-radiation-based technique of near-edge x-ray-absorption fine structure (NEXAFS) as a unique method for investigating oxidation states and structures of low-Z elemental constituents forming the backbones of glass and ceramic host matrices for waste-form materials. In addition, light metal ions in ceramic hosts, such as titanium, are also ideal for investigation by NEXAFS in the soft-x-ray region. Thus, one of the main objectives was to understand outstanding issues in waste-form science via NEXAFS investigations and to translate this understanding into better waste-form materials, followed by eventual capability to investigate “real” waste-form materials by the same methodology. We conducted several detailed structural investigations of both pyrochlore ceramic and borosilicate-glass materials during the project and developed improved capabilities at Beamline 6.3.1 of the Advanced Light Source (ALS) to perform the studies.

Lead iron phosphate glass as a containment medium for disposal of high-level nuclear waste

Science.gov (United States)

Boatner, Lynn A.; Sales, Brian C.

1989-01-01

Lead-iron phosphate glasses containing a high level of Fe.sub.2 O.sub.3 for use as a storage medium for high-level radioactive nuclear waste. By combining lead-iron phosphate glass with various types of simulated high-level nuclear waste, a highly corrosion resistant, homogeneous, easily processed glass can be formed. For corroding solutions at 90.degree. C., with solution pH values in the range between 5 and 9, the corrosion rate of the lead-iron phosphate nuclear waste glass is at least 10.sup.2 to 10.sup.3 times lower than the corrosion rate of a comparable borosilicate nuclear waste glass. The presence of Fe.sub.2 O.sub.3 in forming the lead-iron phosphate glass is critical. Lead-iron phosphate nuclear waste glass can be prepared at temperatures as low as 800.degree. C., since they exhibit very low melt viscosities in the 800.degree. to 1050.degree. C. temperature range. These waste-loaded glasses do not readily devitrify at temperatures as high as 550.degree. C. and are not adversely affected by large doses of gamma radiation in H.sub.2 O at 135.degree. C. The lead-iron phosphate waste glasses can be prepared with minimal modification of the technology developed for processing borosilicate glass nuclear wasteforms.

Glass-bonded iodosodalite waste form for immobilization of 129 I

Energy Technology Data Exchange (ETDEWEB)

Chong, Saehwa; Peterson, Jacob A.; Riley, Brian J.; Tabada, Diana; Wall, Donald; Corkhill, Claire L.; McCloy, John S.

2018-06-01

Immobilization of radioiodine (e.g., 129I, 131I) is an important need for current and future nuclear fuel cycles. For the current work, iodosodalite [Na8(AlSiO4)6I2] was synthesized hydrothermally from metakaolin, NaI, and NaOH. Following hydrothermal treatment, dried unwashed powders were used to make glass-bonded iodosodalite waste forms by heating pressed pellets at 650, 750, or 850 °C with two different types of sodium borosilicate glass binders, i.e., NBS-4 and SA-800. These heat-treated specimens were characterized with X-ray diffraction, Fourier-transform infrared spectroscopy, scanning electron microscopy, energy dispersive spectroscopy, thermal analysis, porosity and density measurements, neutron activation analysis, and inductively-coupled plasma mass spectrometry. The pellets mixed with 10 mass% of NBS-4 or SA-800 and heat-treated at 750 °C contained relatively high percentage iodine retention (~44-47 % of the maximum iodine loading) with relatively low porosities, while other pellets with higher percentages iodine retention either contained higher porosity or were not completely sintered. ASTM C1308 chemical durability tests of monolithic specimens showed a large initial release of Na, Al, Si, and I on the first day, possibly from water-soluble salt crystals or non-durable amorphous phases. Release rates of Na and Si were higher than for Al and I, probably due to a poorly durable Na-Si-O phase from the glass bonding matrix. The cumulative normalized release of iodine was 12.5 g m-2 for the first 10 1-d exchanges, suggestive of coherent dissolution. The average release rate from 10-24 days during the 7-d exchange intervals was 0.2336 g m-2 d-1.

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

International Nuclear Information System (INIS)

Jantzen, C.

2008-01-01

The Savannah River National Laboratory (SRNL) was requested to generate a document for the Washington State Department of Ecology and the U.S. Environmental Protection Agency that would cover the following topics: (1) A description of the mineral structures produced by Fluidized Bed Steam Reforming (FBSR) of Hanford type Low Activity Waste (LAW including LAWR which is LAW melter recycle waste) waste, especially the cage structured minerals and how they are formed. (2) How the cage structured minerals contain some contaminants, while others become part of the mineral structure (Note that all contaminants become part of the mineral structure and this will be described in the subsequent sections of this report). (3) Possible contaminant release mechanisms from the mineral structures. (4) Appropriate analyses to evaluate these release mechanisms. (5) Why the appropriate analyses are comparable to the existing Hanford glass dataset. In order to discuss the mineral structures and how they bond contaminants a brief description of the structures of both mineral (ceramic) and vitreous waste forms will be given to show their similarities. By demonstrating the similarities of mineral and vitreous waste forms on atomic level, the contaminant release mechanisms of the crystalline (mineral) and amorphous (glass) waste forms can be compared. This will then logically lead to the discussion of why many of the analyses used to evaluate vitreous waste forms and glass-ceramics (also known as glass composite materials) are appropriate for determining the release mechanisms of LAW/LAWR mineral waste forms and how the durability data on LAW/LAWR mineral waste forms relate to the durability data for LAW/LAWR glasses. The text will discuss the LAW mineral waste form made by FBSR. The nanoscale mechanism by which the minerals form will be also be described in the text. The appropriate analyses to evaluate contaminant release mechanisms will be discussed, as will the FBSR test results to

Molecular glasses for nuclear waste encapsulation

International Nuclear Information System (INIS)

Ropp, R.C.

1982-01-01

The use of a molecular glass based upon a polymerized phosphate of aluminum (PAP), indium or gallium overcomes all of the prior objections to use of glass as a high-level nuclear waste (HLW) encapsulation agent. This HLW glass product could not be made to devitrify, dissolved all of the oxides found in calcine, including the difficultly soluble ones, did not form microcrystallites in the melt or subsequent glass-casting, and possessed a hydrolytic etching rate to boiling water even lower than that of HLW-ZBS glass. A precursor compound, M(H 2 PO 4 ) 3 , is prepared, where M is a trivalent metal selected from the group consisting of aluminum, indium and gallium. The impurity level is carefully controlled so as not to exceed 300 ppm total. The precursor crystals may be washed to remove excess phosphoric acid as desired. HLW is added to the crystals and the mixture is then heated at a controlled heating rate to induce solid state polymerization and to form a melt at 1350 degrees C in which the HLW oxides dissolve rapidly

Volumetric change of simulated radioactive waste glass irradiated by electron accelerator. [Silica glass

Energy Technology Data Exchange (ETDEWEB)

Sato, Seichi; Furuya, Hirotaka; Inagaki, Yaohiro; Kozaka, Tetsuo; Sugisaki, Masayasu

1987-11-01

Density changes of simulated radioactive waste glasses, silica glass and Pyrex glass irradiated by an electron accelerator were measured by a ''sink-float'' technique. The density changes of the waste and silica glasses were less than 0.05 %, irradiated at 2.0 MeV up to the fluence of 1.7 x 10/sup 17/ ecm/sup 2/, while were remarkably smaller than that of Pyrex glass of 0.18 % shrinkage. Precision of the measurements in the density changes of the waste glass was lower than that of Pyrex glass possibly because of the inhomogeneity of the waste glass

The incorporation of technetium into a representative low-activity waste glass

International Nuclear Information System (INIS)

Ebert, W.L.; Bakel, A.J.; Bowers, D.L.; Buck, E.C.; Emery, J.W.

1997-01-01

A glass that has been tested to understand the corrosion behavior of waste glasses with high soda contents for immobilizing Hanford incidental wastes has been made by melting crushed glass with either TcO 2 or NaTcO 4 at 1,100--1,300 C. Incorporation of technetium in the glass was affected by solubility or kinetic effects. Metallic technetium inclusions formed in all the TcO 2 -doped glasses. Inclusions also formed in glasses with added NaTcO 4 that were melted at 1,100 C, but a glass melted at 1,200 C did not contain detectable inclusions. The presence of Tc-bearing inclusions complicates the interpretation of results from dissolution tests because of the simultaneous release of technetium from more than one phase, the unknown surface areas of each phase, and the possible incorporation of technetium that is released from one phase into another phase. A glass containing about 0.15 mass % Tc dissolved in the glass is being used in dissolution tests to study the release behavior of technetium

Characterization of Savannah River Plant waste glass

International Nuclear Information System (INIS)

Plodinec, M.J.

1985-01-01

The objective of the glass characterization programs at the Savannah River Laboratory (SRL) is to ensure that glass containing Savannah River Plant high-level waste can be permanently stored in a federal repository, in an environmentally acceptable manner. To accomplish this objective, SRL is carrying out several experimental programs, including: fundamental studies of the reactions between waste glass and water, particularly repository groundwater; experiments in which candidate repository environments are simulated as accurately as possible; burial tests of simulated waste glass in candidate repository geologies; large-scale tests of glass durability; and determination of the effects of process conditions on glass quality. In this paper, the strategy and current status of each of these programs is discussed. The results indicate that waste packages containing SRP waste glass will satisfy emerging regulatory criteria

Systems approach to nuclear waste glass development

International Nuclear Information System (INIS)

Jantzen, C.M.

1986-01-01

Development of a host solid for the immobilization of nuclear waste has focused on various vitreous wasteforms. The systems approach requires that parameters affecting product performance and processing be considered simultaneously. Application of the systems approach indicates that borosilicate glasses are, overall, the most suitable glasses for the immobilization of nuclear waste. Phosphate glasses are highly durable; but the glass melts are highly corrosive and the glasses have poor thermal stability and low solubility for many waste components. High-silica glasses have good chemical durability, thermal stability, and mechanical stability, but the associated high melting temperatures increase volatilization of hazardous species in the waste. Borosilicate glasses are chemically durable and are stable both thermally and mechanically. The borosilicate melts are generally less corrosive than commercial glasses, and the melt temperature miimizes excessive volatility of hazardous species. Optimization of borosilicate waste glass formulations has led to their acceptance as the reference nuclear wasteform in the United States, United Kingdom, Belgium, Germany, France, Sweden, Switzerland, and Japan

Thermal conductivity of multibarrier waste form components

International Nuclear Information System (INIS)

Lokken, R.O.

1981-01-01

The multiple barrier concept of radioactive waste immobilization under investigation at Pacific Northwest Laboratory (PNL) uses composite waste forms which exhibit enhanced inertness through improvements in thermal stability, mechanical strength, and leachability by the use of coatings and metal matrices. Since excessive heat may be generated by radioactive decay of the waste, the thermal conductivity of the various barriers, and more importantly of the composite, becomes an important parameter in design criteria. This report presents results of thermal conductivity measurements on 21 various glass, ceramic, metal and composite materials used in multibarrier waste forms development

High-level waste borosilicate glass a compendium of corrosion characteristics. Volume 1

International Nuclear Information System (INIS)

Cunnane, J.C.

1994-03-01

Current plans call for the United States Department of Energy (DOE) to start up facilities for vitrification of high-level radioactive waste (HLW) stored in tanks at the Savannah River Site, Aiken, South Carolina, in 1995; West Valley Demonstration Project, West Valley, New York, in 1996; and at the Hanford Site, Richland, Washington, after the year 2000. The product from these facilities will be canistered HLW borosilicate glass, which will be stored, transported, and eventually disposed of in a geologic repository. The behavior of this glass waste product, under the range of likely service conditions, is the subject of considerable scientific and public interest. Over the past few decades, a large body of scientific information on borosilicate waste glass has been generated worldwide. The intent of this document is to consolidate information pertaining to our current understanding of waste glass corrosion behavior and radionuclide release. The objective, scope, and organization of the document are discussed in Section 1.1, and an overview of borosilicate glass corrosion is provided in Section 1.2. The history of glass as a waste form and the international experience with waste glass are summarized in Sections 1.3 and 1.4, respectively

Decontamination processes for waste glass canisters

International Nuclear Information System (INIS)

Rankin, W.N.

1981-06-01

The process which will be used to decontaminate waste glass canisters at the Savannah River Plant consists of: decontamination (slurry blasting); rinse (high-pressure water); and spot decontamination (high-pressure water plus slurry). No additional waste will be produced by this process because glass frit used in decontamination will be mixed with the radioactive waste and fed into the glass melter. Decontamination of waste glass canisters with chemical and abrasive blasting techniques was investigated. The ability of a chemical technique with HNO 3 -HF and H 2 C 2 O 4 to remove baked-on contamination was demonstrated. A correlation between oxide removal and decontamination was observed. Oxide removal and, thus, decontamination by abrasive blasting techniques with glass frit as the abrasive was proposed and demonstrated

Corrosion of simulated nuclear waste glass

International Nuclear Information System (INIS)

Music, S.; Ristic, M.; Gotic, M.; Foric, J.

1988-01-01

In this study the preparation and characterization of borosilicate glasses of different chemical composition were investigated. Borosilicate glasses were doped with simulated nuclear waste oxides. The chemical corrosion in water of these glasses was followed by measuring the leach rates as a function of time. It was found that a simulated nuclear waste glass with the chemical composition (weight %), 15.61% Na 2 O, 10.39% B 2 O 3 , 45.31% SiO 2 , 13.42% ZnO, 6.61% TiO 2 and 8.66% waste oxides, is characterized by low melting temperature and with good corrosion resistance in water. Influence of passive layers on the leaching behaviour of nuclear waste glasses is discussed. (author) 20 refs.; 7 figs.; 4 tabs

Decontamination processes for waste glass canisters

International Nuclear Information System (INIS)

Rankin, W.N.

1981-01-01

The process which will be used to decontaminate waste glass canisters at the Savannah River Plant consists of: decontamination (slurry blasting); rinse (high-pressure water); and spot decontamination (high-pressure water plus slurry). No additional waste will be produced by this process because glass frit used in decontamination will be mixed with the radioactive waste and fed into the glass melter. Decontamination of waste glass canisters with chemical and abrasive blasting techniques was investigated. The ability of a chemical technique with HNO 3 -HF and H 2 C 2 O 4 to remove baked-on contamination was demonstrated. A correlation between oxide removal and decontamination was observed. Oxide removal and, thus, decontamination by abrasive blasting techniques with glass frit as the abrasive was proposed and demonstrated

Optimization of waste loading in high-level glass in the presence of uncertainty

International Nuclear Information System (INIS)

Hoza, M.; Fann, G.I.; Hopkins, D.F.

1995-02-01

Hanford high-level liquid waste will be converted into a glass form for long-term storage. The glass must meet certain constraints on its composition and properties in order to have desired properties for processing (e.g., electrical conductivity, viscosity, and liquidus temperature) and acceptable durability for long-term storage. The Optimal Waste Loading (OWL) models, based on rigorous mathematical optimization techniques, have been developed to minimize the number of glass logs required and determine glass-former compositions that will produce a glass meeting all relevant constraints. There is considerable uncertainty in many of the models and data relevant to the formulation of high-level glass. In this paper, we discuss how we handle uncertainty in the glass property models and in the high-level waste composition to the vitrification process. Glass property constraints used in optimization are inequalities that relate glass property models obtained by regression analysis of experimental data to numerical limits on property values. Therefore, these constraints are subject to uncertainty. The sampling distributions of the regression models are used to describe the uncertainties associated with the constraints. The optimization then accounts for these uncertainties by requiring the constraints to be satisfied within specified confidence limits. The uncertainty in waste composition is handled using stochastic optimization. Given means and standard deviations of component masses in the high-level waste stream, distributions of possible values for each component are generated. A series of optimization runs is performed; the distribution of each waste component is sampled for each run. The resultant distribution of solutions is then statistically summarized. The ability of OWL models to handle these forms of uncertainty make them very useful tools in designing and evaluating high-level waste glasses formulations

High-Level Waste Glass Formulation Model Sensitivity Study 2009 Glass Formulation Model Versus 1996 Glass Formulation Model

International Nuclear Information System (INIS)

Belsher, J.D.; Meinert, F.L.

2009-01-01

This document presents the differences between two HLW glass formulation models (GFM): The 1996 GFM and 2009 GFM. A glass formulation model is a collection of glass property correlations and associated limits, as well as model validity and solubility constraints; it uses the pretreated HLW feed composition to predict the amount and composition of glass forming additives necessary to produce acceptable HLW glass. The 2009 GFM presented in this report was constructed as a nonlinear optimization calculation based on updated glass property data and solubility limits described in PNNL-18501 (2009). Key mission drivers such as the total mass of HLW glass and waste oxide loading are compared between the two glass formulation models. In addition, a sensitivity study was performed within the 2009 GFM to determine the effect of relaxing various constraints on the predicted mass of the HLW glass.

Inhibitory Effect of Waste Glass Powder on ASR Expansion Induced by Waste Glass Aggregate

Directory of Open Access Journals (Sweden)

Shuhua Liu

2015-10-01

Full Text Available Detailed research is carried out to ascertain the inhibitory effect of waste glass powder (WGP on alkali-silica reaction (ASR expansion induced by waste glass aggregate in this paper. The alkali reactivity of waste glass aggregate is examined by two methods in accordance with the China Test Code SL352-2006. The potential of WGP to control the ASR expansion is determined in terms of mean diameter, specific surface area, content of WGP and curing temperature. Two mathematical models are developed to estimate the inhibitory efficiency of WGP. These studies show that there is ASR risk with an ASR expansion rate over 0.2% when the sand contains more than 30% glass aggregate. However, WGP can effectively control the ASR expansion and inhibit the expansion rate induced by the glass aggregate to be under 0.1%. The two mathematical models have good simulation results, which can be used to evaluate the inhibitory effect of WGP on ASR risk.

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

International Nuclear Information System (INIS)

Jantzen, C

2006-01-01

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

Development and characterization of solidified forms for high-level wastes: 1978. Annual report

International Nuclear Information System (INIS)

Ross, W.A.; Mendel, J.E.

1979-12-01

Development and characterization of solidified high-level waste forms are directed at determining both process properties and long-term behaviors of various solidified high-level waste forms in aqueous, thermal, and radiation environments. Waste glass properties measured as a function of composition were melt viscosity, melt electrical conductivity, devitrification, and chemical durability. The alkali metals were found to have the greatest effect upon glass properties. Titanium caused a slight decrease in viscosity and a significant increase in chemical durability in acidic solutions (pH-4). Aluminum, nickel and iron were all found to increase the formation of nickel-ferrite spinel crystals in the glass. Four multibarrier advanced waste forms were produced on a one-liter scale with simulated waste and characterized. Glass marbles encapsulated in a vacuum-cast lead alloy provided improved inertness with a minimal increase in technological complexity. Supercalcine spheres exhibited excellent inertness when coated with pyrolytic carbon and alumina and put in a metal matrix, but the processing requirements are quite complex. Tests on simulated and actual high-level waste glasses continue to suggest that thermal devitrification has a relatively small effect upon mechanical and chemical durabilities. Tests on the effects radiation has upon waste forms also continue to show changes to be relatively insignificant. Effects caused by decay of actinides can be estimated to saturate at near 10 19 alpha-events/cm 3 in homogeneous solids. Actually, in solidified waste forms the effects are usually observed around certain crystals as radiation causes amorphization and swelling of th crystals

Glass: a candidate engineered material for management of high level nuclear waste

International Nuclear Information System (INIS)

Mishra, R.K.; Kaushik, C.P.

2011-01-01

While the commercial importance of glass is generally recognized, a few people are aware of extremely wide range of glass formulations that can be made and of the versatility of this engineered material. Some of the recent developments in the field of glass leading to various technological applications include glass fiber reinforcement of cement to give new building materials, substrates for microelectronics circuitry in form of semiconducting glasses, nuclear waste immobilization and specific medical applications. The present paper covers fundamental understanding of glass structure and its application for immobilization of high level radioactive liquid waste. High level radioactive liquid waste (HLW) arising during reprocessing of spent fuel are immobilized in sodium borosilicate glass matrix developed indigenously. Glass compositions are modified according to the composition of HLW to meet the criteria of desirable properties in terms. These glass matrices have been characterized for different properties like homogeneity, chemical durability, thermal stability and radiation stability. (author)

The immobilization of High Level Waste Into Glass

International Nuclear Information System (INIS)

Aisyah; Martono, H.

1998-01-01

High level liquid waste is generated from the first step extraction in the nuclear fuel reprocessing. The waste is immobilized with boro-silicate glass. A certain composition of glass is needed for a certain type of waste, so that the properties of waste glass would meet the requirement either for further process or for disposal. The effect of waste loading on either density, thermal expansion, softening point and leaching rate has been studied. The composition of the high level liquid waste has been determined by ORIGEN 2 and the result has been used to prepare simulated high level waste. The waste loading in the waste glass has been set to be 19.48; 22.32; 25.27; and 26.59 weight percent. The result shows that increasing the waste loading has resulted in the higher density with no thermal expansion and softening point significant change. The increase in the waste loading increase that leaching rate. The properties of the waste glass in this research have not shown any deviation from the standard waste glass properties

Electrical resistivities of glass melts containing simulated SRP waste sludges

International Nuclear Information System (INIS)

Wiley, J.R.

1978-08-01

One option for the long-term management of radioactive waste at the Savannah River Plant is to solidify the waste in borosilicate glass by using a continuous, joule-heated, ceramic melter. Electrical resistivities that are needed for melter design were measured for melts of two borosilicate, glass-forming mixtures, each of which was combined with various amounts of several simulated-waste sludges. The simulated sludge spanned the composition range of actual sludges sampled from SRP waste tanks. Resistivities ranged from 6 to 10 ohm-cm at 500 0 C. Melt composition and temperature were correlated with resistivity. Resistivity was not a simple function of viscosity. 15 figures, 4 tables

Factors influencing chemical durability of nuclear waste glasses

International Nuclear Information System (INIS)

Feng, Xiangdong; Bates, J.K.

1993-01-01

A short summary is given of our studies on the major factors that affect the chemical durability of nuclear waste glasses. These factors include glass composition, solution composition, SA/V (ratio of glass surface area to the volume of solution), radiation, and colloidal formation. These investigations have enabled us to gain a better understanding of the chemical durability of nuclear waste glasses and to accumulate.a data base for modeling the long-term durability of waste glass, which will be used in the risk assessment of nuclear waste disposal. This knowledge gained also enhances our ability to formulate optimal waste glass compositions

Waste glass corrosion modeling: Comparison with experimental results

International Nuclear Information System (INIS)

Bourcier, W.L.

1993-11-01

A chemical model of glass corrosion will be used to predict the rates of release of radionuclides from borosilicate glass waste forms in high-level waste repositories. The model will be used both to calculate the rate of degradation of the glass, and also to predict the effects of chemical interactions between the glass and repository materials such as spent fuel, canister and container materials, backfill, cements, grouts, and others. Coupling between the degradation processes affecting all these materials is expected. Models for borosilicate glass dissolution must account for the processes of (1) kinetically-controlled network dissolution, (2) precipitation of secondary phases, (3) ion exchange, (4) rate-limiting diffusive transport of silica through a hydrous surface reaction layer, and (5) specific glass surface interactions with dissolved cations and anions. Current long-term corrosion models for borosilicate glass employ a rate equation consistent with transition state theory embodied in a geochemical reaction-path modeling program that calculates aqueous phase speciation and mineral precipitation/dissolution. These models are currently under development. Future experimental and modeling work to better quantify the rate-controlling processes and validate these models are necessary before the models can be used in repository performance assessment calculations

Predictive modeling of crystal accumulation in high-level waste glass melters processing radioactive waste

Science.gov (United States)

Matyáš, Josef; Gervasio, Vivianaluxa; Sannoh, Sulaiman E.; Kruger, Albert A.

2017-11-01

The effectiveness of high-level waste vitrification at Hanford's Waste Treatment and Immobilization Plant may be limited by precipitation/accumulation of spinel crystals [(Fe, Ni, Mn, Zn)(Fe, Cr)2O4] in the glass discharge riser of Joule-heated ceramic melters during idling. These crystals do not affect glass durability; however, if accumulated in thick layers, they can clog the melter and prevent discharge of molten glass into canisters. To address this problem, an empirical model was developed that can predict thicknesses of accumulated layers as a function of glass composition. This model predicts well the accumulation of single crystals and/or small-scale agglomerates, but excessive agglomeration observed in high-Ni-Fe glass resulted in an underprediction of accumulated layers, which gradually worsened over time as an increased number of agglomerates formed. The accumulation rate of ∼53.8 ± 3.7 μm/h determined for this glass will result in a ∼26 mm-thick layer after 20 days of melter idling.

Basaltic glasses from Iceland and the deep sea: Natural analogues to borosilicate nuclear waste-form glass

International Nuclear Information System (INIS)

Jercinovic, M.J.; Ewing, R.C.

1987-12-01

The report provides a detailed analysis of the alteration process and products for natural basaltic glasses. Information of specific applicability to the JSS project include: * The identification of typical alteration products which should be expected during the long-term corrosion process of low-silica glasses. The leached layers contain a relatively high proportion of crystalline phases, mostly in the form of smectite-type clays. Channels through the layer provide immediate access of solutions to the fresh glass/alteration layer interface. Thus, glasses are not 'protected' from further corrosion by the surface layer. * Corrosion proceeds with two rates - an initial rate in silica-undersaturated environments and a long-term rate in silica-saturated environments. This demonstrates that there is no unexpected change in corrosion rate over long periods of time. The long-term corrosion rate is consistent with that of borosilicate glasses. * Precipitation of silica-containing phases can result in increased alteration of the glass as manifested by greater alteration layer thicknesses. This emphasizes the importance of being able to predict which phases form during the reaction sequence. * For natural basaltic glasses the flow rate of water and surface area of exposed glass are critical parameters in minimizing glass alteration over long periods of time. The long-term stability of basalt glasses is enhanced when silica concentrations in solution are increased. In summary, there is considerable agreement between corrosion phenomena observed for borosilicate glasses in the laboratory and those observed for natural basalt glasses of great age. (With 121 refs.) (authors)

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

International Nuclear Information System (INIS)

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

1996-02-01

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

Coated particle waste form development

International Nuclear Information System (INIS)

Oma, K.H.; Buckwalter, C.Q.; Chick, L.A.

1981-12-01

Coated particle waste forms have been developed as part of the multibarrier concept at Pacific Northwest Laboratory under the Alternative Waste Forms Program for the Department of Energy. Primary efforts were to coat simulated nuclear waste glass marbles and ceramic pellets with low-temperature pyrolytic carbon (LT-PyC) coatings via the process of chemical vapor deposition (CVD). Fluidized bed (FB) coaters, screw agitated coaters (SAC), and rotating tube coaters were used. Coating temperatures were reduced by using catalysts and plasma activation. In general, the LT-PyC coatings did not provide the expected high leach resistance as previously measured for carbon alone. The coatings were friable and often spalled off the substrate. A totally different concept, thermal spray coating, was investigated at PNL as an alternative to CVD coating. Flame spray, wire gun, and plasma gun systems were evaluated using glass, ceramic, and metallic coating materials. Metal plasma spray coatings (Al, Sn, Zn, Pb) provided a two to three orders-of-magnitude increase in chemical durability. Because the aluminum coatings were porous, the superior leach resistance must be due to either a chemical interaction or to a pH buffer effect. Because they are complex, coated waste form processes rank low in process feasibility. Of all the possible coated particle processes, plasma sprayed marbles have the best rating. Carbon coating of pellets by CVD ranked ninth when compared with ten other processes. The plasma-spray-coated marble process ranked sixth out of eleven processes

Examining the role of canister cooling conditions on the formation of nepheline from nuclear waste glasses

Energy Technology Data Exchange (ETDEWEB)

Christian, J. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2015-09-01

Nepheline (NaAlSiO₄) crystals can form during slow cooling of high-level waste (HLW) glass after it has been poured into a waste canister. Formation of these crystals can adversely affect the chemical durability of the glass. The tendency for nepheline crystallization to form in a HLW glass increases with increasing concentrations of Al₂O₃ and Na₂O.

Relative leach behavior of waste glasses and naturally occurring glasses

International Nuclear Information System (INIS)

Adams, P.B.

1979-01-01

Simulated nuclear waste glasses of the sodium-borosilicate type with a low waste loading and of the zinc-borosilicate type with a high waste loading have been compared with obsidians. The resuls indicate that the waste glasses would corrode in normal natural environments at a rate of about 0.1 μm per year at 30 0 C and about 5 μm per year at 90 0 C, compared with obsidians which seem to corrode at, or less than, about 0.01 μm per year at 30 0 C and less than 1 μm per year at 90 0 C. Activation energies for reactions of the two waste glasses with pure water are about 20 kcal/g-mol. 3 figures, 7 tables

Analyses of SRS waste glass buried in granite in Sweden and salt in the United States

International Nuclear Information System (INIS)

Williams, J.P.; Wicks, G.G.; Clark, D.E.; Lodding, A.R.

1991-01-01

Simulated Savannah River Site (SRS) waste glass forms have been buried in the granite geology of the Stirpa mine in Sweden for two years. Analyses of glass surfaces provided a measure of the performance of the waste glasses as a function of time. Similar SRS waste glass compositions have also been buried in salt at the WIPP facility in Carlsbad, New Mexico for a similar time period. Analyses of the SRS waste glasses buried in-situ in granite will be presented and compared to the performance of these same compositions buried in salt at WIPP

Direct conversion of radioactive and chemical waste containing metals, ceramics, amorphous solids, and organics to glass

International Nuclear Information System (INIS)

Forsberg, C.W.; Beahm, E.C.; Parker, G.W.

1994-01-01

The Glass Material Oxidation and Dissolution System (CMODS) is a new process for direct conversion of radioactive, mixed, and chemical wastes to glass. The wastes can be in the chemical forms of metals, ceramics, amorphous solids, and organics. GMODS destroys organics and it incorporates heavy metals and radionuclides into a glass. Processable wastes may include miscellaneous spent fuels (SF), SF hulls and hardware, plutonium wastes in different forms, high-efficiency particulate air (HEPA) filters, ion-exchange resins, failed equipment, and laboratory wastes. Thermodynamic calculations indicate theoretical feasibility. Small-scale laboratory experiments (< 100 g per test) have demonstrated chemical laboratory feasibility for several metals. Additional work is needed to demonstrate engineering feasibility

The quality study of recycled glass phosphor waste for LED

Science.gov (United States)

Tsai, Chun-Chin; Chen, Guan-Hao; Yue, Cheng-Feng; Chen, Cin-Fu; Cheng, Wood-Hi

2017-02-01

To study the feasibility and quality of recycled glass phosphor waste for LED packaging, the experiments were conducted to compare optical characteristics between fresh color conversion layer and that made of recycled waste. The fresh color conversion layer was fabricated through sintering pristine mixture of Y.A.G. powder [yellow phosphor (Y3AlO12 : Ce3+). Those recycled waste glass phosphor re-melted to form Secondary Molten Glass Phosphor (S.M.G.P.). The experiments on such low melting temperature glass results showed that transmission rates of S.M.G.P. are 9% higher than those of first-sintered glass phosphor, corresponding to 1.25% greater average bubble size and 36% more bubble coverage area in S.M.G.P. In the recent years, high power LED modules and laser projectors have been requiring higher thermal stability by using glass phosphor materials for light mixing. Nevertheless, phosphor and related materials are too expensive to expand their markets. It seems a right trend and research goal that recycling such waste of high thermal stability and quality materials could be preferably one of feasible cost-down solutions. This technical approach could bring out brighter future for solid lighting and light source module industries.

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

International Nuclear Information System (INIS)

Hrma, P.

1994-04-01

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

Evaluation of forms for the immobilization of high-level and transuranic wastes

International Nuclear Information System (INIS)

Schuman, R.P.; Cox, N.D.; Gibson, G.W.; Kelsey, P.V. Jr.

1982-08-01

A figure-of-merit (FOM) analysis has been made of a number of waste forms for solidifying both defense and commercial high-level reprocessing waste (HLW) and transuranic (TRU) wastes. The evaluation includes iron-enriched basalt (IEB), a fusion-produced glass-ceramic, which has not been included in other assessments. For HLW, concrete receives the highest FOM, but may not meet regulatory requirements; IEB and glass are the best choices of the materials that should easily meet regulatory requirements. Concrete waste forms are the best choice for TRU wastes, with IEB a close contender. 116 references, 3 figures, 112 tables

Assessment of processes, facilities, and costs for alternative solid forms for immobilization of SRP defense waste

International Nuclear Information System (INIS)

Dunson, J.B. Jr.; Eisenberg, A.M.; Schuyler, R.L. III; Haight, H.G. Jr.; Mello, V.E.; Gould, T.H. Jr.; Butler, J.L.; Pickett, J.B.

1982-03-01

A quantitative merit evaluation which assesses the relative difficulty of remote processing of Savannah River Plant high-level wastes for seven alternative waste forms is presented. The reference borosilicate glass process is rated as the simplest, followed by FUETAP concrete. The other processes evaluated in order of increasing complexity were: glass marbles in a lead matrix, high-silica glass, crystalline ceramic (Synroc-D and tailored ceramic), and coated ceramic particles. Cost appraisals are summarized for the borosilicate glass, high-silica glass, and ceramic waste form processing facilities

Temperature effects on waste glass performance

International Nuclear Information System (INIS)

Mazer, J.J.

1991-02-01

The temperature dependence of glass durability, particularly that of nuclear waste glasses, is assessed by reviewing past studies. The reaction mechanism for glass dissolution in water is complex and involves multiple simultaneous reaction proceeded, including molecular water diffusion, ion exchange, surface reaction, and precipitation. These processes can change in relative importance or dominance with time or changes in temperature. The temperature dependence of each reaction process has been shown to follow an Arrhenius relationship in studies where the reaction process has been isolated, but the overall temperature dependence for nuclear waste glass reaction mechanisms is less well understood, Nuclear waste glass studies have often neglected to identify and characterize the reaction mechanism because of difficulties in performing microanalyses; thus, it is unclear if such results can be extrapolated to other temperatures or reaction times. Recent developments in analytical capabilities suggest that investigations of nuclear waste glass reactions with water can lead to better understandings of their reaction mechanisms and their temperature dependences. Until a better understanding of glass reaction mechanisms is available, caution should be exercised in using temperature as an accelerating parameter. 76 refs., 1 tab

Final Report for Crucible -Scale Radioactive Vitrification and Product Test of Waste Envelope B (AZ-102) Low-Activity Waste Glass

International Nuclear Information System (INIS)

CRAWFORD, CHARLES

2004-01-01

A proof-of-technology demonstration for the Hanford River Protection Project (RPP) Waste treatment and Immobilization Plant (WTP) was performed by the Savannah River Technology Center (SRTC). As part of this demonstration, treated AZ-102 Low-Activity Waste supernate was vitrified using a crucible-scale furnace. Initial glass samples were quench-cooled and characterized for metals and radionuclides. The glass was also durability tested using the American Society for Testing and Materials (ASTM) Product Consistency Test (PCT) protocol. These tests used the AZ-102 glass formulation Low Activity Waste (LAW) B88 that targeted AZ-102 waste loading at 5 wt% Na2O. After these initial results were obtained with the quench-cooled LAWB88 glass, a prototypical container centerline cooling (CCC) program was supplied to SRTC by WTP. A portion of the quench-cooled LAWB88 glass was remelted and centerline cooled. Samples from the CCC low-activity AZ-102 glass waste form were durability tested using the PCT and characterized for crystalline phase identification.This final report documents the characterization and durability of this AZ-102 glass

Equipping a glovebox for waste form testing and characterization of plutonium bearing materials

International Nuclear Information System (INIS)

Noy, M.; Johnson, S.G.; Moschetti, T.L.

1997-01-01

The recent decision by the Department of Energy to pursue a hybrid option for the disposition of weapons plutonium has created the need for additional facilities that can examine and characterize waste forms that contain Pu. This hybrid option consists of the placement of plutonium into stable waste forms and also into mixed oxide fuel for commercial reactors. Glass and glass-ceramic waste forms have a long history of being effective hosts for containing radionuclides, including plutonium. The types of tests necessary to characterize the performance of candidate waste forms include: static leaching experiments on both monolithic and crushed waste forms, microscopic examination, and density determination. Frequently, the respective candidate waste forms must first be produced using elevated temperatures and/or high pressures. The desired operations in the glovebox include, but are not limited to the following: (1) production of vitrified/sintered samples, (2) sampling of glass from crucibles or other vessels, (3) preparing samples for microscopic inspection and monolithic and crushed static leach tests, and (4) performing and analyzing leach tests in situ. This paper will describe the essential equipment and modifications that are necessary to successfully accomplish the goal of outfitting a glovebox for these functions

Glass and nuclear wastes

International Nuclear Information System (INIS)

Sombret, C.

1982-10-01

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

Prediction of waste glass melt rates

International Nuclear Information System (INIS)

Lee, L.

1987-01-01

Under contract to the Department of Energy, the Du Pont Company has begun construction of a Defense Waste Processing Facility to immobilize radioactive wastes now stored as liquids at the Department of Energy's Savannah River Plant. The immobilization process solidifies waste sludge by vitrification into a leach-resistant borosilicate glass. Development of this process has been the responsibility of the Savannah River Laboratory. As part of the development, a simple model was developed to predict the melt rates for the waste glass melter. This model is based on an energy balance for the cold cap and gives very good agreement with melt rate data obtained from experimental campaigns in smaller scale waste glass melters

Long term stability of yttria-stabilized zirconia waste forms. Stability for secular change of partitioned TRU waste composition by disintegration

International Nuclear Information System (INIS)

Kuramoto, Ken-ichi; Banba, Tsunetaka; Mitamura, Hisayoshi; Sakai, Etsuro; Uno, Masayoshi; Kinoshita, H.; Yamanaka, Shinsuke

1999-01-01

In this study, the stability of YSZ waste forms for secular change of partitioned TRU waste composition by disintegration, one of important terms in long-term stability, is the special concern. Designed amount of waste and YSZ powder were mixed and sintered. These TRU waste forms were submitted to tests of phase stability, chemical durability, mechanical property and compactness. The results were compared with those of another YSZ waste forms, non-radioactive Ce and/or Nd doped YSZ samples, and glass and Synroc waste forms. Experimental results show following: (1) Phase stability of (Np+Am)-, (Np+U)-, and (Np+U+Bi)-doped YSZ waste forms could be maintained of that of the initial Np+Am-doped YSZ waste form permanently even when the composition of partitioned TRU waste were changed by disintegration. (2) Secular change also accelerated volume increase of YSZ waste forms as well as alpha-decay damage. (3) Hv, E and K IC of (Np+U)- and (Np+U+Bi)-doped YSZ waste forms were independent of the secular change of the partitioned TRU waste composition by disintegration. (4) Mechanical properties of YSZ waste forms were more than those of a glass and Synroc waste forms. (5) Compactness of YSZ waste forms was good as waste forms for the partitioned TRU wastes. (J.P.N.)

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.

Molecular Environmental Science Using Synchrotron Radiation: Chemistry and Physics of Waste Form Materials. Final Report

International Nuclear Information System (INIS)

Lindle, Dennis W.

2011-01-01

Production of defense-related nuclear materials has generated large volumes of complex chemical wastes containing a mixture of radionuclides. The disposition of these wastes requires conversion of the liquid and solid-phase components into durable, solid forms suitable for long-term immobilization. Specially formulated glass compositions and ceramics such as pyrochlores and apatites are the main candidates for these wastes. An important consideration linked to the durability of waste-form materials is the local structure around the waste components. Equally important is the local structure of constituents of the glass and ceramic host matrix. Knowledge of the structure in the waste-form host matrices is essential, prior to and subsequent to waste incorporation, to evaluate and develop improved waste-form compositions based on scientific considerations. This project used the soft-x-ray synchrotron-radiation-based technique of near-edge x-ray-absorption fine structure (NEXAFS) as a unique method for investigating oxidation states and structures of low-Z elemental constituents forming the backbones of glass and ceramic host matrices for waste-form materials. In addition, light metal ions in ceramic hosts, such as titanium, are also ideal for investigation by NEXAFS in the soft-x-ray region. Thus, one of the main objectives was to understand outstanding issues in waste-form science via NEXAFS investigations and to translate this understanding into better waste-form materials, followed by eventual capability to investigate 'real' waste-form materials by the same methodology. We conducted several detailed structural investigations of both pyrochlore ceramic and borosilicate-glass materials during the project and developed improved capabilities at Beamline 6.3.1 of the Advanced Light Source (ALS) to perform the studies.

Technetium Incorporation in Glass for the Hanford Tank Waste Treatment and Immobilization Plant

Energy Technology Data Exchange (ETDEWEB)

Kruger, Albert A.; Kim, Dong Sang

2015-01-14

A priority of the United States Department of Energy (U.S. DOE) is to dispose of nuclear wastes accumulated in 177 underground tanks at the Hanford Nuclear Reservation in eastern Washington State. These nuclear wastes date from the Manhattan Project of World War II and from plutonium production during the Cold War. The DOE plans to separate high-level radioactive wastes from low activity wastes and to treat each of the waste streams by vitrification (immobilization of the nuclides in glass) for disposal. The immobilized low-activity waste will be disposed of here at Hanford and the immobilized high-level waste at the national geologic repository. Included in the inventory of highly radioactive wastes is large volumes of 99Tc (~9 × 10E2 TBq or ~2.5 × 104 Ci or ~1500 kg). A problem facing safe disposal of Tc-bearing wastes is the processing of waste feed into in a chemically durable waste form. Technetium incorporates poorly into silicate glass in traditional glass melting. It readily evaporates during melting of glass feeds and out of the molten glass, leading to a spectrum of high-to-low retention (ca. 20 to 80%) in the cooled glass product. DOE-ORP currently has a program at Pacific Northwest National Laboratory (PNNL), in the Department of Materials Science and Engineering at Rutgers University and in the School of Mechanical and Materials Engineering at Washington State University that seeks to understand aspects of Tc retention by means of studying Tc partitioning, molten salt formation, volatilization pathways, and cold cap chemistry. Another problem involves the stability of Tc in glass in both the national geologic repository and on-site disposal after it has been immobilized. The major environmental concern with 99Tc is its high mobility in addition to a long half-life (2.1×105 yrs). The pertechnetate ion (TcO4-) is highly soluble in water and does not adsorb well onto the surface of minerals and so migrates nearly at the same velocity as groundwater

Localized chemistry of 99Tc in simulated low activity waste glass

Science.gov (United States)

Weaver, Jamie L.

A priority of the United States Department of Energy (DOE) is to dispose of the nuclear waste accumulated in the underground tanks at the Hanford Nuclear Reservation in Richland, WA. Incorporation and stabilization of technetium (99Tc) from these tanks into vitrified waste forms is a concern to the waste glass community and DOE due to 99Tc's long half-life ( 2.13˙105 y), and its high mobility in the subsurface environment under oxidizing conditions. Working in collaboration with researchers at Pacific Northwest National Laboratory (PNNL) and other national laboratories, plans were formulated to obtain first-of-a-kind chemical structure determination of poorly understood and environmentally relevant technetium compounds that relate to the chemistry of the Tc in nuclear waste glasses. Knowledge of the structure and spectral signature of these compounds aid in refining the understanding of 99Tc incorporation into and release from oxide based waste glass. In this research a first-of-its kind mechanism for the behavior of 99Tc during vitrification is presented, and the structural role of Tc(VII) and (IV) in borosilicate waste glasses is readdressed.

DWPF waste form compliance plan (Draft Revision)

International Nuclear Information System (INIS)

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

1991-01-01

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

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

International Nuclear Information System (INIS)

1993-08-01

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

Diffusion processes in nuclear waste glasses

International Nuclear Information System (INIS)

Serruys, Y.; Limoge, Y.; Brebec, G.

1992-01-01

Problems concerning the containment of nuclear wastes are presented. Different materials which have been considered for this purpose are briefly reviewed and we see why glass is one of the favorite candidates. It is focussed on what is known about diffusion in 'simple enough' glasses. After a recall concerning the structure and possible defects, the main results on diffusion in 'simple' glasses are given and it is shown what these results involve for the mechanisms of diffusion. The diffusion models are presented which can account for transport in random media: percolation and random walk models. Specific phenomena for the nuclear waste glasses are considered: the effect of irradiation on diffusion and leaching (i.e. corrosion by water). Finally diffusion data in nuclear waste glasses are presented. (author). 199 refs., 6 figs., 1 tab

Studies of high-level radioactive waste form performance at Japan Atomic Energy Research Institute

International Nuclear Information System (INIS)

Banba, Tsunetaka; Kamizono, Hiroshi; Mitamura, Hisayoshi

1992-02-01

The recent studies of high-level radioactive waste form at Japan Atomic Energy Research Institute can be classified into the following three categories; (1) Study on the leaching behavior of the nuclear waste glass placing the focus on the alteration layer and the chemical composition of leachant for the prediction of the long-term corrosion of the waste glass. (2) Study on the radiation (alpha-radiation) effects which have relation to the long-term stability of the nuclear waste glass. (3) Study on the long-term self-irradiation damage of a SYNROC waste form using a curium-doped sample. In the present report, the recent results corresponding to the above categories are described. (author)

Review of radiation effects in solid-nuclear-waste forms

International Nuclear Information System (INIS)

Weber, W.J.

1981-09-01

Radiation effects on the stability of high-level nuclear waste (HLW) forms are an important consideration in the development of technology to immobilize high-level radioactive waste because such effects may significantly affect the containment of the radioactive waste. Since the required containment times are long (10 3 to 10 6 years), an understanding of the long-term cumulative effects of radiation damage on the waste forms is essential. Radiation damage of nuclear waste forms can result in changes in volume, leach rate, stored energy, structure/microstructure, and mechanical properties. Any one or combination of these changes might significantly affect the long-term stability of the nuclear waste forms. This report defines the general radiation damage problem in nuclear waste forms, describes the simulation techniques currently available for accelerated testing of nuclear waste forms, and reviews the available data on radiation effects in both glass and ceramic (primarily crystalline) waste forms. 76 references

Projected radionuclide inventories of DWPF glass from current waste at time of production

International Nuclear Information System (INIS)

Plodinec, M.J.

1993-01-01

The Waste Acceptance Preliminary Specifications (WAPS) require that the DWPF estimate the inventory of long-lived radionuclides present in the waste glass, and report the values in the Waste Form Qualification Report. In this report, conservative (biased high) estimates of the radionuclide inventory of glass produced from waste currently in the Tank Farm are provided. In most cases, these calculated values compare favorably with actual data. In those cases where the agreement is not good, the values reported here are conservative

MIIT: International in-situ testing of nuclear waste glasses-performance of SRS simulated waste glass after 5 years of burial at the waste isolation pilot plant (WIPP)

International Nuclear Information System (INIS)

Wicks, G.G.; Lodding, A.R.; Macedo, P.B.; Clark, D.E.

1993-01-01

In July of 1986, the first in-situ test involving burial of simulated high-level waste [HLW] forms conducted in the United States was started. This program, called the Materials Interface Interactions Test or MIIT, comprises the largest, most cooperative field-testing venture in the international waste management community. Included in the study are over 900 waste form samples comprising 15 different systems supplied by seven nations. Also included are about 300 potential canister or overpack metal samples along with more than 500 geologic and backfill specimens. There are almost 2000 relevant interactions that characterize this effort which has been conducted in the bedded salt site at the Waste Isolation Pilot Plant (WIPP), near Carlsbad, New Mexico. The MIIT program represents a joint effort managed by the Savannah River Technology Center (SRTC) in Aiken, S.C., and Sandia National Laboratories (SNL) in Albuquerque, N.M.. and sponsored by the US Department of Energy. Involved in MIIT are participants from national and federal laboratories, universities, and representatives from laboratories in France, Germany, Canada, Belgium, Japan, Sweden, the United Kingdom, and the United States. In July of 1991, the experimental portion of the 5-year MIIT study was completed on schedule. During this time interval, many in-situ measurements were performed, thousands of brine analyses conducted, and hundreds of waste glass and package components exhumed and evaluated after 6 mo., 1 yr., 2 yr. and 5 yr. burial periods. Although analyses are still in progress, the performance of SRS waste glass based on all data currently available has been seen to be excellent thus far. Initial analyses and assessment of Savannah River (SR) waste glass after burial in WIPP at 90 degrees C for 5 years is presented

Modeling a novel glass immobilization waste treatment process using flow

International Nuclear Information System (INIS)

Ferrada, J.J.; Nehls, J.W. Jr.; Welch, T.D.; Giardina, J.L.

1996-01-01

One option for control and disposal of surplus fissile materials is the Glass Material Oxidation and Dissolution System (GMODS), a process developed at ORNL for directly converting Pu-bearing material into a durable high-quality glass waste form. This paper presents a preliminary assessment of the GMODS process flowsheet using FLOW, a chemical process simulator. The simulation showed that the glass chemistry postulated ion the models has acceptable levels of risks

Glass-ceramics: Their production from wastes - a review

Energy Technology Data Exchange (ETDEWEB)

Rawlings, R.D.; Wu, J.P.; Boccaccini, A.R. [University of London, London (United Kingdom). Imperial College of Science & Technology, Dept. of Medicine

2006-02-15

Glass-ceramics are polycrystalline materials of fine microstructure that are produced by the controlled crystallisation (devitrification) of a glass. Numerous silicate based wastes, such as coal combustion ash, slag from steel production, fly ash and filter dusts from waste incinerators, mud from metal hydrometallurgy, different types of sludge as well as glass cullet or mixtures of them have been considered for the production of glass-ceramics. Developments of glass-ceramics from waste using different processing methods are described comprehensively in this review, covering R&D work carried out worldwide in the last 40 years. Properties and applications of the different glass-ceramics produced are discussed. The review reveals that considerable knowledge and expertise has been accumulated on the process of transformation of silicate waste into useful glass-ceramic products. These glass-ceramics are attractive as building materials for usage as construction and architectural components or for other specialised technical applications requiring a combination of suitable thermo-mechanical properties. Previous attempts to commercialise glass-ceramics from waste and to scale-up production for industrial exploitation are also discussed.

NMR characterization of simulated Hanford low-activity waste glasses and its use in understanding waste form chemical durability

International Nuclear Information System (INIS)

Darab, J.G.; Linehan, J.C.; McGrail, B.P.

1999-01-01

Magic Angle Spinning Nuclear Magnetic Resonance (MAS-NMR) spectroscopy has been used to characterize the structural and chemical environments of B, Al, and Si in model Hanford low-activity waste glasses. The average 29 Si NMR peak position was found to systematically change with changing glass composition and structure. From an understanding of the structural roles of Al and B obtained from MAS-NMR experiments, the authors first developed a model that reliably predicts the distribution of structural units and the average 29 Si chemical shift value, δ, based purely on glass composition. A product consistency test (PCT) was used to determine the normalized elemental release (NL) from the prepared glasses. Comparison of the NMR and PCT data obtained from sodium boro-aluminosilicate glasses indicates that a rudimentary exponential relationship exists between the 29 Si chemical shift value, and the boron NL value

Characterization of low and medium level radioactive waste forms

International Nuclear Information System (INIS)

Sambell, R.A.J.

1983-01-01

The work reported was carried out during the first year of the Commission of the European Community's programme on the characterization of low and medium level waste forms. Ten reference waste forms plus others of special national interest have been identified covering PWR, BWR, GCR and reprocessing wastes. The immobilising media include the three main matrices: cement, polymers and bitumen, and a glass. Characterization is viewed as one input to quality assurance of the waste form and covers: waste-matrix compatibility, radiation effects, leaching, microbiological attack, shrinkage and swelling, ageing processes and thermal effects. The aim is a balanced programme of comparative data, predictive modelling and an undserstanding of basic mechanisms

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-01-10

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

Summary and evaluation of nuclear waste forms. Chapter 12

International Nuclear Information System (INIS)

Lutze, W.; Ewing, R.C.

1988-01-01

In this chapter data are compiled from the foregoing contributed chapters into tables. In a few cases additional more recent data not found in the chapters have been included in the tables. The following waste form data are summarized: physical properties, chemical durability, radiation effects and the status of processing techniques. In addition important aspects of the comparison of waste forms and the response of waste forms (glass and ceramic) to corrosion and radiation effects are discussed. (author). 119 refs.; 6 figs.; 5 tabs

Glass Ceramics Composites Fabricated from Coal Fly Ash and Waste Glass

International Nuclear Information System (INIS)

Angjusheva, B.; Jovanov, V.; Srebrenkoska, V.; Fidancevska, E.

2014-01-01

Great quantities of coal ash are produced in thermal power plants which present a double problem to the society: economical and environmental. This waste is a result of burning of coal at temperatures between 1100-14500C. Fly ash available as fine powder presents a source of important oxides SiO2, Al2O3, Fe2O3, MgO, Na2O, but also consist of small amount of ecologically hazardous oxides such as Cr2O3, NiO, MnO. The combination of the fly ash with waste glass under controlled sintering procedure gave bulk glass-ceramics composite material. The principle of this procedure is presented as a multi barrier concept. Many researches have been conducted the investigations for utilization of fly ash as starting material for various glass–ceramics production. Using waste glass ecologically hazardous components are fixed at the molecular level in the silicate phase and the fabricated new glass-ceramic composites possess significantly higher mechanical properties. The aim of this investigation was to fabricate dense glass ceramic composites using fly ash and waste glass with the potential for its utilization as building material

Performance testing of waste forms in a tuff environment

International Nuclear Information System (INIS)

Oversby, V.M.

1983-11-01

This paper describes experimental work conducted to establish the chemical composition of water which will have reacted with Topopah Spring Member tuff prior to contact with waste packages. The experimental program to determine the behavior of spent fuel and borosilicate glass in the presence of this water is then described. Preliminary results of experiments using spent fuel segments with defects in the Zircaloy cladding are presented. Some results from parametric testing of a borosilicate glass with tuff and 304L stainless steel are also discussed. Experiments conducted using Topopah Spring tuff and J-13 well water have been conducted to provide an estimate of the post-emplacement environment for waste packages in a repository at Yucca Mountain. The results show that emplacement of waste packages should cause only small changes in the water chemistry and rock mineralogy. The changes in environment should not have any detrimental effects on the performance of metal barriers or waste forms. The NNWSI waste form testing program has provided preliminary results related to the release rate of radionuclides from the waste package. Those results indicate that release rates from both spent fuel and borosilicate glass should be below 1 part in 10 5 per year. Future testing will be directed toward making release rate testing more closely relevant to site specific conditions. 17 references, 7 figures

Glass Formulation Development for INEEL Sodium-Bearing Waste

International Nuclear Information System (INIS)

Vienna, J.D.; Schweiger, M.J.; Smith, D.E.; Smith, H.D.; Crum, J.V.; Peeler, D.K.; Reamer, I.A.; Musick, C.A.; Tillotson, R.D.

1999-01-01

a standard liquid-fed joule-heated melter. The normalized elemental releases by 7-day PCT are all well below 1 g/m 2 , which is a very conservative set point used in this study. The T L , ignoring sulfate formation, is less than the 1050 C limit. Based on these observations and the reasonable waste loading of 35 mass 0/0, the SBW glass was a prime candidate for further testing. Sulfate salt segregation was observed in all test melts formed from oxidized carbonate precursors. Melts fabricated using SBW simulants suggest that the sulfate-salt segregation seen in oxide and carbonate melts was much less of a problem. The cause for the difference is likely H 2 SO 4 fuming during the boil-down stage of wet-slurry processing. Additionally, some crucible tests with SBW simulant were conducted at higher temperatures (1250 C), which could increase the volatility of sulfate salts. The fate of sulfate during the melting process is still uncertain and should be the topic of future studies. The properties of the simulant glass confirmed those of the oxide and carbonate glass. Corrosion tests on Inconel 690 electrodes and K-3 refractory blocks conducted at INEEL suggest that the glass is not excessively corrosive. Based on the results of this study, the authors recommend that a glass made of 35% SBW simulant (on a mass oxide and halide basis) and 65% of the additive mix (either filled or raw chemical) be used in demonstrating the direct vitrification of INEEL SBW. It is further recommended that a study be conducted to determine the fate of sulfate during glass processing and the tolerance of the chosen melter technology to sulfate salt segregation and corrosivity of the melt

Coupling of Nuclear Waste Form Corrosion and Radionuclide Transports in Presence of Relevant Repository Sediments

International Nuclear Information System (INIS)

Wall, Nathalie A.; Neeway, James J.; Qafoku, Nikolla P.; Ryan, Joseph V.

2015-01-01

decrease the need for expensive engineered barriers.Our current work aims are 1) quantifying and understanding the processes associated with glass alteration in contact with Fe-bearing materials; 2) quantifying and understanding the processes associated with glass alteration in presence of MgO (example of engineered barrier used in WIPP); 3) identifying glass alteration suppressants and the processes involved to reach glass alteration suppression; 4) quantifying and understanding the processes associated with Saltstone and Cast Stone (SRS and Hanford cementitious waste forms) in various representative groundwaters; 5) investigating positron annihilation as a new tool for the study of glass alteration; and 6) quantifying and understanding the processes associated with glass alteration under gamma irradiation.

Coupling of Nuclear Waste Form Corrosion and Radionuclide Transports in Presence of Relevant Repository Sediments

Energy Technology Data Exchange (ETDEWEB)

Wall, Nathalie A. [Washington State Univ., Pullman, WA (United States); Neeway, James J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Qafoku, Nikolla P. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ryan, Joseph V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2015-09-30

decrease the need for expensive engineered barriers.Our current work aims are 1) quantifying and understanding the processes associated with glass alteration in contact with Fe-bearing materials; 2) quantifying and understanding the processes associated with glass alteration in presence of MgO (example of engineered barrier used in WIPP); 3) identifying glass alteration suppressants and the processes involved to reach glass alteration suppression; 4) quantifying and understanding the processes associated with Saltstone and Cast Stone (SRS and Hanford cementitious waste forms) in various representative groundwaters; 5) investigating positron annihilation as a new tool for the study of glass alteration; and 6) quantifying and understanding the processes associated with glass alteration under gamma irradiation.

Glasses obtained from industrial wastes

International Nuclear Information System (INIS)

Bortoluzzi, D.; Oliveira Fillho, J.; Uggioni, E.; Bernardin, A.M.

2009-01-01

This paper deals with the study of the vitrification mechanism as an inertization method for industrial wastes contaminated with heavy metals. Ashes from coal (thermoelectric), wastes from mining (fluorite and feldspar) and plating residue were used to compose vitreous systems planed by mixture design. The chemical composition of the wastes was determined by XRF and the formulations were melted at 1450 deg C for 2h using 10%wt of CaCO 3 (fluxing agent). The glasses were poured into a mold and annealed (600 deg C). The characteristic temperatures were determined by thermal analysis (DTA, air, 20 deg C/min) and the mechanical behavior by Vickers microhardness. As a result, the melting temperature is strongly dependent on silica content of each glass, and the fluorite residue, being composed mainly by silica, strongly affects Tm. The microhardness of all glasses is mainly affected by the plating residue due to the high iron and zinc content of this waste. (author)

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

International Nuclear Information System (INIS)

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

1984-04-01

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

Storage and disposal of radioactive waste as glass in canisters

International Nuclear Information System (INIS)

Mendel, J.E.

1978-12-01

A review of the use of waste glass for the immobilization of high-level radioactive waste glass is presented. Typical properties of the canisters used to contain the glass, and the waste glass, are described. Those properties are used to project the stability of canisterized waste glass through interim storage, transportation, and geologic disposal

Colloid formation during waste glass corrosion

International Nuclear Information System (INIS)

Mertz, C.J.; Buck, E.C.; Fortner, J.A.; Bates, J.K.

1996-01-01

The long-term behavior of nuclear waste glass in a geologic repository may require a technical consideration of the role of colloids in the release and transport of radionuclides. The neglect of colloidal properties in assessing the near- and far-field migration behavior of actinides may lead to significant underestimates and poor predictions of biosphere exposure from high-level waste (HLW) disposal. Existing data on colloid-facilitated transport suggests that radionuclide migration may be enhanced, but the importance of colloids is not adequately assessed. Indeed, the occurrence of radionuclide transport, attributed to colloidal species, has been reported at Mortandad Canyon, Los Alamos and at the Nevada Test Site; both unsaturated regions are similar to the proposed HLW repository at Yucca Mountain. Although some developments have been made on understanding the transport characteristics of colloids, the characterization of colloids generated from the corrosion of the waste form has been limited. Colloids are known to incorporate radionuclides either from hydrolysis of dissolved species (real colloids) or from adsorption of dissolved species onto existing groundwater colloids (pseudocolloids); however, these colloids may be considered secondary and solubility limited when compared to the colloids generated during glass alteration

Role of lead as modifier on the properties of lead iron phosphate nuclear waste glasses

International Nuclear Information System (INIS)

Hazra, G.; Mitra, P.; Das, T.

2011-01-01

Lead-iron phosphate glasses are a promising new waste form for the safe immobilization of both high level defence and high level commercial radioactive waste for long term disposal. Lead iron phosphate glasses have several advantages such as lower aqueous corrosion rate, lower processing temperature etc. (author)

A Review of Iron Phosphate Glasses and Recommendations for Vitrifying Hanford Waste

Energy Technology Data Exchange (ETDEWEB)

Delbert E. Ray; Chandra S. Ray

2013-11-01

This report contains a comprehensive review of the research conducted, world-wide, on iron phosphate glass over the past ~30 years. Special attention is devoted to those iron phosphate glass compositions which have been formulated for the purpose of vitrifying numerous types of nuclear waste, with special emphasis on the wastes stored in the underground tanks at Hanford WA. Data for the structural, chemical, and physical properties of iron phosphate waste forms are reviewed for the purpose of understanding their (a) outstanding chemical durability which meets all current DOE requirements, (b) high waste loadings which can exceed 40 wt% (up to 75 wt%) for several Hanford wastes, (c) low melting temperatures, can be as low as 900°C for certain wastes, and (d) high tolerance for “problem” waste components such as sulfates, halides, and heavy metals (chromium, actinides, noble metals, etc.). Several recommendations are given for actions that are necessary to smoothly integrate iron phosphate glass technology into the present waste treatment plans and vitrification facilities at Hanford.

Alternative solid forms for Savannah River Plant defense waste

International Nuclear Information System (INIS)

Stone, J.A.; Goforth, S.T.; Smith, P.K.

1980-01-01

Solid forms and processes were evaluated for immobilization of SRP high-level radioactive waste, which contains bulk chemicals such as hydrous iron and aluminium oxides. Borosilicate glass currently is the best overall choice. High-silica glass, tailored ceramics, and coated ceramics are potentially superior products, but require more difficult processes

Borosilicate glass as a matrix for immobilization of SRP high-level waste

International Nuclear Information System (INIS)

Wicks, G.G.

1980-01-01

Approximately 22 million gallons of high-level radioactive defense waste are currently being stored in large underground tanks located on the Savannah River Plant (SRP) site in Aiken, South Carolina. One option now being considered for long-term management of this waste involves removing the waste from the tanks, chemically processing the waste, and immobilizing the potentially harmful radionuclides in the waste into a borosilicate glass matrix. The technology for producing waste glass forms is well developed and has been demonstrated on various scales using simulated as well as radioactive SRP waste. Recently, full-scale prototypical equipment has been made operational at SRP. This includes both a joule-heated ceramic melter and an in-can melter. These melters are a part of an integrated vitrification system which is under evaluation and includes a spray calciner, direct liquid feed apparatus, and various elements of an off-gas system. Two of the most important properties of the waste glass are mechanical integrity and leachability. Programs are in progress at SRL aimed at minimizing thermally induced cracking by carefully controlling cooling cycles and using ceramic liners or coatings. The leachability of SRP waste glass has been studied under many different conditions and consistently found to be low. For example, the leachability of actual SRP waste glass was found to be 10 -6 to 10 -5 g/(cm 2 )(day) initially and decreasing to 10 -9 to 10 -8 g/(cm 2 )(day) after 100 days. Waste glass is also being studied under anticipated storage conditions. In brine at 90 0 C, the leachability is about 5 x 10 -8 g/(cm 2 )(day) after 60 days. The effects of other geological media including granite, basalt, shale, and tuff are also being studied as part of the multibarrier isolation system

A review of glass-ceramics for the immobilization of nuclear fuel recycle wastes

International Nuclear Information System (INIS)

Hayward, P.J.

1987-01-01

This report reviews the status of the Canadian, German, U.S., Japanese, U.S.S.R. and Swedish programs for the development of glass-ceramic materials for immobilizing the high-level radioactive wastes arising from the recycling of used nuclear fuel. The progress made in these programs is described, with emphasis on the Canadian program for the development of sphene-based glass-ceramics. The general considerations of product performance and process feasibility for glass-ceramics as a category of waste form material are discussed. 137 refs

Eco-efficient waste glass recycling: Integrated waste management and green product development through LCA

International Nuclear Information System (INIS)

Blengini, Gian Andrea; Busto, Mirko; Fantoni, Moris; Fino, Debora

2012-01-01

Highlights: ► A new eco-efficient recycling route for post-consumer waste glass was implemented. ► Integrated waste management and industrial production are crucial to green products. ► Most of the waste glass rejects are sent back to the glass industry. ► Recovered co-products give more environmental gains than does avoided landfill. ► Energy intensive recycling must be limited to waste that cannot be closed-loop recycled. - Abstract: As part of the EU Life + NOVEDI project, a new eco-efficient recycling route has been implemented to maximise resources and energy recovery from post-consumer waste glass, through integrated waste management and industrial production. Life cycle assessment (LCA) has been used to identify engineering solutions to sustainability during the development of green building products. The new process and the related LCA are framed within a meaningful case of industrial symbiosis, where multiple waste streams are utilised in a multi-output industrial process. The input is a mix of rejected waste glass from conventional container glass recycling and waste special glass such as monitor glass, bulbs and glass fibres. The green building product is a recycled foam glass (RFG) to be used in high efficiency thermally insulating and lightweight concrete. The environmental gains have been contrasted against induced impacts and improvements have been proposed. Recovered co-products, such as glass fragments/powders, plastics and metals, correspond to environmental gains that are higher than those related to landfill avoidance, whereas the latter is cancelled due to increased transportation distances. In accordance to an eco-efficiency principle, it has been highlighted that recourse to highly energy intensive recycling should be limited to waste that cannot be closed-loop recycled.

Recycling of inorganic waste in monolithic and cellular glass-based materials for structural and functional applications.

Science.gov (United States)

Rincón, Acacio; Marangoni, Mauro; Cetin, Suna; Bernardo, Enrico

2016-07-01

The stabilization of inorganic waste of various nature and origin, in glasses, has been a key strategy for environmental protection for the last decades. When properly formulated, glasses may retain many inorganic contaminants permanently, but it must be acknowledged that some criticism remains, mainly concerning costs and energy use. As a consequence, the sustainability of vitrification largely relies on the conversion of waste glasses into new, usable and marketable glass-based materials, in the form of monolithic and cellular glass-ceramics. The effective conversion in turn depends on the simultaneous control of both starting materials and manufacturing processes. While silica-rich waste favours the obtainment of glass, iron-rich wastes affect the functionalities, influencing the porosity in cellular glass-based materials as well as catalytic, magnetic, optical and electrical properties. Engineered formulations may lead to important reductions of processing times and temperatures, in the transformation of waste-derived glasses into glass-ceramics, or even bring interesting shortcuts. Direct sintering of wastes, combined with recycled glasses, as an example, has been proven as a valid low-cost alternative for glass-ceramic manufacturing, for wastes with limited hazardousness. The present paper is aimed at providing an up-to-date overview of the correlation between formulations, manufacturing technologies and properties of most recent waste-derived, glass-based materials. © 2016 The Authors. Journal of Chemical Technology & Biotechnology published by John Wiley & Sons Ltd on behalf of Society of Chemical Industry.

Tests with ceramic waste form materials made by pressureless consolidation

International Nuclear Information System (INIS)

Lewis, M. A.; Hash, M. C.; Hebden, A. S.; Ebert, W. L.

2002-01-01

A multiphase waste form referred to as the ceramic waste form (CWF) will be used to immobilize radioactively contaminated salt wastes recovered after the electrometallurgical treatment of spent sodium-bonded nuclear fuel. The CWF is made by first occluding salt in zeolite and then encapsulating the zeolite in a borosilicate binder glass. A variety of surrogate CWF materials were made using pressureless consolidation (PC) methods for comparison with CWF consolidated using a hot isostatic press (HIP) method and to study the effects of glass/zeolite batching ratio and processing conditions on the physical and chemical properties of the resulting materials. The data summarized in this report will also be used to support qualification of the PC CWF for disposal in the proposed federal high-level radioactive waste repository at Yucca Mountain. The phase composition and microstructure of HIP CWF and PC CWF are essentially identical: both are composed of about 70% sodalite, 25% binder glass, and a 5% total of inclusion phases (halite, nepheline, and various oxides and silicates). The primary difference is that PC CWF materials have higher porosities than HIP CWFs. The product consistency test (PCT) that was initially developed to monitor homogeneous glass waste forms was used to measure the chemical durabilities of the CWF materials. Series of replicate tests with several PC CWF materials indicate that the PCT can be conducted with the same precision with CWF materials as with borosilicate glasses. Short-term (7-day) PCTs were used to evaluate the repeatability of making the PC CWF and the effects of the glass/zeolite mass ratio, process temperature, and processing time on the chemical durability. Long-term (up to 1 year) PCTs were used to compare the durabilities of HIP and PC CWFs and to estimate the apparent solubility limit for the PC CWF that is needed for modeling. The PC and HIP CWF materials had similar disabilities, based on the release of silicon in long

Evaluation and review of alternative waste forms for immobilization of high level radioactive wastes. Report number 3

International Nuclear Information System (INIS)

1981-01-01

A discussion of the relative strengths and weaknesses of the alternative forms and recommendations for future program directions are presented in the body of this report. In addition to the relative ranking, the Peer Review Panel makes the following observations and recommendations: (1) Differences in overall performance of most of the uncoated waste forms are relatively small when compared under approximately equivalent conditions. (2) The increased scientific basis for this class of waste forms has not yet been sufficient to achieve reliably large improvements in waste form performance over the best borosilicate glasses. (3) The increased leach rates at elevated temperatures and the uncertainty regarding mechanisms of leaching under repository conditions continue to indicate that surface temperatures of waste canisters and especially any waste form-water interfaces should be restricted to less than 100 0 C, until more data is available to indicate otherwise. (4) Improvements are noteworthy, but there is still a need for adopting additional standardized tests, standard reference materials, common units and standardized methods of data presentation in the nuclear waste program. (5) Comparative data on leach rates in waters equilibrated with candidate rocks and potential geologic environments are almost non-existent and are essential to establish relevant long term extrapolation of waste form performance.(6) Understanding radiation damage effects on the microstructure and leaching mechanisms of polycrystalline ceramics is still insufficient to judge long term reliability of this class of waste forms. (7) More extensive data on rates and mechanisms of leaching of all waste forms under radiation and repository conditions are needed. (8) Additional studies of fundamental mechanisms controlling long term reliability of glass and alternative waste forms are strongly encouraged

Material interactions relating to long-term geologic disposal of nuclear waste glass

International Nuclear Information System (INIS)

Bibler, N.E.; Jantzen, C.M.

1986-01-01

This review paper systematizes the additional interactions that materials in a geologic repository will impose on the borosilicate glass waste form-groundwater interactions. These materials are the steel canister that holds the glass, the steel overpack over the canister, backfill materials that may be used, and last, the repository host rock. The repository geologies reviewed are tuff, salt, basalt, and granite. The interactions emphasized are those appropriate to conditions expected after repository closure, e.g., oxic vs anoxic conditions. Whenever possible, the effect of radiation from the waste form on the interaction(s) is examined. The interactions are evaluated based on their effect on the release and speciation of various elements including radionuclides from the glass. Repository relevant interactions testing that requires further study before long-term predictions can be made are noted. 62 refs

Glass to contain wastes

International Nuclear Information System (INIS)

Moncouyoux, M.; Jacquet-Francillon, M.

1994-01-01

Here are the tables and figures presented during the conference on the glass to confine high level radioactive wastes: definition, fabrication, storage and disposal. The composition of glasses are detailed, their properties and the vitrification proceeding. The behaviour of these glasses in front of water, irradiation and heat are shown. The characteristics of parcels are given according to the radiation protection rule, ALARA principle, the concept of multi-barriers and the geological stability

Preliminary evaluation of alternative forms for immobilization of Hanford high-level defense wastes

International Nuclear Information System (INIS)

Schulz, W.W.; Beary, M.M.; Gallagher, S.A.; Higley, B.A.; Johnston, R.G.; Jungfleisch, F.M.; Kupfer, M.J.; Palmer, R.A.; Watrous, R.A.; Wolf, G.A.

1980-09-01

A preliminary evaluation of solid waste forms for immobilization of Hanford high-level radioactive defense wastes is presented. Nineteen different waste forms were evaluated and compared to determine their applicability and suitability for immobilization of Hanford salt cake, sludge, and residual liquid. This assessment was structured to address waste forms/processes for several different leave-retrieve long-term Hanford waste management alternatives which give rise to four different generic fractions: (1) sludge plus long-lived radionuclide concentrate from salt cake and residual liquid; (2) blended wastes (salt cake plus sludge plus residual liquid); (3) residual liquid; and (4) radionuclide concentrate from residual liquid. Waste forms were evaluated and ranked on the basis of weighted ratings of seven waste form and seven process characteristics. Borosilicate Glass waste forms, as marbles or monoliths, rank among the first three choices for fixation of all Hanford high-level wastes (HLW). Supergrout Concrete (akin to Oak Ridge National Laboratory Hydrofracture Process concrete) and Bitumen, low-temperature waste forms, rate high for bulk disposal immobilization of high-sodium blended wastes and residual liquid. Certain multi-barrier (e.g., Coated Ceramic) and ceramic (SYNROC Ceramic, Tailored Ceramics, and Supercalcine Ceramic) waste forms, along with Borosilicate Glass, are rated as the most satisfactory forms in which to incorporate sludges and associated radionuclide concentrates. The Sol-Gel process appears superior to other processes for manufacture of a generic ceramic waste form for fixation of Hanford sludge. Appropriate recommendations for further research and development work on top ranking waste forms are made

Preliminary flowsheet for the conversion of Hanford high-level waste to glass

International Nuclear Information System (INIS)

Beary, M.M.; Chick, L.A.; Ely, P.C.; Gott, S.A.

1977-06-01

The flowsheets describe a process for converting waste removed from the Hanford underground waste tanks to more immobile form. The process involves a chemical separation of the radionuclides from industrial chemicals, and then making glass from the resulting small volume of highly radioactive waste. Removal of Sr, actinides, cesium, and technetium is discussed

Glass Formulation For The Hanford Tank Waste Treatment And Immobilization Plant (WTP)

International Nuclear Information System (INIS)

Kruger, A.A.; Jain, V.

2009-01-01

A computational method for formulating Hanford HLW glasses was developed that is based on empirical glass composition-property models, accounts for all associated uncertainties, and can be solved in Excel R in minutes. Calculations for all waste form processing and compliance requirements included. Limited experimental validation performed.

Standard test method for measuring waste glass or glass ceramic durability by vapor hydration test

CERN Document Server

American Society for Testing and Materials. Philadelphia

2009-01-01

1.1 The vapor hydration test method can be used to study the corrosion of a waste forms such as glasses and glass ceramics upon exposure to water vapor at elevated temperatures. In addition, the alteration phases that form can be used as indicators of those phases that may form under repository conditions. These tests; which allow altering of glass at high surface area to solution volume ratio; provide useful information regarding the alteration phases that are formed, the disposition of radioactive and hazardous components, and the alteration kinetics under the specific test conditions. This information may be used in performance assessment (McGrail et al, 2002 (1) for example). 1.2 This test method must be performed in accordance with all quality assurance requirements for acceptance of the data. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practice...

HIGH ALUMINUM HLW (HIGH LEVEL WASTE) GLASSES FOR HANFORD'S WTP (WASTE TREATMENT PROJECT)

International Nuclear Information System (INIS)

Kruger, A.A.; Bowan, B.W.; Joseph, I.; Gan, H.; Kot, W.K.; Matlack, K.S.; Pegg, I.L.

2010-01-01

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

Vitrified waste form performance modeling applied to the treatment and disposal of mixed-waste sludge at the Savannah River Site

International Nuclear Information System (INIS)

Whited, A.R.; Fjeld, R.A.

1998-01-01

Vitrification, the conversion of source components into a solid amorphous glass matrix, has emerged as a viable treatment technology for low-level radioactive waste and mixed waste. To dispose of vitrified low-level waste at US Department of Energy facilities, site-specific radiological performance assessments must be conducted to demonstrate that waste glass satisfies performance objectives for environmental protection. More than 2,500 m 3 of F0006-listed low-level mixed-waste sludge stored in the Reactor Materials Department (M-Area) at the Savannah River Site (SRS) is scheduled for vitrification. This study evaluates the feasibility of on-site disposal of vitrified M-Area waste at SRS. Laboratory leaching tests that accelerate the glass corrosion process are currently the best indicators of vitrified waste form durability. A method to incorporate laboratory leaching data into performance assessments is presented. A screening-level performance assessments code is used to model trench disposal of M-Area waste glass. The allowable leach rate for vitrified M-Area waste is determined based on both a maximum radiological dose equivalent of 4 mrem/yr for the drinking water pathway and a maximum uranium concentration of 20 microg/ell in groundwater. The allowable leach rate is compared with published long-term leaching data for a wide range of waste glass compositions and test conditions. This analysis demonstrates that trench disposal of the waste glass is likely to meet applicable performance objectives if the glass is of above average durability compared with the reference set of glasses

Chemical analysis of simulated high level waste glasses to support stage III sulfate solubility modeling

Energy Technology Data Exchange (ETDEWEB)

Fox, K. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-03-17

The U.S. Department of Energy (DOE), Office of Environmental Management (EM) is sponsoring an international, collaborative project to develop a fundamental model for sulfate solubility in nuclear waste glass. The solubility of sulfate has a significant impact on the achievable waste loading for nuclear waste forms within the DOE complex. These wastes can contain relatively high concentrations of sulfate, which has low solubility in borosilicate glass. This is a significant issue for low-activity waste (LAW) glass and is projected to have a major impact on the Hanford Tank Waste Treatment and Immobilization Plant (WTP). Sulfate solubility has also been a limiting factor for recent high level waste (HLW) sludge processed at the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF). The low solubility of sulfate in glass, along with melter and off-gas corrosion constraints, dictate that the waste be blended with lower sulfate concentration waste sources or washed to remove sulfate prior to vitrification. The development of enhanced borosilicate glass compositions with improved sulfate solubility will allow for higher waste loadings and accelerate mission completion.The objective of the current scope being pursued by SHU is to mature the sulfate solubility model to the point where it can be used to guide glass composition development for DWPF and WTP, allowing for enhanced waste loadings and waste throughput at these facilities. A series of targeted glass compositions was selected to resolve data gaps in the model and is identified as Stage III. SHU fabricated these glasses and sent samples to SRNL for chemical composition analysis. SHU will use the resulting data to enhance the sulfate solubility model and resolve any deficiencies. In this report, SRNL provides chemical analyses for the Stage III, simulated HLW glasses fabricated by SHU in support of the sulfate solubility model development.

Post-test evaluations of Waste Isolation Pilot Plant - Savannah River simulated defense HLW canisters and waste form

International Nuclear Information System (INIS)

Molecke, M.A.; Sorensen, N.R.; Harbour, J.R.; Ferrara, D.M.

1993-01-01

Eighteen nonradioactive defense high-level waste (DHLW) canisters were emplaced in and subjected to accelerated overtest thermal conditions for about three years at the bedded salt Waste Isolation Pilot Plant (WIPP) facility. Post-test laboratory corrosion results of several stainless steel 304L waste canisters, cast steel overpacks, and associated instruments ranged from negligible to moderate. We found appreciable surface corrosion and corrosion products on the cast steel overpacks. Pieces of both 304L and 316 stainless steel test apparatus underwent extensive stress-corrosion cracking failure and nonuniform attack. One of the retrieved test packages contained nonradioactive glass waste form from the Savannah River Site. We conducted post-test analyses of this glass to determine the degree of resultant glass fracturing, and whether any respirable fines were present. Linear glass fracture density ranged from about 1 to 8 fractures intersecting every 5 cm (2 inch) segment along a diameter line of the canister cross-section. Glass fines between 1 and 10 microns in diameter were detected, but were not quantified

Compositional Dependence of Solubility/Retention of Molybdenum Oxides in Aluminoborosilicate-Based Model Nuclear Waste Glasses.

Science.gov (United States)

Brehault, Antoine; Patil, Deepak; Kamat, Hrishikesh; Youngman, Randall E; Thirion, Lynn M; Mauro, John C; Corkhill, Claire L; McCloy, John S; Goel, Ashutosh

2018-02-08

Molybdenum oxides are an integral component of the high-level waste streams being generated from the nuclear reactors in several countries. Although borosilicate glass has been chosen as the baseline waste form by most of the countries to immobilize these waste streams, molybdate oxyanions (MoO 4 2- ) exhibit very low solubility (∼1 mol %) in these glass matrices. In the past three to four decades, several studies describing the compositional and structural dependence of molybdate anions in borosilicate and aluminoborosilicate glasses have been reported in the literature, providing a basis for our understanding of fundamental science that governs the solubility and retention of these species in the nuclear waste glasses. However, there are still several open questions that need to be answered to gain an in-depth understanding of the mechanisms that control the solubility and retention of these oxyanions in glassy waste forms. This article is focused on finding answers to two such questions: (1) What are the solubility and retention limits of MoO 3 in aluminoborosilicate glasses as a function of chemical composition? (2) Why is there a considerable increase in the solubility of MoO 3 with incorporation of rare-earth oxides (for example, Nd 2 O 3 ) in aluminoborosilicate glasses? Accordingly, three different series of aluminoborosilicate glasses (compositional complexity being added in a tiered approach) with varying MoO 3 concentrations have been synthesized and characterized for their ability to accommodate molybdate ions in their structure (solubility) and as a glass-ceramic (retention). The contradictory viewpoints (between different research groups) pertaining to the impact of rare-earth cations on the structure of aluminoborosilicate glasses are discussed, and their implications on the solubility of MoO 3 in these glasses are evaluated. A novel hypothesis explaining the mechanism governing the solubility of MoO 3 in rare-earth containing aluminoborosilicate

Effects of composition on waste glass properties

International Nuclear Information System (INIS)

Mellinger, G.B.; Chick, L.A.

1979-01-01

The electrical conductivity, viscosity, chemical durability, devitrification, and crystallinity of a defense waste glass were measured. Each oxide component in the glass was varied to determine its effect on these properties. A generic study is being developed which will determine the effects of 26 oxides on the above and additional properties of a wide field of possible waste glasses. 5 figures, 2 tables

Evaluation of conditioned high-level waste forms

International Nuclear Information System (INIS)

Mendel, J.E.; Turcotte, R.P.; Chikalla, T.D.; Hench, L.L.

1983-01-01

The evaluation of conditioned high-level waste forms requires an understanding of radiation and thermal effects, mechanical properties, volatility, and chemical durability. As a result of nuclear waste research and development programs in many countries, a good understanding of these factors is available for borosilicate glass containing high-level waste. The IAEA through its coordinated research program has contributed to this understanding. Methods used in the evaluation of conditioned high-level waste forms are reviewed. In the US, this evaluation has been facilitated by the definition of standard test methods by the Materials Characterization Center (MCC), which was established by the Department of Energy (DOE) in 1979. The DOE has also established a 20-member Materials Review Board to peer-review the activities of the MCC. In addition to comparing waste forms, testing must be done to evaluate the behavior of waste forms in geologic repositories. Such testing is complex; accelerated tests are required to predict expected behavior for thousands of years. The tests must be multicomponent tests to ensure that all potential interactions between waste form, canister/overpack and corrosion products, backfill, intruding ground water and the repository rock, are accounted for. An overview of the status of such multicomponent testing is presented

Reuse of ground waste glass as aggregate for mortars.

Science.gov (United States)

Corinaldesi, V; Gnappi, G; Moriconi, G; Montenero, A

2005-01-01

This work was aimed at studying the possibility of reusing waste glass from crushed containers and building demolition as aggregate for preparing mortars and concrete. At present, this kind of reuse is still not common due to the risk of alkali-silica reaction between the alkalis of cement and silica of the waste glass. This expansive reaction can cause great problems of cracking and, consequently, it can be extremely deleterious for the durability of mortar and concrete. However, data reported in the literature show that if the waste glass is finely ground, under 75mum, this effect does not occur and mortar durability is guaranteed. Therefore, in this work the possible reactivity of waste glass with the cement paste in mortars was verified, by varying the particle size of the finely ground waste glass. No reaction has been detected with particle size up to 100mum thus indicating the feasibility of the waste glass reuse as fine aggregate in mortars and concrete. In addition, waste glass seems to positively contribute to the mortar micro-structural properties resulting in an evident improvement of its mechanical performance.

Joule-heated glass-furnace system for the incineration of low-level radioactive wastes

International Nuclear Information System (INIS)

Armstrong, K.M.; Klingler, L.M.; Doty, J.W.; Kramer, D.P.

1982-01-01

For the past 1-1/2 years, Mound has been preparing and evaluating a commercially available joule-heated glass furnace unit, coupled with a wet scrubbing system. The purpose of the glass furnace evaluation is to advance and document incinerator technology for such combustibles as solids, resins, and sludges, and to develop a stable waste form for subsequent disposal. Four (4) waste nonradioactive types were selected to determine the combustion efficiency of the furnace unit: (1) dry solid waste composed of paper, plastics, rubber, and cloth, (2) ion exchange resin of both the anionic and cationic type, (3) filter sludge composed of diatomaceous earth, organic cellulosic filter aid, and powdered ion exchange resin, and (4) cartridge filters having glass and plastic filter surfaces and nonmetallic cores. When completed, the combustion efficiency experiments for the proposed nonradioactive waste-types revealed the ability of the furnace to easily incinerate waste at feedrates of up to 150 lb/hr. During the course of the experiments, combustibles in the offgas remained consistently low, suggesting excellent combustion efficiency. Furthermore, ash produced by the combustion process was effectively incorporated into the melt by convective currents in the glass. Future work on the glass furnace incinerator will include spiking the waste to determine radioisotope behavior in the furnace

Multibarrier waste forms. Part I. Development

International Nuclear Information System (INIS)

Rusin, J.M.; Lokken, R.O.; Lukacs, J.M.; Sump, K.R.; Browning, M.F.; McCarthy, G.J.

1978-09-01

The multibarrier concept produces a composite waste form with enhanced inertness through improvements in thermal stability, mechanical strength, and leachability by the use of coatings and metal matrices. This report describes research and development activities resulting in the demonstration of the multibarrier concept for nonradioactive simulated waste compositions. The multibarrier concept is to utilize up to three barriers to isolate radionuclides from the environment: a solid waste inner core, an impervious coating, and a metal matrix. Two inner core materials, sintered supercalcine and glass marbles, have been demonstrated. The coating barrier provides enhanced leach, impact, and oxidation resistance as well as thermal protection during encapsulation in the metal matrix. Py/Al 2 O 3 coatings deposited by chemical vapor deposition (CVD) and glass coatings have been applied to supercalcine cores to improve inertness. The purpose of the metal matrix is to improve impact resistance, protect the inner core rom any adverse environments, provide radiation shielding, and increase thermal conductivity, yielding lower internal temperatures. The development of gravity sintering and vacuum casting techniques for matrix encapsulation are discussed. Four multibarrier products were demonstrated: (1) Glass marbles encapsulated in vacuum-cast Pb-10Sn; (2) uncoated, sintered supercalcine pellets encapsulated in vacuum-cast Al-12Si; (3) glass-coated, sintered supercalcine pellets encapsulated in vacuum-cast Al-12Si; and (4) PyC/Al 2 O 3 -coated supercalcine encapsulated in gravity-sintered Cu. 23 figs., 20 tables

GLASS FORMULATION FOR THE HANFORD TANK WASTE TREATMENT AND IMMOBILIZATION PLANT (WTP)

Energy Technology Data Exchange (ETDEWEB)

KRUGER AA; VIENNA JD; KIM DS; JAIN V

2009-05-27

A computational method for formulating Hanford HLW glasses was developed that is based on empirical glass composition-property models, accounts for all associated uncertainties, and can be solved in Excel{sup R} in minutes. Calculations for all waste form processing and compliance requirements included. Limited experimental validation performed.

Production of highly porous glass-ceramics from metallurgical slag, fly ash and waste glass

OpenAIRE

Mangutova Bianka V.; Fidancevska Emilija M.; Milosevski Milosav I.; Bossert Joerg H.

2004-01-01

Glass-ceramics composites were produced based on fly-ash obtained from coal power stations, metallurgical slag from ferronickel industry and waste glass from TV monitors, windows and flasks. Using 50% waste flask glass in combination with fly ash and 20% waste glass from TV screens in combination with slag, E-modulus and bending strength values of the designed systems are increased (system based on fly ash: E-modulus from 6 to 29 GPa, and bending strength from 9 to 75 MPa). The polyurethane f...

Alternative Electrochemical Salt Waste Forms, Summary of FY/CY2011 Results

International Nuclear Information System (INIS)

Riley, Brian J.; McCloy, John S.; Crum, Jarrod V.; Rodriguez, Carmen P.; Windisch, Charles F.; Lepry, William C.; Matyas, Josef; Westman, Matthew P.; Rieck, Bennett T.; Lang, Jesse B.; Pierce, David A.

2011-01-01

This report summarizes the 2011 fiscal+calendar year efforts for developing waste forms for a spent salt generated in reprocessing nuclear fuel with an electrochemical separations process. The two waste forms are tellurite (TeO2-based) glasses and sol-gel-derived high-halide mineral analogs to stable minerals found in nature.

Alternative Electrochemical Salt Waste Forms, Summary of FY/CY2011 Results

Energy Technology Data Exchange (ETDEWEB)

Riley, Brian J.; McCloy, John S.; Crum, Jarrod V.; Rodriguez, Carmen P.; Windisch, Charles F.; Lepry, William C.; Matyas, Josef; Westman, Matthew P.; Rieck, Bennett T.; Lang, Jesse B.; Pierce, David A.

2011-12-01

This report summarizes the 2011 fiscal+calendar year efforts for developing waste forms for a spent salt generated in reprocessing nuclear fuel with an electrochemical separations process. The two waste forms are tellurite (TeO2-based) glasses and sol-gel-derived high-halide mineral analogs to stable minerals found in nature.

Process description and plant design for preparing ceramic high-level waste forms

International Nuclear Information System (INIS)

Grantham, L.F.; McKisson, R.L.; Guon, J.; Flintoff, J.F.; McKenzie, D.E.

1983-01-01

The ceramics process flow diagram has been simplified and upgraded to utilize only two major processing steps - fluid-bed calcination and hot isostatic press consolidating. Full-scale fluid-bed calcination has been used at INEL to calcine high-level waste for 18 y; and a second-generation calciner, a fully remotely operated and maintained calciner that meets ALARA guidelines, started calcining high-level waste in 1982. Full-scale hot isostatic consolidation has been used by DOE and commercial enterprises to consolidate radioactive components and to encapsulate spent fuel elements for several years. With further development aimed at process integration and parametric optimization, the operating knowledge of full-scale demonstration of the key process steps should be rapidly adaptable to scale-up of the ceramic process to full plant size. Process flowsheets used to prepare ceramic and glass waste forms from defense and commercial high-level liquid waste are described. Preliminary layouts of process flow diagrams in a high-level processing canyon were prepared and used to estimate the preliminary cost of the plant to fabricate both waste forms. The estimated costs for using both options were compared for total waste management costs of SRP high-level liquid waste. Using our design, for both the ceramic and glass plant, capital and operating costs are essentially the same for both defense and commercial wastes, but total waste management costs are calculated to be significantly less for defense wastes using the ceramic option. It is concluded from this and other studies that the ceramic form may offer important advantages over glass in leach resistance, waste loading, density, and process flexibility. Preliminary economic calculations indicate that ceramics must be considered a leading candidate for the form to immobilize high-level wastes

Preparation techniques for ceramic waste form powder

International Nuclear Information System (INIS)

Hash, M.C.; Pereira, C.; Lewis, M.A.

1997-01-01

The electrometallurgical treatment of spent nuclear fuels result in a chloride waste salt requiring geologic disposal. Argonne National Laboratory (ANL) is developing ceramic waste forms which can incorporate this waste. Currently, zeolite- or sodalite-glass composites are produced by hot isostatic pressing (HIP) techniques. Powder preparations include dehydration of the raw zeolite powders, hot blending of these zeolite powders and secondary additives. Various approaches are being pursued to achieve adequate mixing, and the resulting powders have been HIPed and characterized for leach resistance, phase equilibria, and physical integrity

Silica based gel as a potential waste form for high level waste from fuel reprocessing

International Nuclear Information System (INIS)

Ford, C.E.; Dempster, T.J.; Melling, P.J.

1983-10-01

To assess the feasibility of safe disposal of high-level radioactive waste as synthetic clay, or material that would react with ground water to form clay, experiments have been carried out to determine the hydrothermal crystallisation and leaching behaviour of silica based gels fired at 900 deg C. Crystallisation rates at a pressure of 500 bars and at temperatures below 400 deg C are negligible and this more or less precludes pre-disposal production of synthetic clay on the scale required. Leaching experiments suggest that the leach rates of Cs from gels by distilled water are higher than those of boro-silicate glasses and SYNROC at the lower temperatures that would be preferred for geological storage. However, amounts of bulk dissolution of gels may be lower than those of boro-silicate glasses. The initial leaching behaviour of gels might be considerably improved by hot compaction at 900 to 1000 deg C. Consideration of likely waste form dissolution behaviour in a repository environment suggests that gels of appropriate composition might perform as well as, or better than, boro-silicate glasses. A novel hypothetical plant is described that could produce the gel waste form on the scale required on a more or less continuous basis. (author)

Treatment of copper industry waste and production of sintered glass-ceramic.

Science.gov (United States)

Coruh, Semra; Ergun, Osman Nuri; Cheng, Ta-Wui

2006-06-01

Copper waste is iron-rich hazardous waste containing heavy metals such as Cu, Zn, Co, Pb. The results of leaching tests show that the concentration of these elements exceeds the Turkish and EPA regulatory limits. Consequently, this waste cannot be disposed of in its present form and therefore requires treatment to stabilize it or make it inert prior to disposal. Vitrification was selected as the technology for the treatment of the toxic waste under investigation. During the vitrification process significant amounts of the toxic organic and inorganic chemical compounds could be destroyed, and at the same time, the metal species are immobilized as they become an integral part of the glass matrix. The copper flotation waste samples used in this research were obtained from the Black Sea Copper Works of Samsun, Turkey. The samples were vitrified after being mixed with other inorganic waste and materials. The copper flotation waste and their glass-ceramic products were characterized by X-ray analysis (XRD), scanning electron microscopy and by the toxicity characteristic leaching procedure test. The products showed very good chemical durability. The glass-ceramics fabricated at 850 degrees C/2 h have a large application potential especially as construction and building materials.

Advanced High-Level Waste Glass Research and Development Plan

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-01

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

Thermodynamic model of natural, medieval and nuclear waste glass durability

International Nuclear Information System (INIS)

Jantzen, C.M.; Plodinec, M.J.

1983-01-01

A thermodynamic model of glass durability based on hydration of structural units has been applied to natural glass, medieval window glasses, and glasses containing nuclear waste. The relative durability predicted from the calculated thermodynamics correlates directly with the experimentally observed release of structural silicon in the leaching solution in short-term laboratory tests. By choosing natural glasses and ancient glasses whose long-term performance is known, and which bracket the durability of waste glasses, the long-term stability of nuclear waste glasses can be interpolated among these materials. The current Savannah River defense waste glass formulation is as durable as natural basalt from the Hanford Reservation (10 6 years old). The thermodynamic hydration energy is shown to be related to the bond energetics of the glass. 69 references, 2 figures, 1 table

Relationship between reaction layer thickness and leach rate for nuclear waste glasses

International Nuclear Information System (INIS)

Chick, L.A.; Pederson, L.R.

1984-02-01

Three leaching tests, devised to distinguish among several proposed nuclear waste glass leaching mechanisms, were carried out for four different waste glasses. In the first test, the influence of a pre-formed reaction layer on elemental release was evaluated. In the second test, glass specimens were replaced with fresh samples halfway through the leaching experiment, to evaluate the influence of the concentration of glass components in leaching. Finally, regular replacement of the leachant at fixed time intervals essentially removed the variable changing solution concentration, and allowed an assessment of the influence of reaction layer thickness on the leaching rate. Results for all glasses tested indicated that the reaction layer presented little or no barrier to leaching, and that most of the retardation on leaching rates generally observed are attributable to saturation effects. 20 references, 6 figures, 1 table

Eco-efficient waste glass recycling: Integrated waste management and green product development through LCA.

Science.gov (United States)

Blengini, Gian Andrea; Busto, Mirko; Fantoni, Moris; Fino, Debora

2012-05-01

As part of the EU Life + NOVEDI project, a new eco-efficient recycling route has been implemented to maximise resources and energy recovery from post-consumer waste glass, through integrated waste management and industrial production. Life cycle assessment (LCA) has been used to identify engineering solutions to sustainability during the development of green building products. The new process and the related LCA are framed within a meaningful case of industrial symbiosis, where multiple waste streams are utilised in a multi-output industrial process. The input is a mix of rejected waste glass from conventional container glass recycling and waste special glass such as monitor glass, bulbs and glass fibres. The green building product is a recycled foam glass (RFG) to be used in high efficiency thermally insulating and lightweight concrete. The environmental gains have been contrasted against induced impacts and improvements have been proposed. Recovered co-products, such as glass fragments/powders, plastics and metals, correspond to environmental gains that are higher than those related to landfill avoidance, whereas the latter is cancelled due to increased transportation distances. In accordance to an eco-efficiency principle, it has been highlighted that recourse to highly energy intensive recycling should be limited to waste that cannot be closed-loop recycled. Copyright © 2011 Elsevier Ltd. All rights reserved.