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Sample records for waste containment applications

  1. Poly-urea spray elastomer for waste containment applications

    International Nuclear Information System (INIS)

    Miller, C.J.; Cheng, S.C.J.; Tanis, R.

    1997-01-01

    Geomembrane usage in environmental applications has increased dramatically following the promulgation of federal regulations resulting from the Resource Conservation and Recovery Act of 1976 (RCRA). Subtitle D rules, formulated under the authority of RCRA, call for minimum performance standards to limit adverse effects of a solid waste disposal facility on human health or the environment (40 CFR 257,258, August 30, 1988). These rules set minimum standards requiring new landfill designs to include liner systems and final cover systems. Each state has the responsibility to develop rules that are at least as stringent as the Subtitle D rules. There are several types of geomembranes currently available for landfill applications, each offering particular advantages and disadvantages. For example, PVC does not show the yield point (point of instability) that HDPE shows, HDPE has a higher puncture resistance than PVC, and PVC will deform much more than HDPE before barrier properties of the geomembrane are lost. Because each geomembrane material exhibits its own particular characteristics the material selected should be chosen based on the individual project requirements. It is preferable to select a design that uses the least expensive material and meets the performance specifications of the project

  2. Waste container and method for containing waste

    International Nuclear Information System (INIS)

    Ono, Akira; Matsushita, Mitsuhiro; Doi, Makoto; Nakatani, Seiichi.

    1990-01-01

    In a waste container, water-proof membranes and rare earth element layers are formed on the inner surface of a steel plate concrete container in which steel plates are embedded. Further, rear earth element detectors are disposed each from the inner side of the steel plate concrete container by way of a pressure pipe to the outer side of the container. As a method for actually containing wastes, when a plurality of vessels in which wastes are fixed are collectively enhoused to the waste container, cussioning materials are attached to the inner surface of the container and wastes fixing containers are stacked successively in a plurality of rows in a bag made of elastic materials. Subsequently, fixing materials are filled and tightly sealed in the waste container. When the waste container thus constituted is buried underground, even if it should be deformed to cause intrusion of rain water to the inside of the container, the rare earth elements in the container dissolved in the rain water can be detected by the detectors, the containers are exchanged before the rain water intruding to the inner side is leached to the surrounding ground, to previously prevent the leakage of radioactive nuclides. (K.M.)

  3. Application of fuel cell for pyrite and heavy metal containing mining waste

    Science.gov (United States)

    Keum, H.; Ju, W. J.; Jho, E. H.; Nam, K.

    2015-12-01

    Once pyrite and heavy metal containing mining waste reacts with water and air it produces acid mine drainage (AMD) and leads to the other environmental problems such as contamination of surrounding soils. Pyrite is the major source of AMD and it can be controlled using a biological-electrochemical dissolution method. By enhancing the dissolution of pyrite using fuel cell technology, not only mining waste be beneficially utilized but also be treated at the same time by. As pyrite-containing mining waste is oxidized in the anode of the fuel cell, electrons and protons are generated, and electrons moves through an external load to cathode reducing oxygen to water while protons migrate to cathode through a proton exchange membrane. Iron-oxidizing bacteria such as Acidithiobacillus ferrooxidans, which can utilize Fe as an electron donor promotes pyrite dissolution and hence enhances electrochemical dissolution of pyrite from mining waste. In this study mining waste from a zinc mine in Korea containing 17 wt% pyrite and 9% As was utilized as a fuel for the fuel cell inoculated with A. ferrooxidans. Electrochemically dissolved As content and chemically dissolved As content was compared. With the initial pH of 3.5 at 23℃, the dissolved As concentration increased (from 4.0 to 13 mg/L after 20 d) in the fuel cell, while it kept decreased in the chemical reactor (from 12 to 0.43 mg/L after 20 d). The fuel cell produced 0.09 V of open circuit voltage with the maximum power density of 0.84 mW/m2. Dissolution of As from mining waste was enhanced through electrochemical reaction. Application of fuel cell technology is a novel treatment method for pyrite and heavy metals containing mining waste, and this method is beneficial for mining environment as well as local community of mining areas.

  4. WASTE CONTAINMENT OVERVIEW

    Science.gov (United States)

    BSE waste is derived from diseased animals such as BSE (bovine spongiform encepilopothy, also known as Mad Cow) in cattle and CWD (chronic wasting disease) in deer and elk. Landfilling is examined as a disposal option and this presentation introduces waste containment technology...

  5. Nonradioactive Air Emissions Notice of Construction (NOC) Application for the Central Waste Complex (CSC) for Storage of Vented Waste Containers

    International Nuclear Information System (INIS)

    KAMBERG, L.D.

    2000-01-01

    This Notice of Construction (NOC) application is submitted for the storage and management of waste containers at the Central Waste Complex (CWC) stationary source. The CWC stationary source consists of multiple sources of diffuse and fugitive emissions, as described herein. This NOC is submitted in accordance with the requirements of Washington Administrative Code (WAC) 173-400-110 (criteria pollutants) and 173-460-040 (toxic air pollutants), and pursuant to guidance provided by the Washington State Department of Ecology (Ecology). Transuranic (TRU) mixed waste containers at CWC are vented to preclude the build up of hydrogen produced as a result of radionuclide decay, not as safety pressure releases. The following activities are conducted within the CWC stationary source: Storage and inspection; Transfer and staging; Packaging; Treatment; and Sampling. This NOC application is intended to cover all existing storage structures within the current CWC treatment, storage, and/or disposal (TSD) boundary, as well as any storage structures, including waste storage pads and staging areas, that might be constructed in the future within the existing CWC boundary

  6. Radioactive waste processing container

    International Nuclear Information System (INIS)

    Ishizaki, Kanjiro; Koyanagi, Naoaki; Sakamoto, Hiroyuki; Uchida, Ikuo.

    1992-01-01

    A radioactive waste processing container used for processing radioactive wastes into solidification products suitable to disposal such as underground burying or ocean discarding is constituted by using cements. As the cements, calcium sulfoaluminate clinker mainly comprising calcium sulfoaluminate compound; 3CaO 3Al 2 O 3 CaSO 4 , Portland cement and aqueous blast furnace slug is used for instance. Calciumhydroxide formed from the Portland cement is consumed for hydration of the calcium sulfoaluminate clinker. According, calcium hydroxide is substantially eliminated in the cement constituent layer of the container. With such a constitution, damages such as crackings and peelings are less caused, to improve durability and safety. (I.N.)

  7. A review of the Hanford Site soil corrosion applicable to solid waste containers

    International Nuclear Information System (INIS)

    Divine, J.R.

    1991-05-01

    The first phase of the assessment of the soil corrosion in the solid waste burial grounds of the 200 Areas at the Hanford Site is completed with this review of both existing information developed at the site and relevant offsite information. Detailed soil corrosion data are needed for several reasons: (1) the possibility of predicting the damage to the containers of the retrievable stored transuranic waste that are under soil cover, (2) the feasibility of forecasting the state of waste containers being retrieved in remedial investigation/feasibility studies, (3) the capability of predicting subsidence of the soil over the waste containers, and (4) the capability of forecasting when stored lead shielding or hazardous chemicals might be exposed to the environment. Because corrosion in soils is dependent on the soil type, site-specific data are required even though offsite data can provide guidance on the type and the approximate extent of corrosion to expect. These data permit rough estimations of the corrosion rates of a variety of materials -- including carbon steels, cast irons, stainless steels, and lead -- in the Hanford Site soils. This report attempts to compile these data to facilitate current estimates of waste container longevity. However, because of the lack of well-documented, site-specific data, it is difficult to provide a definite life expectancy for waste containers and other structures. Consequently, additional data are essential for reliable container life estimates. 36 refs., 10 figs., 7 tabs

  8. Electrokinetic applications for environmental restoration, waste volume reduction, and contaminant containment systems

    International Nuclear Information System (INIS)

    Lomasney, H.L.; Lomasney, C.A.

    1996-01-01

    In the US and all over the world, following over 50 years of nuclear arms production operations, the magnitude of resultant environmental damage is only beginning to surface. The US Department of Energy estimates that by the year 2070, the total volume of high-level waste, transuranic waste, low-level waste, and low-level mixed waste, generated as a result of past and current nuclear activities, will exceed 20 million cubic meters. In Russia, it is reported that more than 30% of all groundwater is contaminated with agricultural and industrial chemical waste. Government agencies today are faced with the responsibility of developing technologies that are suitable for dealing with severe environmental contamination and accumulating waste inventories. In response to this demand, applications of electrokinetics have emerged in the field of environmental waste management as alternatives for environmental decontamination and ecological protection. Electrokinetics involves the movement of charged species under the influence of an applied electric field and is applicable in several areas of environmental waste management, including cleanup of soil and groundwater, barrier detection, and emergency or protective fencing. The worldwide interest in this technology has steadily escalated over the past decade. Today, state-of-the-art applications of electrokinetics have been demonstrated in the US, The Netherlands, Russia, The Ukraine, and India. This paper addresses the latest advances in the various applications of this technology as well as the most significant breakthroughs in the history of electrokinetics

  9. Application of PINS and GNAT to the assay of 55-gal containers of radioactive waste

    International Nuclear Information System (INIS)

    Gehrke, R.J.; Aryaeinejad, R.; Watts, K.D.; Staples, D.R.; Akers, D.W.

    1994-03-01

    The Portable Isotropic Neutron Spectroscopy (PINS) and Gamma Neutron Assay Technique (GNAT) assay systems that were developed with funding from the office of Research and Development (NN20), were taken to the Stored Waste Examination Pilot Plant (SWEPP) facility at the Radioactive Waste Management Complex (RWMC) and applied to the assay of surrogate and Rocky Flats Plant waste contained in 55-gal drums. PINS, a portable prompt γ neutron activation analysis technique, was able to identify key elements in both the surrogate and real waste so that three-main waste categories: metal, combustible material, and cemented chlorinated sludge wastes could be identified. GNAT, a γ, neutron assay technique for the identification and quantification of fissioning isotopes, was able to identify 240 Pu in surrogate waste in which nine 1-g nuclear accident dosimeters were inserted. GNAT was also able to identify 24O Pu in real 55-gal waste drums containing 15- and 40-g of plutonium even in the presence of high activity concentrations of 241 Am

  10. Radioactive waste sealing container

    International Nuclear Information System (INIS)

    Tozawa, S.; Kitamura, T.; Sugimoto, S.

    1984-01-01

    A low- to medium-level radioactive waste sealing container is constructed by depositing a foundation coating consisting essentially of zinc, cadmium or a zinc-aluminum alloy over a steel base, then coating an organic synthetic resin paint containing a metal phosphate over the foundation coating, and thereafter coating an acryl resin, epoxy resin, and/or polyurethane paint. The sealing container can consist of a main container body, a lid placed over the main body, and fixing members for clamping and fixing the lid to the main body. Each fixing member may consist of a material obtained by depositing a coating consisting essentially of cadmium or a zinc-aluminum alloy over a steel base

  11. Iron and aluminium oxides containing industrial wastes as adsorbents of heavy metals: Application possibilities and limitations.

    Science.gov (United States)

    Jacukowicz-Sobala, Irena; Ociński, Daniel; Kociołek-Balawejder, Elżbieta

    2015-07-01

    Industrial wastes with a high iron or aluminium oxide content are produced in huge quantities as by-products of water treatment (water treatment residuals), bauxite processing (red mud) and hard and brown coal burning in power plants (fly ash). Although they vary in their composition, the wastes have one thing in common--a high content of amorphous iron and/or aluminium oxides with a large specific surface area, whereby this group of wastes shows very good adsorbability towards heavy metals, arsenates, selenates, etc. But their physical form makes their utilisation quite difficult, since it is not easy to separate the spent sorbent from the solution and high bed hydraulic resistances occur in dynamic regime processes. Nevertheless, because of the potential benefits of utilising the wastes in industrial effluent treatment, this issue attracts much attention today. This study describes in detail the waste generation processes, the chemical structure of the wastes, their physicochemical properties, and the mechanisms of fixing heavy metals and semimetals on the surface of iron and aluminium oxides. Typical compositions of wastes generated in selected industrial plants are given. A detailed survey of the literature on the adsorption applications of the wastes, including methods of their thermal and chemical activation, as well as regeneration of the spent sorbents, is presented. The existing and potential ways of modifying the physical form of the discussed group of wastes, making it possible to overcome the basic limitation on their practical use, are discussed. © The Author(s) 2015.

  12. Plutonium waste container identification

    International Nuclear Information System (INIS)

    Schmierer, T.J.

    1979-01-01

    The purpose of this paper is to define the parameters of a method for identifying plutonium waste containers. This information will form the basis for a permanent committee to develop a complete identification program for use throughout the world. Although a large portion of the information will be on handwritten notebooks and may not be as extensive as is desired, it will all be helpful. The final information will be programmed into computer language and be available to all interested parties as well as a central control committee which will have the expertise to provide each government with advice on the packaging, storage, and measurement of the waste for which it is responsible. As time progresses, this central control committee should develop permanent storage sites and establish a system of records which will last for hundreds of years

  13. Application of biotechnology in management of industrial wastes containing toxic metals

    Energy Technology Data Exchange (ETDEWEB)

    Barton, L L; Fekete, F A; Huybrechts, M M.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Biology; Sillerud, L O [Los Alamos National Lab., NM (United States); Blacke, II, R C [Meharry Medical Coll., Nashville, TN (United States); Pigg, C J [Sandia National Labs., Albuquerque, NM (United States)

    1994-01-01

    The role of microorganisms in transformation and absorption of metals in the environment is examined in this article. Understanding the metabolic processes by which microorganisms interact with toxic metals is paramount for any bioremediation process dealing with restoration of toxic waste site. Bacteria, fungi, and yeast that displayed resistance to lead, mercury, or chromium were isolated from the environment. Cotolerance studies revealed that many of these organisms could grow in high concentrations of several different toxic elements. Transformation of chromium, mercury, and lead was displayed by means of the isolated bacterial strains. Data regarding the activities of these organisms can provide a basis for use of metal/tolerant organisms in bioremediation of toxic wastes containing mercury, chromium, and lead. (author) 1 fig., 7 tabs., 28 refs.

  14. Applicability of Activated Carbon to Treatment of Waste Containing Iodine-Labeled Compounds

    International Nuclear Information System (INIS)

    Gad, H.M.H.; El-Mouhty, N.R.A.; Aly, H.F.

    2008-01-01

    The applicability of activated carbon prepared from sawdust (SD) by one-step chemical activation process using H 3 PO 4 (H) to treatment of aqueous waste contaminated with iodine-labeled prolactin (I-PRL) has been investigated. Treatment processes were performed under the varying conditions; contact time, temperature, carbon type, carbon dosage, and different particle size of activated carbon (SDH). Effect of aqueous waste volume has been investigated to calculate the batch factor (V/M) and the distribution coefficient (K d ). The used activated carbon (SDH) was characterized by N 2 adsorption, FTIR, density, ph, point of zero charge ph p zc, moisture and ash content. Methylene blue (MB) and iodine number was calculated by adsorption from solution. In order to investigate the mechanism of sorption and potential rate controlling steps, pseudo first- and second-order equations, intra particle diffusion equation and the Elovich equation have been used to test experimental data. Kinetic analysis of the four models has been carried out for system variables in order to assess which model provides the best fit predicted data with experimental results. 7 M NaOH can be used for regeneration of spent SDH activated carbon with the efficiency of 99.6% and the regenerated carbon can be reused for five cycles effectively. The prospect of applying the SDH activated carbon prepared from agricultural by-product, sawdust, to treatment of aqueous waste contaminated with I-PRL appears promising and is considered highly applicable because of its high adsorption capacity, available at low cost, easily regenerated and reused

  15. Investigation into the application of polyetherimide to nuclear waste storage containers

    Energy Technology Data Exchange (ETDEWEB)

    Saboui, Y.; Bonin, H.W.; Bui, V.T. [Royal Military College of Canada, Kingston, Ontario (Canada)

    2009-07-01

    The procedure of the analysis of the effects of irradiation on the mechanical and chemical properties of the polyetherimide (PEI) is outlined. Previous research in this field at the Royal Military College of Canada is presented. Samples of PEI will be exposed to a mixed radiation field, in the pool of a SLOWPOKE-2 nuclear reactor, then changes in mechanical properties, degradation product formation, and physical property changes will be assessed. Additionally, the heat transfer in the sample will be calculated in order to model the heat transfer rate and heat diffusion profile of PEI. The purpose of the proposed research is to determine the feasibility of using PEI for spent CANDU nuclear fuel and nuclear waste storage containers. (author)

  16. Investigation into the application of polyetherimide to nuclear waste storage containers

    Energy Technology Data Exchange (ETDEWEB)

    Saboui, Y.; Bonin, H.W.; Bui, V.T. [Royal Military College, Kingston, Ontario (Canada)

    2010-07-01

    The procedure of the analysis of the effects of irradiation on the mechanical and chemical properties of the polyetherimide (PEI) is outlined. Previous research in this field at the Royal Military College of Canada is presented. Samples of PEI will be exposed to a mixed radiation field, in the pool of a SLOWPOKE-2 nuclear reactor, then changes in mechanical properties, degradation product formation, and physical property changes will be assessed. Additionally, the heat transfer in the sample will be calculated in order to model the heat transfer rate and heat diffusion profile of PEI. The purpose of the proposed research is to determine the feasibility of using PEI for spent CANDU nuclear fuel and nuclear waste storage containers. (author)

  17. Investigation into the application of polyetherimide to nuclear waste storage containers

    International Nuclear Information System (INIS)

    Saboui, Y.; Bonin, H.W.; Bui, V.T.

    2009-01-01

    The procedure of the analysis of the effects of irradiation on the mechanical and chemical properties of the polyetherimide (PEI) is outlined. Previous research in this field at the Royal Military College of Canada is presented. Samples of PEI will be exposed to a mixed radiation field, in the pool of a SLOWPOKE-2 nuclear reactor, then changes in mechanical properties, degradation product formation, and physical property changes will be assessed. Additionally, the heat transfer in the sample will be calculated in order to model the heat transfer rate and heat diffusion profile of PEI. The purpose of the proposed research is to determine the feasibility of using PEI for spent CANDU nuclear fuel and nuclear waste storage containers. (author)

  18. 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

  19. Airborne microorganisms from waste containers.

    Science.gov (United States)

    Jedlicka, Sabrina S; Stravitz, David M; Lyman, Charles E

    2012-01-01

    In physician's offices and biomedical labs, biological waste is handled every day. This waste is disposed of in waste containers designed for holding red autoclave bags. The containers used in these environments are closed hands-free containers, often with a step pedal. While these containers protect the user from surface-borne microorganisms, the containers may allow airborne microorganisms to escape via the open/close mechanism because of the air current produced upon open/close cycles. In this study, the air current was shown to be sufficient to allow airborne escape of microorganisms held in the container, including Aspergillus niger. However, bacterial cultures, such as Escherichia coli and Lactococcus lactis did not escape. This may be due to the choice of bacterial cultures and the absence of solid waste, such as dust or other particulate matter in the waste containers, that such strains of bacteria could travel on during aerosolization. We compared these results to those obtained using a re-designed receptacle, which mimimizes air currents, and detected no escaping microorganisms. This study highlights one potential source of airborne contamination in labs, hospitals, and other environments that dispose of biological waste.

  20. Corrosion studies on selected metallic materials for application in nuclear waste disposal containers

    International Nuclear Information System (INIS)

    Smailos, E.; Fiehn, B.; Gago, J.A.; Azkarate, I.

    1994-03-01

    In previous corrosion studies, carbon steels and the alloy Ti 99.8-Pd were identified as promising materials for heat-generating nuclear waste containers acting as a radionuclide barrier in a rock-salt repository. To characterize the long-term corrosion behaviour of these materials in more detail, a research programme including laboratory-scale and in-situ corrosion studies has been undertaken jointly by KfK and ENRESA/INASMET. In the period under review, gamma irradiation corrosion studies of up to about 6 months at 10 Gy/h and stress corrosion cracking studies at slow strain rates (10 -4 -10 -7 s -1 ) were performed on three preselected carbon steels in disposal relevant brines (NaCl-rich, MgCl 2 -rich) at 90 C and 150 C (TStE 355, TStE 460, 15 MnNi 6.3). Moreover, results were obtained from long-term in-situ corrosion studies (maximum test duration 9 years) conducted on carbon steel, Ti 99.8-Pd, Hastelloy C4, Ni-resist D4, and Si-cast iron in boreholes in the Asse salt mine. (orig./MM) [de

  1. Corrosion on reinforced concrete structures. An application for the intermediate level radioactive waste container

    International Nuclear Information System (INIS)

    Arva, Alejandro; Alvarez, Marta G.; Duffo, Gustavo S.

    2003-01-01

    The behavior of steel reinforcement bars (rebars) for a high performance reinforced concrete made of sulfate resistant portland cement was evaluated from the rebars corrosion point of view. The results from the present work will be used to evaluate the materials properties to be used in the construction of the intermediate level radioactive waste disposal containers. The study is carried out evaluating the incidence of chloride and sulfate ions, as well as, concrete carbonation in the rebar corrosion process. The electrochemical parameters that characterize the corrosion process (corrosion potential [E corr ], polarisation resistance [Rp] and concrete electrical resistivity [ρ]) were monitored on specially designed reinforced concrete specimens. The results up to date (about 1000 days of exposure) reveal that the concrete under study provides to the steel reinforcement bars of a passive state against corrosion under the test conditions. An increasing tendency as a function of time of ρ is observed that corroborates the continuous curing process of concrete. The chloride and carbonation diffusion coefficients were also determined, and their values are comparable with those of high quality concrete. (author)

  2. Radioactive waste containment

    International Nuclear Information System (INIS)

    Beranger, J.-C.

    1978-01-01

    The problem of confining the radioactive wastes produced from the nuclear industry, after the ore concentration stage, is envisaged. These residues being not released into the environment are to be stored. The management policy consists in classifying them in view of adapting to each type of treatment, the suitable conditioning and storage. This classification is made with taking account of the following data: radioactivity (weak, medium or high) nature and lifetime of this radioactivity (transuranians) physical nature and volume. The principles retained are those of volume reduction and shaping into insoluble solids (vitrification) [fr

  3. Treatment of mercury containing waste

    Science.gov (United States)

    Kalb, Paul D.; Melamed, Dan; Patel, Bhavesh R; Fuhrmann, Mark

    2002-01-01

    A process is provided for the treatment of mercury containing waste in a single reaction vessel which includes a) stabilizing the waste with sulfur polymer cement under an inert atmosphere to form a resulting mixture and b) encapsulating the resulting mixture by heating the mixture to form a molten product and casting the molten product as a monolithic final waste form. Additional sulfur polymer cement can be added in the encapsulation step if needed, and a stabilizing additive can be added in the process to improve the leaching properties of the waste form.

  4. Methane from waste containing paper

    Energy Technology Data Exchange (ETDEWEB)

    1981-12-24

    Waste solids containing paper are biologically treated in a system by: fermentation with lactobacilli, separation of the solids, ion exchange of the supernatant from the separation, anaerobic digestion of the ion-exchanged liquor, separation of a liquor from the fermentation, and digestion of the liquor. Thus, a municipal waste containing paper and water was inoculated with Aspergillus niger and lactobacilli for 2 days; the mixture was anaerobically treated and centrifuged; the clear liquor was ion exchanged; and the solid waste was filter pressed. The filter cake was treated with Trichoderma nigricaus and filtered. The filtrate and the ion-exchanged liquor were digested for CH/sub 4/ production.

  5. Beverage containers in municipal waste

    Energy Technology Data Exchange (ETDEWEB)

    Enhoerning, B

    1979-01-01

    The composition of containers (cans, glass) in Sweden's wastes is given. Recycling and reclamation of these containers are discussed. The energy demand of fabricating the containers is analyzed for recycling rates of 0 and 100%. The free forces of the market cannot be depended on to direct the containers back to the manufacturer; only a well-functioning deposit system can do this. 7 figures. (DLC)

  6. Properties of radioactive wastes and waste containers

    International Nuclear Information System (INIS)

    Arora, H.S.; Dayal, R.

    1984-01-01

    Major tasks in this NRC-sponsored program include: (1) an evaluation of the acceptability of low-level solidified wastes with respect to minimizing radionuclide releases after burial; and (2) an assessment of the influence of pertinent environmental stresses on the performance of high-integrity radwaste container (HIC) materials. The waste form performance task involves studies on small-scale laboratory specimens to predict and extrapolate: (1) leachability for extended time periods; (2) leach behavior of full-size forms; (3) performance of waste forms under realistic leaching conditions; and (4) leachability of solidified reactor wastes. The results show that leach data derived from testing of small-scale specimens can be extrapolated to estimate leachability of a full-scale specimen and that radionuclide release data derived from testing of simulants can be employed to predict the release behavior of reactor wastes. Leaching under partially saturated conditions exhibits lower releases of radionuclides than those observed under the conventional IAEA-type or ANS 16.1 leach tests. The HIC assessment task includes the characterization of mechanical properties of Marlex CL-100, a candidate radwaste high density polyethylene material. Tensile strength and creep rupture tests have been carried out to determine the influence of specific waste constituents as well as gamma irradiation on material performance. Emphasis in ongoing tests is being placed on studying creep rupture while the specimens are in contact with a variety of chemicals including radiolytic by-products of irradiated resin wastes. 12 references 6 figures, 2 tables

  7. Polymeric radioactive waste disposal containers: an investigation into the application of polymers vice metals to house low and intermediate level radioactive waste

    International Nuclear Information System (INIS)

    Walker, M.W.; Bonin, H.W.; Bui, V.T.

    2001-01-01

    The research carried out in Canada in the design of containers for the disposal of radioactive waste has focussed on spent nuclear fuel, even though the quantities of other currently stored radioactive wastes are substantially greater. Research carried out at the Royal Military College of Canada on the effects of mixed fields of radiation on high polymer adhesives and composite materials has shown that some polymers are quite resistant to radiation and could well serve in the fabrication of radioactive waste disposal containers. The purpose of this research was to determine if thermoplastic polymers could be used as superior materials to replace metals in the application of low and intermediate level radioactive waste disposal containers. Polymers have the advantage that they do not corrode like metals. The experimental methods, used in this research, focused on the effects of radiation on the properties of the materials. Polypropylene, Nylon 66, Polycarbonate, and Polyurethane, with and without glass fibre reinforcement, were studied. The method involved irradiating injection moulded tensile test bars with the SLOWPOKE-2 reactor to accumulate doses ranging from 0.5 to 3.0 MGy. To determine the effects of the various doses on the materials, density, tensile, differential scanning calorimetry, and scanning electron microscopy tests were run. For each polymer, the test methods supported predominant crosslinking of polymeric chains severed by radiation. This was evident from observed changes in the mechanical and chemical properties of the polymers, typical of crosslinking. The mechanical changes included an overall increase in density, an increase in Young's modulus, a decrease in strain at break, and only minor changes in strength. The chemical changes included differences in chemical transition temperatures characteristic of radiation damage. The test methods also evidenced minor radiation degradation at the fibre/matrix interfaces in the glass fibre reinforced

  8. Polymeric radioactive waste disposal containers: an investigation into the application of polymers vice metals to house low and intermediate level radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Walker, M.W.; Bonin, H.W.; Bui, V.T. [Royal Military College of Canada, Kingston, Ontario (Canada)

    2001-07-01

    The research carried out in Canada in the design of containers for the disposal of radioactive waste has focussed on spent nuclear fuel, even though the quantities of other currently stored radioactive wastes are substantially greater. Research carried out at the Royal Military College of Canada on the effects of mixed fields of radiation on high polymer adhesives and composite materials has shown that some polymers are quite resistant to radiation and could well serve in the fabrication of radioactive waste disposal containers. The purpose of this research was to determine if thermoplastic polymers could be used as superior materials to replace metals in the application of low and intermediate level radioactive waste disposal containers. Polymers have the advantage that they do not corrode like metals. The experimental methods, used in this research, focused on the effects of radiation on the properties of the materials. Polypropylene, Nylon 66, Polycarbonate, and Polyurethane, with and without glass fibre reinforcement, were studied. The method involved irradiating injection moulded tensile test bars with the SLOWPOKE-2 reactor to accumulate doses ranging from 0.5 to 3.0 MGy. To determine the effects of the various doses on the materials, density, tensile, differential scanning calorimetry, and scanning electron microscopy tests were run. For each polymer, the test methods supported predominant crosslinking of polymeric chains severed by radiation. This was evident from observed changes in the mechanical and chemical properties of the polymers, typical of crosslinking. The mechanical changes included an overall increase in density, an increase in Young's modulus, a decrease in strain at break, and only minor changes in strength. The chemical changes included differences in chemical transition temperatures characteristic of radiation damage. The test methods also evidenced minor radiation degradation at the fibre/matrix interfaces in the glass fibre

  9. 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)

  10. Nuclear waste storage container with metal matrix

    International Nuclear Information System (INIS)

    Sump, K.R.

    1978-01-01

    The invention relates to a storage container for high-level waste having a metal matrix for the high-level waste, thereby providing greater impact strength for the waste container and increasing heat transfer properties

  11. Nuclear waste storage container with metal matrix

    Science.gov (United States)

    Sump, Kenneth R.

    1978-01-01

    The invention relates to a storage container for high-level waste having a metal matrix for the high-level waste, thereby providing greater impact strength for the waste container and increasing heat transfer properties.

  12. Method and article for primary containment of cesium wastes. [DOE patent application

    Science.gov (United States)

    Angelini, P.; Lackey, W.J.; Stinton, D.P.; Blanco, R.E.; Bond, W.D.; Arnold, W.D. Jr.

    1981-09-03

    A method for producing a cesium-retentive waste form, characterized by a high degree of compositional stability and mechanical integrity, is provided by subjecting a cesium-loaded zeolite to heat under conditions suitable for stabilizing the zeolite and immobilizing the cesium, and coating said zeolite for sufficient duration within a suitable environment with at least one dense layer of pyrolytic carbon to seal therein said cesium to produce a final, cesium-bearing waste form. Typically, the zolite is stabilized and the cesium immobilized in less than four hours by confinement within an air environment maintained at about 600/sup 0/C. Coatings are thereafter applied by confining the calcined zeolite within a coating environment comprising inert fluidizing and carbon donor gases maintained at 1000/sup 0/C for a suitable duration.

  13. Storage container for radioactive wastes

    International Nuclear Information System (INIS)

    Catalayoud, L.; Gerard, M.

    1990-01-01

    Tightness, shock resistance and corrosion resistance of containers for storage of radioactive wastes it obtained by complete fabrication with concrete reinforced with metal fibers. This material is used for molding the cask, the cover and the joint connecting both parts. Dovetail grooves are provided on the cask and the cover for the closure [fr

  14. Application of autonomous robotics to surveillance of waste storage containers for radioactive surface contamination

    International Nuclear Information System (INIS)

    Sweeney, F.J.; Beckerman, M.; Butler, P.L.; Jones, J.P.; Reister, D.B.

    1991-01-01

    This paper describes a proof-of-principal demonstration performed with the HERMIES-III mobile robot to automate the inspection of waste storage drums for radioactive surface contamination and thereby reduce the human burden of operating a robot and worker exposure to potentially hazardous environments. Software and hardware for the demonstration were developed by a team consisting of Oak Ridge National Laboratory, and the Universities of Florida, Michigan, Tennessee, and Texas. Robot navigation, machine vision, manipulator control, parallel processing and human-machine interface techniques developed by the team were demonstrated utilizing advanced computer architectures. The demonstration consists of over 100,000 lines of computer code executing on nine computers

  15. Application of a passive electrochemical noise technique to localized corrosion of candidate radioactive waste container materials

    International Nuclear Information System (INIS)

    Korzan, M.A.

    1994-05-01

    One of the key engineered barriers in the design of the proposed Yucca Mountain repository is the waste canister that encapsulates the spent fuel elements. Current candidate metals for the canisters to be emplaced at Yucca Mountain include cast iron, carbon steel, Incoloy 825 and titanium code-12. This project was designed to evaluate passive electrochemical noise techniques for measuring pitting and corrosion characteristics of candidate materials under prototypical repository conditions. Experimental techniques were also developed and optimized for measurements in a radiation environment. These techniques provide a new method for understanding material response to environmental effects (i.e., gamma radiation, temperature, solution chemistry) through the measurement of electrochemical noise generated during the corrosion of the metal surface. In addition, because of the passive nature of the measurement the technique could offer a means of in-situ monitoring of barrier performance

  16. Mass transfer from penetrations in waste containers

    International Nuclear Information System (INIS)

    Pescatore, C.; Sastre, C.

    1987-01-01

    Recent studies have indicated that localized corrosion of a relatively small area of a waste container may impair the containment function to such an extent that larger releases may be possible than from the bare waste form. This would take place when a large number of holes coexist on the container while their concentration fields do not interact significantly with each other. After performing a steady state analysis of the release from a hole, it is shown that much fewer independent holes can coexist on a container surface than previously estimated. The calculated radionuclide release from multiple independent holes must be changed accordingly. Previous analyses did not proceed to a correct application of the linear superposition principle. This resulted in unacceptable physical conclusions and undue strain on the performance assessment necessary for a container licensing procedure. The paper also analyzes the steady state release from penetrations of finite length and whose concentration fields interact with one another. The predicted release from these penetrations is lower than the previously calculated release from holes of zero thickness. It is concluded here that the steady-state release from multiple holes on a waste container can not exceed the release from the bare waste form and that multiple perforations need not be a serious liability to container performance. 8 refs., 3 figs., 1 tab

  17. Generalized waste package containment model

    International Nuclear Information System (INIS)

    Liebetrau, A.M.; Apted, M.J.

    1985-02-01

    The US Department of Energy (DOE) is developing a performance assessment strategy to demonstrate compliance with standards and technical requirements of the Environmental Protection Agency (EPA) and the Nuclear Regulatory Commission (NRC) for the permanent disposal of high-level nuclear wastes in geologic repositories. One aspect of this strategy is the development of a unified performance model of the entire geologic repository system. Details of a generalized waste package containment (WPC) model and its relationship with other components of an overall repository model are presented in this paper. The WPC model provides stochastically determined estimates of the distributions of times-to-failure of the barriers of a waste package by various corrosion mechanisms and degradation processes. The model consists of a series of modules which employ various combinations of stochastic (probabilistic) and mechanistic process models, and which are individually designed to reflect the current state of knowledge. The WPC model is designed not only to take account of various site-specific conditions and processes, but also to deal with a wide range of site, repository, and waste package configurations. 11 refs., 3 figs., 2 tabs

  18. 1994 Solid waste forecast container volume summary

    International Nuclear Information System (INIS)

    Templeton, K.J.; Clary, J.L.

    1994-09-01

    This report describes a 30-year forecast of the solid waste volumes by container type. The volumes described are low-level mixed waste (LLMW) and transuranic/transuranic mixed (TRU/TRUM) waste. These volumes and their associated container types will be generated or received at the US Department of Energy Hanford Site for storage, treatment, and disposal at Westinghouse Hanford Company's Solid Waste Operations Complex (SWOC) during a 30-year period from FY 1994 through FY 2023. The forecast data for the 30-year period indicates that approximately 307,150 m 3 of LLMW and TRU/TRUM waste will be managed by the SWOC. The main container type for this waste is 55-gallon drums, which will be used to ship 36% of the LLMW and TRU/TRUM waste. The main waste generator forecasting the use of 55-gallon drums is Past Practice Remediation. This waste will be generated by the Environmental Restoration Program during remediation of Hanford's past practice sites. Although Past Practice Remediation is the primary generator of 55-gallon drums, most waste generators are planning to ship some percentage of their waste in 55-gallon drums. Long-length equipment containers (LECs) are forecasted to contain 32% of the LLMW and TRU/TRUM waste. The main waste generator forecasting the use of LECs is the Long-Length Equipment waste generator, which is responsible for retrieving contaminated long-length equipment from the tank farms. Boxes are forecasted to contain 21% of the waste. These containers are primarily forecasted for use by the Environmental Restoration Operations--D ampersand D of Surplus Facilities waste generator. This waste generator is responsible for the solid waste generated during decontamination and decommissioning (D ampersand D) of the facilities currently on the Surplus Facilities Program Plan. The remaining LLMW and TRU/TRUM waste volume is planned to be shipped in casks and other miscellaneous containers

  19. Decontamination of organic wastes containing radionuclides

    International Nuclear Information System (INIS)

    Unsworth, T.J.; Pimblott, S.M.; Brown, N.W.

    2015-01-01

    An electrochemical oxidation treatment has been developed by Arvia Technology for organic wastes containing radionuclides, in which GIC-bisulphate is used as an adsorbent and electrode. Significant work has been carried out in the irradiation of graphite for medical and nuclear applications and in the use of carbonaceous adsorbents but knowledge of the applicability of graphite intercalation compounds (GICs) in these roles is limited. This project will attempt to fill this gap. It will investigate the suitability of GIC-bisulphate as an adsorbent in an electrochemical treatment process for radioactive organic liquids. The process was initially used to treat waste-water from non-nuclear operations and now requires technical knowledge and research to adapt the treatment for the nuclear industry. Adsorption processes involving organic wastes containing mobile radionuclides such as 137 Cs are difficult to understand. The effects of gamma radiation on the chemistry of water and organics could complicate the treatment process further. To ensure the suitability and effectiveness of the electrochemical oxidation treatment for radioactive organic wastes, the following effects are being investigated: -) radiolytic degradation of GIC-bisulphate in solution, -) leaching of intercalated ions due to gamma radiation, -) effect of gamma radiation on the adsorption of organics by GIC-bisulphate, -) changes in the sorption behaviour of radioactive contaminants, -) distribution coefficients of contaminants in organic and aqueous phases, and -) selective or competitive adsorption on graphite surface sites

  20. High integrity container evaluation for solid waste disposal burial containers

    International Nuclear Information System (INIS)

    Josephson, W.S.

    1996-01-01

    In order to provide radioactive waste disposal practices with the greatest measure of public protection, Solid Waste Disposal (SWD) adopted the Nuclear Regulatory Commission (NRC) requirement to stabilize high specific activity radioactive waste prior to disposal. Under NRC guidelines, stability may be provided by several mechanisms, one of which is by placing the waste in a high integrity container (HIC). During the implementation process, SWD found that commercially-available HICs could not accommodate the varied nature of weapons complex waste, and in response developed a number of disposal containers to function as HICs. This document summarizes the evaluation of various containers that can be used for the disposal of Category 3 waste in the Low Level Burial Grounds. These containers include the VECTRA reinforced concrete HIC, reinforced concrete culvert, and the reinforced concrete vault. This evaluation provides justification for the use of these containers and identifies the conditions for use of each

  1. Method for processing radioactive wastes containing sodium

    International Nuclear Information System (INIS)

    Kubota, Takeshi.

    1975-01-01

    Object: To bake, solidify and process even radioactive wastes highly containing sodium. Structure: H and or NH 4 zeolites of more than 90g per chemical equivalent of sodium present in the waste is added to and left in radioactive wastes containing sodium, after which they are fed to a baker such as rotary cylindrical baker, spray baker and the like to bake and solidify the wastes at 350 to 800 0 C. Thereby, it is possible to bake and solidify even radioactive wastes highly containing sodium, which has been impossible to do so previously. (Kamimura, M.)

  2. Buried waste containment system materials. Final Report

    International Nuclear Information System (INIS)

    Weidner, J.R.; Shaw, P.G.

    1997-10-01

    This report describes the results of a test program to validate the application of a latex-modified cement formulation for use with the Buried Waste Containment System (BWCS) process during a proof of principle (POP) demonstration. The test program included three objectives. One objective was to validate the barrier material mix formulation to be used with the BWCS equipment. A basic mix formula for initial trials was supplied by the cement and latex vendors. The suitability of the material for BWCS application was verified by laboratory testing at the Idaho National Engineering and Environmental Laboratory (INEEL). A second objective was to determine if the POP BWCS material emplacement process adversely affected the barrier material properties. This objective was met by measuring and comparing properties of material prepared in the INEEL Materials Testing Laboratory (MTL) with identical properties of material produced by the BWCS field tests. These measurements included hydraulic conductivity to determine if the material met the US Environmental Protection Agency (EPA) requirements for barriers used for hazardous waste sites, petrographic analysis to allow an assessment of barrier material separation and segregation during emplacement, and a set of mechanical property tests typical of concrete characterization. The third objective was to measure the hydraulic properties of barrier material containing a stop-start joint to determine if such a feature would meet the EPA requirements for hazardous waste site barriers

  3. Properties of radioactive wastes and waste containers

    International Nuclear Information System (INIS)

    Morcos, N.; Dayal, R.

    1982-01-01

    This program is sponsored by the Nuclear Regulatory Commission to address basic concerns in assessing the performance of solidified radwaste. Experiments were initiated to address these concerns. In particular, leachability of solidified radwastes and the physical stability of the ensuing waste forms were evaluated. In addition, leaching experiments designed to address the effects of alternating wet/dry cycles and of varying the length of these cycles on the leach behavior of waste forms were initiated

  4. Containment and stabilization technologies for mixed hazardous and radioactive wastes

    International Nuclear Information System (INIS)

    Buelt, J.L.

    1993-05-01

    A prevalent approach to the cleanup of waste sites contaminated with hazardous chemicals and radionuclides is to contain and/or stabilize wastes within the site. Stabilization involves treating the wastes in some fashion, either in situ or above ground after retrieval, to reduce the leachability and release rate of waste constituents to the environment. This approach is generally reserved for radionuclide contaminants, inorganic hazardous contaminants such as heavy metals, and nonvolatile organic contaminants. This paper describes the recent developments in the technical options available for containing and stabilizing wastes. A brief description of each technology is given along with a discussion of the most recent developments and examples of useful applications

  5. Solid waste containing method and solid waste container

    International Nuclear Information System (INIS)

    Sawai, Takeshi.

    1997-01-01

    Solid wastes are filled in a sealed vessel, and support spacers are inserted to the gap between the inner wall of a vessel main body and the solid wastes. The solid wastes comprise shorn pieces (crushed pieces) of spent fuel rod cladding tubes, radioactively contaminated metal pieces and miscellaneous solids pressed into a disk-like shape. The sealed vessel comprises, for example, a stainless steel. The solid wastes are filled while being stacked in a plurality of stages. A solidifying filler is filled into the gap between the inner wall and the solid wastes in the vessel main body by way of an upper opening, and the upper opening is closed by a closing lid to provide an entirely sealed state. Alumina particles having high heat conductivity and excellent heat durability are used for the solid filler. It is preferable to fill an inert gas such as a dried nitrogen gas in the sealed vessel. (I.N.)

  6. Cermets for high level waste containment

    International Nuclear Information System (INIS)

    Aaron, W.S.; Quinby, T.C.; Kobisk, E.H.

    1978-01-01

    Cermet materials are currently under investigation as an alternate for the primary containment of high level wastes. The cermet in this study is an iron--nickel base metal matrix containing uniformly dispersed, micron-size fission product oxides, aluminosilicates, and titanates. Cermets possess high thermal conductivity, and typical waste loading of 70 wt % with volume reduction factors of 2 to 200 and low processing volatility losses have been realized. Preliminary leach studies indicate a leach resistance comparable to other candidate waste forms; however, more quantitative data are required. Actual waste studies have begun on NFS Acid Thorex, SRP dried sludge and fresh, unneutralized SRP process wastes

  7. Containers for packaging of solid and intermediate level radioactive wastes

    International Nuclear Information System (INIS)

    1993-01-01

    Low and intermediate level radioactive wastes are generated at all stages in the nuclear fuel cycle and also from the medical, industrial and research applications of radiation. These wastes can potentially present risks to health and the environment if they are not managed adequately. Their effective management will require the wastes to be safely stored, transported and ultimately disposed of. The waste container, which may be defined as any vessel, drum or box, made from metals, concrete, polymers or composite materials, in which the waste form is placed for interim storage, for transport and/or for final disposal, is an integral part of the whole package for the management of low and intermediate level wastes. It has key roles to play in several stages of the waste management process, starting from the storage of raw wastes and ending with the disposal of conditioned wastes. This report provides an overview of the various roles that a container may play and the factors that are important in each of these roles. This report has two main objectives. The first is to review the main requirements for the design of waste containers. The second is to provide advice on the design, fabrication and handling of different types of containers used in the management of low and intermediate level radioactive solid wastes. Recommendations for design and testing are given, based on the extensive experience available worldwide in waste management. This report is not intended to have any regulatory status or objectives. 56 refs, 16 figs, 10 tabs

  8. Materials for high-level waste containment

    International Nuclear Information System (INIS)

    Marsh, G.P.

    1982-01-01

    The function of the high-level radioactive waste container in storage and of a container/overpack combination in disposal is considered. The consequent properties required from potential fabrication materials are discussed. The strategy adopted in selecting containment materials and the experimental programme underway to evaluate them are described. (U.K.)

  9. Waste management of ENM-containing solid waste in Europe

    DEFF Research Database (Denmark)

    Heggelund, Laura Roverskov; Boldrin, Alessio; Hansen, Steffen Foss

    2015-01-01

    the Danish nanoproduct inventory (www.nanodb.dk) to get a general understanding of the fate of ENM during waste management in the European context. This was done by: 1. assigning individual products to an appropriate waste material fraction, 2. identifying the ENM in each fraction, 3. comparing identified...... waste fractions with waste treatment statistics for Europe, and 4. illustrating the general distribution of ENM into incineration, recycling and landfilling. Our results indicate that ╲plastic from used product containers╡ is the most abundant and diverse waste fraction, comprising a variety of both...... nanoproducts and materials. While differences are seen between individual EU countries/regions according to the local waste management system, results show that all waste treatment options are significantly involved in nanowaste handling, suggesting that research activities should cover different areas...

  10. Radioactive wastes processing and disposing container

    International Nuclear Information System (INIS)

    Wada, Jiro; Kato, Hiroaki.

    1987-01-01

    Purpose: To obtain a processing and disposing container at low level radioactive wastes, excellent in corrosion and water resistance, as well as impact shock resistance for the retrieval storage over a long period of time. Constitution: The container is constituted with sands and pebbles as aggregates and glass fiber-added unsaturated polyester resins as binders. The container may entirely be formed with such material or only the entire inner surface may be formed with the material as liners. A container having excellent resistance to water, chemicals, freezing or melting, whether impact shock, etc. can be obtained, thereby enabling retrieval storage for radioactive wastes at the optimum low level. (Takahashi, M.)

  11. Storage container for radioactive wastes

    International Nuclear Information System (INIS)

    Catalayoud, L.; Gerard, M.

    1990-01-01

    The container comprises a cask and a cover made of concrete reinforced with metal fibers and closed with a seal poured in a dovetailed annular space provided between the cask and the cover. The seal is made with the material used for the cask [fr

  12. Plasma methods for metals recovery from metal-containing waste.

    Science.gov (United States)

    Changming, Du; Chao, Shang; Gong, Xiangjie; Ting, Wang; Xiange, Wei

    2018-04-27

    Metal-containing waste, a kind of new wastes, has a great potential for recycling and is also difficult to deal with. Many countries pay more and more attention to develop the metal recovery process and equipment of this kind of waste as raw material, so as to solve the environmental pollution and comprehensively utilize the discarded metal resources. Plasma processing is an efficient and environmentally friendly way for metal-containing waste. This review mainly discuss various metal-containing waste types, such as printed circuit boards (PCBs), red mud, galvanic sludge, Zircon, aluminium dross and incinerated ash, and the corresponding plasma methods, which include DC extended transferred arc plasma reactor, DC non-transferred arc plasma torch, RF thermal plasma reactor and argon and argon-hydrogen plasma jets. In addition, the plasma arc melting technology has a better purification effect on the extraction of useful metals from metal-containing wastes, a great capacity of volume reduction of waste materials, and a low leaching toxicity of solid slag, which can also be used to deal with all kinds of metal waste materials, having a wide range of applications. Copyright © 2018 Elsevier Ltd. All rights reserved.

  13. Characterization of Hanford tank wastes containing ferrocyanides

    International Nuclear Information System (INIS)

    Tingey, J.M.; Matheson, J.D.; McKinley, S.G.; Jones, T.E.; Pool, K.H.

    1993-02-01

    Currently, 17 storage tanks on the Hanford site that are believed to contain > 1,000 gram moles (465 lbs) of ferrocyanide compounds have been identified. Seven other tanks are classified as ferrocyanide containing waste tanks, but contain less than 1,000 gram moles of ferrocyanide compounds. These seven tanks are still included as Hanford Watch List Tanks. These tanks have been declared an unreviewed safety question (USQ) because of potential thermal reactivity hazards associated with the ferrocyanide compounds and nitrate and nitrite. Hanford tanks with waste containing > 1,000 gram moles of ferrocyanide have been sampled. Extensive chemical, radiothermical, and physical characterization have been performed on these waste samples. The reactivity of these wastes were also studied using Differential Scanning Calorimetry (DSC) and Thermogravimetric analysis. Actual tank waste samples were retrieved from tank 241-C-112 using a specially designed and equipped core-sampling truck. Only a small portion of the data obtained from this characterization effort will be reported in this paper. This report will deal primarily with the cyanide and carbon analyses, thermal analyses, and limited physical property measurements

  14. Vitrification of organics-containing wastes

    International Nuclear Information System (INIS)

    Bickford, D.F.

    1997-01-01

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

  15. Predicting the Lifetimes of Nuclear Waste Containers

    Science.gov (United States)

    King, Fraser

    2014-03-01

    As for many aspects of the disposal of nuclear waste, the greatest challenge we have in the study of container materials is the prediction of the long-term performance over periods of tens to hundreds of thousands of years. Various methods have been used for predicting the lifetime of containers for the disposal of high-level waste or spent fuel in deep geological repositories. Both mechanical and corrosion-related failure mechanisms need to be considered, although until recently the interactions of mechanical and corrosion degradation modes have not been considered in detail. Failure from mechanical degradation modes has tended to be treated through suitable container design. In comparison, the inevitable loss of container integrity due to corrosion has been treated by developing specific corrosion models. The most important aspect, however, is to be able to justify the long-term predictions by demonstrating a mechanistic understanding of the various degradation modes.

  16. Cadmium-containing waste and recycling possibilities

    International Nuclear Information System (INIS)

    Wiegand, V.; Rauhut, A.

    1981-01-01

    To begin with, the processes of cadmium production from zinc ores in smelting plants or from intermediates of other metal works are described. A considerable amount of the cadmium is obtained in the recycling process in zinc, lead, and copper works. The way of the cadmium-containing intermediaries, processing, enrichment, and disposal of cadmium waste are described. Uses of cadmium and its compounds are mentioned, and cadmium consumption in the years 1973-1977 in West Germany is presented in a table. Further chapters discuss the production and the way of waste during production and processing of cadmium-containing products, the problem of cadmium in household refuse and waste incineration plants, and the problem of cadmium emissions. (IHOE) [de

  17. Container for processing and disposing radioactive wastes and industrial wastes

    International Nuclear Information System (INIS)

    Araki, Kunio; Kasahara, Yuko; Kasai, Noboru; Sudo, Giichi; Ishizaki, Kanjiro.

    1978-01-01

    Purpose: To improve the performance of containers for radioactive wastes for ocean disposal and on-land disposal such as impact strength, chemical resistance, fire resistance, corrosion resistance, water impermeability and the like. Constitution: Steel fiber-reinforced concrete previously molded in a shape of a container is impregnated with polymerizable impregnating agent selected from the group consisting of a polymerizable monomer, liquid mixture of a polymerizable monomer and an oligomer, a polymer solution, a copolymer solution and the liquid mixture thereof. Then, the polymerizable impregnating agent is polymerized to solidify in the concrete by way of heat-polymerization or radiation-induced polymerization to form a waste container. The container thus obtained can be improved with the impact resistance and wear resistance and further improved with salt water resistance, acid resistance, corrosion resistance and solidity by the impregnation of the polymer, as well as can effectively be prevented from leaching out of radioactive substances. (Furukawa, Y.)

  18. Radioactive waste containment - a literature study

    International Nuclear Information System (INIS)

    Mohiuddin, G.

    1985-01-01

    One of the basic requirements of safe radioactive waste disposal is isolation of the radioactive substances to prevent leakage into the biosphere. The multi-barrier concept has been developed to meet this requirement. Within the framework of the concept, barriers can be either natural or man-made. Natural barriers, i.e. geologic formations,have been investigated for their suitability, with host rock and their different properties being determined and compared. It has been found that the qualification of a proposed repository medium cannot be defined on the basis of physical, chemical, and mineralogical criteria alone, but that these data have to be completed by a global evaluation of the entire system consisting of waste products and waste forms, host rock, and surrounding rock. The study in hand reviews the reports and also lists the studies made on engineered barriers, as e.g. immobilisation barriers, container and package barriers, of various waste forms. A review of the studies dealing with the various waste disposal techniques shows that the sub-surface waste disposal and the deep underground disposal in mines are the best developed techniques currently. A review of ultimate disposal concepts adopted abroad shows that most countries favour the mining technology approach, with the exception of Denmark where R and D work in this field is focused on deep well disposal. (orig./HP) [de

  19. Treatment of radioactive wastes containing plutonium

    International Nuclear Information System (INIS)

    Orlando, O.S.; Aparicio, G.; Greco, L.; Orosco, E.H.; Cassaniti, P.; Salguero, D.; Toubes, B.; Perez, A.E.; Menghini, J.E.; Esteban, A.; Adelfang, P.

    1987-01-01

    The radioactive wastes generated in the process of manufacture and control of experimental fuel rods of mixed oxides, (U,Pu)O 2 , require an specific treatment due to the plutonium content. The composition of liquid wastes, mostly arising from chemical checks, is variable. The salt content, the acidity, and the plutonium and uranium content are different, which makes necessary a chemical treatment before the inclusion in concrete. The solid waste, such as neoprene gloves, PVC sleeves, filter paper, disposable or broken laboratory material, etc. are also included in concrete. In this report the methods used to dispose of wastes at Alpha Facility are described. With regard to the liquid wastes, the glove box built to process them is detailed, as well as the applied chemical treatment, including neutralization, filtration and later solidification. As for the solid wastes, it is described the cementation method consisting in introducing them into an expanded metal matrix, of the basket type, that contains as a concentric drum of 200 liter capacity which is smaller than the matrix, and the filling with wet cement mortar. (Author)

  20. Waste-to-energy: Dehalogenation of plastic-containing wastes.

    Science.gov (United States)

    Shen, Yafei; Zhao, Rong; Wang, Junfeng; Chen, Xingming; Ge, Xinlei; Chen, Mindong

    2016-03-01

    The dehalogenation measurements could be carried out with the decomposition of plastic wastes simultaneously or successively. This paper reviewed the progresses in dehalogenation followed by thermochemical conversion of plastic-containing wastes for clean energy production. The pre-treatment method of MCT or HTT can eliminate the halogen in plastic wastes. The additives such as alkali-based metal oxides (e.g., CaO, NaOH), iron powders and minerals (e.g., quartz) can work as reaction mediums and accelerators with the objective of enhancing the mechanochemical reaction. The dehalogenation of waste plastics could be achieved by co-grinding with sustainable additives such as bio-wastes (e.g., rice husk), recyclable minerals (e.g., red mud) via MCT for solid fuels production. Interestingly, the solid fuel properties (e.g., particle size) could be significantly improved by HTT in addition with lignocellulosic biomass. Furthermore, the halogenated compounds in downstream thermal process could be eliminated by using catalysts and adsorbents. Most dehalogenation of plastic wastes primarily focuses on the transformation of organic halogen into inorganic halogen in terms of halogen hydrides or salts. The integrated process of MCT or HTT with the catalytic thermal decomposition is a promising way for clean energy production. The low-cost additives (e.g., red mud) used in the pre-treatment by MCT or HTT lead to a considerable synergistic effects including catalytic effect contributing to the follow-up thermal decomposition. Copyright © 2015 Elsevier Ltd. All rights reserved.

  1. Nanoporous Glasses for Nuclear Waste Containment

    OpenAIRE

    Woignier, Thierry; Primera, Juan; Reynes, Jerôme

    2016-01-01

    Research is in progress to incorporate nuclear waste in new matrices with high structural stability, resistance to thermal shock, and high chemical durability. Interactions with water are important for materials used as a containment matrix for the radio nuclides. It is indispensable to improve their chemical durability to limit the possible release of radioactive chemical species, if the glass structure is attacked by corrosion. By associating high structural stability and high chemical dura...

  2. Treatment of cyanide-contained Waste Water

    International Nuclear Information System (INIS)

    Scheglov, M.Y.

    1999-01-01

    This work contains results of theoretical and experimental investigations of possibility to apply industrial ionites of different kinds for recovering complex cyanide of some d-elements (Cu, Zn, an dso on) and free CN-ions with purpose to develop technology and unit for plating plant waste water treatment. Finally, on basis of experimental data about equilibrium kinetic and dynamic characteristic of the sorption in model solutions, strong base anionite in CN- and OH-forms was chosen. This anionite has the best values of operational sorption uptake. Recommendations of using the anionite have been developed for real cyanide-contained wastewater treatment

  3. The application of biotechnology to the treatment of wastes produced from the nuclear fuel cycle: Biodegradation and bioaccumulation as a means of treating radionuclide-containing streams

    International Nuclear Information System (INIS)

    Macaskie, L.E.

    1991-01-01

    Recent concerns on the radiotoxicity and longevity of nuclides have prompted the development of new technologies for their removal from wastes produced from nuclear power programs and nuclear fuel reprocessing activities. Alongside developments from traditional chemical treatment processes, interest has also centered on the application of biotechnology for efficient waste treatment. Many biological techniques have relied on empirical approaches in simple model systems, with scant regard to the nature and volume of actual target wastes; such considerations may limit the application of the new technologies in practice. This review aims to identify some of the likely problems, to discuss the various approaches under current consideration, and to evaluate ways in which either the target waste or the detoxifying biomass may be modified or presented for the most efficient treatment. 278 references

  4. Apparatus for filling a container with radioactive solid wastes

    International Nuclear Information System (INIS)

    Adachi, T.; Hiratake, S.

    1984-01-01

    In apparatus for filling a container suitable for storage with radioactive solid wastes arising from atomic power plants or the like, a plasma arc is irradiated toward a portion of the wastes to melt the portion of the wastes; portions of the wastes are successively moved so as to be subjected to irradiation of the plasma arc to continuously melt the wastes; and the melts obtained by melting the wastes are permitted to flow down toward the bottom of the container

  5. Hospital wastes management containing in radioactive refusals

    International Nuclear Information System (INIS)

    Campi, F.

    1999-01-01

    In large hospitals, featuring a nuclear medicine department, diagnostic examinations and metabolic therapies are performed using an amount of radio drugs per day averaging around some hundreds mCi. Part of these drugs are disposed in the conventional patient related waste and collected within the hospital itself. Before directing the wastes to the disposal, it is necessary verify that they do not contain radioactive materials. This article refers a study on the possibility to perform this verification by means of an automatic radio-metric system, in order to improve the efficiency, the speed and the safety of the control. Measures devoted to determined the minimum detectable activities for the main radionuclides used in the hospitals have been executed, and it has been designed a comprehensive device able to operate automatically, and unattended by any operator, the selection of radioactive refusals [it

  6. Testing protocols for evaluating monolithic waste forms containing mixed wastes

    International Nuclear Information System (INIS)

    Gilliam, T.M.; Sams, T.L.; Pitt, W.W.

    1986-01-01

    Test protocols have been presented which can be used as a guide in cement-based grout formulation development studies. Based on experience at ORNL, these six tests are generally sufficient to develop a grout product which will meet all applicable DOE, NRC, and EPA performance criteria. As such, these tests can be used to minimize the time required to tailor a grout to be compatible with both the waste stream and the process disposal scenario. 9 refs

  7. Multipurpose container for low-level radioactive waste

    International Nuclear Information System (INIS)

    Anderson, R.T.; Pearson, S.D.

    1993-01-01

    A method is described for disposing of low-level radioactive waste, comprising the steps of (a) introducing the waste into a multipurpose container, the multipurpose container comprising a polymeric inner container disposed within a concrete outer shell, the shape of the inner container conforming substantially to the shape of the outer shell's inner surface, (b) transporting the waste in the same multipurpose container to a storage location, and (c) storing the container at the storage location

  8. Testing waste forms containing high radionuclide loadings

    International Nuclear Information System (INIS)

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

    1986-01-01

    The Low-Level Waste Data Base Development - EPICOR-II Resin/Liner Investigation Program funded by the US Nuclear Regulatory Commission (NRC) is obtaining information on radioactive waste during NRC-prescribed tests and in a disposal environment. This paper describes the resin solidification task of that program, including the present status and results to date. An unusual aspect of this investigation is the use of commercial grade, ion exchange resins that have been loaded with over five times the radioactivity normally seen in a commercial application. That dramatically increases the total radiation dose to the resins. The objective of the resin solidification task is to determine the adequacy of test procedures specified by NRC for ion exchange resins having high radionuclide loadings

  9. Cellulose-containing Waste and Bituminized Salts

    International Nuclear Information System (INIS)

    Valcke, E.

    2005-01-01

    In Belgium, Medium-Level radioactive Waste (MLW) would be eventually disposed off in an underground repository in a geological formation such as the Boom Clay, which is studied as a reference host rock formation. MLW contains large quantities of non-radioactive chemicals that are released upon contact with pore water. It could be the case, for instance, for plutonium bearing cellulosic waste - such as paper tissues used to clean alpha glove boxes - issued from nuclear fuel fabrication (Belgonucleaire). At high pH, as in a disposal gallery backfilled with cement, the chemical degradation of cellulose will generate water-soluble products that may form strong complexes with actinides such as Am, Pu, Np, and U. This could lower the sorption of these elements onto the clay minerals, and hence increase their migration through the clay barrier. Another chemical perturbation could occur from the 3000 m 3 of so-called Eurobitum bituminised MLW, with precipitation sludges from the chemical treatment of spent nuclear fuel, and containing about 750 tons of NaNO 3 . The presence of NaNO 3 in this waste will give rise to several processes susceptible to affect the safety of the disposal system. Amongst others, it is necessary to verify that the swelling pressure of bitumen on the gallery wall and the osmotic pressure within the near-field are not too high to induce a fissuration of the host rock, leading to the formation of preferential migration pathways. The major objective of our work is to obtain a broad understanding of the different processes induced by the release of non-radioactive chemicals in the clay formation, to assess the chemical compatibility of different MLW forms with the clay

  10. Monitoring of wastes containing plutonium. Necessity and method

    International Nuclear Information System (INIS)

    Sousselier, Y.; Pottier, P.

    1979-01-01

    Importance of problems set by wastes containing plutonium is rapidly growing. Plutonium is not a waste, recycling limits heavily the quantity of plutonium to be stored with wastes. Optimized waste management must take definitive storage and economical limits of plutonium recovery into account. Waste monitoring is a must for safety, economy and waste management. Methods used require reliability, simplicity, sensibility and accuracy particularly for threshold detection [fr

  11. Liquid effluent retention facility dangerous waste permit application

    International Nuclear Information System (INIS)

    1991-06-01

    This appendix to the Liquid Effluent Retention Facility Dangerous Waste Permit Application contains pumps, piping, leak detection systems, geomembranes, leachate collection systems, earthworks and floating cover systems

  12. Long-term behaviour of concrete: development of operational model to predict the evolution of its containment performance. Application to cemented waste packages

    International Nuclear Information System (INIS)

    Peycelon, H.; Le Bescop, P.; Richet, C.; Adenot, F.

    2001-01-01

    In order to describe the main phenomena during different stages of cement waste packages life-time and to predict the long-term behaviour (containment performance) of concrete, coupled experiments and modelling studies are achieved. With respect to logical methodology, improvement of these studies is accomplished. Degradation of concrete in low mineralized, carbonated and sulfated water lead to an evolution of chemical characteristics (dissolution/precipitation of solid phases) and of transport properties which must be included or coupled in retention/transport modelling of radio nuclides to predict containment performance. (author)

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

    International Nuclear Information System (INIS)

    1991-10-01

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

  14. Nanoporous Glasses for Nuclear Waste Containment

    Directory of Open Access Journals (Sweden)

    Thierry Woignier

    2016-01-01

    Full Text Available Research is in progress to incorporate nuclear waste in new matrices with high structural stability, resistance to thermal shock, and high chemical durability. Interactions with water are important for materials used as a containment matrix for the radio nuclides. It is indispensable to improve their chemical durability to limit the possible release of radioactive chemical species, if the glass structure is attacked by corrosion. By associating high structural stability and high chemical durability, silica glass optimizes the properties of a suitable host matrix. According to an easy sintering stage, nanoporous glasses such as xerogels, aerogels, and composite gels are alternative ways to synthesize silica glass at relatively low temperatures (≈1,000–1,200°C. Nuclear wastes exist as aqueous salt solutions and we propose using the open pore structure of the nanoporous glass to enable migration of the solution throughout the solid volume. The loaded material is then sintered, thereby trapping the radioactive chemical species. The structure of the sintered materials (glass ceramics is that of nanocomposites: actinide phases (~100 nm embedded in a vitreous silica matrix. Our results showed a large improvement in the chemical durability of glass ceramic over conventional nuclear glass.

  15. APPLICATION OF CHEMICAL METHODS TO THE SOLID WASTE MANAGEMENT

    Directory of Open Access Journals (Sweden)

    C. P. Bulimaga

    2008-12-01

    Full Text Available The present article is a synthesis analysis of application of chemical methods for the development of technologies of hazardous waste management. Here are offered some technologies of neutralization of the waste containing hexacyanofferates, galvanic wastes and those with contain of vanadium, which are collected at Power Thermoelectric Plants.

  16. Treatment for hydrazine-containing waste water solution

    Science.gov (United States)

    Yade, N.

    1986-01-01

    The treatment for waste solutions containing hydrazine is presented. The invention attempts oxidation and decomposition of hydrazine in waste water in a simple and effective processing. The method adds activated charcoal to waste solutions containing hydrazine while maintaining a pH value higher than 8, and adding iron salts if necessary. Then, the solution is aerated.

  17. Method of burying vessel containing radioactive waste

    International Nuclear Information System (INIS)

    Koga, Yoshihito.

    1989-01-01

    A float having an inert gas sealed therein is attached to a tightly closed vessel containing radioactive wastes. The vessel is inserted and kept in a small hole for burying the tightly closed vessel in an excavated shaft in rocks such as of granite or rock salts, while filling bentonite as shielding material therearound. In this case, the float is so adjusted that the apparent specific gravity is made equal or nearer between the tightly closed vessel and the bentonite, so that the rightly closed vessel does not sink and cause direct contact with the rocks even if bentonite flows due to earthquakes, etc. This can prevent radioactivity contamination through water in the rocks. (S.K.)

  18. MAVL wastes containers functional demonstration and associated tests program

    International Nuclear Information System (INIS)

    Templier, J.C.

    2002-01-01

    In the framework of studies on the MAVL wastes, the CEA develops containers for middle time wastes storage. This program aims to realize a ''B wastes containers'' demonstrator. A demonstrator is a container, parts of a container or samples which must validate the tests. This document presents the state of the study in the following three chapters: functions description, base data and design choices; presentation of the functional demonstrators; demonstration tests description. (A.L.B.)

  19. PRODUCTION OF NEW BIOMASS/WASTE-CONTAINING SOLID FUELS

    Energy Technology Data Exchange (ETDEWEB)

    David J. Akers; Glenn A. Shirey; Zalman Zitron; Charles Q. Maney

    2001-04-20

    CQ Inc. and its team members (ALSTOM Power Inc., Bliss Industries, McFadden Machine Company, and industry advisors from coal-burning utilities, equipment manufacturers, and the pellet fuels industry) addressed the objectives of the Department of Energy and industry to produce economical, new solid fuels from coal, biomass, and waste materials that reduce emissions from coal-fired boilers. This project builds on the team's commercial experience in composite fuels for energy production. The electric utility industry is interested in the use of biomass and wastes as fuel to reduce both emissions and fuel costs. In addition to these benefits, utilities also recognize the business advantage of consuming the waste byproducts of customers both to retain customers and to improve the public image of the industry. Unfortunately, biomass and waste byproducts can be troublesome fuels because of low bulk density, high moisture content, variable composition, handling and feeding problems, and inadequate information about combustion and emissions characteristics. Current methods of co-firing biomass and wastes either use a separate fuel receiving, storage, and boiler feed system, or mass burn the biomass by simply mixing it with coal on the storage pile. For biomass or biomass-containing composite fuels to be extensively used in the U.S., especially in the steam market, a lower cost method of producing these fuels must be developed that includes both moisture reduction and pelletization or agglomeration for necessary fuel density and ease of handling. Further, this method of fuel production must be applicable to a variety of combinations of biomass, wastes, and coal; economically competitive with current fuels; and provide environmental benefits compared with coal. Notable accomplishments from the work performed in Phase I of this project include the development of three standard fuel formulations from mixtures of coal fines, biomass, and waste materials that can be used in

  20. Application of two-barrier model of radioactive agent transport in sea water for analyzing artificial radionuclide release from containers with radioactive waste dumped in Kara Sea

    Energy Technology Data Exchange (ETDEWEB)

    Grishin, Denis S.; Laykin, Andrey I.; Kuchin, Nickolay L.; Platovskikh, Yuri A. [Krylov State Research Center, Saint Petersburg, 44 Moskovskoe shosse, 196158 (Russian Federation)

    2014-07-01

    Modeling of artificial radionuclide transport in sea water is crucial for prognosis of radioecological situation in regions where dumping of radioactive waste had been made and/or accidents with nuclear submarines had taken place. Distribution of artificial radionuclides in bottom sediments can be a detector of radionuclide release from dumped or sunk objects to marine environment. Proper model can determine the dependence between radionuclide distribution in sediments and radionuclide release. Following report describes two-barrier model of radioactive agent transport in sea water. It was tested on data from 1994 - 2013 expeditions to Novaya Zemlya bays, where regular dumping of solid radioactive waste was practiced by the former USSR from the early 1960's until 1990. Two-barrier model agrees with experimental data and allows more accurate determination of time and intensity of artificial radionuclide release from dumped containers. (authors)

  1. Hanford Site Solid Waste Landfill permit application

    International Nuclear Information System (INIS)

    1991-01-01

    Daily activities at the Hanford Site generate sanitary solid waste (nonhazardous and nonradioactive) that is transported to and permanently disposed of at the Hanford Site Solid Waste Landfill. This permit application describes the manner in which the solid Waste Landfill will be operated under Washington State Department of Ecology Minimum Functional Standards for Solid Waste Handling, Washington Administrative Code 173-304. The solid Waste Landfill is owned by the US Department of Energy -- Richland Operations Office and is used for disposal of solid waste generated at the US Department of Energy Hanford Site. The jurisdictional health department's permit application form for the Solid Waste Landfill is provided in Chapter 1.0. Chapter 2.0 provides a description of the Hanford Site and the Solid Waste Landfill and reviews applicable locational, general facility, and landfilling standards. Chapter 3.0 discusses the characteristics and quantity of the waste disposed of in the Solid Waste Landfill. Chapter 4.0 reviews the regional and site geology and hydrology and the groundwater and vadose zone quality beneath the landfill. Chapters 5.0, 6.0, and 7.0 contain the plan of operation, closure plan, and postclosure plan, respectively. The plan of operation describes the routine operation and maintenance of the Solid Waste Landfill, the environmental monitoring program, and the safety and emergency plans. Chapter 5.0 also addresses the operational cover, environmental controls, personnel requirements, inspections, recordkeeping, reporting, and site security. The postclosure plan describes requirements for final cover maintenance and environmental monitoring equipment following final closure. Chapter 8.0 discusses the integration of closure and postclosure activities between the Solid Waste Landfill and adjacent Nonradioactive Dangerous Waste Landfill. 76 refs., 48 figs, 15 tabs

  2. Development of polymer concrete radioactive waste management containers

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.; Lee, M. S.; Ahn, D. H.; Won, H. J.; Kang, H. S.; Lee, H. S.; Lim, S.P.; Kim, Y. E.; Lee, B. O.; Lee, K. P.; Min, B. Y.; Lee, J.K.; Jang, W. S.; Sim, W. B.; Lee, J. C.; Park, M. J.; Choi, Y. J.; Shin, H. E.; Park, H. Y.; Kim, C. Y

    1999-11-01

    A high-integrity radioactive waste container has been developed to immobilize the spent resin wastes from nuclear power plants, protect possible future, inadvertent intruders from damaging radiation. The polymer concrete container is designed to ensure safe and reliable disposal of the radioactive waste for a minimum period of 300 years. A built-in vent system for each container will permit the release of gas. An experimental evaluation of the mechanical, chemical, and biological tests of the container was carried out. The tests showed that the polymer concrete container is adequate for safe disposal of the radioactive wastes. (author)

  3. NEW CRITERIA FOR ASSIGNING WASTE CONTAINING TECH-NOGENIC RADIONUCLIDES TO THE RADIOACTIVE WASTE

    Directory of Open Access Journals (Sweden)

    I. K. Romanovich

    2010-01-01

    Full Text Available The article contains detailed description of criteria for assigning of liquid and gaseous industrial waste containing technogenicradionuclides to the radioactive waste, presented in the new Basic Sanitary Rulesof Radiation Safety (OSPORB-99/2010. The analysisof shortcomings and discrepancies of the previously used in Russia system of criteria for assigning waste to the radioactive waste is given.

  4. Defense High Level Waste Disposal Container System Description Document

    International Nuclear Information System (INIS)

    Pettit, N. E.

    2001-01-01

    The Defense High Level Waste Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the accesses using a rail mounted transporter, and emplaced in emplacement drifts. The defense high level waste (HLW) disposal container provides long-term confinement of the commercial HLW and defense HLW (including immobilized plutonium waste forms [IPWF]) placed within disposable canisters, and withstands the loading, transfer, emplacement, and retrieval loads and environments. US Department of Energy (DOE)-owned spent nuclear fuel (SNF) in disposable canisters may also be placed in a defense HLW disposal container along with commercial HLW waste forms, which is known as co-disposal. The Defense High Level Waste Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container/waste package maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual canister temperatures after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Defense HLW disposal containers for HLW disposal will hold up to five HLW canisters. Defense HLW disposal containers for co-disposal will hold up to five HLW canisters arranged in a ring and one DOE SNF canister inserted in the center and/or one or more DOE SNF canisters displacing a HLW canister in the ring. Defense HLW disposal containers also will hold two Multi-Canister Overpacks (MCOs) and two HLW canisters in one disposal container. The disposal container will include outer and inner cylinders, outer and inner cylinder lids, and may include a canister guide. An exterior label will provide a means by

  5. Electrochemical corrosion studies on a selected carbon steel for application in nuclear waste disposal containers: Influence of chemical species in brines on corrosion

    International Nuclear Information System (INIS)

    Farvaque-Bera, A.M.; Smailos, E.

    1994-04-01

    In previous corrosion studies, carbon steels were identified as promising materials for the manufacture of long-lived high-level waste containers that could act as an engineered barrier in a rock-salt repository. In this paper, the influence of chemical species, potentially present in salt brines, on the electrochemical corrosion behavior of the preselected fine-grained steel TStE 355 was studied. The steel was examined at 90 C in a disposal relevant NaCl-rich brine containing various species (Br - , I - , Cu 2+ , Mn 2+ , S 2- , B(OH )4 - and Fe 3+ ) at concentrations between 10 -5 M/I and 10 -1 M/I. (orig.) [de

  6. Corrosion of metal containers containing cemented radioactive wastes

    International Nuclear Information System (INIS)

    Duffo, G.S.; Farina, S.B.; Schulz, F.M.; Marotta, F

    2010-01-01

    Nuclear activities generate different kinds of radioactive wastes. In the case of Argentina, wastes classified as low and medium level are conditioned in metal drums for final disposal in a repository whose design is based on the use of multiple and independent barriers. Nuclear energy plants generate a large volume of mid-level radioactive wastes, consisting mainly of ion-exchange resins contaminated by fission products. Other contaminated products such as gloves, papers, clothing, rubber and plastic tubing, can be incinerated and the ashes from the combustion also constitute wastes that must be disposed of. These wastes (resins and ashes) must be immobilized in order to avoid the release of radionuclides into the environment. The wastes usually undergo a process of cementing to immobilize them. This work aims to systematically study the process of degradation by corrosion of the steel drums in contact with the cemented resins and with the ashes cemented with the addition of different types and concentrations of aggressive compounds (chloride and sulfate). The specimens are configured so that the parameters of interest for the steel in contact with the cemented materials can be measured. The variables of corrosion potential, electric resistivity of the matrix and polarization resistance (PR) were monitored and show that the presence of chloride increases the susceptibility to corrosion of the drum steel that is in contact with the cement resin matrix

  7. Processing of nuclear power plant waste streams containing boric acid

    International Nuclear Information System (INIS)

    1996-10-01

    Boric acid is used in PWR type reactor's primary coolant circuit to control the neutron flux. However, boric acid complicates the control of water chemistry of primary coolant and the liquid radioactive waste produced from NPP. The purpose of this report is to provide member states with up-to-date information and guidelines for the treatment and conditioning of boric acid containing wastes. It contains chapters on: (a) characteristics of waste streams; (b) options for management of boric acid containing waste; (c) treatment/decontamination of boric acid containing waste; (d) concentration and immobilization of boric acid containing waste; (e) recovery and re-use of boric acid; (f) selected industrial processes in various countries; and (g) the influence of economic factors on process selection. 72 refs, 23 figs, 5 tabs

  8. Treatment of low and intermediate aqueous waste containing Cs-137 by chemical precipitation

    International Nuclear Information System (INIS)

    Valdezco, E.M.; Marcelo, E.A.; Alamares, A.L.; Junio, J.B.; Dela Cruz, J.M.

    1996-01-01

    The use of radioactive materials in various applications has been increasing since its introduction in the early sixties. The Philippine Nuclear Research Institute has established a centralized facility for treating radioactive wastes i.e. aqueous wastes with assistance from the International Atomic Energy Agency - Technical Cooperation Programme. Liquid wastes containing Cs-137 are generated from aqueous wastes containing Cs-137 by nickel ferrocyanide precipitation will be presented. The aim of this study is to investigate the efficiency treatment in removing Cs-137 from an aqueous effluent. Actual aqueous wastes known to contain Cs-137 were used in the experiments. Low cost and simple nickel ferrocyanide precipitation method with the aid of a flocculant has been selected for the separation of Cs-137 from low and intermediate aqueous waste. By varying the chemical dosage added into the aqueous waste, different decontamination factors were obtained. Hence, the optimum dosage of the chemicals that give the highest decontamination factor can be determined. (author)

  9. Method of vitrificating fine-containing liquid waste

    International Nuclear Information System (INIS)

    Hagiwara, Minoru; Matsunaka, Kazuhisa.

    1989-01-01

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

  10. Application of EPA regulations to low-level radioactive waste

    International Nuclear Information System (INIS)

    Bowerman, B.S.; Piciulo, P.L.

    1985-01-01

    The survey reported here was conducted with the intent of identifying categories of low-level radioactive wastes which would be classified under EPA regulations 40 CFR Part 261 as hazardous due to the chemical properties of the waste. Three waste types are identified under these criteria as potential radioactive mixed wastes: wastes containing organic liquids; wastes containing lead metal; and wastes containing chromium. The survey also indicated that certain wastes, specific to particular generators, may also be radioactive mixed wastes. Ultimately, the responsibility for determining whether a facility's wastes are mixed wastes rest with the generator. However, the uncertainties as to which regulations are applicable, and the fact that no legal definition of mixed wastes exists, make such a determination difficult. In addition to identifying mixed wastes, appropriate methods for the management of mixed wastes must be defined. In an ongoing study, BNL is evaluating options for the management of mixed wastes. These options will include segregation, substitution, and treatments to reduce or eliminate chemical hazards associated with the wastes listed above. The impacts of the EPA regulations governing hazardous wastes on radioactive mixed waste cannot be assessed in detail until the applicability of these regulations is agreed upon. This issue is still being discussed by EPA and NRC and should be resolved in the near future. Areas of waste management which may affect generators of mixed wastes include: monitoring/tracking of wastes before shipment; chemical testing of wastes; permits for treatment of storage of wastes; and additional packaging requirements. 3 refs., 1 fig., 2 tabs

  11. Process development for treatment of fluoride containing wastes

    Energy Technology Data Exchange (ETDEWEB)

    Singh, Mahesh; Kanvinde, V Y [Chemical Engineering Division, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    Many chemical and metallurgical industries generate liquid wastes containing high values of fluorides in association of nitrates and other metals. Due to harmful effects of fluorides these type of wastes can not be disposed off in the environment without proper treatment. Bench-scale laboratory experiments were conducted to develop a process scheme to fix the fluorides as non-leachable solid waste and fluoride free treated liquid waste for their disposal. To optimize the important parameters, simulated synthetic and actual wastes were used. For this study, three waste streams were collected from Nuclear Fuel Complex, Hyderabad. (author). 6 tabs., 1 fig.

  12. Corrosion process studies in a nuclear waste container

    International Nuclear Information System (INIS)

    Guasp, Ruben A.; Lanzani, Liliana A.; Coronel, Pascual; Bruzzoni, Pablo; Semino, Carlos J.

    1999-01-01

    Latest results on corrosion behavior studies on high activity nuclear waste container are reported. Corrosion evaluation on lead base alloys and modeling to predict carbon steel external container cover generalized corrosion, are the main issues of these studies. (author)

  13. Coating of waste containing ceramic granules

    International Nuclear Information System (INIS)

    Neumann, W.; Kofler, O.

    1979-01-01

    Simulated high-level waste granules produced by fluidized-bed calcination were overcoated by chemical vapor deposition (CVD) with pyrocarbon and nickel in laboratory-scale experiments. Successful development enables pyrocrbon deposition at temperatures of 600 to 800 0 K. The coated granules have excellent properties for long-term waste storage

  14. Testing waste forms containing high radionuclide loadings

    International Nuclear Information System (INIS)

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

    1986-01-01

    The Low-Level Waste Data Base Development - EPICOR-II Resin/Liner Investigation Program of the US Nuclear Regulatory Commission (NRC) is obtaining information on radioactive waste during NRC-prescribed tests and in a disposal environment. This paper describes the resin solidification task of that program, including the present status and results to date

  15. Method for treating waste containing stainless steel

    International Nuclear Information System (INIS)

    Kujawa, S.T.; Battleson, D.M.; Rademacher, E.L. Jr.; Cashell, P.V.; Filius, K.D.; Flannery, P.A.; Whitworth, C.G.

    1999-01-01

    A centrifugal plasma arc furnace is used to vitrify contaminated soils and other waste materials. An assessment of the characteristics of the waste is performed prior to introducing the waste into the furnace. Based on the assessment, a predetermined amount of iron is added to each batch of waste. The waste is melted in an oxidizing atmosphere into a slag. The added iron is oxidized into Fe 3 O 4 . Time of exposure to oxygen is controlled so that the iron does not oxidize into Fe 2 O 3 . Slag in the furnace remains relatively non-viscous and consequently it pours out of the furnace readily. Cooled and solidified slag produced by the furnace is very resistant to groundwater leaching. The slag can be safely buried in the earth without fear of contaminating groundwater. 3 figs

  16. Defense High Level Waste Disposal Container System Description Document

    International Nuclear Information System (INIS)

    2000-01-01

    The Defense High Level Waste Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the accesses using a rail mounted transporter, and emplaced in emplacement drifts. The defense high level waste (HLW) disposal container provides long-term confinement of the commercial HLW and defense HLW (including immobilized plutonium waste forms (IPWF)) placed within disposable canisters, and withstands the loading, transfer, emplacement, and retrieval loads and environments. U.S. Department of Energy (DOE)-owned spent nuclear fuel (SNF) in disposable canisters may also be placed in a defense HLW disposal container along with commercial HLW waste forms, which is known as 'co-disposal'. The Defense High Level Waste Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container/waste package maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual canister temperatures after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Defense HLW disposal containers for HLW disposal will hold up to five HLW canisters. Defense HLW disposal containers for co-disposal will hold up to five HLW canisters arranged in a ring and one DOE SNF canister in the ring. Defense HLW disposal containers also will hold two Multi-Canister Overpacks (MCOs) and two HLW canisters in one disposal container. The disposal container will include outer and inner cylinders, outer and inner cylinder lids, and may include a canister guide. An exterior label will provide a means by which to identify the disposal container and its contents. Different materials

  17. Corrosion of radioactive waste containers, case of a container made of low allow steel

    International Nuclear Information System (INIS)

    Bataillon, C.; Musy, C.; Roy, M.

    2001-01-01

    The following topics were dealt with: radioactive waste concept ANDRA, low alloy steel (XC38) container corrosion under representative storage conditions, corrosion rate and passivation effects, micrographic investigations

  18. Electrochemical corrosion studies on a selected carbon steel for application in nuclear waste disposal containers: Influence of radiolytic products on corrosion in brines

    International Nuclear Information System (INIS)

    Farvaque-Bera, A.M.; Smailos, E.

    1994-07-01

    In previous corrosion studies, carbon steels were identified as promising materials for the manufacturing of long-lived high-level waste containers that could act as a radionuclide barrier in a rock-salt repository. In the present work, the influence of some important oxidizing radiolytic products generated in gamma irradiated brines on the electrochemical corrosion behaviour of the preselected fine-grained steel TStE 355 was studied. The steel was examined by potentiodynamic and potentiostatic polarization methods at 90 C in a disposal relevant NaCl-rich brine containing radiolytic products such as H 2 O 2 , ClO - , ClO 3 - and ClO 4 - at concentrations between 10 -4 and 10 -2 M/l. The significance of the radiolytic products to steel corrosion depends on their concentration at the metal-brine interface, which in turn, depends on many factors such as the dose rate, the amount of water present in the disposal area, the escape of gases (e.g. H 2 )

  19. Grout treatment facility dangerous waste permit application

    International Nuclear Information System (INIS)

    1988-01-01

    This section briefly describes the Hanford Site, provides a general description of the site operations and administration, provides an overview of the contents of this Grout Treatment Facility (GTF) Permit Application, and gives a list of acronyms and abbreviations used in the document. The decision was made to use the checklist as a locator reference instead of using the checklist section numbers as paragraph section numbers because several different types of waste management units, some of which are not addressed in the checklists, are part of the GTF. The GTF is a waste management unit within the Hanford Site facility. In May 1988, a permit application was filed that identified the GTF as an existing facility. The GTF mixes dry cementitious solids with liquid mixed wastes (containing both dangerous and radioactive constituents) produced by Hanford Site operations. In addition to the design and operating features of the GTF that are intended to meet the requirements of dangerous waste regulations, many additional design and operating features are necessary to comply with radioactive waste management practices. The GTF design features and practices are intended to keep operational exposure to radionuclides and dangerous substances ''as low as reasonably achievable'' (ALARA) and to provide a disposal system that protects the environment for at least 10,000 yr. In some instances, ALARA practices present difficulties when complying with requirements of dangerous waste regulations. This volume contains 2 appendices covering engineering drawings and operating procedures

  20. Grout treatment facility dangerous waste permit application

    International Nuclear Information System (INIS)

    1988-01-01

    This section briefly describes the Hanford Site, provides a general description of the site operations and administration, provides an overview of the contents of this Grout Treatment Facility (GTF) Permit Application, and gives a list of acronyms and abbreviations used in the document. The decision was made to use the checklist as a locator reference instead of using the checklist section numbers as paragraph section numbers because several different types of waste management units, some of which are not addressed in the checklists, are part of the GTF. The GTF is a waste management unit within the Hanford Site facility. In May 1988, a permit application was filed that identified the GTF as an existing facility. The GTF mixes dry cementitious solids with liquid mixed wastes (containing both dangerous and radioactive constitutents) produced by Hanford Site operations. In addition to the design and operating features of the GTF that are intended to meet the requirements of dangerous waste regulations, many additional design and operating features are necessary to comply with radioactive waste management practices. The GTF design features and practices are intended to keep operational exposure to radionuclides and dangerous substances ''as low as reasonably achievable'' (ALARA) and to provide a disposal system that protects the environment for at least 10,000 yr. In some instances, ALARA practices present difficulties when complying with requirements of dangerous waste regulations. This volume contains 2 Appendices covering engineering drawings and operating procedures

  1. Grout Treatment Facility dangerous waste permit application

    International Nuclear Information System (INIS)

    1988-01-01

    This section briefly describes the Hanford Site, provides a general description of the site operations and administration, provides an overview of the contents of this Grout Treatment Facility (GTF) Permit Application, and gives a list of acronyms and abbreviations used in the document. The decision was made to use the checklist as a locator reference instead of using the checklist section numbers as paragraph section numbers because several different types of waste management units, some of which are not addressed in the checklists, are part of the GTF. The GTF is a waste management unit within the Hanford Site facility. In May 1988, a permit application was filed that identified the GTF as an existing facility. The GTF mixes dry cementitious solids with liquid wastes (containing both dangerous and radioactive constituents) produced by Hanford Site operations. In addition to the design and operating features of the GTF that are intended to meet the requirements of dangerous waste regulations, many additional design and operating features are necessary to comply with radioactive waste management practices. The GTF design features and practices are intended to keep operational exposure to radionuclides and dangerous substances ''as low as reasonably achievable'' (ALARA) and to provide a disposal system that protects the environment for at least 10,000 yr. In some instances, ALARA practices present difficulties when complying with requirements of dangerous waste regulations. This volume contains 14 Appendices. Topics include Engineering Drawings, Maps, Roads, Toxicity Testing, and Pilot-Scale Testing

  2. Shielded transport containers for reactor waste

    International Nuclear Information System (INIS)

    Grundfelt, B.; Eriksson, E.

    The report presents that part of risk analysis which deals with the frequency of breakdowns and the damage on containers. The report focusses on shielded containers made of reinforced concrete. Also a container made of steel is referred to the cases of breakdown are closely allied to collisions with ships. The frequency of breakdowns which might damage the containers is low in all respects, namely 1.10 -5 per year or lower for the shielded container. (G.B.)

  3. Stabilization Using Phosphate Bonded Ceramics. Salt Containing Mixed Waste Treatment. Mixed Waste Focus Area. OST Reference No. 117

    International Nuclear Information System (INIS)

    1999-01-01

    Throughout the Department of Energy (DOE) complex there are large inventories of homogeneous mixed waste solids, such as wastewater treatment residues, fly ashes, and sludges that contain relatively high concentrations (greater than 15% by weight) of salts. The inherent solubility of salts (e.g., nitrates, chlorides, and sulfates) makes traditional treatment of these waste streams difficult, expensive, and challenging. One alternative is low-temperature stabilization by chemically bonded phosphate ceramics (CBPCs). The process involves reacting magnesium oxide with monopotassium phosphate with the salt waste to produce a dense monolith. The ceramic makes a strong environmental barrier, and the metals are converted to insoluble, low-leaching phosphate salts. The process has been tested on a variety of surrogates and actual mixed waste streams, including soils, wastewater, flyashes, and crushed debris. It has also been demonstrated at scales ranging from 5 to 55 gallons. In some applications, the CBPC technology provides higher waste loadings and a more durable salt waste form than the baseline method of cementitious grouting. Waste form test specimens were subjected to a variety of performance tests. Results of waste form performance testing concluded that CBPC forms made with salt wastes meet or exceed both RCRA and recommended Nuclear Regulatory Commission (NRC) low-level waste (LLW) disposal criteria. Application of a polymer coating to the CBPC may decrease the leaching of salt anions, but continued waste form evaluations are needed to fully assess the deteriorating effects of this leaching, if any, over time.

  4. Seal welded cast iron nuclear waste container

    International Nuclear Information System (INIS)

    Filippi, A.M.; Sprecace, R.P.

    1987-01-01

    An article of manufacture is described comprising a cast iron container having an opening at one end and a cast iron plug; a first nickel-carbon alloy fusion weldable insert surrounding the opening and metallurgically bonded to the cast iron container at the one end of the container; a second nickel-carbon alloy insert metallurgically bonded to the cast iron plug located within the opening and surrounded by the first insert the inserts being jointed by a fusion bond in the opening without heating the cast iron container to an austenite formation temperature thereby sealing the interior of the container from the exterior ambient outside the opening; the nickel-carbon alloy containing about 2 to 5 w% carbon; and both the nickel-carbon alloy insert and the cast iron container have a microstructure containing a graphite phase

  5. LCA comparison of container systems in municipal solid waste management

    International Nuclear Information System (INIS)

    Rives, Jesus; Rieradevall, Joan; Gabarrell, Xavier

    2010-01-01

    The planning and design of integrated municipal solid waste management (MSWM) systems requires accurate environmental impact evaluation of the systems and their components. This research assessed, quantified and compared the environmental impact of the first stage of the most used MSW container systems. The comparison was based on factors such as the volume of the containers, from small bins of 60-80 l to containers of 2400 l, and on the manufactured materials, steel and high-density polyethylene (HDPE). Also, some parameters such as frequency of collections, waste generation, filling percentage and waste container contents, were established to obtain comparable systems. The methodological framework of the analysis was the life cycle assessment (LCA), and the impact assessment method was based on CML 2 baseline 2000. Results indicated that, for the same volume, the collection systems that use HDPE waste containers had more of an impact than those using steel waste containers, in terms of abiotic depletion, global warming, ozone layer depletion, acidification, eutrophication, photochemical oxidation, human toxicity and terrestrial ecotoxicity. Besides, the collection systems using small HDPE bins (60 l or 80 l) had most impact while systems using big steel containers (2400 l) had less impact. Subsequent sensitivity analysis about the parameters established demonstrated that they could change the ultimate environmental impact of each waste container collection system, but that the comparative relationship between systems was similar.

  6. ERG review of waste package container materials selection and corrosion

    International Nuclear Information System (INIS)

    Moak, D.P.; Perrin, J.S.

    1986-07-01

    The Engineering Review Group (ERG) was established by the Office of Nuclear Waste Isolation (ONWI) to help evaluate engineering-related issues in the US Department of Energy's nuclear waste repository program. The October 1984 meeting of the ERG reviewed the waste package container materials selection and corrosion. This report documents the ERG's comments and recommendations on these subjects and the ONWI response to the specific points raised by the ERG

  7. Conditioning of uranium-containing technological radioactive waste

    International Nuclear Information System (INIS)

    Smodis, B.; Tavcar, G.; Stepisnik, M.; Pucelj, B.

    2006-01-01

    Conditioning of mostly liquid uranium containing technological radioactive waste emerging from the past research activities at the Jozef Stefan Institute is described. The waste was first thoroughly characterised, then the radionuclides present solidified by appropriate chemical treatment, and the final product separated and prepared for storage in compliance with the legislation. The activities were carried out within the recently renewed Hot Cells Facility of the Jozef Stefan Institute and the overall process resulted in substantial volume reduction of the waste initially present. (author)

  8. Program for certification of waste from contained firing facility: Establishment of waste as non-reactive and discussion of potential waste generation problems

    International Nuclear Information System (INIS)

    Green, L.; Garza, R.; Maienschein, J.; Pruneda, C.

    1997-01-01

    Debris from explosives testing in a shot tank that contains 4 weight percent or less of explosive is shown to be non-reactive under the specified testing protocol in the Code of Federal Regulations. This debris can then be regarded as a non-hazardous waste on the basis of reactivity, when collected and packaged in a specified manner. If it is contaminated with radioactive components (e.g. depleted uranium), it can therefore be disposed of as radioactive waste or mixed waste, as appropriate (note that debris may contain other materials that render it hazardous, such as beryllium). We also discuss potential waste generation issues in contained firing operations that are applicable to the planned new Contained Firing Facility (CFF). The goal of this program is to develop and document conditions under which shot debris from the planned Contained Firing Facility (CFF) can be handled, shipped, and accepted for waste disposal as non-reactive radioactive or mixed waste. This report fulfills the following requirements as established at the outset of the program: 1. Establish through testing the maximum level of explosive that can be in a waste and still have it certified as non-reactive. 2. Develop the procedure to confirm the acceptability of radioactive-contaminated debris as non-reactive waste at radioactive waste disposal sites. 3. Outline potential disposal protocols for different CFF scenarios (e.g. misfires with scattered explosive)

  9. Method for primary containment of cesium wastes

    International Nuclear Information System (INIS)

    Angelini, P.; Arnold, W.D.; Blanco, R.E.; Bond, W.D.; Lackey, W.J.; Stinton, D.P.

    1983-01-01

    A method for producing a cesium-retentive waste form, characterized by a high degree of compositional stability and mechanical integrity, is provided by subjecting a cesium-loaded zeolite to heat under conditions suitable for stabilizing the zeolite and immobilizing the cesium, and coating said zeolite for sufficient duration within a suitable environment with at least one dense layer of pyrolytic carbon to seal therein said cesium to produce a final, cesium-bearing waste form. Typically, the zeolite is stabilized and the cesium immobilized in less than four hours by confinement within an air environment maintained at about 600 0 C. Coatings are thereafter applied by confining the calcined zeolite within a coating environment comprising inert fluidizing and carbon donor gases maintained at 1,000* C. For a suitable duration

  10. Neutron scattering applied to environmental waste containment

    International Nuclear Information System (INIS)

    Elcombe, M.M.; Studer, A.J.; Waring, C.L.

    1998-01-01

    Full text: A major environmental problem in Australia occurs at mine sites, where rock dumps and tailings dams are still causing problems many years after the mines have ceased operation. ANSTO has developed a method of producing a neutral barrier in-situ, which reduces water flow through the waste material. This in turn prevents water carrying waste products out into the wider environment. Both the loose grained sand substrate and the Neutral Barrier produced are crystalline and therefore amenable to diffraction techniques. In recent laboratory experiments neutron scattering has been used to confirm the presence of the barrier and measure the amount of calcite forming the barrier, at centimetre depths below the surface. The results of these measurements will be presented

  11. Removal of organic wastes containing tributyl phosphate

    International Nuclear Information System (INIS)

    Drobnik, S.

    TBP in dodecane and kerosene is one of the waste solutions from the reprocessing of spent nuclear fuels by the Purex process. The following methods were investigated for removing the organic solvents: adsorption on suitable solids, extraction, reaction with neutral salts, and saponification with acids or alkalis. Results showed that the best method of TBP removal is saponification with alkali hydroxides, either with dibutyl phosphate or with ortho-phosphate

  12. Treatment of organic waste solutions containing tributyl phosphate

    International Nuclear Information System (INIS)

    Drobnik, S.

    The two processes developed in the laboratory for treating waste solutions containing TBP, namely TBP separation with phosphoric acid and saponification were tested on a semi-industrial scale. A waste solution from the first phase of the Karlsruhe reprocessing plant was used

  13. Discharge of water containing waste emanating from land to the ...

    African Journals Online (AJOL)

    containing waste (wastewater), which emanates from land-based sources and which directly impact on the marine environment. These sources include sea outfalls, storm water drains, canals, rivers and diffuse sources of pollution. To date ...

  14. How reliable does the waste package containment have to be

    International Nuclear Information System (INIS)

    Wick, E.A.

    1985-01-01

    The final rule (10 CFR Part 60) for Disposal of High-Level Radioactive Wastes in Geologic Repositories specifies that the engineered barrier system shall be designed so that, assuming anticipated processes and events, containment of high-level radioactive wastes (HLW) will be substantially complete during the period when radiation and thermal conditions in the engineered barrier system are dominated by fission product decay. This requirement leads to the Nuclear Regulatory Commission (NRC) being asked the following questions: What is meant by ''substantially complete''. How reliable does waste package containment have to be. How many waste packages can fail. Although the NRC has not defined quantitatively the term ''substantially complete'', a numerical concept for acceptable release during the containment period is discussed. The number of containment failures that could be tolerated under the rule would depend upon the acceptable release, the time at which failure occurs and the rate of release from a failed package

  15. Grout treatment facility dangerous waste permit application

    International Nuclear Information System (INIS)

    1988-01-01

    This section briefly describes the Hanford Site, provides a general description of the site operations and administration, provides an overview of the contents of this Grout Treatment Facility (GTF) Permit Application, and gives a list of acronyms and abbreviations used in the document. The decision was made to use the checklist as a locator reference instead of using the checklist section numbers as paragraph section numbers because several different types of waste management units, some of which are not addressed in the checklists, are part of the GTF. The GTF is a waste management unit within the Hanford Site facility. In May 1988, permit application was filed that identified the GTF as an existing facility. The GTF mixes dry cementitious solids with liquid mixed wastes (containing both dangerous and radioactive constituents) produced by Hanford Site operations. In addition to the design and operating features of the GTF that are intended to meet the requirements of dangerous waste regulations, many additional design and operating features are necessary to comply with radioactive waste management practices. The GTF design features and practices are intended to keep operational exposure to radionuclides and dangerous substances ''as low as reasonably achievable'' (ALARA) and to provide a disposal system that protects the environment for at least 10,000 yr. In some instances, ALARA practices present difficulties when complying with requirements of dangerous waste regulations

  16. A probabilistic approach to assessing radioactive waste container lifetimes

    International Nuclear Information System (INIS)

    Porter, F.M.; Naish, C.C.; Sharland, S.M.

    1994-01-01

    A general methodology has been developed to make assessments of the lifetime of specific radioactive waste container designs in a repository environment. The methodology employs a statistical approach, which aims to reflect uncertainty in the corrosion rates, and the evolution of the environmental conditions. In this paper, the methodology is demonstrated for an intermediate-level waste (ILW) container in the anticipated UK repository situation

  17. Working towards a universal container for category B waste

    International Nuclear Information System (INIS)

    Tallec, M.

    2002-01-01

    Long-lived, intermediate-level waste, known as category 8 waste, accounts for most long lived waste (> 90%), although it only accounts for a very small fraction of radiotoxicity (< 10%). It comes in a wide variety of forms. The first step to be taken is to classify it into a few families and define a standard management mode for each one. Research teams are therefore seeking to propose a range of universal containers for existing packages and waste still to be conditioned. (author)

  18. Organic waste process containing at least one radioactive element

    International Nuclear Information System (INIS)

    Le Roy, F.

    1977-01-01

    The description is given of an organic waste process containing at least one element from the group comprising strontium, cesium, iodine and ruthenium. It comprises the introduction of the organic waste and gaseous oxygen in a bath of melted salt containing an alkaline carbonate, the bath being maintained at a high temperature between 400 and 1000 0 C and at a pressure of 0.5 to 10 bars, so that the organic waste is burnt and oxidised at least partly, the element selected being retained by the bath of melted salt [fr

  19. Process for treatment of detergent-containing radioactive liquid wastes

    International Nuclear Information System (INIS)

    Kamiya, K.; Chino, K.; Funabashi, K.; Horiuchi, S.; Motojima, K.

    1984-01-01

    A detergent-containing radioactive liquid waste originating from atomic power plants is concentrated to have about 10 wt. % detergent concentration, then dried in a thin film evaporator, and converted into powder. Powdered activated carbon is added to the radioactive waste in advance to prevent the liquid waste from foaming in the evaporator by the action of surface active agents contained in the detergent. The activated carbon is added in accordance with the COD concentration of the radioactive liquid waste to be treated, and usually at a concentration 2-4 times as large as the COD concentration of the liquid waste to be treated. A powdery product having a moisture content of not more than 15 wt. % is obtained from the evaporator, and pelletized and then packed into drums to be stored for a predetermined period

  20. Processing method and device for radioactive waste containing surfactant

    International Nuclear Information System (INIS)

    Yukita, Atsushi; Yoshikawa, Ryozo; Izumida, Tatsuo; Nishi, Takashi; Hattori, Yasuo.

    1997-01-01

    Washing liquid wastes generated in washing facilities in a nuclear power plant are collected in a liquid waste collecting tank. A suspension containing a powdery active carbon is supplied to the liquid waste collecting tank. Organic ingredients such as of a surfactant, oil ingredients and radioactive materials in the form of ions contained in the washing liquid wastes are adsorbed to the powdery active carbon. The washing liquid wastes containing the powdery active carbon and granular radioactive materials are led into an active carbon separating and drying device. The powdery active carbon and granular radioactive materials contained in the washing liquid wastes are filtered and separated by a filtering plate, and accumulated as filtered materials on the surface of the filtering plate. The purified washing liquid wastes are discharged to the outside. The filtered materials are dried by hot steams (or hot water) and dried air. The filtered materials are peeled from the filtering plate. The filtered materials, in other word, dried powdery active carbon and granular radioactive materials are transported to and burnt in an incinerator. (I.N.)

  1. Alternative containers for low-level wastes containing large amounts of tritium

    International Nuclear Information System (INIS)

    Gause, E.P.; Lee, B.S.; MacKenzie, D.R.; Wiswall, R. Jr.

    1984-11-01

    High-activity tritiated waste generated in the United States is mainly composed of tritium gas and tritium-contaminated organic solvents sorbed onto Speedi-Dri which are packaged in small glass bulbs. Low-activity waste consists of solidified and adsorbed liquids. In this report, current packages for high-activity gaseous and low-activity adsorbed liquid wastes are emphasized with regard to containment potential. Containers for low-level radioactive waste containing large amounts of tritium need to be developed. An integrity may be threatened by: physical degradation due to soil corrosion, gas pressure build-up (due to radiolysis and/or biodegradation), rapid permeation of tritium through the container, and corrosion from container contents. Literature available on these points is summarized in this report. 136 references, 20 figures, 40 tables

  2. Test plan for buried waste containment system materials

    International Nuclear Information System (INIS)

    Weidner, J.; Shaw, P.

    1997-03-01

    The objectives of the FY 1997 barrier material work at the Idaho National Engineering and Environmental Laboratory are to (1) select a waste barrier material and verify that it is compatible with the Buried Waste Containment System Process, and (2) determine if, and how, the Buried Waste Containment System emplacement process affects the material properties and performance (on proof of principle scale). This test plan describes a set of measurements and procedures used to validate a waste barrier material for the Buried Waste Containment System. A latex modified proprietary cement manufactured by CTS Cement Manufacturing Company will be tested. Emplacement properties required for the Buried Waste Containment System process are: slump between 8 and 10 in., set time between 15 and 30 minutes, compressive strength at set of 20 psi minimum, and set temperature less than 100 degrees C. Durability properties include resistance to degradation from carbonate, sulfate, and waste-site soil leachates. A set of baseline barrier material properties will be determined to provide a data base for comparison with the barrier materials when tested in the field. The measurements include permeability, petrographic analysis to determine separation and/or segregation of mix components, and a set of mechanical properties. The measurements will be repeated on specimens from the field test material. The data will be used to determine if the Buried Waste Containment System equipment changes the material. The emplacement properties will be determined using standard laboratory procedures and instruments. Durability of the barrier material will be evaluated by determining the effect of carbonate, sulfate, and waste-site soil leachates on the compressive strength of the barrier material. The baseline properties will be determined using standard ASTM procedures. 9 refs., 1 fig., 2 tabs

  3. Method of processing nitrate-containing radioactive liquid wastes

    International Nuclear Information System (INIS)

    Ogawa, Norito; Nagase, Kiyoharu; Otsuka, Katsuyuki; Ouchi, Jin.

    1983-01-01

    Purpose: To efficiently concentrate nitrate-containing low level radioactive liquid wastes by electrolytically dialyzing radioactive liquid wastes to decompose the nitrate salt by using an electrolytic cell comprising three chambers having ion exchange membranes and anodes made of special materials. Method: Nitrate-containing low level radioactive liquid wastes are supplied to and electrolytically dialyzed in a central chamber of an electrolytic cell comprising three chambers having cationic exchange membranes and anionic exchange membranes made of flouro-polymer as partition membranes, whereby the nitrate is decomposed to form nitric acid in the anode chamber and alkali hydroxide compound or ammonium hydroxide in the cathode chamber, as well as concentrate the radioactive substance in the central chamber. Coated metals of at least one type of platinum metal is used as the anode for the electrolytic cell. This enables efficient industrial concentration of nitrate-containing low level radioactive liquid wastes. (Yoshihara, H.)

  4. Transuranic contaminated waste container characterization and data base. Revision I

    International Nuclear Information System (INIS)

    Kniazewycz, B.G.

    1980-05-01

    The Nuclear Regulatory Commission (NRC) is developing regulations governing the management, handling and disposal of transuranium (TRU) radioisotope contaminated wastes as part of the NRC's overall waste management program. In the development of such regulations, numerous subtasks have been identified which require completion before meaningful regulations can be proposed, their impact evaluated and the regulations implemented. This report was prepared to assist in the development of the technical data base necessary to support rule-making actions dealing with TRU-contaminated wastes. An earlier report presented the waste sources, characteristics and inventory of both Department of Energy (DOE) generated and commercially generated TRU waste. In this report a wide variety of waste sources as well as a large TRU inventory were identified. The purpose of this report is to identify the different packaging systems used and proposed for TRU waste and to document their characteristics. This document then serves as part of the data base necessary to complete preparation and initiate implementation of TRU waste container and packaging standards and criteria suitable for inclusion in the present TRU waste management program. It is the purpose of this report to serve as a working document which will be used as appropriate in the TRU Waste Management Program. This report, and those following, will be compatible not only in format, but also in reference material and direction

  5. Management of radioactive wastes with regard to radioisotope application

    International Nuclear Information System (INIS)

    1989-01-01

    The report contains the abstracts of lectures held for the instruction of members of developing countries concerned with waste processing from isotope applications in research, medicine and industry. (HP)

  6. Conceptual TRU waste container design specifications

    International Nuclear Information System (INIS)

    Doty, J.W.; Peterson, J.B.

    1978-01-01

    A contractor questionnaire was used to gather pertinent data. Site visits were made to formulate an integrated contractor consensus; a packaging meeting was held to examine, discuss, and integrate packaging philosophies; and data collected from these activities and from Task Force meetings were consolidated to provide input to the Basic Application Checklist and Criteria Checklists. Conceptual Design Criteria were developed from an analysis and evaluation of the application data against federal regulations and interim/terminal storage constraints

  7. Radiolytic gas production from concrete containing Savannah River Plant waste

    International Nuclear Information System (INIS)

    Bibler, N.E.

    1978-01-01

    To determine the extent of gas production from radiolysis of concrete containing radioactive Savannah River Plant waste, samples of concrete and simulated waste were irradiated by 60 Co gamma rays and 244 Cm alpha particles. Gamma radiolysis simulated radiolysis by beta particles from fission products in the waste. Alpha radiolysis indicated the effect of alpha particles from transuranic isotopes in the waste. With gamma radiolysis, hydrogen was the only significant product; hydrogen reached a steady-state pressure that increased with increasing radiation intensity. Hydrogen was produced faster, and a higher steady-state pressure resulted when an organic set retarder was present. Oxygen that was sealed with the wastes was depleted. Gamma radiolysis also produced nitrous oxide gas when nitrate or nitrite was present in the concrete. With alpha radiolysis, hydrogen and oxygen were produced. Hydrogen did not reach a steady-state pressure at 137 Cs and 90 Sr), hydrogen will reach a steady-state pressure of 8 to 28 psi, and oxygen will be partially consumed. These predictions were confirmed by measurement of gas produced over a short time in a container of concrete and actual SRP waste. The tests with simulated waste also indicated that nitrous oxide may form, but because of the low nitrate or nitrite content of the waste, the maximum pressure of nitrous oxide after 300 years will be 238 Pu and 239 Pu will predominate; the hydrogen and oxygen pressures will increase to >200 psi

  8. Assessment of gas flammability in transuranic waste container

    International Nuclear Information System (INIS)

    Connolly, M.J.; Loehr, C.A.; Djordjevic, S.M.; Spangler, L.R.

    1995-01-01

    The Safety Analysis Report for the TRUPACT-II Shipping Package [Transuranic Package Transporter-II (TRUPACT-II) SARP] set limits for gas generation rates, wattage limits, and flammable volatile organic compound (VOC) concentrations in transuranic (TRU) waste containers that would be shipped to the Waste Isolation Pilot Plant (WIPP). Based on existing headspace gas data for drums stored at the Idaho National Engineering Laboratory (INEL) and the Rocky Flats Environmental Technology Site (RFETS), over 30 percent of the contact-handled TRU waste drums contain flammable VOC concentrations greater than the limit. Additional requirements may be imposed for emplacement of waste in the WIPP facility. The conditional no-migration determination (NMD) for the test phase of the facility required that flame tests be performed if significant levels of flammable VOCs were present in TRU waste containers. This paper describes an approach for investigating the potential flammability of TRU waste drums, which would increase the allowable concentrations of flammable VOCS. A flammability assessment methodology is presented that will allow more drums to be shipped to WIPP without treatment or repackaging and reduce the need for flame testing on drums. The approach includes experimental work to determine mixture lower explosive limits (MLEL) for the types of gas mixtures observed in TRU waste, a model for predicting the MLEL for mixtures of VOCS, hydrogen, and methane, and revised screening limits for total flammable VOCs concentrations and concentrations of hydrogen and methane using existing drum headspace gas data and the model predictions

  9. Hanford facility dangerous waste permit application

    International Nuclear Information System (INIS)

    1991-01-01

    This document, Set 2, the Hanford Facility Dangerous Waste Part B Permit Application, consists of 15 chapters that address the content of the Part B checklists prepared by the Washington State Department of Ecology (Ecology 1987) and the US Environmental Protection Agency (40 CFR 270), with additional information requirements mandated by the Hazardous and Solid Waste Amendments of 1984 and revisions of WAC 173-303. For ease of reference, the Washington State Department of Ecology checklist section numbers, in brackets, follow the chapter headings and subheadings. This permit application contains ''umbrella- type'' documentation with overall application to the Hanford Facility. This documentation is broad in nature and applies to all TSD units that have final status under the Hanford Facility Permit

  10. Containment of solidified liquid hazardous waste in domal salt

    International Nuclear Information System (INIS)

    Domenico, P.A.; Lerman, A.

    1992-01-01

    In recent years, the solidification of hazardous liquid waste has become a viable option in waste management. The solidification process results in an increased volume but more stable waste form that must be disposed of or stored in a dry environment. An environment of choice in south central Texas is domal salt. The salt dome currently under investigation has a water content of 0.002 percent by weight and a permeability less than one nanodarcy. A question that must be addressed is whether a salt dome has a particular set of attributes that will prevent the release of contaminants to the environment. From a regulatory perspective, a ''no migration'' petition must be approved by the U.S.E.P.A. for the containment facility. By ''no migration'' it is implied that the waste must be contained for 10,000 years. A demonstration that this condition will be met will require model calculations and such models must be based on the physical and chemical characteristics of the waste form and the geologic environment. In particular, the models must address the rate of brine infiltration into the caverns, providing information on how fast an immobile solid waste form could convert to a more mobile liquid state. Additionally, the potential for migration by both diffusion and advection is of concern. Lastly, given a partially saturated cavern, the question of how far gaseous waste will be transported over the 10,000 year containment period must also be addressed. Results indicate that the containment capabilities of domal salt are exceptional. A nominal volume of brine will seep into the cavern and most voids between the injected solidified waste pellets will remain unsaturated. Very small quantities of hazardous constituents will be leached from the waste pellets

  11. Design and testing of wood containers for radioactive waste

    International Nuclear Information System (INIS)

    Roberts, R.S.; Barry, P.E.

    1981-01-01

    A wood container for shipping and storing radioactive waste was designed to eliminate the problems caused by the weight, cost, and shape of the steel containers previously used. Tests specified by federal regulations (compression, free-drop, penetration, and vibration) were conducted on two of the containers, one loaded to 2500 lb and one loaded to 5000 lb. The 5000-lb container failed the free-drop test, but the 2500-lb container easily passed the tests and therefore qualifies as a Type A container. Its simplicity of design, low weight, and ease in handling have proved to be time-saving and cost-effective

  12. Concrete as secondary containment for interior wall embedded waste lines

    International Nuclear Information System (INIS)

    Porter, C.L.

    1993-01-01

    Throughout the Department of Energy (DOE) complex are numerous facilities that handle hazardous waste solutions. Secondary containment of tank systems and their ancillary piping is a major concern for existing facilities. The Idaho Division of Environmental Quality was petitioned in 1990 for an Equivalent Device determination regarding secondary containment of waste lines embedded in interior concrete walls. The petition was granted, however it expires in 1996. To address the secondary containment issue, additional studies were undertaken. One study verified the hypothesis that an interior wall pipe leak would follow the path of least resistance through the naturally occurring void found below a rigidly supported pipe and pass into an adjacent room where detection could occur, before any significant deterioration of the concrete takes place. Other tests demonstrated that with acidic waste solutions rebar and cold joints are not an accelerated path to the environment. The results from these latest studies confirm that the subject configuration meets all the requirements of secondary containment

  13. Destruction of high explosives and wastes containing high explosives using the molten salt destruction process

    International Nuclear Information System (INIS)

    Upadhye, R.S.; Brummond, W.A.; Pruneda, C.O.

    1992-01-01

    This paper reports the Molten Salt Destruction (MSD) Process which has been demonstrated for the destruction of HE and HE-containing wastes. MSD has been used by Rockwell International and by Anti-Pollution Systems to destroy hazardous wastes. MSD converts the organic constituents (including the HE) of the waste into non-hazardous substances such as carbon dioxide, nitrogen and water. In the case of HE-containing mixed wastes, any actinides in the waste are retained in the molten salt, thus converting the mixed wastes into low-level wastes. (Even though the MSD process is applicable to mixed wastes, this paper will emphasize HE-treatment.) The destruction of HE is accomplished by introducing it, together with oxidant gases, into a crucible containing a molten salt, such as sodium carbonate, or a suitable mixture of the carbonates of sodium, potassium, lithium and calcium. The temperature of the molten salt can be between 400 to 900 degrees C. The combustible organic components of the waste react with oxygen to produce carbon dioxide, nitrogen and steam

  14. Sulphate in Liquid Nuclear Waste: from Production to Containment

    Energy Technology Data Exchange (ETDEWEB)

    Lenoir, M.; Grandjean, A.; Ledieu, A.; Dussossoy, J.L.; Cau Dit Coumes, C.; Barre, Y.; Tronche, E. [CEA Marcoule, DEN/DTCD/SECM/LDMC, Batiment 208 BP17171, Bagnols sur Ceze, 30207 (France)

    2009-06-15

    Nuclear industry produces a wide range of low and intermediate level liquid radioactive wastes which can include different radionuclides such as {sup 90}Sr. In La Hague reprocessing plant and in the nuclear research centers of CEA (Commissariat a l'Energie Atomique), the coprecipitation of strontium with barium sulphate is the technique used to treat selectively these contaminated streams with the best efficiency. After the decontamination process, low and intermediate level activity wastes incorporating significant quantities of sulphate are obtained. The challenge is to find a matrix easy to form and with a good chemical durability which is able to confine this kind of nuclear waste. The current process used to contain sulphate-rich nuclear wastes is bituminization. However, in order to improve properties of containment matrices and simplify the process, CEA has chosen to supervise researches on other materials such as cements or glasses. Indeed, cements are widely used for the immobilization of a variety of wastes (low and intermediate level wastes) and they may be an alternative matrix to bitumen. Even if Portland cement, which is extensively used in the nuclear industry, presents some disadvantages for the containment of sulphate-rich nuclear wastes (risk of swelling and cracking due to delayed ettringite formation), other cement systems, such as calcium sulfo-aluminate binders, may be valuable candidates. Another matrix to confine sulphate-rich waste could be the glass. One of the advantages of this material is that it could also immobilize sulphate containing high level nuclear waste which is present in some countries. This waste comes from the use of ferrous sulfamate as a reducing agent for the conversion of Pu{sup 4+} to Pu{sup 3+} in the partitioning stage of the actinides during reprocessing. Sulphate solubility in borosilicate glasses has already been studied in CEA at laboratory and pilot scales. At a pilot scale, low level liquid waste has been

  15. LLNL/YMP Waste Container Fabrication and Closure Project

    International Nuclear Information System (INIS)

    1990-10-01

    The Department of Energy's Office of Civilian Radioactive Waste Management (OCRWM) Program is studying Yucca Mountain, Nevada as a suitable site for the first US high-level nuclear waste repository. Lawrence Livermore National Laboratory (LLNL) has the responsibility for designing and developing the waste package for the permanent storage of high-level nuclear waste. This report is a summary of the technical activities for the LLNL/YMP Nuclear Waste Disposal Container Fabrication and Closure Development Project. Candidate welding closure processes were identified in the Phase 1 report. This report discusses Phase 2. Phase 2 of this effort involved laboratory studies to determine the optimum fabrication and closure processes. Because of budget limitations, LLNL narrowed the materials for evaluation in Phase 2 from the original six to four: Alloy 825, CDA 715, CDA 102 (or CDA 122) and CDA 952. Phase 2 studies focused on evaluation of candidate material in conjunction with fabrication and closure processes

  16. Waste container weighing data processing to create reliable information of household waste generation.

    Science.gov (United States)

    Korhonen, Pirjo; Kaila, Juha

    2015-05-01

    Household mixed waste container weighing data was processed by knowledge discovery and data mining techniques to create reliable information of household waste generation. The final data set included 27,865 weight measurements covering the whole year 2013 and it was selected from a database of Helsinki Region Environmental Services Authority, Finland. The data set contains mixed household waste arising in 6m(3) containers and it was processed identifying missing values and inconsistently low and high values as errors. The share of missing values and errors in the data set was 0.6%. This provides evidence that the waste weighing data gives reliable information of mixed waste generation at collection point level. Characteristic of mixed household waste arising at the waste collection point level is a wide variation between pickups. The seasonal variation pattern as a result of collective similarities in behaviour of households was clearly detected by smoothed medians of waste weight time series. The evaluation of the collection time series against the defined distribution range of pickup weights on the waste collection point level shows that 65% of the pickups were from collection points with optimally dimensioned container capacity and the collection points with over- and under-dimensioned container capacities were noted in 9.5% and 3.4% of all pickups, respectively. Occasional extra waste in containers occurred in 21.2% of the pickups indicating the irregular behaviour of individual households. The results of this analysis show that processing waste weighing data using knowledge discovery and data mining techniques provides trustworthy information of household waste generation and its variations. Copyright © 2015 Elsevier Ltd. All rights reserved.

  17. Management of hazardous waste containers and container storage areas under the Resource Conservation and Recovery Act

    International Nuclear Information System (INIS)

    1993-08-01

    DOE's Office of Environmental Guidance, RCRA/CERCLA Division, has prepared this guidance document to assist waste management personnel in complying with the numerous and complex regulatory requirements associated with RCRA hazardous waste and radioactive mixed waste containers and container management areas. This document is designed using a systematic graphic approach that features detailed, step-by-step guidance and extensive references to additional relevant guidance materials. Diagrams, flowcharts, reference, and overview graphics accompany the narrative descriptions to illustrate and highlight the topics being discussed. Step-by-step narrative is accompanied by flowchart graphics in an easy-to-follow, ''roadmap'' format

  18. Management of hazardous waste containers and container storage areas under the Resource Conservation and Recovery Act

    Energy Technology Data Exchange (ETDEWEB)

    1993-08-01

    DOE`s Office of Environmental Guidance, RCRA/CERCLA Division, has prepared this guidance document to assist waste management personnel in complying with the numerous and complex regulatory requirements associated with RCRA hazardous waste and radioactive mixed waste containers and container management areas. This document is designed using a systematic graphic approach that features detailed, step-by-step guidance and extensive references to additional relevant guidance materials. Diagrams, flowcharts, reference, and overview graphics accompany the narrative descriptions to illustrate and highlight the topics being discussed. Step-by-step narrative is accompanied by flowchart graphics in an easy-to-follow, ``roadmap`` format.

  19. Structural optimization of reinforced concrete container for radioactive wastes

    International Nuclear Information System (INIS)

    Tamura, M.

    1984-01-01

    A structural optimization study of reinforced concrete container for transportation and disposal of the low level radioactive waste generated in Brazilian nuclear power plants. The code requires the structural integrity of these containers when subjected to fall from specified height, avoiding environmental contamination. The structural optimization allows material and transportation cost reduction by container wall thickness reduction. The structural analysis is performed by tridimensional mathematical model using finite element method. (Author) [pt

  20. PIC-container for containment and disposal of low and intermediate level radioactive wastes

    International Nuclear Information System (INIS)

    Araki, Kunio; Shinji, Yoshimasa; Maki, Yasuro; Ishizaki, Kanjiro; Minegishi, Keiichi; Sudoh, Giichi.

    1981-03-01

    Steel fiber reinforced polymer-impregnated concrete (SFPIC) has been investigated for low and intermediate level radioactive waste containers. The present study has been carried out by the following stages. A) Preliminary evaluation: 60 L size container for cold and hot tests. B) Evaluation of size effect: 200 L size container for cold tests. The 60 L and 200 L containers were designed as pressure-container (without equalizer) for 500 kg/cm 2 and 700 kg/cm 2 . Polymerization of impregnated methylmethacrylate monomer for stage-A and B were performed by 60 Co-γ ray radiation and thermal catalytic polymerization, respectively. Under the loading of 500 kg/cm 2 and 700 kg/cm 2 -outside hydraulic pressure, these containers were kept in their good condition. The observed maximum strains were about 1380 x 10 -6 and 3950 x 10 -6 at the outside central position of container body for circumferential direction of the 60 L and 200 L container, respectively. An accelerated leaching test was performed by charging the concentrate of the liquid radioactive waste from JMTR in JAERI into the container. Although they were immersed in deionized water for 400 days, nuclides were not leached from the container. From results of various tests, it was evaluated that the SFPIC-container was suitable for containment and disposal of low and intermediate level radioactive wastes. There was not any great difference between the two size containers for the physical and chemical properties except in their preparation process. (author)

  1. Differential lead retention in zircons: implications for nuclear waste containment.

    Science.gov (United States)

    Gentry, R V; Sworski, T J; McKown, H S; Smith, D H; Eby, R E; Christie, W H

    1982-04-16

    An innovative ultrasensitive technique was used for lead isotopic analysis of individual zircons extracted from granite core samples at depths of 960, 2170, 2900, 3930, and 4310 meters. The results show that lead, a relatively mobile element compared to the nuclear waste-related actinides uranium and thorium, has been highly retained at elevated temperatures (105 degrees to 313 degrees C) under conditions relevant to the burial of synthetic rock waste containers in deep granite holes.

  2. Parametric Criticality Safety Calculations for Arrays of TRU Waste Containers

    Energy Technology Data Exchange (ETDEWEB)

    Gough, Sean T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-10-26

    The Nuclear Criticality Safety Division (NCSD) has performed criticality safety calculations for finite and infinite arrays of transuranic (TRU) waste containers. The results of these analyses may be applied in any technical area onsite (e.g., TA-54, TA-55, etc.), as long as the assumptions herein are met. These calculations are designed to update the existing reference calculations for waste arrays documented in Reference 1, in order to meet current guidance on calculational methodology.

  3. Process for treating waste water containing hydrazine from power stations

    International Nuclear Information System (INIS)

    Hoffmann, W.

    1982-01-01

    A process for treating waste water containing hydrazine from nuclear power stations is proposed, characterized by the fact that the water is taken continuously through a water decomposition cell. If the water does not have sufficient conductivity itself, a substance raising the electrical conductivity is added to the water to be treated. The electrolysis is situated in the waste water tank. (orig./RB) [de

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

    International Nuclear Information System (INIS)

    1997-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-07-01

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

  6. Preliminary assessment of the controlled release of radionuclides from waste packages containing borosilicate waste glass

    International Nuclear Information System (INIS)

    Strachan, D.M.; McGrail, B.P.; Apted, M.J.; Engle, D.W.; Eslinger, P.W.

    1990-06-01

    The purpose of this report is to provide a preliminary assessment of the release-rate for an engineered barriers subsystem (EBS) containing waste packages of defense high-level waste borosilicate glass at geochemical and hydrological conditions similar to the those at Yucca Mountain. The relationship between the proposed Waste Acceptance Preliminary Specifications (WAPS) test of glass- dissolution rate and compliance with the NRC's release-rate criterion is also evaluated. Calculations are reported for three hierarchical levels: EBS analysis, waste-package analysis, and waste-glass analysis. The following conclusions identify those factors that most acutely affect the magnitude of, or uncertainty in, release-rate performance

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

    International Nuclear Information System (INIS)

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

    1993-03-01

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

  8. Acid-digestion treatment of plutonium-containing waste

    International Nuclear Information System (INIS)

    Wieczorek, H.; Kemmler, G.; Krause, H.

    1981-01-01

    The Radioactive Acid-Digestion Test Unit (RADTU) has been constructed at Hanford to demonstrate the application of the acid-digestion process for treating combustible transuranic wastes and scrap materials. The RADTU, with its original tray digestion vessel, has recently completed a six-month campaign processing potentially contaminated non-glovebox wastes from a Hanford plutonium facility. During this campaign, it processed 2100 kg largely cellulosic wastes at an average sustained processing rate of 3 kg/h as limited by the acid-waste contact and the water boil-off rate from the acid feeds. The on-line operating efficiency was nearly 50% on a twelve-hour day, five-day week basis. Following this campaign, a new annular high-rate digester has been installed for testing. In preliminary tests with simulated wastes, the new digester demonstrated a sustained capacity of 10 kg/h with greatly improved intimacy of contact between the digestion acid and the waste. The new design also doubles the heat-transfer surface, which is expected to provide at least twice the water boil-off rate of the previous tray digester design. Following shakedown testing with simulated and low-level wastes, the new unit will be used to process combustible plutonium scrap and waste from Hanford plutonium facilities for the purposes of volume reduction, plutonium recovery, and stabilization of the final waste form. (author)

  9. Detection of free liquid in containers of solidified radioactive waste

    Science.gov (United States)

    Greenhalgh, W.O.

    Nondestructive detection of the presence of free liquid within a sealed enclosure containing solidified waste is accomplished by measuring the levels of waste at two diametrically opposite locations while slowly tilting the enclosure toward one of said locations. When the measured level remains constant at the other location, the measured level at said one location is noted and any measured difference of levels indicates the presence of liquid on the surface of the solifified waste. The absence of liquid in the enclosure is verified when the measured levels at both locations are equal.

  10. Containment of Solid Wastes in some Large Scandinavian Cities

    DEFF Research Database (Denmark)

    Du-Thinh, Kien

    1998-01-01

    to the geological-geotechnical characteristics of the subsoil of the waste sites which determine to a large extent the risks of infiltration and transport of leachates. The role of the barrier, its design and construction or the consequences arising from the lack of abarrier are dealt with herein. The monitoring......Two kinds of containment of solid wastes - one in the vicinity of Copenhagen, the capital of Denmark, another on the outskirts of Gothenburg, the second largest city of sweden - are reviewed in this article. They represent two different approaches to waste management. Special attention is given...

  11. Detection of free liquid in containers of solidified radioactive waste

    Science.gov (United States)

    Greenhalgh, Wilbur O.

    1985-01-01

    A method of nondestructively detecting the presence of free liquid within a sealed enclosure containing solidified waste by measuring the levels of waste at two diametrically opposite locations while slowly tilting the enclosure toward one of said locations. When the measured level remains constant at the other location, the measured level at said one location is noted and any measured difference of levels indicates the presence of liquid on the surface of the solidified waste. The absence of liquid in the enclosure is verified when the measured levels at both locations are equal.

  12. Hydrothermal waste package interactions with methane-containing basalt groundwater

    International Nuclear Information System (INIS)

    McGrail, B.P.

    1984-01-01

    Hydrothermal waste package interaction tests were conducted with a mixture of crushed glass, basalt, and steel in methane-containing synthetic basalt groundwater. In the absence of gamma radiolysis, methane was found to have little influence on the corrosion behavior of the waste package constituents. Under gamma radiolysis, methane was found to significantly lower the solution oxidation potential when compared to identical tests without methane. In addition, colloidal hydrocarbon polymers that have been produced under the irradiation conditions of these experiments were not formed. The presence of the waste package constituents apparently inhibited the formation of the polymers. However, the mechanism which prevented their formation was not determined

  13. Treatment and minimization of heavy metal-containing wastes 1995

    International Nuclear Information System (INIS)

    Hager, J.P.; Mishra, B.; Litz, J.L.

    1995-01-01

    This symposium was held in conjunction with the 1995 Annual Meeting of the Minerals, Metals and Materials Society in Las Vegas, Nevada, February 12--16, 1995. The purpose of this meeting was to provide a forum for exchange of state-of-the-art information on treating and minimizing heavy metal-containing wastes. Papers were categorized under the following broad headings: aqueous processing; waste water treatment; thermal processing and stabilization; processing of fly ash, flue dusts, and slags; and processing of lead, mercury, and battery wastes. Individual papers have been processed separately for inclusion in the appropriate data bases

  14. Development of high integrity containers for rad-waste treatment

    Energy Technology Data Exchange (ETDEWEB)

    Song, Yung Chul; Cho, Myung Sug; Jung, Yun Sub [Korea Electric Power Corp. (KEPCO), Taejon (Korea, Republic of). Research Center

    1995-12-31

    Nuclear power plants are generating rad waste such as solid wastes, concentrated liquid wastes, spent resins and spent filters, and various types of imported containers which have different specifications and material properties are employed to handle the rad wastes according to facility characteristics of the plants or the type of wastes. These containers are stored at the intermediate storage facilities at the plant site due to the construction delay of permanent disposal site, and the additional construction of storage and disposal sites become more difficult with increase of the numbers and the operation time of the plants. In order to solve these difficulties, rad wastes volume reduction facilities such as High Pressure Compression Facility or Drying Facility are being installed and use of High Integrity Containers(HIC) are increasing. Therefore, we decide quality and technology standards required for the HIC, and then develop the HIC which satisfies the standards with new composite material called Steel Fiber Polymer Impregnated Concrete(SFPIC) (author). 84 refs., 118 figs.

  15. Development of high integrity containers for rad-waste treatment

    Energy Technology Data Exchange (ETDEWEB)

    Song, Yung Chul; Cho, Myung Sug; Jung, Yun Sub [Korea Electric Power Corp. (KEPCO), Taejon (Korea, Republic of). Research Center

    1996-12-31

    Nuclear power plants are generating rad waste such as solid wastes, concentrated liquid wastes, spent resins and spent filters, and various types of imported containers which have different specifications and material properties are employed to handle the rad wastes according to facility characteristics of the plants or the type of wastes. These containers are stored at the intermediate storage facilities at the plant site due to the construction delay of permanent disposal site, and the additional construction of storage and disposal sites become more difficult with increase of the numbers and the operation time of the plants. In order to solve these difficulties, rad wastes volume reduction facilities such as High Pressure Compression Facility or Drying Facility are being installed and use of High Integrity Containers(HIC) are increasing. Therefore, we decide quality and technology standards required for the HIC, and then develop the HIC which satisfies the standards with new composite material called Steel Fiber Polymer Impregnated Concrete(SFPIC) (author). 84 refs., 118 figs.

  16. 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

  17. Reliability evaluation methodologies for ensuring container integrity of stored transuranic (TRU) waste

    International Nuclear Information System (INIS)

    Smith, K.L.

    1995-06-01

    This report provides methodologies for providing defensible estimates of expected transuranic waste storage container lifetimes at the Radioactive Waste Management Complex. These methodologies can be used to estimate transuranic waste container reliability (for integrity and degradation) and as an analytical tool to optimize waste container integrity. Container packaging and storage configurations, which directly affect waste container integrity, are also addressed. The methodologies presented provide a means for demonstrating Resource Conservation and Recovery Act waste storage requirements

  18. Method for solidification of radioactive iodine-containing solid wastes

    International Nuclear Information System (INIS)

    Ozawa, Yoshihiro; Funabashi, Kiyomi; Uetake, Naoto.

    1987-01-01

    Purpose: To process radioactive iodine containing solid wastes as non-leaching solidified wastes with no risk of iodine release. Method: It has been known for the thermal stability of CuI, PbI 2 or adsorbents containing the same that they do not release iodine in an inert gas atmosphere or in a reducing atmosphere at a temperature lower than 480 deg C. In view of the above, adsorbents containing iodine in the chemical form of CuI or PbI 2 , or CuI or powdery PbI 2 per se are sealed and solidified into low melting glass at a temperature of lower than 480 deg C at which no iodine release occurs in a non-oxidative atmosphere. Since the products are vitrified wastes, they scarcely show leaching property and are excellent in durability and stability. (Takahashi, M.)

  19. Gas flow in and out of a nuclear waste container

    International Nuclear Information System (INIS)

    Zwahlen, E.D.; Pigford, T.H.; Chambre, P.L.; Lee, W.W.L.

    1989-05-01

    We analyze the flow of gases out of and into a high-level-waste container in the unsaturated tuff of Yucca Mountain. Containers are expected to fail eventually by localized cracks and penetrations. Even though the penetrations may be small, argon gas initially in the hot container can leak out. As the waste package cools, the pressure inside the container can become less than atmospheric, and air can leak in. 14 C released from the hot fuel-cladding surface can leak out of penetrations, and air inleakage can mobilize additional 14 C and other volatile radioactive species as it oxidizes the fuel cladding and the spent fuel. In an earlier paper we studied the gas flow through container penetrations occurring at the time of emplacement. Here we analyze the flow of gas for various penetration sizes occurring at 300 years. 3 refs., 2 figs

  20. Double-shell tank system dangerous waste permit application

    International Nuclear Information System (INIS)

    1991-06-01

    This Double-Shell Tank System Dangerous Waste Permit Application should be read in conjunction with the 242-A Evaporator Dangerous Waste Permit Application and the Liquid Effluent Retention Facility Dangerous Waste Permit Application, also submitted on June 28, 1991. Information contained in the Double-Shell Tank System permit application is referenced in the other two permit applications. The Double-Shell Tank System stores and treats mixed waste received from a variety of sources on the Hanford Site. The 242-A Evaporator treats liquid mixed waste received from the double-shell tanks. The 242-A Evaporator returns a mixed-waste slurry to the double-shell tanks and generates the dilute mixed-waste stream stored in the Liquid Effluent Retention Facility. This report contains information on the following topics: Facility Description and General Provisions; Waste Characteristics; Process Information; Groundwater Monitoring; Procedures to Prevent Hazards; Contingency Plan; Personnel Training; Exposure Information Report; Waste Minimization Plan; Closure and Postclosure Requirements; Reporting and Recordkeeping; other Relevant Laws; and Certification. 150 refs., 141 figs., 118 tabs

  1. Russian Containers for Transportation of Solid Radioactive Waste

    International Nuclear Information System (INIS)

    Petrushenko, V. G.; Baal, E. P.; Tsvetkov, D. Y.; Korb, V. R.; Nikitin, V. S.; Mikheev, A. A.; Griffith, A.; Schwab, P.; Nazarian, A.

    2002-01-01

    The Russian Shipyard ''Zvyozdochka'' has designed a new container for transportation and storage of solid radioactive wastes. The PST1A-6 container is cylindrical shaped and it can hold seven standard 200-liter (55-gallon) drums. The steel wall thickness is 6 mm, which is much greater than standard U.S. containers. These containers are fully certified to the Russian GOST requirements, which are basically identical to U.S. and IAEA standards for Type A containers. They can be transported by truck, rail, barge, ship, or aircraft and they can be stacked in 6 layers in storage facilities. The first user of the PST1A-6 containers is the Northern Fleet of the Russian Navy, under a program sponsored jointly by the U.S. DoD and DOE. This paper will describe the container design and show how the first 400 containers were fabricated and certified

  2. Grout Treatment Facility dangerous waste permit application

    International Nuclear Information System (INIS)

    1992-07-01

    The Grout Treatment Facility (GTF) is an existing treatment, storage, and/or disposal (TSD) unit located in the 200 East Area and the adjacent 600 Area of the Hanford Site. The GTF mixes dry cementitious solids with liquid mixed waste (containing both dangerous and radioactive constituents) produced by Hanford Site operations. The GTF consists of the following: The 241-AP-02D and 241-AP-04D waste pump pits and transfer piping; Dry Materials Facility (DMF); Grout Disposal Facility (GDF), consisting of the disposal vault and support and monitoring equipment; and Grout Processing Facility (GPF) and Westinghouse Hanford Company on the draft Hanford Facility Dangerous Waste Permit and may not be read to conflict with those comments. The Grout Treatment Facility Dangerous Waste Permit Application consists of both a Part A and a Part B permit application. An explanation of the Part A revisions associated with this TSD unit, including the current revision, is provided at the beginning of the Part A section. The Part B consists of 15 chapters addressing the organization and content of the Part B checklist prepared by the Washington State Department of Ecology (Ecology 1987). For ease of reference, the checklist section numbers, in brackets, follow chapter headings and subheadings

  3. Concrete containers in radioactive waste management: a review

    International Nuclear Information System (INIS)

    Tavares, Bárbara L.; Tello, Clédola Cássia O. de

    2017-01-01

    Nuclear power is considered a clean energy, because it does not produce the gases responsible for greenhouse effect. However, like all human activities, it is susceptible to waste generation. With increasing demand for energy in Brazil, the use of nuclear power is being expanded, as a result, the implementation of correct treatment and disposal are a necessity, in order to ensure the non-contamination of the public or environment and that exposure doses are lower than limits by legislation. Most of waste produced in Brazil are classified as low and intermediate radiation level; consequently, the national repository will be near surface, in accordance with the legislation. Considering the multi-barrier concept for the repository, the radioactive waste product is the first barrier. To have a qualified radioactive waste product, it should be solid or solidified using an inert material. With the intention of standardize the disposal process, all radioactive waste products will be placed in concrete containers. These containers will be settled in a concrete cell, the final engineered barrier of the repository. The state of the art is the first part of the study of the concrete containers and its specific criteria acceptation. Since the repository’s operational and surveillance period is 60 and 300 years, respectively, tests still need to be fulfilled in order to ensure the stability and resistance of the material. (author)

  4. Concrete containers in radioactive waste management: a review

    Energy Technology Data Exchange (ETDEWEB)

    Tavares, Bárbara L.; Tello, Clédola Cássia O. de, E-mail: barbaralacerdat@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte/MG (Brazil)

    2017-07-01

    Nuclear power is considered a clean energy, because it does not produce the gases responsible for greenhouse effect. However, like all human activities, it is susceptible to waste generation. With increasing demand for energy in Brazil, the use of nuclear power is being expanded, as a result, the implementation of correct treatment and disposal are a necessity, in order to ensure the non-contamination of the public or environment and that exposure doses are lower than limits by legislation. Most of waste produced in Brazil are classified as low and intermediate radiation level; consequently, the national repository will be near surface, in accordance with the legislation. Considering the multi-barrier concept for the repository, the radioactive waste product is the first barrier. To have a qualified radioactive waste product, it should be solid or solidified using an inert material. With the intention of standardize the disposal process, all radioactive waste products will be placed in concrete containers. These containers will be settled in a concrete cell, the final engineered barrier of the repository. The state of the art is the first part of the study of the concrete containers and its specific criteria acceptation. Since the repository’s operational and surveillance period is 60 and 300 years, respectively, tests still need to be fulfilled in order to ensure the stability and resistance of the material. (author)

  5. Salt removal from tanks containing high-level radioactive waste

    International Nuclear Information System (INIS)

    Kiser, D.L.

    1981-01-01

    At the Savannah River Plant (SRP), there are 23 waste storage tanks containing high-level radioactive wastes that are to be retired. These tanks contain about 23 million liters of salt and about 10 million liters of sludge, that are to be relocated to new Type III, fully stress-relieved tanks with complete secondary containment. About 19 million liters of salt cake are to be dissolved. Steam jet circulators were originally proposed for the salt dissolution program. However, use of steam jet circulators raised the temperature of the tank contents and caused operating problems. These included increased corrosion risk and required long cooldown periods prior to transfer. Alternative dissolution concepts were investigated. Examination of mechanisms affecting salt dissolution showed that the ability of fresh water to contact the cake surface was the most significant factor influencing dissolution rate. Density driven and mechanical agitation techniques were developed on a bench scale and then were demonstrated in an actual waste tank. Actual waste tank demonstrations were in good agreement with bench-scale experiments at 1/85 scale. The density driven method utilizes simple equipment, but leaves a cake heel in the tank and is hindered by the presence of sludge or Zeolite in the salt cake. Mechanical agitation overcomes the problems found with both steam jet circulators and the density driven technique and is the best method for future waste tank salt removal

  6. Stability and Volumetric Properties of Asphalt Mixture Containing Waste Plastic

    Directory of Open Access Journals (Sweden)

    Abd Kader Siti Aminah

    2017-01-01

    Full Text Available The objectives of this study are to determine the optimum bitumen content (OBC for every percentage added of waste plastics in asphalt mixtures and to investigate the stability properties of the asphalt mixtures containing waste plastic. Marshall stability and flow values along with density, air voids in total mix, voids in mineral aggregate, and voids filled with bitumen were determined to obtain OBC at different percentages of waste plastic, i.e., 4%, 6%, 8%, and 10% by weight of bitumen as additive. Results showed that the OBC for the plastic-modified asphalt mixtures at 4%, 6%, 8%, and 10% are 4.98, 5.44, 5.48, and 5.14, respectively. On the other hand, the controlled specimen’s shows better volumetric properties compared to plastic mixes. However, 4% additional of waste plastic indicated better stability than controlled specimen.

  7. Method of processing liquid waste containing fission product

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  8. POLYMER COMPOSITES MODIFIED BY WASTE MATERIALS CONTAINING WOOD FIBRES

    Directory of Open Access Journals (Sweden)

    Bernardeta Dębska

    2016-11-01

    Full Text Available In recent years, the idea of sustainable development has become one of the most important require-ments of civilization. Development of sustainable construction involves the need for the introduction of innovative technologies and solutions that will combine beneficial economic effects with taking care of the health and comfort of users, reducing the negative impact of the materials on the environment. Composites obtained from the use of waste materials are part of these assumptions. These include modified epoxy mortar containing waste wood fibres, described in this article. The modification consists in the substitution of sand by crushed waste boards, previously used as underlays for panels, in quantities of 0%, 10%, 20%, 35% and 50% by weight, respectively. Composites containing up to 20% of the modifier which were characterized by low water absorption, and good mechanical properties, also retained them after the process of cyclic freezing and thawing.

  9. 224-T Transuranic Waste Storage and Assay Facility dangerous waste permit application

    International Nuclear Information System (INIS)

    1992-01-01

    Westinghouse Hanford Company is a major contractor to the US Department of Energy Richland Field Office and serves as cooperator of the 224-T Transuranic Waste Storage and Assay Facility, the storage unit addressed in this permit application. At the time of submission of this portion of the Hanford Facility. Dangerous Waste Permit Application covering the 224-T Transuranic Waste Storage and Assay Facility, many issues identified in comments to the draft Hanford Facility Dangerous Waste Permit remain unresolved. This permit application reflects the positions taken by the US Department of Energy, Company on the draft Hanford Facility Dangerous Waste Permit and may not be read to conflict with those comments. The 224-T Transuranic Waste Storage and Assay Facility Dangerous Waste Permit Application (Revision 0) consists of both a Part A and Part B permit application. An explanation of the Part A revisions associated with this unit, including the Part A revision currently in effect, is provided at the beginning of the Part A section. The Part B consists of 15 chapters addressing the organization and content of the Part B Checklist prepared by the Washington State Department of Ecology (Ecology 1987). The 224-T Transuranic Waste Storage and Assay Facility Dangerous Waste Permit Application contains information current as of March 1, 1992

  10. Safety of systems for the retention of wastes containing radionuclides

    International Nuclear Information System (INIS)

    1980-11-01

    Information and minimal requirements demanded by CNEN for the emission of the Approval Certificate of the Safety Analysis Report related to system for the retention of wastes containing radionuclide, are established, aiming to assure low radioactivity levels to the environment. (E.G.) [pt

  11. Effect on localized waste-container failure on radionuclide transport from an underground nuclear waste vault

    International Nuclear Information System (INIS)

    Cheung, S.C.H.; Chan, T.

    1983-07-01

    In the geological disposal of nuclear fuel waste, one option is to emplace the waste container in a borehole drilled into the floor of the underground vault. In the borehole, the waste container is surrounded by a compacted soil material known as the buffer. A finite-element simulation has been performed to study the effect of localized partial failure of the waste container on the steady-state radionuclide transport by diffusion from the container through the buffer to the surrounding rock and/or backfill. In this study, the radionuclide concentration at the buffer-backfill interface is assumed to be zero. Two cases are considered at the interface between the buffer and the rock. In case 1, a no-flux boundary condition is used to simulate intact rock. In case 2, a constant radionuclide concentration condition is used to simulate fractured rock with groundwater flow. The results show that the effect of localized partial failure of the waste container on the total flux is dependent on the boundary condition at the buffer-rock interface. For the intact rock condition, the total flux is mainly dependent on the location of the failure. The total flux increases as the location changes from the bottom to the top of the emplaced waste container. For a given localized failure of the waste container, the total flux remains unaffected by the area of failed surface below the top of the failure. For fractured rock, the total flux is directly proportional to the failed surface area of the waste container regardless of the failure location

  12. Hanford Waste Vitrification Plant Dangerous Waste Permit Application

    International Nuclear Information System (INIS)

    1991-10-01

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

  13. Biofilm treatment of soil for waste containment and remediation

    International Nuclear Information System (INIS)

    Turner, J.P.; Dennis, M.L.; Osman, Y.A.; Chase, J.; Bulla, L.A.

    1997-01-01

    This paper examines the potential for creating low-permeability reactive barriers for waste treatment and containment by treating soils with Beijerinckia indica, a bacterium which produces an exopolysaccharide film. The biofilm adheres to soil particles and causes a decrease in soil hydraulic conductivity. In addition, B. Indica biodegrades a variety of polycyclic aromatic hydrocarbons and chemical carcinogens. The combination of low soil hydraulic conductivity and biodegradation capabilities creates the potential for constructing reactive biofilm barriers from soil and bacteria. A laboratory study was conducted to evaluate the effects of B. Indica on the hydraulic conductivity of a silty sand. Soil specimens were molded with a bacterial and nutrient solution, compacted at optimum moisture content, permeated with a nutrient solution, and tested for k sat using a flexible-wall permeameter. Saturated hydraulic conductivity (k sat ) was reduced from 1 x 10 -5 cm/sec to 2 x 10 -8 cm/sec: by biofilm treatment. Permeation with saline, acidic, and basic solutions following formation of a biofilm was found to have negligible effect on the reduced k sat , for up to three pore volumes of flow. Applications of biofilm treatment for creating low-permeability reactive barriers are discussed, including compacted liners for bottom barriers and caps and creation of vertical barriers by in situ treatment

  14. Characteristics of solidified products containing radioactive molten salt waste.

    Science.gov (United States)

    Park, Hwan-Seo; Kim, In-Tae; Cho, Yong-Zun; Eun, Hee-Chul; Kim, Joon-Hyung

    2007-11-01

    The molten salt waste from a pyroprocess to recover uranium and transuranic elements is one of the problematic radioactive wastes to be solidified into a durable wasteform for its final disposal. By using a novel method, named as the GRSS (gel-route stabilization/solidification) method, a molten salt waste was treated to produce a unique wasteform. A borosilicate glass as a chemical binder dissolves the silicate compounds in the gel products to produce one amorphous phase while most of the phosphates are encapsulated by the vitrified phase. Also, Cs in the gel product is preferentially situated in the silicate phase, and it is vitrified into a glassy phase after a heat treatment. The Sr-containing phase is mainly phosphate compounds and encapsulated by the glassy phase. These phenomena could be identified by the static and dynamic leaching test that revealed a high leach resistance of radionuclides. The leach rates were about 10(-3) - 10(-2) g/m2 x day for Cs and 10(-4) - 10(-3) g/m2 x day for Sr, and the leached fractions of them were predicted to be 0.89% and 0.39% at 900 days, respectively. This paper describes the characteristics of a unique wasteform containing a molten salt waste and provides important information on a newly developed immobilization technology for salt wastes, the GRSS method.

  15. Criticality Potential of Waste Packages Containing DOE SNF Affected by Igneous Intrusion

    International Nuclear Information System (INIS)

    D.S. Kimball; C.E. Sanders

    2006-01-01

    The Department of Energy (DOE) is currently preparing an application to submit to the U.S. Nuclear Regulatory Commission for a construction authorization for a monitored geologic repository. The repository will contain spent nuclear fuel (SNF) and defense high-level waste (DHLW) in waste packages placed in underground tunnels, or drifts. The primary objective of this paper is to perform a criticality analysis for waste packages containing DOE SNF affected by a disruptive igneous intrusion event in the emplacement drifts. The waste packages feature one DOE SNF canister placed in the center and surrounded by five High-Level Waste (HLW) glass canisters. The effective neutron multiplication factor (k eff ) is determined for potential configurations of the waste package during and after an intrusive igneous event. Due to the complexity of the potential scenarios following an igneous intrusion, finding conservative and bounding configurations with respect to criticality requires some additional considerations. In particular, the geometry of a slumped and damaged waste package must be examined, drift conditions must be modeled over a range of parameters, and the chemical degradation of DOE SNF and waste package materials must be considered for the expected high temperatures. The secondary intent of this calculation is to present a method for selecting conservative and bounding configurations for a wide range of end conditions

  16. Preliminary selection criteria for the Yucca Mountain Project waste package container material

    International Nuclear Information System (INIS)

    Halsey, W.G.

    1991-01-01

    The Department of Energy's Yucca Mountain Project (YMP) is evaluating a site at Yucca Mountain in Nevada for construction of a geologic repository for the storage of high-level nuclear waste. Lawrence Livermore National Laboratory's (LLNL) Nuclear Waste Management Project (NWMP) has the responsibility for design, testing, and performance analysis of the waste packages. The design is performed in an iterative manner in three sequential phases (conceptual design, advanced conceptual design, and license application design). An important input to the start of the advanced conceptual design is the selection of the material for the waste containers. The container material is referred to as the 'metal barrier' portion of the waste package, and is the responsibility of the Metal Barrier Selection and Testing task at LLNL. The selection will consist of several steps. First, preliminary, material-independent selection criteria will be established based on the performance goals for the container. Second, a variety of engineering materials will be evaluated against these criteria in a screening process to identify candidate materials. Third, information will be obtained on the performance of the candidate materials, and final selection criteria and quantitative weighting factors will be established based on the waste package design requirements. Finally, the candidate materials will be ranked against these criteria to determine whether they meet the mandated performance requirements, and to provide a comparative score to choose the material for advanced conceptual design activities. This document sets forth the preliminary container material selection criteria to be used in screening candidate materials. 5 refs

  17. Possible combustion hazards in 3013 plutonium waste container

    International Nuclear Information System (INIS)

    Sherman, M.P.

    1999-01-01

    Are there combustion hazards in plutonium-contaminated waste containers caused by combustible gas generation? Current gas generation models in which the only reaction considered is radiolysis must inevitably predict eventual complete dissociation of any water present into hydrogen and oxygen. Waste prepared for the 3013 container should be less subject to this problem because organic material and most of the absorbed water should have been removed. Depending on the waste form, moisture content, organic content, temperature, and container material, the pressure rise due to gas generation will be bounded by backreactions, recombination of the hydrogen and oxygen, absorption of the oxygen by plutonium oxide, and possibly other chemical reactions. Examination of a variety of food pack waste containers at Los Alamos National Laboratory (LANL) has shown little pressure rise, indeed often subatmospheric pressures. In a few cases large hydrogen concentrations up to 47% mole fraction were observed, but with negligible oxygen content. The only fuel seen in significant quantities was H 2 and, in one case, CO; the only oxidizer seen in significant quantities was O 2 . Considerable work on measuring gas generation is being done at Westinghouse Savannah River Company and LANL. In a mixture of H 2 , O 2 , and other diluent gases, if the hydrogen concentration is below the value at the lean flammability limit, or if the oxygen concentration is below that at the rich flammability limit, a flame will not propagate from an ignition source. Assuming H 2 is the only fuel present in significant quantities, a mixture leaner than the lean limit will get only leaner if mixed with air and is therefore no combustion hazard. However, when a mixture containing large amounts of H 2 is nonflammable because there is insufficient O 2 , there is a hazard. If the mixture should leak into a volume containing O 2 , or the container is opened into the surrounding air, the mixture will pass through the

  18. Disposal containers for radioactive waste materials and separation systems for radioactive waste materials

    International Nuclear Information System (INIS)

    Rubin, L.S.

    1986-01-01

    A separation system for dewatering radioactive waste materials includes a disposal container, drive structure for receiving the container, and means for releasably attaching the container to the drive structure. The separation structure disposed in the container adjacent the inner surface of the side wall structure retains solids while allowing passage of liquids. The inlet port structure in the container top wall is normally closed by first valve structure that is centrifugally actuated to open the inlet port and the discharge port structure at the container periphery receives liquid that passes through the separation structure and is normally closed by a second valve structure that is centrifugally actuated to open the discharge ports. The container also includes a coupling structure for releasable engagement with the centrifugal drive structure. The centrifugal force produced when the container is driven in rotation by the drive structure opens the valve structures, and radioactive waste material introduced into the container through the open inlet port is dewatered, and the waste is compacted. The ports are automatically closed by the valves when the container drum is not subjected to centrifugal force such that containment effectiveness is enhanced and exposure of personnel to radioactive materials is minimized. (author)

  19. A program to assess microbial impacts on nuclear waste containment

    International Nuclear Information System (INIS)

    Horn, J.; Meike, A.

    1996-01-01

    In this paper we discuss aspects of a comprehensive program to identify and bound potential effects of microorganisms on long-term nuclear waste containment, using as examples, studies conducted within the Yucca Mountain Project. A comprehensive program has been formulated which cuts across standard disciplinary lines to address the specific concerns of microbial activity in a radioactive waste repository. Collectively, this program provides bounding parameters of microbial activities that modify the ambient geochemistry and hydrology, modify corrosion rates, and transport and transform radionuclides under conditions expected to be encountered after geological waste emplacement. This program is intended to provide microbial reaction rates and bounding conditions in a form that can be integrated into existing chemical and hydrological models. The inclusion of microbial effects will allow those models to more accurately assess long term repository integrity

  20. Hydrothermal waste package interactions with methane-containing basalt groundwater

    International Nuclear Information System (INIS)

    McGrail, B.P.

    1984-11-01

    Hydrothermal waste package interaction tests with methane-containing synthetic basalt groundwater have shown that in the absence of gamma radiolysis, methane has little influence on the glass dissolution rate. Gamma radiolysis tests at fluxes of 5.5 x 10 5 and 4.4 x 10 4 R/hr showed that methane-saturated groundwater was more reducing than identical experiments where Ar was substituted for CH 4 . Dissolved methane, therefore, may be beneficial to the waste package in limiting the solubility of redox sensitive radionuclides such a 99 Tc. Hydrocarbon polymers known to form under the irradiation conditions of these tests were not produced. The presence of the waste package constituents apparently inhibited the formation of the polymers, however, the mechanism which prevented their formation was not determined

  1. Demonstration of close-coupled barriers for subsurface containment of buried waste

    International Nuclear Information System (INIS)

    Dwyer, B.P.; Heiser, J.; Stewart, W.

    1996-01-01

    The primary objective of this project is to develop and demonstrate a close-coupled barrier for the containment of subsurface waste or contaminant migration. A close-coupled barrier is produced by first installing a conventional cement grout curtain followed by a thin inner lining of a polymer grout. The resultant barrier is a cement polymer composite that has economic benefits derived from the cement and performance benefits from the durable and resistant polymer layer. Close-coupled barrier technology is applicable for final, interim, or emergency containment of subsurface waste forms. Consequently, when considering the diversity of technology application, the construction emplacement and material technology maturity, general site operational requirements, and regulatory compliance incentives, the close-coupled barrier system provides an alternative for any hazardous or mixed waste remediation plan. This paper discusses the installation of a close-coupled barrier and the subsequent integrity verification

  2. Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

    International Nuclear Information System (INIS)

    Bullen, D.B.; Gdowski, G.E.; Weiss, H.

    1988-06-01

    Three copper-based alloys, CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni), are being considered along with three austenitic candidates as possible materials for fabrication of containers for disposal of high-level radioactive waste. The waste will include spent fuel assemblies from reactors as well as high-level reprocessing wastes in borosilicate glass and will be sent to the prospective repository at Yucca Mountain, Nevada, for disposal. The containers must maintain mechanical integrity for 50 yr after emplacement to allow for retrieval of waste during the preclosure phase of repository operation. Containment is required to be substantially complete for up to 300 to 1000 yr. During the early period, the containers will be exposed to high temperatures and high gamma radiation fields from the decay of high-level waste. The final closure joint will be critical to the integrity of the containers. This volume surveys the available data on the metallurgy of the copper-based candidate alloys and the welding techniques employed to join these materials. The focus of this volume is on the methods applicable to remote-handling procedures in a hot-cell environment with limited possibility of postweld heat treatment. The three copper-based candidates are ranked on the basis of the various closure techniques. On the basis of considerations regarding welding, the following ranking is proposed for the copper-based alloys: CDA 715 (best) > CDA 102 > CDA 613 (worst). 49 refs., 15 figs., 1 tab

  3. Assessing the disposal of wastes containing NORM in nonhazardous waste landfills

    International Nuclear Information System (INIS)

    Smith, K. P.; Blunt, D. L.; Williams, G. P.; Arnish, J. J.; Pfingston, M. R.; Herbert, J.

    1999-01-01

    In the past few years, many states have established specific regulations for the management of petroleum industry wastes containing naturally occurring radioactive material (NORM) above specified thresholds. These regulations have limited the number of disposal options available for NORM-containing wastes, thereby increasing the related waste management costs. In view of the increasing economic burden associated with NORM management, industry and regulators are interested in identifying cost-effective disposal alternatives that still provide adequate protection of human health and the environment. One such alternative being considered is the disposal of NORM-containing wastes in landfills permitted to accept only nonhazardous wastes. The disposal of petroleum industry wastes containing radium-226 and lead-210 above regulated levels in nonhazardous landfills was modeled to evaluate the potential radiological doses and associated health risks to workers and the general public. A variety of scenarios were considered to evaluate the effects associated with the operational phase (i.e., during landfill operations) and future use of the landfill property. Doses were calculated for the maximally exposed receptor for each scenario. This paper presents the results of that study and some conclusions and recommendations drawn from it

  4. Corrosion aspects of steel radioactive waste containers in cementitious materials

    International Nuclear Information System (INIS)

    Smart, Nick

    2012-01-01

    Nick Smart from Serco, UK, gave an overview of the effects of cementitious materials on the corrosion of steel during storage and disposal of various low- and intermediate-level radioactive wastes. Steel containers are often used as an overpack for the containment of radioactive wastes and are routinely stored in an open atmosphere. Since this is an aerobic and typically humid environment, the steel containers can start to corrode whilst in storage. Steel containers often come into contact with cementitious materials (e.g. grout encapsulants, backfill). An extensive account of different steel container designs and of steel corrosion mechanisms was provided. Steel corrosion rates under conditions buffered by cementitious materials have been evaluated experimentally. The main conclusion was that the cementitious environment generally facilitates the passivation of steel materials. Several general and localised corrosion mechanisms need to be considered when evaluating the performance of steel containers in cementitious environments, and environmental thresholds can be defined and used with this aim. In addition, the consequences of the generation of gaseous hydrogen by the corrosion of carbon steel under anoxic conditions must be taken into account. Discussion of the paper included: Is crevice corrosion really significant in cementitious systems? Crevice corrosion is unlikely in the cementitious backfill considered because it will tend to neutralise any acidic conditions in the crevice. What is the role of microbially-induced corrosion (MIC) in cementitious systems? Microbes are likely to be present in a disposal facility but their effect on corrosion is uncertain

  5. Method of processing liquid wastes containing radioactive materials

    International Nuclear Information System (INIS)

    Matsumoto, Kaname; Shirai, Takamori; Nemoto, Kuniyoshi; Yoshikawa, Jun; Matsuda, Takeshi.

    1983-01-01

    Purpose: To reduce the number of solidification products by removing, particularly, Co-60 that is difficult to remove in a radioactive liquid wastes containing a water-soluble chelating agent, by adsorbing Co-60 to a specific chelating agent. Method: Liquid wastes containing radioactive cobalt and water-soluble chelating agent are passed through the layer of less water-soluble chelating agent that forms a complex compound with cobalt in an acidic pH region. Thus, the chelating compound of radioactive cobalt (particularly Co-60) is eliminated by adsorbing the same on a specific chelating agent layer. The chelating agent having Co-60 adsorbed thereon is discarded as it is through the cement- or asphalt-solidification process, whereby the number of solidification products to be generated can significantly be suppressed. (Moriyama, K.)

  6. Technology applications for radioactive waste minimization

    International Nuclear Information System (INIS)

    Devgun, J.S.

    1994-01-01

    The nuclear power industry has achieved one of the most successful examples of waste minimization. The annual volume of low-level radioactive waste shipped for disposal per reactor has decreased to approximately one-fifth the volume about a decade ago. In addition, the curie content of the total waste shipped for disposal has decreased. This paper will discuss the regulatory drivers and economic factors for waste minimization and describe the application of technologies for achieving waste minimization for low-level radioactive waste with examples from the nuclear power industry

  7. Investigation of Properties of Asphalt Concrete Containing Boron Waste as Mineral Filler

    Directory of Open Access Journals (Sweden)

    Cahit GÜRER

    2016-05-01

    Full Text Available During the manufacture of compounds in the boron mining industry a large quantity of waste boron is produced which has detrimental effects on the environment. Large areas have to be allocated for the disposal of this waste. Today with an increase in infrastructure construction, more efficient use of the existing sources of raw materials has become an obligation and this involves the recycling of various waste materials. Road construction requires a significant amount of raw materials and it is possible that substantial amounts of boron-containing waste materials can be recycled in these applications. This study investigates the usability of boron wastes as filler in asphalt concrete. For this purpose, asphalt concrete samples were produced using mineral fillers containing 4%, 5%, 6%, 7% and 8% boron waste as well as a 6% limestone filler (6%L as the control sample. The Marshall Design, mechanical immersion and Marshall Stability test after a freeze-thaw cycle and indirect tensile stiffness modulus (ITSM test were performed for each of the series. The results of this experimental study showed that boron waste can be used in medium and low trafficked asphalt concrete pavements wearing courses as filler.

  8. Corrosion models for predictions of performance of high-level radioactive-waste containers

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, J.C.; McCright, R.D. [Lawrence Livermore National Lab., CA (United States); Gdowski, G.E. [KMI Energy Services, Livermore, CA (United States)

    1991-11-01

    The present plan for disposal of high-level radioactive waste in the US is to seal it in containers before emplacement in a geologic repository. A proposed site at Yucca Mountain, Nevada, is being evaluated for its suitability as a geologic repository. The containers will probably be made of either an austenitic or a copper-based alloy. Models of alloy degradation are being used to predict the long-term performance of the containers under repository conditions. The models are of uniform oxidation and corrosion, localized corrosion, and stress corrosion cracking, and are applicable to worst-case scenarios of container degradation. This paper reviews several of the models.

  9. Long-term leach rates of glasses containing actual waste

    International Nuclear Information System (INIS)

    Wiley, J.R.; LeRoy, J.H.

    1979-01-01

    Leach rates of borosilicate glasses that contained actual Savannah River Plant waste were measured. Leaching was done by water and by buffer solutions of pH 4, 7, and 9. Leach rates were then determined from the amount of 137 Cs, 90 Sr, and Pu released into the leach solutions. The cumulative fractions leached were fit to a mathematical model that included leaching by diffusion and glass dissolution

  10. Long-term leach rates of glasses containing actual waste

    International Nuclear Information System (INIS)

    Wiley, J.R.; LeRoy, J.H.

    1979-01-01

    Leach rates of borosilicate glasses that contained actual Savannah River Plant waste were measured. Leaching was done by water and by buffer solutions of pH 4, 7, and 9. Leach rates were then determined from the amount of 137 Cs, 90 Sr, and plutonium released into the leach solutions. The cumulative fractions leached were fit to a mathematical model that included leaching by diffusion and glass dissolution. 5 figures, 3 tables

  11. Cement-Polymer Composite Containers for Radioactive Wastes Disposal

    International Nuclear Information System (INIS)

    Ghattas, N.K.; Eskander, S.B.; Bayoumi, T.A.; Saleh, H.M.

    2009-01-01

    Improving cement-composite containers using polymer as organic additives was studied extensively. Both unsaturated styrenated polyester (SPE) and polymethyl methacrylate (PMMA) were used to fill the pores in cement containers that used for disposal of radioactive wastes. Two different techniques were adopted for the addition of organic polymers based on their viscosity. The low density PMMA was added using impregnation technique. On the other hand high density SPE was mixed with cement paste as a premix process. Predetermined weight of dried borate radioactive powder waste simulate was introduced into the Cement-polymer composite (CPC) container and then closed before subjecting it to leaching characterization. The effect of the organic polymers on the hydration of cement matrix and on the properties of the obtained CPC container has been studied using X-ray diffraction, IR-analysis, thermal effects and weight loss. Porosity, pore parameters and rate of release were also determined. The results obtained showed that for the candidate CPC container positive effect of polymer dominates and an improvement in the retardation rate of PMMA release radionuclides was observed

  12. Candidate container materials for Yucca Mountain waste package designs

    International Nuclear Information System (INIS)

    McCright, R.D.; Halsey, W.G.; Gdowski, G.E.; Clarke, W.L.

    1991-09-01

    Materials considered as candidates for fabricating nuclear waste containers are reviewed in the context of the Conceptual Design phase of a potential repository located at Yucca Mountain. A selection criteria has been written for evaluation of candidate materials for the next phase -- Advanced Conceptual Design. The selection criteria is based on the conceptual design of a thin-walled container fabricated from a single metal or alloy; the criteria consider the performance requirements on the container and the service environment in which the containers will be emplaced. A long list of candidate materials is evaluated against the criteria, and a short list of materials is proposed for advanced characterization in the next design phase

  13. Ceramic waste forms for fuel-containing masses at Chernobyl

    International Nuclear Information System (INIS)

    Oversby, V.M.

    1994-05-01

    The fuel materials originally in the core of the Chernobyl Unit 4 reactor are now present within the Ukrytie in three major forms: (1) very fine particles of fuel dispersed as dust (about 10 tonnes), (2) fragments of the destroyed core, and (3) lavas containing fuel, cladding, and other materials. All of these materials will need to be immobilized into waste forms suitable for final disposal. We propose a ceramic waste form system that could accommodate all three waste types with a single set of processing equipment. The waste form would include the mineral zirconolite for immobilization of actinide materials (including uranium), perovskite, nepheline, spinel, and other phases as dictated by the chemistry of the lava masses. Waste loadings as high as 50% U can be achieved if pyrochlore, a close relative of zirconolite, is used as the U host. The ceramic immobilization could be achieved with low additions of inert chemicals to minimize the final disposal volume while ensuring a durable product. The sequence of processing would be to collect and immobilize the fuel dust first. This material will require minimal preprocessing and will provide experience in the handling of the fuel materials. Core fragments would be processed next, using a cryogenic crushing stage to reduce the size prior to adding ceramic additives. The lavas would be processed last, which is compatible with the likely sequence of availability of materials and with the complexity of the operations. The lavas will require more adjustment of chemical additive composition than the other streams to ensure that the desired phases are produced in the waste form

  14. Buffer lining manufacturing method for radioactive waste container

    International Nuclear Information System (INIS)

    Kawakami, Susumu; Sugino, Hiroyuki

    1998-01-01

    A recessed portion is formed on an upper surface of a filler layer made of a buffer powder filled into a container main body, the upper portion of the vessel main body is closed by a shrinkable liquid tight film. It is placed in a pressurizing container and pressed to mold a buffer lining base material integrated with the vessel main body. A flat upper surface and a containing space are formed by shaving to form a buffer lining. A disposing vessel containing radioactive wastes is inserted into the containing space, and the containing space is closed by a buffer block. The upper surface is sealed by a lid. With such a constitution, since a buffer lining integrated with the vessel main body can be formed easily inside the vessel main body, the disposing vessel can be contained in the containing vessel in a state surrounded by the buffer easily and stably without laying or piling over a large quantity of buffer blocks. (T.M.)

  15. A review of mechanochemistry applications in waste management

    International Nuclear Information System (INIS)

    Guo Xiuying; Xiang Dong; Duan Guanghong; Mou Peng

    2010-01-01

    Mechanochemistry is defined to describe the chemical and physicochemical transformation of substances during the aggregation caused by the mechanical energy. Mechanochemical technology has several advantages, such as simple process, ecological safety and the possibility of obtaining a product in the metastable state. It potentially has a prospective application in pollution remediation and waste management. Therefore, this paper aims to give an overall review of the mechanochemistry applications in waste management and the related mechanisms. Based on our study, the modification of fly ash and asbestos-containing wastes (ACWs) can be achieved by mechanochemical technology. Waste metal oxides can be transformed into easily recyclable sulfide by mechanochemical sulfidization. Besides, the waste plastics and rubbers, which are usually very difficult to be recycled, can also be recycled by mechanochemical technology.

  16. Leaching and geochemical behavior of fired bricks containing coal wastes.

    Science.gov (United States)

    Taha, Yassine; Benzaazoua, Mostafa; Edahbi, Mohamed; Mansori, Mohammed; Hakkou, Rachid

    2018-03-01

    High amounts of mine wastes are continuously produced by the mining industry all over the world. Recycling possibility of some wastes in fired brick making has been investigated and showed promising results. However, little attention is given to the leaching behavior of mine wastes based fired bricks. The objective of this paper is to evaluate the geochemical behavior of fired bricks containing different types of coal wastes. The leachates were analyzed for their concentration of As, Ba, Cd, Co, Cr, Cu, Mo, Ni, Pb, Zn and sulfates using different leaching tests; namely Tank Leaching tests (NEN 7375), Toxicity Characteristic Leaching Procedure (TCLP) and pH dependence test (EPA, 1313). The results showed that the release of constituents of potential interest was highly reduced after thermal treatment and were immobilized within the glassy matrix of the fired bricks. Moreover, it was also highlighted that the final pH of all fired samples changed and stabilized around 8-8.5 when the initial pH of leaching solution was in the range 2.5-11.5. The release of heavy metals and metalloids (As) tended to decrease with the increase of pH from acidic to alkaline solutions while Mo displayed a different trend. Copyright © 2017 Elsevier Ltd. All rights reserved.

  17. Recycling of Metal Containing Waste by Liquid-Liquid Extraction

    International Nuclear Information System (INIS)

    Reinhardt, H.

    1999-01-01

    Through the years, a large number of liquid-liquid extraction have been proposed for metal waste recovery and recycling(1,2). However, few of them have achieved commercial application. In fact, relatively little information is available on practical operation and economic feasibility. This presentation will give complementary information by describing and comparing three processes, based on the Am MAR hydrometallurgical concept and representing three different modes of operation

  18. Treatment technology analysis for mixed waste containers and debris

    International Nuclear Information System (INIS)

    Gehrke, R.J.; Brown, C.H.; Langton, C.A.; Askew, N.M.; Kan, T.; Schwinkendorf, W.E.

    1994-03-01

    A team was assembled to develop technology needs and strategies for treatment of mixed waste debris and empty containers in the Department of Energy (DOE) complex, and to determine the advantages and disadvantages of applying the Debris and Empty Container Rules to these wastes. These rules issued by the Environmental Protection Agency (EPA) apply only to the hazardous component of mixed debris. Hazardous debris that is subjected to regulations under the Atomic Energy Act because of its radioactivity (i.e., mixed debris) is also subject to the debris treatment standards. The issue of treating debris per the Resource Conservation and Recovery Act (RCRA) at the same time or in conjunction with decontamination of the radioactive contamination was also addressed. Resolution of this issue requires policy development by DOE Headquarters of de minimis concentrations for radioactivity and release of material to Subtitle D landfills or into the commercial sector. The task team recommends that, since alternate treatment technologies (for the hazardous component) are Best Demonstrated Available Technology (BDAT): (1) funding should focus on demonstration, testing, and evaluation of BDAT on mixed debris, (2) funding should also consider verification of alternative treatments for the decontamination of radioactive debris, and (3) DOE should establish criteria for the recycle/reuse or disposal of treated and decontaminated mixed debris as municipal waste

  19. In-situ containment and stabilization of buried waste

    International Nuclear Information System (INIS)

    Allan, M.L.; Kukacka, L.E.

    1993-10-01

    In FY 1993 research continued on development and testing of grout materials for in-situ containment and stabilization of buried waste. Specifically, the work was aimed at remediation of the Chemical Waste Landfill (CWL) at Sandia National Laboratories (SNL) in Albuquerque, New Mexico as part of the Mixed Waste Landfill Integrated Demonstration (MWLID). The work on grouting materials was initiated in FY 1992 and the accomplishments for that year are documented in the previous annual report (Allan, Kukacka and Heiser, 1992). The remediation plan involves stabilization of the chromium plume, placement of impermeable vertical and horizontal barriers to isolate the landfill and installation of a surface cap. The required depth of subsurface barriers is approximately 33 m (100 ft). The work concentrated on optimization of grout formulations for use as grout and soil cement barriers and caps. The durability of such materials was investigated, in addition to shrinkage cracking resistance, compressive and flexural strength and permeability. The potential for using fibers in grouts to control cracking was studied. Small scale field trials were conducted to test the practicality of using the identified formulations and to measure the long term performance. Large scale trials were conducted at Sandia as part of the Subsurface Barrier Emplacement Technology Program. Since it was already determined in FY 1992 that cementitious grouts could effectively stabilize the chromium plume at the CWL after pre-treatment is performed, the majority of the work was devoted to the containment aspect

  20. Lead corrosion evaluation in high activity nuclear waste container (Argentina)

    International Nuclear Information System (INIS)

    Guasp, R.; Lanzani, L.; Bruzzoni, P.; Cufre, W.; Semino, C.J.

    2000-01-01

    This report describes a study of high activity nuclear waste canister corrosion in a deep geological disposal. In this canister design, the vitrified nuclear waste stainless steel container is shielded by a 100 mm thick lead wall. For mechanical resistance, the canister will also have a thin carbon steel external liner. Experimental and mathematical modeling studies are aimed to asses the corrosion kinetics of the carbon steel liner in first instance and then, once this liner has been corroded away, the corrosion kinetics of the main lead barrier. Being that oxygen reduction is the main cathodic reaction that supports the anodic oxidation of iron, a model is described predicting the rate of oxygen consumption in a sealed deep nuclear waste disposal vault as a result of the canister corrosion. Oxidation processes other than container corrosion, and that can account also for oxygen depletion, are not taken into consideration. Corrosion experimental studies on lead and its alloys in groundwater are also reported. These experiments are aimed to improve the corrosion resistance of commercial lead in groundwater. (author)

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

    International Nuclear Information System (INIS)

    Jouan, A.

    1993-01-01

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

  2. Characterization of cement and bitumen waste forms containing simulated low-level waste incinerator ash

    International Nuclear Information System (INIS)

    Westsik, J.H. Jr.

    1984-08-01

    Incinerator ash from the combustion of general trash and ion exchange resins was immobilized in cement and bitumen. Tests were conducted on the resulting waste forms to provide a data base for the acceptability of actual low-level waste forms. The testing was done in accordance with the US Nuclear Regulatory Commission Technical Position on Waste Form. Bitumen had a measured compressive strength of 130 psi and a leachability index of 13 as measured with the ANS 16.1 leach test procedure. Cement demonstrated a compressive strength of 1400 psi and a leachability index of 7. Both waste forms easily exceed the minimum compressive strength of 50 psi and leachability index of 6 specified in the Technical Position. Irradiation to 10 8 Rad and exposure to 31 thermal cycles ranging from +60 0 ) to -30 0 C did not significantly impact these properties. Neither waste form supported bacterial or fungal growth as measured with ASTM G21 and G22 procedures. However, there is some indication of biodegradation due to co-metabolic processes. Concentration of organic complexants in leachates of the ash, cement and bitumen were too low to significantly affect the release of radionuclides from the waste forms. Neither bitumen nor cement containing incinerator ash caused any corrosion or degradation of potential container materials including steel, polyethylene and fiberglass. However, moist ash did cause corrosion of the steel

  3. Containment of transuranic contamination at the early waste retrieval project

    International Nuclear Information System (INIS)

    Harness, J.L.; McKinney, J.D.

    1977-01-01

    On July 26, 1976, while retrieving buried transuranic waste under the Early Waste Retrieval Program, a corroded 55-gallon 17H drum was retrieved. When uprighted, several liters of liquid escaped from the drum. This liquid was contaminated with transuranics, principally Pu-239, Am-241, and some Pu-238. As a result of the spread of this contamination in the Operating Area Confinement, six working days were required to decontaminate the area. At no time did the contamination escape the interior of the Operating Area Confinement building, and no contamination to personnel resulted from this occurrence, nor was a hazard presented to the general public. The facility was designed and constructed to contain the transuranic contamination resulting from such an occurrence. Proper prior planning and personnel training prevented the contamination occurrence from becoming a major event. This report details the occurrence, the recovery, and the information obtained from this event

  4. Alcohol-free alkoxide process for containing nuclear waste

    Science.gov (United States)

    Pope, James M.; Lahoda, Edward J.

    1984-01-01

    Disclosed is a method of containing nuclear waste. A composition is first prepared of about 25 to about 80%, calculated as SiO.sub.2, of a partially hydrolyzed silicon compound, up to about 30%, calculated as metal oxide, of a partially hydrolyzed aluminum or calcium compound, about 5 to about 20%, calculated as metal oxide, of a partially hydrolyzed boron or calcium compound, about 3 to about 25%, calculated as metal oxide, of a partially hydrolyzed sodium, potassium or lithium compound, an alcohol in a weight ratio to hydrolyzed alkoxide of about 1.5 to about 3% and sufficient water to remove at least 99% of the alcohol as an azeotrope. The azeotrope is boiled off and up to about 40%, based on solids in the product, of the nuclear waste, is mixed into the composition. The mixture is evaporated to about 25 to about 45% solids and is melted and cooled.

  5. 40 CFR 273.4 - Applicability-Mercury-containing equipment.

    Science.gov (United States)

    2010-07-01

    ... equipment. 273.4 Section 273.4 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) STANDARDS FOR UNIVERSAL WASTE MANAGEMENT General § 273.4 Applicability—Mercury...-containing components have been removed. (c) Generation of waste mercury-containing equipment. (1) Used...

  6. Technical considerations for evaluating substantially complete containment of high-level waste within the waste package

    International Nuclear Information System (INIS)

    Manaktala, H.K.; Interrante, C.G.

    1990-12-01

    This report deals with technical information that is considered essential for demonstrating the ability of the high-level radioactive waste package to provide ''substantially complete containment'' of its contents (vitrified waste form or spent light-water reactor fuel) for a period of 300 to 1000 years in a geological repository environment. The discussion is centered around technical considerations of the repository environment, materials and fabrication processes for the waste package components, various degradation modes of the materials of construction of the waste packages, and inspection and monitoring of the waste package during the preclosure and retrievability period, which could begin up to 50 years after initiation of waste emplacement. The emphasis in this report is on metallic materials. However, brief references have been made to other materials such as ceramics, graphite, bonded ceramic-metal systems, and other types of composites. The content of this report was presented to an external peer review panel of nine members at a workshop held at the Center for Nuclear Waste Regulatory Analyses (CNWRA), Southwest Research Institute, San Antonio, Texas, April 2--4, 1990. The recommendations of the peer review panel have been incorporated in this report. There are two companion reports; the second report in the series provides state-of-the-art techniques for uncertainty evaluations. 97 refs., 1 fig

  7. Technical considerations for evaluating substantially complete containment of high-level waste within the waste package

    Energy Technology Data Exchange (ETDEWEB)

    Manaktala, H.K. (Southwest Research Inst., San Antonio, TX (USA). Center for Nuclear Waste Regulatory Analyses); Interrante, C.G. (Nuclear Regulatory Commission, Washington, DC (USA). Div. of High-Level Waste Management)

    1990-12-01

    This report deals with technical information that is considered essential for demonstrating the ability of the high-level radioactive waste package to provide substantially complete containment'' of its contents (vitrified waste form or spent light-water reactor fuel) for a period of 300 to 1000 years in a geological repository environment. The discussion is centered around technical considerations of the repository environment, materials and fabrication processes for the waste package components, various degradation modes of the materials of construction of the waste packages, and inspection and monitoring of the waste package during the preclosure and retrievability period, which could begin up to 50 years after initiation of waste emplacement. The emphasis in this report is on metallic materials. However, brief references have been made to other materials such as ceramics, graphite, bonded ceramic-metal systems, and other types of composites. The content of this report was presented to an external peer review panel of nine members at a workshop held at the Center for Nuclear Waste Regulatory Analyses (CNWRA), Southwest Research Institute, San Antonio, Texas, April 2--4, 1990. The recommendations of the peer review panel have been incorporated in this report. There are two companion reports; the second report in the series provides state-of-the-art techniques for uncertainty evaluations. 97 refs., 1 fig.

  8. 242-A evaporator dangerous waste permit application

    International Nuclear Information System (INIS)

    1991-01-01

    The 242-A Evaporator is a waste management unit within the Hanford Facility that consists of process vessels and support systems for heating, evaporating, and condensing double-shell tank (DST) waste generated by Hanford Site operations. Operation of the 242-A Evaporator serves to reduce the volume of waste solutions within the DSTs that do not self-boil, while separating inorganic and radionuclide constituents from organic constituents. This operation reduces the number of underground DSTs required for waste storage and also makes the mixed waste more suitable for future treatment and disposal (i.e., grouting and vitrification). The 242-A Evaporator receives mixed-waste streams from the DSTs that contain organic and inorganic constituents and radionuclides. The waste is a dangerous waste (DW) because of corrosivity, reactivity, and toxicity characteristics, and is an extremely hazardous waste (EHW) as a result of toxicity (state criteria only), carcinogenicity, and persistence under the state mixture rule. The waste also contains spent nonhalogenated solvents

  9. Ferrocyanide-containing waste tanks: Ferrocyanide chemistry and reactivity

    International Nuclear Information System (INIS)

    Scheele, R.D.; Burger, L.L.; Tingey, J.M.; Bryan, S.A.; Borsheim, G.L.; Simpson, B.C.; Cash, R.J.; Cady, H.H.

    1991-09-01

    The complexing constant for hexacyano-iron complexes, both Fe(2) and Fe(3), are exceptionally large. The derived transition metal salts or double salts containing alkali metal ions are only slightly soluble. The various nickel compounds examined in this study, i.e., those predicted to have been formed in the Hanford waste scavenging program, are typical examples. In spite of their relative stability towards most reagents under ambient conditions, they are all thermodynamically unstable towards oxidation and react explosively with oxidants such as nitrate or nitrate salts when heated to temperatures in excess of 200 degree C. 42 refs., 5 figs., 3 tabs

  10. Multivariate methods in nuclear waste remediation: Needs and applications

    International Nuclear Information System (INIS)

    Pulsipher, B.A.

    1992-05-01

    The United States Department of Energy (DOE) has developed a strategy for nuclear waste remediation and environmental restoration at several major sites across the country. Nuclear and hazardous wastes are found in underground storage tanks, containment drums, soils, and facilities. Due to the many possible contaminants and complexities of sampling and analysis, multivariate methods are directly applicable. However, effective application of multivariate methods will require greater ability to communicate methods and results to a non-statistician community. Moreover, more flexible multivariate methods may be required to accommodate inherent sampling and analysis limitations. This paper outlines multivariate applications in the context of select DOE environmental restoration activities and identifies several perceived needs

  11. Plasma technologies: applications to waste processing

    International Nuclear Information System (INIS)

    Fauchais, P.

    2007-01-01

    Since the 1990's, plasma technologies have found applications in the processing of toxic wastes of military and industrial origin, like the treatment of contaminated solids and low level radioactive wastes, the decontamination of soils etc.. Since the years 2000, this development is becoming exponential, in particular for the processing of municipal wastes and the recovery of their synthesis gas. The advantage of thermal plasmas with respect to conventional combustion techniques are: a high temperature (more than 6000 K), a pyrolysis capability (CO formation instead of CO 2 ), about 90% of available energy above 1500 K (with respect to 23% with flames), a greater energy density, lower gas flow rates, and plasma start-up and shut-down times of only few tenth of seconds. This article presents: 1 - the present day situation of thermal plasmas development; 2 - some general considerations about plasma waste processing; 3 - the plasma processes: liquid toxic wastes, solid wastes (contaminated soils and low level radioactive wastes, military wastes, vitrification of incinerators fly ash, municipal wastes processing, treatment of asbestos fibers, treatment of chlorinated industrial wastes), metallurgy wastes (dusts, aluminium slags), medical and ship wastes, perspectives; 4 -conclusion. (J.S.)

  12. Grout treatment facility dangerous waste permit application

    International Nuclear Information System (INIS)

    1992-07-01

    The Grout Treatment Facility (GTF) will provide permanent disposal for approximately 43 Mgal of radioactive liquid waste currently being stored in underground tanks on the Hanford Site. The first step in permanent disposal is accomplished by solidifying the low-level liquid waste with cementitious dry materials. The resulting grout is cast within underground vaults. This report on the GTF contains information on the following: Hanford Site Maps, road evaluation for the grout treatment facility, Department of Ecology certificate of non-designation for centralia fly ash, double-shell tank waste compositional modeling, laboratory analysis reports for double-shell tank waste, stored in tanks 241-AN-103, 241-AN-106, and 241-AW-101, grout vault heat transfer results for M-106 grout formulation, test results for extraction procedure toxicity testing, test results for toxicity testing of double-shell tank grout, pilot-scale grout production test with a simulated low-level waste, characterization of simulated low-level waste grout produced in a pilot-scale test, description of the procedure for sampling nonaging waste storage tanks, description of laboratory procedures, grout campaign waste composition verification, variability in properties of grouted phosphate/sulfate N-reactor waste, engineering drawings, description of operating procedures, equipment list--transportable grout equipment, grout treatment facility--tank integrity assessment plan, long-term effects of waste solutions on concrete and reinforcing steel, vendor information, grout disposal facilities construction quality assurance plan, and flexible membrane liner/waste compatibility test results

  13. Radionuclide transport through perforations in nuclear waste containers

    International Nuclear Information System (INIS)

    Aidun, C.K.; Bloom, S.G.; Raines, G.E.

    1987-11-01

    Previous analytical models for the steady-state radionuclide release rate through perforations in nuclear waste containers into the surrounding medium are based on a zero wall thickness assumption. In this paper we investigate the effect of the wall thickness on the mass transfer rate through isolated cylindrical holes. We solve the steady-state diffusion equation for the concentration field and derive a model based on the analytical solution. By direct comparison, we show that the zero wall thickness model overpredicts the mass transfer rate by about 1300 percent for a circular hole with 1-cm radius and a wall thickness of 10 cm. As expected, the zero-thickness model becomes even less accurate as the hole radius decreases; it predicts a greater release rate from a large number of small holes than the mass transfer rate from an uncontained waste form cylinder. In contrast, the results predicted by our model remain bounded for isolated holes and never exceed the mass transfer from an uncontained waste form. 6 refs., 9 figs., 3 tabs

  14. Container materials for isolation of radioactive waste in salt

    International Nuclear Information System (INIS)

    Streicher, M.A.; Andrews, A.

    1987-10-01

    The workshop reviewed the extensive data on the corrosion resistance of low-carbon steel in simulated salt repository environments, determined whether these data were sufficient to recommend low-carbon steel for fabrication of the container, and assessed the suitability of other materials under consideration in the SRP. The panelists determined the need for testing and research programs, recommended experimental approaches, and recommended materials based on existing technology. On the first day of the workshop, presentations were made on waste package requirements; the expected corrosion environment; degradation processes, including a review of data from corrosion tests on carbon steel; and rationales for container design and materials, modeling studies, and planned future work. The second day was devoted to a panel caucus, presentation of workshop findings, and open discussion. 76 refs., 2 figs., 3 tabs

  15. Waste minimization applications at a remediation site

    International Nuclear Information System (INIS)

    Allmon, L.A.

    1995-01-01

    The Fernald Environmental Management Project (FEMP) owned by the Department of Energy was used for the processing of uranium. In 1989 Fernald suspended production of uranium metals and was placed on the National Priorities List (NPL). The site's mission has changed from one of production to environmental restoration. Many groups necessary for producing a product were deemed irrelevant for remediation work, including Waste Minimization. Waste Minimization does not readily appear to be applicable to remediation work. Environmental remediation is designed to correct adverse impacts to the environment from past operations and generates significant amounts of waste requiring management. The premise of pollution prevention is to avoid waste generation, thus remediation is in direct conflict with this premise. Although greater amounts of waste will be generated during environmental remediation, treatment capacities are not always available and disposal is becoming more difficult and costly. This creates the need for pollution prevention and waste minimization. Applying waste minimization principles at a remediation site is an enormous challenge. If the remediation site is also radiologically contaminated it is even a bigger challenge. Innovative techniques and ideas must be utilized to achieve reductions in the amount of waste that must be managed or dispositioned. At Fernald the waste minimization paradigm was shifted from focusing efforts on source reduction to focusing efforts on recycle/reuse by inverting the EPA waste management hierarchy. A fundamental difference at remediation sites is that source reduction has limited applicability to legacy wastes but can be applied successfully on secondary waste generation. The bulk of measurable waste reduction will be achieved by the recycle/reuse of primary wastes and by segregation and decontamination of secondary wastestreams. Each effort must be measured in terms of being economically and ecologically beneficial

  16. Microbial control on decomposition of radionuclides-containing oily waste in soil

    Science.gov (United States)

    Selivanovskaya, Svetlana; Galitskaya, Polina

    2014-05-01

    The oily wastes are formed annually during extraction, refinement, and transportation of the oil and may cause pollution of the environment. These wastes contain different concentrations of waste oil (40-60%), waste water (30-90%), and mineral particles (5-40%). Some oily wastes also contain naturally occurring radionuclides which were incorporated by water that was pumped up with the oil. For assessment of the hazard level of waste treated soil, not only measurements of contaminants content are needed, because bioavailability of oily components varies with hydrocarbon type, and soil properties. As far as namely microbial communities control the decomposition of organic contaminants, biological indicators have become increasingly important in hazard assessment and the efficiency of remediation process. In this study the decomposition of radionuclides-containing oily waste by soil microbial communities were estimated. Waste samples collected at the Tikchonovskii petroleum production yard (Tatarstan, Russia) were mixed with Haplic greyzem soil at ratio 1:4 and incubated for 120 days. During incubation period, the total hydrocarbon content of the soil mixed with the waste reduced from 156 ± 48 g kg-1 to 54 ± 8 g kg-1 of soil. The concentrations of 226Ra and 232Th were found to be 643 ± 127, 254 ± 56 Bq kg-1 and not changed significantly during incubation. Waste application led to a soil microbial biomass carbon decrease in comparison to control (1.9 times after 1 day and 1.3 times after 120 days of incubation). Microbial respiration increased in the first month of incubation (up to 120% and 160% of control after 1 and 30 days, correspondingly) and decreased to the end of incubation period (74% of control after 120 days). Structure of bacterial community in soil and soil/waste mixture was estimated after 120 days of incubation using SSCP method. The band number decreased in contaminated soil in comparison to untreated soil. Besides, several new dominant DNA

  17. Lining materials for waste disposal containment and waste storage facilities. (Latest citations from the NTIS bibliographic database). Published Search

    International Nuclear Information System (INIS)

    1993-11-01

    The bibliography contains citations concerning the design characteristics, performance, and materials used to make liners for the waste disposal and storage industry. Liners made of concrete, polymeric materials, compacted clays, asphalt, and in-situ glass are discussed. The use of these liners to contain municipal wastes, hazardous waste liquids, and both low-level and high-level radioactive wastes is presented. Liner permeability, transport, stability, construction, and design are studied. Laboratory field measurements for specific wastes are included. (Contains a minimum of 213 citations and includes a subject term index and title list.)

  18. Thermosyphon evaporator for nuclear waste management application

    Energy Technology Data Exchange (ETDEWEB)

    Menon, Rajani; Singh, A K; Rana, D S [Waste Management Projects Division, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    Nuclear plant equipment are associated with radioactive material which needs to be safely contained under all conditions of operation. Because of large radioactivity associated with the operations of nuclear waste management plants, the equipment are not accessible to human intervention. Hence, the design of the equipment needs to incorporate features for high reliability and safety so as to avoid unnecessary outage. As far as possible the equipment must be maintenance free. Wherever maintenance is inevitable, it has to be designed to be carried out without exposure of personnel to radiation, preventing spread of radiation or contamination. This paper outlines the design features of a thermosyphon evaporator for nuclear application. (author). 2 figs., 1 tab.

  19. Experiences in the emptying of waste silos containing solid nuclear waste from graphite- moderated reactors

    International Nuclear Information System (INIS)

    Wall, S.; Schwarz, T.

    2003-01-01

    Before reactor sites can be handed over for ultimate decommissioning, at some sites silos containing waste from operations need to be emptied. The form and physical condition of the waste demands sophisticated retrieval technologies taking into account the onsite situation in terms of infrastructure and silo geometry. Furthermore, in the case of graphite moderated reactors, this waste usually includes several tonnes of graphite waste requiring special HVAC and dust handling measures. RWE NUKEM Group has already performed several contracts dealing with such emptying tasks. Of particular interest for the upcoming decommissioning projects in France might be the activities at Vandellos, Spain and Trawsfynnyd, UK. Retrieval System for Vandellos NPP is discussed. Following an international competitive tender exercise, RWE NUKEM won the contract to provide a turn-key retrieval system. This involved the design, manufacture and installation of a system built around the modules of a 200 kg capacity version of the ARTISAN manipulator system. The ARTISAN 200 manipulator, with remote slave arm detach facility, was deployed on a telescopic mast inserted into the silos through the roof penetrations. The manipulator deployed a range of tools to gather the waste and load it into a transfer basket, deployed through an adjacent penetration. After commissioning, the system cleared the vaults in less than the scheduled period with no failures. At the Trawsfynnyd Magnox plants two types of intermediate level waste (ILW) accumulated on site; namely Miscellaneous Activated Components (MAC) and Fuel Element Debris (FED). MAC is predominantly components that have been activated by the reactor core and then discharged. FED mainly consists of fuel cladding produced when fuel elements were prepared for dispatch to the reprocessing facility. RWE NUKEM Ltd. was awarded a contract to design, supply, commission and operate equipment to retrieve, pack and immobilize the two waste streams. Major

  20. Properties of slag concrete for low-level waste containment

    International Nuclear Information System (INIS)

    Langton, C.A.; Wong, P.B.

    1991-01-01

    Ground granulated blast furnace slag was incorporated in the concrete mix used for construction of low-level radioactive waste disposal vaults. The vaults were constructed as six 100 x 100 x 25 ft cells with each cell sharing internal walls with the two adjacent cells. The vaults were designed to contain a low-level radioactive wasteform called saltstone and to isolate the saltstone from the environment until the landfill is closed. Closure involves backfilling with native soil, installation of clay cap, and run-off control. The design criteria for the slag-substituted concrete included compressive strength, 4000 psi after 28 days; slump, 6 inch; permeability, less than 10 -7 cm/sec; and effective nitrate, chromium and technetium diffusivities of 10 -8 , 10 -12 and 10 -12 cm 2 /sec, respectively. The reducing capacity of the slag resulted in chemically reducing Cr +6 to Cr +3 and Tc +7 to Tc +4 and subsequent precipitation of the respective hydroxides in the alkaline pore solution. Consequently, the concrete vault enhances containment of otherwise mobile waste ions and contributes to the overall protection of the groundwater at the disposal site

  1. Sealing method and sealing device for radioactive waste containing vessel

    International Nuclear Information System (INIS)

    Ishiwatari, Koji; Otsuki, Akira

    1998-01-01

    A radioactive waste-containing body is hoisted down into a strong-material vessel opened upwardly, and a strong-material lid is hoisted down to the opening of the strong-material-vessel and welded. The strong material vessel is hoisted up and loaded on a corrosion resistant-material bottom plate placed horizontally. A corrosion resistant-material vessel having one opening end and having a corrosion resistant-material flange on the other end and previously agreed with the strong material-vessel main body is hoisted up by a hoisting device having an inserting device so that the opening of the corrosion resistant vessel is directed downwardly. The corrosion resistant vessel is press-fitted to the outside of the strong material-vessel by the inserting device while being heated by a preheater to shrink. Subsequently, the lower end of the corrosion resistant-material vessel and the corrosion resistant-material bottom plate are welded to constitute a corrosion resistant-material vessel. Then, the radioactive waste containing body can be sealed in a sealing vessel comprising the strong-material vessel and the corrosion resistant-material vessel. (N.H.)

  2. Wool-waste as organic nutrient source for container-grown plants

    Energy Technology Data Exchange (ETDEWEB)

    Zheljazkov, Valtcho D. [Mississippi State University, North Mississippi Research and Extension Center, Verona, MS 38879 (United States)], E-mail: vj40@pss.msstate.edu; Stratton, Glenn W [Department of Plant and Animal Sciences and Department of Environmental Sciences, Nova Scotia Agricultural College, Truro, NS, B2N 5E3 (Canada); Pincock, James [Department of Chemistry, Dalhousie University, Halifax, NS, B3H 4J3 (Canada); Butler, Stephanie [Department of Plant and Animal Sciences and Department of Environmental Sciences, Nova Scotia Agricultural College, Truro, NS, B2N 5E3 (Canada); Jeliazkova, Ekaterina A [Mississippi State University, Department of Plant and Soil Sciences, Mississippi State, MS 39762 (United States); Nedkov, Nedko K [Research Institute for Roses and Aromatic Crops, 49 Osvobojdenie Blv., Kazanluk (Bulgaria); Gerard, Patrick D [Department of Applied Economics and Statistics, Clemson University, Clemson, SC 29634 (United States)

    2009-07-15

    A container experiment was conducted to test the hypothesis that uncomposted wool wastes could be used as nutrient source and growth medium constituent for container-grown plants. The treatments were: (1) rate of wool-waste application (0 or unamended control, 20, 40, 80, and 120 g of wool per 8-in. pot), (2) growth medium constituents [(2.1) wool plus perlite, (2.2) wool plus peat, and (2.3) wool plus peat plus perlite], and (3) plant species (basil and Swiss chard). A single addition of 20, 40, 80, or 120 g of wool-waste to Swiss chard (Beta vulgaris L.) and basil (Ocimum basilicum L.) in pots with growth medium provided four harvests of Swiss chard and five harvests of basil. Total basil yield from the five harvests was 1.6-5 times greater than the total yield from the unamended control, while total Swiss chard yield from the four harvests was 2-5 times greater relative to the respective unamended control. The addition of wool-waste to the growth medium increased Swiss chard and basil tissue N, and NO{sub 3}-N and NH{sub 4}-N in growth medium relative to the unamended control. Scanning electron microscopy (SEM) and energy dispersive X-ray (EDX) microanalysis of wool fibers sampled at the end of the experiments indicated various levels of decomposition, with some fibers retaining their original surface structure. Furthermore, most of the wool fibers' surfaces contained significant concentrations of S and much less N, P, or K. SEM/EDX revealed that some plant roots grow directly on wool-waste fibers suggesting either (1) root directional growth towards sites with greater nutrient concentration and/or (2) a possible role for roots or root exudates in wool decomposition. Results from this study suggest that uncomposted wool wastes can be used as soil amendment, growth medium constituent, and nutrient source for container-grown plants.

  3. Wool-waste as organic nutrient source for container-grown plants

    International Nuclear Information System (INIS)

    Zheljazkov, Valtcho D.; Stratton, Glenn W.; Pincock, James; Butler, Stephanie; Jeliazkova, Ekaterina A.; Nedkov, Nedko K.; Gerard, Patrick D.

    2009-01-01

    A container experiment was conducted to test the hypothesis that uncomposted wool wastes could be used as nutrient source and growth medium constituent for container-grown plants. The treatments were: (1) rate of wool-waste application (0 or unamended control, 20, 40, 80, and 120 g of wool per 8-in. pot), (2) growth medium constituents [(2.1) wool plus perlite, (2.2) wool plus peat, and (2.3) wool plus peat plus perlite], and (3) plant species (basil and Swiss chard). A single addition of 20, 40, 80, or 120 g of wool-waste to Swiss chard (Beta vulgaris L.) and basil (Ocimum basilicum L.) in pots with growth medium provided four harvests of Swiss chard and five harvests of basil. Total basil yield from the five harvests was 1.6-5 times greater than the total yield from the unamended control, while total Swiss chard yield from the four harvests was 2-5 times greater relative to the respective unamended control. The addition of wool-waste to the growth medium increased Swiss chard and basil tissue N, and NO 3 -N and NH 4 -N in growth medium relative to the unamended control. Scanning electron microscopy (SEM) and energy dispersive X-ray (EDX) microanalysis of wool fibers sampled at the end of the experiments indicated various levels of decomposition, with some fibers retaining their original surface structure. Furthermore, most of the wool fibers' surfaces contained significant concentrations of S and much less N, P, or K. SEM/EDX revealed that some plant roots grow directly on wool-waste fibers suggesting either (1) root directional growth towards sites with greater nutrient concentration and/or (2) a possible role for roots or root exudates in wool decomposition. Results from this study suggest that uncomposted wool wastes can be used as soil amendment, growth medium constituent, and nutrient source for container-grown plants.

  4. Applications to waste management operations

    International Nuclear Information System (INIS)

    Paine, D.; Uresk, V.; Schreckhise, R.G.

    1977-01-01

    Ecological studies of the 200 Area plateau waste management environs have provided preliminary answers to questions concerning the environmental health of associated biota, potential for radionuclide transport through the biotic system and risk to man. More importantly creation of this ecological data base provides visible evidence of environmental expertise so essential for maintenance of continued public confidence in waste management operations

  5. The stress corrosion cracking of copper nuclear waste containers

    International Nuclear Information System (INIS)

    King, F.; Litke, C.D.; Ikeda, B.M.

    1999-01-01

    The extent of stress corrosion cracking (SCC) of copper nuclear waste containers is being predicted on the basis of a 'limited propagation' argument. In this argument, it is accepted that crack initiation may occur, but it is argued that the environmental conditions and material properties required for a through-wall crack to propagate will not be present. In this paper, the effect of one environmental parameter, the supply of oxidant (J ox ), on the crack growth rate is examined. Experiments have been conducted on two grades of Cu in NANO 2 environments using two loading techniques. The supply of oxidant has been varied either electrochemically in bulk solution using different applied current densities or by embedding the loaded test specimens in compacted buffer material containing O 2 as the oxidant. Measured and theoretical crack growth rates as a function of J ox are compared with the predicted oxidant flux to the containers in a disposal vault and an estimate of the maximum crack depth on a container obtained. (author)

  6. The stress corrosion cracking of copper nuclear waste containers

    International Nuclear Information System (INIS)

    King, F.; Litke, C.D.; Ikeda, B.M.

    1999-01-01

    The extent of stress corrosion cracking (SCC) of copper nuclear waste containers is being predicted on the basis of a limited propagation argument. In this argument, it is accepted that crack initiation may occur, but it is argued that the environmental conditions and material properties required for a through-wall crack to propagate will not be present. In this paper, the effect of one environmental parameter, the supply of oxidant (J OX ), on the crack growth rate is examined. Experiments have been conducted on two grades of Cu in NaNO 2 environments using two loading techniques. The supply of oxidant has been varied either electrochemically in bulk solution using different applied current densities or by embedding the loaded test specimens in compacted buffer material containing O 2 as the oxidant. Measured and theoretical crack growth rates as a function of J OX are compared with the predicted oxidant flux to the containers in a disposal vault and an estimate of the maximum crack depth on a container obtained

  7. Hanford facility dangerous waste permit application, general information portion

    International Nuclear Information System (INIS)

    Hays, C.B.

    1998-01-01

    The Hanford Facility Dangerous Waste Permit Application is considered to be a single application organized into a General Information Portion (document number DOE/RL-91-28) and a Unit-Specific Portion. Both the General Information and Unit-Specific portions of the Hanford Facility Dangerous Waste Permit Application address the content of the Part B permit application guidance prepared by the Washington State Department of Ecology (Ecology 1996) and the U.S. Environmental Protection Agency (40 Code of Federal Regulations 270), with additional information needed by the Hazardous and Solid Waste Amendments and revisions of Washington Administrative Code 173-303. Documentation contained in the General Information Portion is broader in nature and could be used by multiple treatment, storage, and/or disposal units (e.g., the glossary provided in this report)

  8. French processes for waste embedding. The use of epoxy resin for waste containment

    International Nuclear Information System (INIS)

    Augustin, X.; Gauthey, J.C.

    1993-01-01

    The low- and medium-level wastes generated by nuclear facilities when operating as well as during their decommissioning (dismantling, decontamination, etc.) are embedded for the purpose of obtaining a product suitable for disposal. Due to the varieties of waste produced, it was necessary to resort to multi-purpose techniques to solve problems relating to their embedding. The process for waste embedding in thermosetting polymer (polyester, epoxy) developed by the French Atomic Energy Commission (CEA) and its subsidiary TECHNICATOME is easy to operate and yields excellent results having regard to volume reduction and containment of radioisotopes (particularly caesium). The industrial development of this process has led to the design of small, flexible, fixed or mobile, embedding stations. Examples illustrating the increasing use of this process during facility dismantling are described

  9. In situ vitrification: Application to buried waste

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  10. Elaboration of new ceramic composites containing glass fibre production wastes

    International Nuclear Information System (INIS)

    Rozenstrauha, I.; Sosins, G.; Krage, L.; Sedmale, G.; Vaiciukyniene, D.

    2013-01-01

    Two main by-products or waste from the production of glass fibre are following: sewage sludge containing montmorillonite clay as sorbent material and ca 50 % of organic matter as well as waste glass from aluminium borosilicate glass fibre with relatively high softening temperature (> 600 degree centigrade). In order to elaborate different new ceramic products (porous or dense composites) the mentioned by-products and illitic clay from two different layers of Apriki deposit (Latvia) with illite content in clay fraction up to 80-90 % was used as a matrix. The raw materials were investigated by differential-thermal (DTA) and XRD analysis. Ternary compositions were prepared from mixtures of 15 - 35 wt % of sludge, 20 wt % of waste glass and 45 - 65 wt % of clay and the pressed green bodies were thermally treated in sintering temperature range from 1080 to 1120 degree centigrade in different treatment conditions. Materials produced in temperature range 1090 - 1100 degree centigrade with the most optimal properties - porosity 38 - 52 %, water absorption 39 -47 % and bulk density 1.35 - 1.67 g/cm 3 were selected for production of porous ceramics and materials showing porosity 0.35 - 1.1 %, water absorption 0.7 - 2.6 % and bulk density 2.1 - 2.3 g/cm 3 - for dense ceramic composites. Obtained results indicated that incorporation up to 25 wt % of sewage sludge is beneficial for production of both ceramic products and glass-ceramic composites according to the technological properties. Structural analysis of elaborated composite materials was performed by scanning electron microscopy(SEM). By X-ray diffraction analysis (XRD) the quartz, diopside and anorthite crystalline phases were detected. (Author)

  11. Processing method for contaminated water containing radioactive waste

    International Nuclear Information System (INIS)

    Tahara, Toshiaki; Fukagawa, Ken-ichiro.

    1994-01-01

    For absorbing contaminated water containing radioactive substances, a sheet is prepared by covering water absorbing pulps carrying an organic water absorbent having an excellent water absorbability is semi-solidified upon absorption water with a water permeable cloth, such as a non-woven fabric having a shape stability. As the organic water absorbent, a hydrophilic polymer which retains adsorbed water as it is used. In particular, a starch-grafted copolymer having an excellent water absorbability also for reactor water containing boric acid is preferred. The organic water absorbent can be carried on the water absorbing pulps by scattering a granular organic water absorbent to the entire surface of the water absorbing cotton pulp extended thinly to carry it uniformly and putting them between thin absorbing paper sheets. If contaminated water containing radioactive materials are wiped off by using such a sheet, the entire sheet is semi-solidified along with the absorption with no leaching of the contaminated water, thereby enabling to move the wastes to a furnace for applying combustion treatment. (T.M.)

  12. Gamma radiation induced changes in nuclear waste glass containing Eu

    Science.gov (United States)

    Mohapatra, M.; Kadam, R. M.; Mishra, R. K.; Kaushik, C. P.; Tomar, B. S.; Godbole, S. V.

    2011-10-01

    Gamma radiation induced changes were investigated in sodium-barium borosilicate glasses containing Eu. The glass composition was similar to that of nuclear waste glasses used for vitrifying Trombay research reactor nuclear waste at Bhabha Atomic Research Centre, India. Photoluminescence (PL) and electron paramagnetic resonance (EPR) techniques were used to study the speciation of the rare earth (RE) ion in the matrix before and after gamma irradiation. Judd-Ofelt ( J- O) analyses of the emission spectra were done before and after irradiation. The spin counting technique was employed to quantify the number of defect centres formed in the glass at the highest gamma dose studied. PL data suggested the stabilisation of the trivalent RE ion in the borosilicate glass matrix both before and after irradiation. It was also observed that, the RE ion distributes itself in two different environments in the irradiated glass. From the EPR data it was observed that, boron oxygen hole centre based radicals are the predominant defect centres produced in the glass after irradiation along with small amount of E’ centres. From the spin counting studies the concentration of defect centres in the glass was calculated to be 350 ppm at 900 kGy. This indicated the fact that bulk of the glass remained unaffected after gamma irradiation up to 900 kGy.

  13. Leach rate studies on glass containing actual radioactive waste

    International Nuclear Information System (INIS)

    Walker, D.D.; Wiley, J.R.; Dukes, M.D.; LeRoy, J.H.

    1980-01-01

    Borosilicate glass containing radioactive wastes from the Savannah River Plant have been leached for 900 days. The International Standards Organization's (ISO) static leach test procedure was used on glass buttons in various leachants. Leach rates based on 90 Sr and 137 Cs analyses were similar: 2 x 10 -8 to 3 x 10 -8 g/(cm 2 )(d) in distilled water, 1 x 10 -8 to 3 x 10 -7 g/(cm 2 )(d) in pH 7 buffer, 3 x 10 -7 to 7 x 10 -7 g/(cm 2 )(d) in pH 9 buffer, and 7 x 10 -6 to 8 x 10 -5 g/(cm 2 )(d) in pH 4 buffer. Rates based on Pu analyses were the same as above in distilled water and pH 9 buffer, but were lower by an order of magnitude in pH 4 and pH 7 buffers. Almost all leach rates remained constant between 200 and 900 days of leaching. Increasing the concentration of the buffering agents had no effect on the leach rates at pH 7 (phosphate) and pH 9 (carbonate), but dramatically increased the rates at pH 4 (acetate). Leach rates did not differ significantly between high aluminum and high iron waste glasses

  14. Radioactivity concentration measuring device for radiation waste containing vessel

    International Nuclear Information System (INIS)

    Goto, Tetsuo.

    1994-01-01

    The device of the present invention can precisely and accurately measure a radioactive concentration of radioactive wastes irrespective of the radioactivity concentration distribution. Namely, a Ge detector having a collimator and a plurality of radiation detectors are placed at the outside of the radioactive waste containing vessel in such a way that it can rotate and move vertically relative to the vessel. The plurality of radiation detectors detect radiation coefficient signals at an assumed segment unit of a predetermined length in vertical direction and for every predetermined angle unit in the rotational direction. A weight measuring device determines the weight of the vessel. A computer calculates an average density of radioactivity for the region filled with radioactivity based on the determined net weight and radiation coefficient signals assuming that the volume of the radioactivity is constant. In addition, the computer calculates the amount of radioactivity in the assumed segment by conducting γ -ray absorption compensation calculation for the material in the vessel. Each of the amount of radioactivity is integrated to determine the amount of radioactivity in the vessel. (I.S.)

  15. Method of detecting water leakage in radioactive waste containing vessel

    International Nuclear Information System (INIS)

    Ishioka, Hitoshi; Takao, Yoshiaki; Hayakawa, Kiyoshige.

    1989-01-01

    Lower level radioactive wastes formed upon operation of nuclear facilities are processed by underground storage. In this case, a plurality of drum cans packed with radioactive wastes are contained in a vessel and a water soluble dye material is placed at the inside of the vessel. The method of placing the water soluble dye material at the inside of the vessel includes a method of coating the material on the inner surface of the vessel and a method of mixing the material in sands to be filled between each of the drum cans. Then, leakage of water soluble dye material is detected when water intruding from the outside into the vessel is again leached out of the vessel, to detect the water leakage from the inside of the vessel. In this way, it is possible to find a water-invaded vessel before corrosion of the drum can by water intruded into the vessel and leakage of nuclides in the drum can. Accordingly, it is possible to apply treatment such as repair before occurrence of accident and can maintain the safety of radioactive water processing facilities. (I.S.)

  16. Immobilization of calcium sulfate contained in demolition waste

    International Nuclear Information System (INIS)

    Ambroise, J.; Pera, J.

    2008-01-01

    This paper presents the results of a laboratory study undertaken to examine the treatment of demolition waste containing calcium sulfate by means of calcium sulfoaluminate clinker (CSA). The quantity of CSA necessary to entirely consume calcium sulfate was determined. Using infrared spectrometry analysis and X-ray diffraction, it was shown that calcium sulfate was entirely consumed when the ratio between CSA and calcium sulfate was 4. Standard sand was polluted by 4% calcium sulfate. Two solutions were investigated: ·either global treatment of sand by CSA, ·or immobilization of calcium sulfate by CSA, followed by the introduction of this milled mixture in standard sand. Regardless of the type of treatment, swelling was almost stabilized after 28 days of immersion in water

  17. In situ vitrification applications to hazardous wastes

    International Nuclear Information System (INIS)

    Liikala, S.

    1989-01-01

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

  18. Management of radioactive waste from nuclear applications

    International Nuclear Information System (INIS)

    1997-01-01

    Radioactive waste arises from the generation of nuclear energy and from the production of radioactive materials and their applications in industry, agriculture, research and medicine. The importance of safe management of radioactive waste for the protection of human health and the environment has long been recognized and considerable experience has been gained in this field. Technical expertise is a prerequisite for safe and cost-effective management of radioactive waste. A training course is considered an effective tool for providing technical expertise in various aspects of waste management. The IAEA, in co-operation with national authorities concerned with radioactive waste management, has organized and conducted a number of radioactive waste management training courses. The results of the courses conducted by the IAEA in 1991-1995 have been evaluated at consultants meetings held in December 1995 and May 1996. This guidance document for use by Member States in arranging national training courses on the management of low and intermediate level radioactive waste from nuclear applications has been prepared as the result of that effort. The report outlines the various requirements for the organization, conduct and evaluation of training courses in radioactive waste management and proposes an annotated outline of a reference training course

  19. Grout treatment facility dangerous waste permit application

    International Nuclear Information System (INIS)

    1992-07-01

    The Grout Treatment Facility (GTF) will provide permanent disposal for approximately 43 Mgal of radioactive liquid waste currently being stored in underground tanks on the Hanford Site. The first step in permanent disposal is accomplished by solidifying the liquid waste with cementitious dry materials. The resulting grout is cast within underground vaults. This report on the GTF contains information on the following: Vault design, run-on/run-off control design, and asphalt compatibility with 90-degree celsius double-shell slurry feed

  20. Wet air oxidation of seedcorn wastes containing pesticides and insecticides

    Energy Technology Data Exchange (ETDEWEB)

    Sievers, M.; Schlaefer, O.; Onyeche, T.I.; Schroeder, C.; Bormann, H.; Schaefer, S. [CUTEC-Inst. GmbH (Clausthal Environment Technology Inst.), Clausthal-Zellerfeld (Germany)

    2003-07-01

    Wet air oxidation as an alternative treatment process to pyrolysis and combustion of seedcorn wastes was investigated in lab-scale experiments. Due to solid condition of the seed corn waste, the process has been adapted by repeated spraying of water on the seed corn bulk to avoid the production of sludge and its subsequent dewatering. Original seed corns from industrial production plants were used for a degradation kinetic study under smooth wet air oxidation conditions. The temperatures were between 80 and 150 C, the pressure from 1 to 4.5 bar and the pH at different values from 3 to 13. Degradation rates for five different compounds of pesticides and insecticides, namely Imidacloprid, Thiram, Hymexazol, Carbofuran and Tefluthrin were conducted. These compounds represent the recently used in agricultural seedcorn applications. The degradation rate depends linearly on temperature between 80 and 150 C. At 120 C the lowest degradation rate was found for Tefluthrin by 25 mg/h per L reaction volume while the highest degradation rate to be conducted was for Imidacloprid at 363 mg/h L. (orig.)

  1. 616 Nonradioactive Dangerous Waste Storage Facility dangerous waste permit application

    International Nuclear Information System (INIS)

    1991-10-01

    The 616 Nonradioactive Dangerous Waste Storage Facility Dangerous Waste Permit Application consists of both a Part A and a Part B permit application. An explanation of the Part A revisions associated with this storage unit, including the Part A included with this document, is provided at the beginning of the Part A Section. The Part B consists of 15 chapters addressing the organization and content of the Part B Checklist prepared by the Washington State Department of Ecology (Ecology 1987). For ease of reference, the checklist section numbers, in brackets, follow chapter headings and subheadings. The 616 Nonradioactive Dangerous Waste Storage Facility Dangerous Waste Permit Application (Revision 0) was submitted to the Washington State Department of Ecology and the US Environmental Protection Agency on July 31, 1989. Revision 1, addressing Washington State Department of Ecology review comments made on Revision 0 dated November 21, 1989, and March 23, 1990, was submitted on June 22, 1990. This submittal, Revision 2, addresses Washington State Department of Ecology review comments made on Revision 1, dated June 22, 1990, August 30, 1990, December 18, 1990, and July 8, 1991

  2. Application of Epoxy Based Coating Instacote on Waste Tank Tops

    International Nuclear Information System (INIS)

    Pike, J.A.

    1998-01-01

    This evaluation examines the compatibility of coating Instacote with existing High-Level Waste facilities and safety practices. No significant incompatibilities are identified. The following actions need to be completed as indicated when applying Instacote on waste tank tops:(1) Prior to application in ITP facilities, the final product should be tested for chemical resistance to sodium tetraphenylborate solutions or sodium titanate slurries.(2) Any waste contaminated with Part A or B that can not be removed by the vendor such as for radiological contamination, HLW must hold the waste until HLW completes a formal assessment of the waste, disposal criteria, and impact.(3) Prior to the start of any application of the coating, each riser needs to be evaluated for masking and masking applied if needed.(4) At the conclusion of an application actual total weight of material applied to a waste tank needs to documented and sent to the tank top loading files for reference purposes.(5) Verify that the final product contains less than 250 ppm chloride

  3. Proposed Objective Odor Control Test Methodology for Waste Containment

    Science.gov (United States)

    Vos, Gordon

    2010-01-01

    The Orion Cockpit Working Group has requested that an odor control testing methodology be proposed to evaluate the odor containment effectiveness of waste disposal bags to be flown on the Orion Crew Exploration Vehicle. As a standardized "odor containment" test does not appear to be a matter of record for the project, a new test method is being proposed. This method is based on existing test methods used in industrial hygiene for the evaluation of respirator fit in occupational settings, and takes into consideration peer reviewed documentation of human odor thresholds for standardized contaminates, industry stardnard atmostpheric testing methodologies, and established criteria for laboratory analysis. The proposed methodology is quantitative, though it can readily be complimented with a qualitative subjective assessment. Isoamyl acetate (IAA - also known at isopentyl acetate) is commonly used in respirator fit testing, and there are documented methodologies for both measuring its quantitative airborne concentrations. IAA is a clear, colorless liquid with a banana-like odor, documented detectable smell threshold for humans of 0.025 PPM, and a 15 PPB level of quantation limit.

  4. Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

    International Nuclear Information System (INIS)

    Gdowski, G.E.; Bullen, D.B.

    1988-08-01

    Three copper-based alloys and three iron- to nickel-based austenitic alloys are being considered as possible materials for fabrication of containers for disposal of high-level radioactive waste. This waste will include spent fuel assemblies from reactors as well as high-level waste in borosilicate glass and will be sent to the prospective site at Yucca Mountain, Nevada, for disposal. The containers must maintain substantially complete containment for at least 300 yr and perhaps as long as 1000 yr. During the first 50 yr after emplacement, they must be retrievable from the disposal site. Shortly after the containers are emplaced in the repository, they will be exposed to high temperatures and high gamma radiation fields from the decay of the high-level waste. This volume surveys the available data on oxidation and corrosion of the iron- to nickel-based austenitic materials (Types 304L and 316L stainless steels and Alloy 825) and the copper-based alloy materials [CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni)], which are the present candidates for fabrication of the containers. Studies that provided a large amount of data are highlighted, and those areas in which little data exists are identified. Examples of successful applications of these materials are given. On the basis of resistance to oxidation and general corrosion, the austenitic materials are ranked as follows: Alloy 825 (best), Type 316L stainless steel, and then Type 304L stainless steel (worst). For the copper-based materials, the ranking is as follows: CDA 715 and CDA 613 (both best), and CDA 102 (worst). 110 refs., 30 figs., 13 tabs

  5. Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

    Energy Technology Data Exchange (ETDEWEB)

    Gdowski, G.E.; Bullen, D.B. (Science and Engineering Associates, Inc., Pleasanton, CA (USA))

    1988-08-01

    Three copper-based alloys and three iron- to nickel-based austenitic alloys are being considered as possible materials for fabrication of containers for disposal of high-level radioactive waste. This waste will include spent fuel assemblies from reactors as well as high-level waste in borosilicate glass and will be sent to the prospective site at Yucca Mountain, Nevada, for disposal. The containers must maintain substantially complete containment for at least 300 yr and perhaps as long as 1000 yr. During the first 50 yr after emplacement, they must be retrievable from the disposal site. Shortly after the containers are emplaced in the repository, they will be exposed to high temperatures and high gamma radiation fields from the decay of the high-level waste. This volume surveys the available data on oxidation and corrosion of the iron- to nickel-based austenitic materials (Types 304L and 316L stainless steels and Alloy 825) and the copper-based alloy materials (CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni)), which are the present candidates for fabrication of the containers. Studies that provided a large amount of data are highlighted, and those areas in which little data exists are identified. Examples of successful applications of these materials are given. On the basis of resistance to oxidation and general corrosion, the austenitic materials are ranked as follows: Alloy 825 (best), Type 316L stainless steel, and then Type 304L stainless steel (worst). For the copper-based materials, the ranking is as follows: CDA 715 and CDA 613 (both best), and CDA 102 (worst). 110 refs., 30 figs., 13 tabs.

  6. Radioactive waste disposal: Regulations and Application

    International Nuclear Information System (INIS)

    Hebert, Jean.

    1977-01-01

    The regulation of radioactive discharges, i.e. solid radioactive waste resulting from operation of nuclear installations and liquid and gazeous effluents released by them may be dealt with from two angles: the receiving environment and the polluting agent. French law covers both. Law on atmospheric pollution is based mainly on the Act of 2 August 1961 while the Act of 16 December 1964 governs water pollution. Both Acts have been the subject of a great number of implementing decrees, certain of which contain standards specific to radioactive pollution. Regulations on the polluting agent, namely its activity, comply with the generally established distinction between large nuclear installations and others. There again, there are many applicable texts, in particular, the Act of 19 July 1976 for classified installations, and the Decree of 11 December 1963, supplemented by the Decrees of 6 November 1974 and 31 December 1974 for large nuclear installations. This detailed analysis of national regulations is followed by a presentation of the applicable provisions in the Communities law and in international public law. (N.E.A.) [fr

  7. Investigations into encapsulation of intermediate level wastes containing organic components

    International Nuclear Information System (INIS)

    Palmer, J.

    1988-01-01

    A product evaluation programme was set up to investigate the properties of a variety of matrix-waste formulations prior to their encapsulation. The waste/matrix forms were defined and characterised and waste pretreatments studied. Potential encapsulation matrices were investigated for their suitability for individual waste streams. The physical, chemical and thermal properties, radiation stability and leaching behaviour of the formulations were studied. Operational and design limits for the encapsulation plant were defined. (U.K.)

  8. Hanford Central Waste Complex: Waste Receiving and Processing Facility dangerous waste permit application

    International Nuclear Information System (INIS)

    1991-10-01

    The Hanford Central Waste Complex is an existing and planned series of treatment, and/or disposal (TSD) unites that will centralize the management of solid waste operations at a single location on the Hanford Facility. The Complex includes two units: the WRAP Facility and the Radioactive Mixed Wastes Storage Facility (RMW Storage Facility). This Part B permit application addresses the WRAP Facility. The Facility will be a treatment and storage unit that will provide the capability to examine, sample, characterize, treat, repackage, store, and certify radioactive and/or mixed waste. Waste treated and stored will include both radioactive and/or mixed waste received from onsite and offsite sources. Certification will be designed to ensure and demonstrate compliance with waste acceptance criteria set forth by onsite disposal units and/or offsite facilities that subsequently are to receive waste from the WRAP Facility. This permit application discusses the following: facility description and general provisions; waste characterization; process information; groundwater monitoring; procedures to prevent hazards; contingency plant; personnel training; exposure information report; waste minimization plan; closure and postclosure requirements; reporting and recordkeeping; other relevant laws; certification

  9. Value Engineering Study for Closing Waste Packages Containing TAD Canisters

    International Nuclear Information System (INIS)

    Colleen Shelton-Davis

    2005-01-01

    The Office of Civilian Radioactive Waste Management announced their intention to have the commercial utilities package spent nuclear fuel in shielded, transportable, ageable, and disposable containers prior to shipment to the Yucca Mountain repository. This will change the conditions used as a basis for the design of the waste package closure system. The environment is now expected to be a low radiation, low contamination area. A value engineering study was completed to evaluate possible modifications to the existing closure system using the revised requirements. Four alternatives were identified and evaluated against a set of weighted criteria. The alternatives are (1) a radiation-hardened, remote automated system (the current baseline design); (2) a nonradiation-hardened, remote automated system (with personnel intervention if necessary); (3) a nonradiation-hardened, semi-automated system with personnel access for routine manual operations; and (4) a nonradiation-hardened, fully manual system with full-time personnel access. Based on the study, the recommended design is Alternative 2, a nonradiation-hardened, remote automated system. It is less expensive and less complex than the current baseline system, because nonradiation-hardened equipment can be used and some contamination control equipment is no longer needed. In addition, the inclusion of remote automation ensures throughput requirements are met, provides a more reliable process, and provides greater protection for employees from industrial accidents and radiation exposure than the semi-automated or manual systems. Other items addressed during the value engineering study as requested by OCRWM include a comparison to industry canister closure systems and corresponding lessons learned; consideration of closing a transportable, ageable, and disposable canister; and an estimate of the time required to perform a demonstration of the recommended closure system

  10. Value Engineering Study for Closing Waste Packages Containing TAD Canisters

    Energy Technology Data Exchange (ETDEWEB)

    Colleen Shelton-Davis

    2005-11-01

    The Office of Civilian Radioactive Waste Management announced their intention to have the commercial utilities package spent nuclear fuel in shielded, transportable, ageable, and disposable containers prior to shipment to the Yucca Mountain repository. This will change the conditions used as a basis for the design of the waste package closure system. The environment is now expected to be a low radiation, low contamination area. A value engineering study was completed to evaluate possible modifications to the existing closure system using the revised requirements. Four alternatives were identified and evaluated against a set of weighted criteria. The alternatives are (1) a radiation-hardened, remote automated system (the current baseline design); (2) a nonradiation-hardened, remote automated system (with personnel intervention if necessary); (3) a nonradiation-hardened, semi-automated system with personnel access for routine manual operations; and (4) a nonradiation-hardened, fully manual system with full-time personnel access. Based on the study, the recommended design is Alternative 2, a nonradiation-hardened, remote automated system. It is less expensive and less complex than the current baseline system, because nonradiation-hardened equipment can be used and some contamination control equipment is no longer needed. In addition, the inclusion of remote automation ensures throughput requirements are met, provides a more reliable process, and provides greater protection for employees from industrial accidents and radiation exposure than the semi-automated or manual systems. Other items addressed during the value engineering study as requested by OCRWM include a comparison to industry canister closure systems and corresponding lessons learned; consideration of closing a transportable, ageable, and disposable canister; and an estimate of the time required to perform a demonstration of the recommended closure system.

  11. Development of Models to Predict the Redox State of Nuclear Waste Containment Glass

    Energy Technology Data Exchange (ETDEWEB)

    Pinet, O.; Guirat, R.; Advocat, T. [Commissariat a l' Energie Atomique (CEA), Departement de Traitement et de Conditionnement des Dechets, Marcoule, BP 71171, 30207 Bagnols-sur-Ceze Cedex (France); Phalippou, J. [Universite de Montpellier II, Laboratoire des Colloides, Verres et Nanomateriaux, 34095 Montpellier Cedex 5 (France)

    2008-07-01

    Vitrification is one of the recommended immobilization routes for nuclear waste, and is currently implemented at industrial scale in several countries, notably for high-level waste. To optimize nuclear waste vitrification, research is conducted to specify suitable glass formulations and develop more effective processes. This research is based not only on experiments at laboratory or technological scale, but also on computer models. Vitrified nuclear waste often contains several multi-valent species whose oxidation state can impact the properties of the melt and of the final glass; these include iron, cerium, ruthenium, manganese, chromium and nickel. Cea is therefore also developing models to predict the final glass redox state. Given the raw materials and production conditions, the model predicts the oxygen fugacity at equilibrium in the melt. It can also estimate the ratios between the oxidation states of the multi-valent species contained in the molten glass. The oxidizing or reductive nature of the atmosphere above the glass melt is also taken into account. Unlike the models used in the conventional glass industry based on empirical methods with a limited range of application, the models proposed are based on the thermodynamic properties of the redox species contained in the waste vitrification feed stream. The thermodynamic data on which the model is based concern the relationship between the glass redox state and the oxygen fugacity in the molten glass. The model predictions were compared with oxygen fugacity measurements for some fifty glasses. The experiments carried out at laboratory and industrial scale with a cold crucible melter. The oxygen fugacity of the glass samples was measured by electrochemical methods and compared with the predicted value. The differences between the predicted and measured oxygen fugacity values were generally less than 0.5 Log unit. (authors)

  12. PUREX Storage Tunnels dangerous waste permit application

    International Nuclear Information System (INIS)

    1991-12-01

    This report is part of a dangerous waste permit application for the storage of wastes from the Purex process at Hanford. Appendices are presented on the following: construction drawings; HSW-5638, specifications for disposal facility for failed equipment, Project CA-1513-A; HWS-8262, specification for Purex equipment disposal, Project CGC 964; storage tunnel checklist; classification of residual tank heels in Purex storage tunnels; emergency plan for Purex facility; training course descriptions; and the Purex storage tunnels engineering study

  13. Demonstration of close-coupled barriers for subsurface containment of buried waste

    International Nuclear Information System (INIS)

    Heiser, J.; Dwyer, B.

    1995-01-01

    The primary objective of this project is to develop and demonstrate a close-coupled barrier for the containment of subsurface waste or contaminant migration. A close-coupled barrier is produced by first installing a conventional cement grout curtain followed by a thin lining of a polymer grout. The resultant barrier is a cement polymer composite that has economic benefits derived from the cement and performance benefits from the durable and resistant polymer layer. Close-coupled barrier technology is applicable for final, interim, or emergency containment of subsurface waste forms. Consequently, when considering the diversity of technology application, the construction emplacement and material technology maturity, general site operational requirements, and regulatory compliance incentives, the close-coupled barrier system provides an alternative for any hazardous or mixed waste remediation plan. This paper will discuss the installation of a close-coupled barrier and the subsequent integrity verification. The demonstration will take place at a cold site at the Hanford Geotechnical Test Facility, 400 Area, Hanford, Washington

  14. Test plan for immobilization of salt-containing surrogate mixed wastes using polyester resins

    International Nuclear Information System (INIS)

    Biyani, R.K.; Douglas, J.C.; Hendrickson, D.W.

    1997-01-01

    Past operations at many Department of Energy (DOE) sites have resulted in the generation of several waste streams with high salt content. These wastes contain listed and characteristic hazardous constituents and are radioactive. The salts contained in the wastes are primarily chloride, sulfate, nitrate, metal oxides, and hydroxides. DOE has placed these types of wastes under the purview of the Mixed Waste Focus Area (MWFA). The MWFA has been tasked with developing and facilitating the implementation of technologies to treat these wastes in support of customer needs and requirements. The MWFA has developed a Technology Development Requirements Document (TDRD), which specifies performance requirements for technology owners and developers to use as a framework in developing effective waste treatment solutions. This project will demonstrate the use of polyester resins in encapsulating and solidifying DOE's mixed wastes containing salts, as an alternative to conventional and other emerging immobilization technologies

  15. Test plan for immobilization of salt-containing surrogate mixed wastes using polyester resins

    Energy Technology Data Exchange (ETDEWEB)

    Biyani, R.K.; Douglas, J.C.; Hendrickson, D.W.

    1997-07-07

    Past operations at many Department of Energy (DOE) sites have resulted in the generation of several waste streams with high salt content. These wastes contain listed and characteristic hazardous constituents and are radioactive. The salts contained in the wastes are primarily chloride, sulfate, nitrate, metal oxides, and hydroxides. DOE has placed these types of wastes under the purview of the Mixed Waste Focus Area (MWFA). The MWFA has been tasked with developing and facilitating the implementation of technologies to treat these wastes in support of customer needs and requirements. The MWFA has developed a Technology Development Requirements Document (TDRD), which specifies performance requirements for technology owners and developers to use as a framework in developing effective waste treatment solutions. This project will demonstrate the use of polyester resins in encapsulating and solidifying DOE`s mixed wastes containing salts, as an alternative to conventional and other emerging immobilization technologies.

  16. Utilization of crushed radioactive concrete for mortar to fill waste container void space

    International Nuclear Information System (INIS)

    Ishikura, Takeshi; Ohnishi, Kazuhiko; Oguri, Daiichiro; Ueki, Hiroyuki

    2004-01-01

    Minimizing the volume of radioactive waste generated during dismantling of nuclear power plants is a matter of great importance. In Japan waste forms buried in a shallow burial disposal facility as low level radioactive waste must be solidified by cement or other materials with adequate strength and must provide no harmful opening. The authors have developed an improved method to minimize radioactive waste volume by utilizing radioactive concrete for fine aggregate for mortars to fill void space in waste containers. Tests were performed with pre-placed concrete waste and with filling mortar using recycled fine aggregate produced from concrete. It was estimated that the improved method substantially increases the waste fill ratio in waste containers, thereby decreasing the total volume of disposal waste. (author)

  17. Study of deuterons induced nuclear reactions on light elements (N, Al and Si): Application to containment materials of radioactive wastes; Etude des reactions nucleaires induites par des deuterons sur des elements legers (N, Al, Si): application aux materiaux de confinement des dechets radioactifs

    Energy Technology Data Exchange (ETDEWEB)

    Pellegrino, St

    2004-03-01

    Nuclear reaction analysis is well adapted to the quantification of light element. Profiles of concentration in order to follow elements migration into materials can be undertaken. This technique is used to study the behavior of the future matrices for nuclear waste containment. This technique is isotopic, characterized by a good signal-to-background ratio and a very low detection limit. The probability of a nuclear reaction is linked to a parameter called 'cross section' we have to know in order to carry out quantitative analysis. We have determined excitation curves for nitrogen, aluminium and silicon. These experiments were done with deuterons from 0.5 to 2 MeV. Two methods for the cross section characterization are presented and are in agreement with each other. The second one reduces uncertainty. Data are incorporated in the simulation software SIMNRA. We have compared the results obtained on different samples when we use data in literature or data of the study. We have noticed a great fit improvement with the data of this study. The new cross sections of this work will be integrated in the general data base SIGMABASE. Applications on materials such as Si{sub 3}N{sub 4}, nano-metric powders, WCN and nuclear glass YLaMgSiAlON studied for radioactive waste containment are also presented. (author)

  18. TECHNICAL EVALUATION OF THE SAFE TRANSPORTATION OF WASTE CONTAINERS COATED WITH POLYUREA

    Energy Technology Data Exchange (ETDEWEB)

    VAIL, T.S.

    2007-03-30

    This technical report is to evaluate and establish that the transportation of waste containers (e.g. drums, wooden boxes, fiberglass-reinforced plywood (FRP) or metal boxes, tanks, casks, or other containers) that have an external application of polyurea coating between facilities on the Hanford Site can be achieved with a level of onsite safety equivalent to that achieved offsite. Utilizing the parameters, requirements, limitations, and controls described in the DOE/RL-2001-36, ''Hanford Sitewide Transportation Safety Document'' (TSD) and the Department of Energy Richland Operations (DOE-RL) approved package specific authorizations (e.g. Package Specific Safety Documents (PSSDs), One-Time Requests for Shipment (OTRSs), and Special Packaging Authorizations (SPAS)), this evaluation concludes that polyurea coatings on packages does not impose an undue hazard for normal and accident conditions. The transportation of all packages on the Hanford Site must comply with the transportation safety basis documents for that packaging system. Compliance with the requirements, limitations, or controls described in the safety basis for a package system will not be relaxed or modified because of the application of polyurea. The inspection criteria described in facility/projects procedures and work packages that ensure compliance with Container Management Programs and transportation safety basis documentation dictate the need to overpack a package without consideration for polyurea. This technical report reviews the transportation of waste packages coated with polyurea and does not credit the polyurea with enhancing the structural, thermal, containment, shielding, criticality, or gas generating posture of a package. Facilities/Projects Container Management Programs must determine if a container requires an overpack prior to the polyurea application recognizing that circumstances newly discovered surface contamination or loss of integrity may require a previously

  19. TECHNICAL EVALUATION OF THE SAFE TRANSPORTATION OF WASTE CONTAINERS COATED WITH POLYUREA

    International Nuclear Information System (INIS)

    VAIL, T.S.

    2007-01-01

    This technical report is to evaluate and establish that the transportation of waste containers (e.g. drums, wooden boxes, fiberglass-reinforced plywood (FRP) or metal boxes, tanks, casks, or other containers) that have an external application of polyurea coating between facilities on the Hanford Site can be achieved with a level of onsite safety equivalent to that achieved offsite. Utilizing the parameters, requirements, limitations, and controls described in the DOE/RL-2001-36, ''Hanford Sitewide Transportation Safety Document'' (TSD) and the Department of Energy Richland Operations (DOE-RL) approved package specific authorizations (e.g. Package Specific Safety Documents (PSSDs), One-Time Requests for Shipment (OTRSs), and Special Packaging Authorizations (SPAS)), this evaluation concludes that polyurea coatings on packages does not impose an undue hazard for normal and accident conditions. The transportation of all packages on the Hanford Site must comply with the transportation safety basis documents for that packaging system. Compliance with the requirements, limitations, or controls described in the safety basis for a package system will not be relaxed or modified because of the application of polyurea. The inspection criteria described in facility/projects procedures and work packages that ensure compliance with Container Management Programs and transportation safety basis documentation dictate the need to overpack a package without consideration for polyurea. This technical report reviews the transportation of waste packages coated with polyurea and does not credit the polyurea with enhancing the structural, thermal, containment, shielding, criticality, or gas generating posture of a package. Facilities/Projects Container Management Programs must determine if a container requires an overpack prior to the polyurea application recognizing that circumstances newly discovered surface contamination or loss of integrity may require a previously un

  20. Modification of clay-based waste containment materials

    International Nuclear Information System (INIS)

    Adu-Wusu, K.; Whang, J.M.; McDevitt, M.F.

    1997-01-01

    Bentonite clays are used extensively for waste containment barriers to help impede the flow of water in the subsurface because of their low permeability characteristics. However, they do little to prevent diffusion of contaminants, which is the major transport mechanism at low water flows. A more effective way of minimizing contaminant migration in the subsurface is to modify the bentonite clay with highly sorptive materials. Batch sorption studies were conducted to evaluate the sorptive capabilities of organo-clays and humic- and iron-based materials. These materials proved to be effective sorbents for the organic contaminants 1,2,4-trichlorobenzene, nitrobenzene, and aniline in water, humic acid, and methanol solution media. The sorption capacities were several orders of magnitude greater than that of unmodified bentonite clay. Modeling results indicate that with small amounts of these materials used as additives in clay barriers, contaminant flux through walls could be kept very small for 100 years or more. The cost of such levels of additives can be small compared to overall construction costs

  1. Dry-distillation of experimental animal wastes containing radioisotopes

    International Nuclear Information System (INIS)

    Miyatake, Hideo; Saito, Kazumi; Kurihara, Norio

    1988-01-01

    Mice were dry-distilled at 800deg C for 10 min, after [ 32 P]orthophosphate or L-[4,5- 3 H]leucine was intraperitoneally administered. Phosphorus-32 was quantitatively recovered in the residual solid, whereas 95% of 3 H was found in the distillate (condensed water). When 14 C (L-[U- 14 C]malic acid or L-[U- 14 C]leucine) was administered to mice and they were dry-distilled, 14 C-radioactivity was distributed into two fractions; residual solid and exhaust gas. In these cases, the recovery percentage of 14 C in residual solid was not very high but increased as the treating temperature decreased. It reached about 40% of the administered 14 C at 400deg C for 120 min. By the dry-distillation of animals, their weight was reduced to about 10% in every animal tested (mice, rats and rabbits). The volume was reduced to about 20% in cases of mice and rats, and about 30% in case of rabbits. It was concluded that the dry-distillation can be useful as a pre-treatment for disposal of animal wastes containing radioisotopes. (author)

  2. Exposure rates from concrete covered cylindrical units containing radioactive waste

    International Nuclear Information System (INIS)

    Hedemann Jensen, P.

    1983-03-01

    Exposure rates from cylindrical waste units containing the nuclides 60 Co, 134 Cs and 137 Cs homogeneously mixed in a solidification product have been calculated. Analyses have been made for single drums and for two disposal geometries, one with the units placed below ground near the surface in a circular geometry, and one with the units placed on the ground in a pile behind a concrete wall. Due to self-shielding of the units, the exposure rate from the two geometries will be a factor of only 10 - 20 higher than from a single unit, even without soil or wall shielding. With one meter of soil above the circular pile below ground, a reduction factor of 5.10 3 to 5.10 4 can be achieved, depending on the nuclide considered. Placing a one-meter concrete wall in front of the drum pile on the ground gives rise to a reduction factor in the range of 5.10 5 to 2.10 7 . (author)

  3. Radioactive waste from non-power applications in Sweden

    International Nuclear Information System (INIS)

    Haegg, Ann-Christin; Lindbom, Gunilla; Persson, Monica

    2001-01-01

    Full text: The system for handling of radioactive waste from the Nuclear Fuel Cycle in Sweden is well established and has been in use for many years. Radioactive waste from other sources is not always handled as rigorously. The Swedish Radiation Protection Institute, SSI has identified the issue and therefore initiated a study with the aim to achieve a sufficient system for handling and disposal of radioactive waste from all sources of radioactive waste. In this paper we discuss some of the sources of radioactive waste and the specific problems they represent. We give a brief description on how they are regulated and handled today and identify some interesting issues. Conventional industry, hospitals, research and education: In the conventional industry the use of different types of radioactive sources is common. The size and type of radioactive source depends on the application (from some megaBq up to thousands of terraBq). The radioactive waste from hospitals, research institutions and pharmaceutical or bio-technical industries consists mainly of very short-lived radionuclides. Also most sealed sources used in the medical field contains short-lived radionuclides. According to the Swedish Radiation Protection Act a licence is needed for the use of sealed sources exceeding 50 kiloBq. For hospitals and research institutes the SSI issues one license covering all radioactive sources below 500 megaBq up to a summary limit depending on the application. All sources with activity exceeding 500 megaBq require a separate license. SSI has issued about 2500 licences. For each licence an annual fee is paid to the SSI. When the radioactive source has fulfilled its purpose the licensee is obliged to inform the SSI that the source is no longer in use and show a certificate from the recognised waste facility. Not until this has been done the licensee is released from its responsibilities. SSI has issued regulations on Radioactive Waste Not Associated with Nuclear Energy. These

  4. Container for waste, identification code reading device thereof, method and system for controlling waste by using them

    International Nuclear Information System (INIS)

    Kikuchi, Takashi; Yoshida, Tomiji; Omote, Tatsuyuki.

    1991-01-01

    In the conventional method of controlling waste containers by labels attached thereto, the data relevant to wastes contained in the waste containers are limited. Further, if the label should be peeled off, there is a possibility that the wastes therein can no more be identified. Then, in the present invention, an identification plate is previously attached, to which mechanically readable codes or visually readable letters or numerical figures are written. Then, the identification codes can be read in a remote control manner at high speed and high reliability and the waste containers can be managed only by the identification codes of the containers. Further, the identification codes on the container are made so as to be free from aging degradation, thereby enabling to manage waste containers for long time storage. With such a constitution, since data can be inputted from an input terminal and a great amount of data such as concerning the source of wastes can be managed collectively on a software, the data can be managed easily. (T.M.)

  5. Study on hazardous substances contained in radioactive waste

    International Nuclear Information System (INIS)

    Kuroki, Ryoichiro; Takahashi, Kuniaki

    2008-01-01

    It is necessary that the technical criteria is established concerning waste package for disposal of the TRU waste generated in Japan Atomic Energy Agency. And it is important to consider the criteria not only in terms of radioactivity but also in terms of chemical hazard and criticality. Therefore the environmental impact of hazardous materials and possibility of criticality were investigated to decide on technical specification of radioactive waste packages. The contents and results are as following. (1) Concerning hazardous materials included in TRU waste, regulations on disposal of industrial wastes and on environmental preservation were investigated. (2) The assessment methods for environmental impact of hazardous materials included in radioactive waste in U.K, U.S.A. and France were investigated. (3) The parameters for mass transport assessment about migration of hazardous materials in waste packages around disposal facilities were compiled. And the upper limits of amounts of hazardous materials in waste packages to satisfy the environmental standard were calculated with mass transport assessment for some disposal concepts. (4) It was suggested from criticality analysis for waste packages in disposal facility that the occurrence of criticality was almost impossible under the realistic conditions. (author)

  6. Lime treatment of liquid waste containing heavy metals, radionuclides and organics

    International Nuclear Information System (INIS)

    DuPont, A.

    1990-01-01

    This paper reports on lime treatment of liquid waste containing heavy metals, radio nuclides and organics. Lime is wellknown for its use in softening drinking water the treatment of municipal wastewaters. It is becoming important in the treatment of industrial wastewater and liquid inorganic hazardous waste; however, there are many questions regarding the use of lime for the treatment of liquid hazardous waste

  7. Hanford Facility Dangerous Waste Permit Application, 222-S Laboratory Complex

    International Nuclear Information System (INIS)

    WILLIAMS, J.F.

    2000-01-01

    The Hanford Facility Dangerous Waste Permit Application is considered to be a single application organized into a General Information Portion (document number DOE/RL-91-28) and a Unit-Specific Portion. The scope of the Unit-Specific Portion is limited to Part B permit application documentation submitted for individual, operating treatment, storage, and/or disposal units, such as the 222-S Laboratory Complex (this document, DOE/RL-91-27). Both the General Information and Unit-Specific portions of the Hanford Facility Dangerous Waste Permit Application address the content of the Part B permit application guidance prepared by the Washington State Department of Ecology (Ecology 1987 and 1996) and the U.S. Environmental Protection Agency (40 Code of Federal Regulations 270), with additional information needs defined by the Hazardous and Solid Waste Amendments and revisions of Washington Administrative Code 173-303. For ease of reference, the Washington State Department of Ecology alpha-numeric section identifiers from the permit application guidance documentation (Ecology 1996) follow, in brackets, the chapter headings and subheadings. Documentation contained in the General Information Portion is broader in nature and could be used by multiple treatment, storage, and/or disposal units (e.g., the glossary provided in the General Information Portion). Wherever appropriate, the 222-S Laboratory Complex permit application documentation makes cross-reference to the General Information Portion, rather than duplicating text. Information provided in this 222-S Laboratory Complex permit application documentation is current as of August 2000

  8. Characterizing the biotransformation of sulfur-containing wastes in simulated landfill reactors.

    Science.gov (United States)

    Sun, Wenjie; Sun, Mei; Barlaz, Morton A

    2016-07-01

    Landfills that accept municipal solid waste (MSW) in the U.S. may also accept a number of sulfur-containing wastes including residues from coal or MSW combustion, and construction and demolition (C&D) waste. Under anaerobic conditions that dominate landfills, microbially mediated processes can convert sulfate to hydrogen sulfide (H2S). The presence of H2S in landfill gas is problematic for several reasons including its low odor threshold, human toxicity, and corrosive nature. The objective of this study was to develop and demonstrate a laboratory-scale reactor method to measure the H2S production potential of a range of sulfur-containing wastes. The H2S production potential was measured in 8-L reactors that were filled with a mixture of the target waste, newsprint as a source of organic carbon required for microbial sulfate reduction, and leachate from decomposed residential MSW as an inoculum. Reactors were operated with and without N2 sparging through the reactors, which was designed to reduce H2S accumulation and toxicity. Both H2S and CH4 yields were consistently higher in reactors that were sparged with N2 although the magnitude of the effect varied. The laboratory-measured first order decay rate constants for H2S and CH4 production were used to estimate constants that were applicable in landfills. The estimated constants ranged from 0.11yr(-1) for C&D fines to 0.38yr(-1) for a mixed fly ash and bottom ash from MSW combustion. Copyright © 2016 Elsevier Ltd. All rights reserved.

  9. A multiobjective modeling approach to locate multi-compartment containers for urban-sorted waste

    International Nuclear Information System (INIS)

    Tralhao, Lino; Coutinho-Rodrigues, Joao; Alcada-Almeida, Luis

    2010-01-01

    The location of multi-compartment sorted waste containers for recycling purposes in cities is an important problem in the context of urban waste management. The costs associated with those facilities and the impacts placed on populations are important concerns. This paper introduces a mixed-integer, multiobjective programming approach to identify the locations and capacities of such facilities. The approach incorporates an optimization model in a Geographical Information System (GIS)-based interactive decision support system that includes four objectives. The first objective minimizes the total investment cost; the second one minimizes the average distance from dwellings to the respective multi-compartment container; the last two objectives address the 'pull' and 'push' characteristics of the decision problem, one by minimizing the number of individuals too close to any container, and the other by minimizing the number of dwellings too far from the respective multi-compartment container. The model determines the number of facilities to be opened, the respective container capacities, their locations, their respective shares of the total waste of each type to be collected, and the dwellings assigned to each facility. The approach proposed was tested with a case study for the historical center of Coimbra city, Portugal, where a large urban renovation project, addressing about 800 buildings, is being undertaken. This paper demonstrates that the models and techniques incorporated in the interactive decision support system (IDSS) can be used to assist a decision maker (DM) in analyzing this complex problem in a realistically sized urban application. Ten solutions consisting of different combinations of underground containers for the disposal of four types of sorted waste in 12 candidate sites, were generated. These solutions and tradeoffs among the objectives are presented to the DM via tables, graphs, color-coded maps and other graphics. The DM can then use this

  10. Open site tests on corrosion of carbon steel containers for radioactive waste forms

    International Nuclear Information System (INIS)

    Barinov, A.S.; Ojovan, M.I.; Ojovan, N.V.; Startceva, I.V.; Chujkova, G.N.

    1999-01-01

    Testing of waste containers under open field conditions is a component part of the research program that is being carried out at SIA Radon for more than 20 years to understand the long-term behavior of radioactive waste forms and waste packages. This paper presents the preliminary results of these ongoing studies. The authors used a typical NPP operational waste, containing 137 Cs, 134 Cs, and 60 Co as the dominant radioactive constituents. Bituminized and vitrified waste samples with 30--50 wt.% waste loading were prepared. Combined effects of climatic factors on corrosion behavior of carbon steel containers were estimated using gravimetric and chemical analyses. The observations suggest that uniform corrosion of containers prevails under open field conditions. The upper limits for the lifetime of containers were derived from calculations based on the model of atmospheric steel corrosion. Estimated lifetime values range from 300 to 600 years for carbon steel containers with the wall thickness of 2 mm containing vitrified waste, and from 450 to 500 years for containers with the wall thickness of 2.5 mm that were used for bituminized waste. However, following the most conservative method, pitting corrosion may cause container integrity failure after 60 to 90 years of exposure

  11. WIPP's Hazardous Waste Facility Permit Renewal Application

    International Nuclear Information System (INIS)

    Most, W.A.; Kehrman, R.F.

    2009-01-01

    Hazardous waste permits issued by the New Mexico Environment Department (NMED) have a maximum term of 10-years from the permit's effective date. The permit condition in the Waste Isolation Pilot Plant's (WIPP) Hazardous Waste Facility Permit (HWFP) governing renewal applications, directs the Permittees to submit a permit application 180 days prior to expiration of the Permit. On October 27, 1999, the Secretary of the NMED issued to the United States Department of Energy (DOE), the owner and operator of WIPP, and to Washington TRU Solutions LLC (WTS), the Management and Operating Contractor and the cooperator of WIPP, a HWFP to manage, store, and dispose hazardous waste at WIPP. The DOE and WTS are collectively known as the Permittees. The HWFP is effective for a fixed term not to exceed ten years from the effective date of the Permit. The Permittees may renew the HWFP by submitting a new permit application at least 180 calendar days before the expiration date, of the HWFP. The Permittees are not proposing any substantial changes in the Renewal Application. First, the Permittees are seeking the authority to dispose of Contact-Handled and Remote-Handled TRU mixed waste in Panel 8. Panels 4 through 7 have been approved in the WIPP Hazardous Waste Facility Permit as it currently exists. No other change to the facility or to the manner in which hazardous waste is characterized, managed, stored, or disposed is being requested. Second, the Permittees also seek to include the Mine Ventilation Rate Monitoring Plan, as Attachment Q in the HWFP. This Plan has existed as a separate document since May 2000. The NMED has requested that the Plan be submitted as part of the Renewal Application. The Permittees have been operating to the Mine Ventilation Rate Monitoring Plan since the Plan was submitted. Third, some information submitted in the original WIPP RCRA Part B Application has been updated, such as demographic information. The Permittees will submit this information in the

  12. Demonstration of close-coupled barriers for subsurface containment of buried waste

    International Nuclear Information System (INIS)

    Dwyer, B.P.

    1996-05-01

    A close-coupled barrier is produced by first installing a conventional cement grout curtain followed by a thin inner lining of a polymer grout. The resultant barrier is a cement polymer composite that has economic benefits derived from the cement and performance benefits from the durable and resistant polymer layer. Close-coupled barrier technology is applicable for final, interim, or emergency containment of subsurface waste forms. Consequently, when considering the diversity of technology application, the construction emplacement and material technology maturity, general site operational requirements, and regulatory compliance incentives, the close-coupled barrier system provides an alternative for any hazardous or mixed waste remediation plan. This paper discusses the installation of a close-coupled barrier and the subsequent integrity verification. The demonstration was installed at a benign site at the Hanford Geotechnical Test Facility, 400 Area, Hanford, Washington. The composite barrier was emplaced beneath a 7,500 liter tank. The tank was chosen to simulate a typical DOE Complex waste form. The stresses induced on the waste form were evaluated during barrier construction. The barrier was constructed using conventional jet grouting techniques. Drilling was completed at a 45 degree angle to the ground, forming a conical shaped barrier with the waste form inside the cone. Two overlapping rows of cylindrical cement columns were grouted in a honeycomb fashion to form the secondary backdrop barrier layer. The primary barrier, a high molecular weight polymer manufactured by 3M Company, was then installed providing a relatively thin inner liner for the secondary barrier. The primary barrier was emplaced by panel jet grouting with a dual wall drill stem, two phase jet grouting system

  13. Treatment of liquid waste containing alpha nuclides by adsorption

    Energy Technology Data Exchange (ETDEWEB)

    Jishu, Zeng; Xiguang, Su; Dejing, Xia; Sianhua, Fan [China Inst. of Atomic Energy, Beijing (China). Radiochemistry Dept.

    1997-02-01

    In this paper, experimental investigations on the removal of actinides from a decontaminating waste stream by using adsorption technique following the cementation of a resultant absorbent sludge are described. One kind of apatites was selected as an actinide absorbent from a number of indigenous materials by batch equilibrium tests. The influence of contact time, temperature, particle size and pH variables on the adsorption of actinides is given. The removal of total alpha activity is higher tan 97% by absorbent precipitation process when the absorbent addition percentage of the liquid waste is more than 3.25 wt%, making alpha-activity level of the primary waste stream below 3.7 x 10{sup 3} Bq/L, which can meet the acceptance requirements of the Low Level Radwaste Treatment Plant. The studies on the cementation of the absorbent sludge included the selection of cements used for solidification, formulation and characterization of the selected cemented waste forms. The results obtained have shown that both 525 type Portland cement and 325 type Portland pozzolana cement were compatible with the absorbent sludge. The selected cemented waste forms meet the requirements of the Chinese National Standard (GB 14569.1-93): Characteristic Requirements for Solidified Waste of Low and Intermediate Level Radioactive Waste - Cement Solidified Waste. (author). 9 refs, 3 figs, 14 tabs.

  14. Method of separating tritium contained in gaseous wastes

    International Nuclear Information System (INIS)

    Hashimoto, Yasuo; Oozono, Hideaki.

    1981-01-01

    Purpose: To decrease tritium concentration in gaseous wastes to less than the allowable level by removing tritium in gaseous wastes generated upon combustion of radioactive wastes by using a plurality of heat exchangers. Method: Gaseous wastes at high temperature generated upon combustion of radioactive wastes in an incinerator are removed with radioactive solid substances through filters, cooled down to a temperature below 10 0 C by the passage through first and second heat exchangers and decreased with tritium content to less than the allowable concentration by the gaseous wastes at low temperature from the second heat exhcanger. The gaseous wastes at low temperature are used as the cooling medium for the first heat exchanger. They are heat exchanged at the upstream of the second heat exchanger with the cooling water from the third heat exchanger and cooled at the downstream by the cooling water cooled by the cooling medium. The gaseous wastes at low temperature thus cooled below 10 0 C are heated to about 350 0 C in the first heat exchanger and discharged. (Moriyama, K.)

  15. Treatment of liquid waste containing alpha nuclides by adsorption

    International Nuclear Information System (INIS)

    Zeng Jishu; Su Xiguang; Xia Dejing; Fan Sianhua

    1997-01-01

    In this paper, experimental investigations on the removal of actinides from a decontaminating waste stream by using adsorption technique following the cementation of a resultant absorbent sludge are described. One kind of apatites was selected as an actinide absorbent from a number of indigenous materials by batch equilibrium tests. The influence of contact time, temperature, particle size and pH variables on the adsorption of actinides is given. The removal of total alpha activity is higher tan 97% by absorbent precipitation process when the absorbent addition percentage of the liquid waste is more than 3.25 wt%, making alpha-activity level of the primary waste stream below 3.7 x 10 3 Bq/L, which can meet the acceptance requirements of the Low Level Radwaste Treatment Plant. The studies on the cementation of the absorbent sludge included the selection of cements used for solidification, formulation and characterization of the selected cemented waste forms. The results obtained have shown that both 525 type Portland cement and 325 type Portland pozzolana cement were compatible with the absorbent sludge. The selected cemented waste forms meet the requirements of the Chinese National Standard (GB 14569.1-93): Characteristic Requirements for Solidified Waste of Low and Intermediate Level Radioactive Waste - Cement Solidified Waste. (author). 9 refs, 3 figs, 14 tabs

  16. Remote automated material handling of radioactive waste containers

    International Nuclear Information System (INIS)

    Greager, T.M.

    1994-09-01

    To enhance personnel safety, improve productivity, and reduce costs, the design team incorporated a remote, automated stacker/retriever, automatic inspection, and automated guidance vehicle for material handling at the Enhanced Radioactive and Mixed Waste Storage Facility - Phase V (Phase V Storage Facility) on the Hanford Site in south-central Washington State. The Phase V Storage Facility, scheduled to begin operation in mid-1997, is the first low-cost facility of its kind to use this technology for handling drums. Since 1970, the Hanford Site's suspect transuranic (TRU) wastes and, more recently, mixed wastes (both low-level and TRU) have been accumulating in storage awaiting treatment and disposal. Currently, the Hanford Site is only capable of onsite disposal of radioactive low-level waste (LLW). Nonradioactive hazardous wastes must be shipped off site for treatment. The Waste Receiving and Processing (WRAP) facilities will provide the primary treatment capability for solid-waste storage at the Hanford Site. The Phase V Storage Facility, which accommodates 27,000 drum equivalents of contact-handled waste, will provide the following critical functions for the efficient operation of the WRAP facilities: (1) Shipping/Receiving; (2) Head Space Gas Sampling; (3) Inventory Control; (4) Storage; (5) Automated/Manual Material Handling

  17. Application of visible spectroscopy in waste sorting

    Science.gov (United States)

    Spiga, Philippe; Bourely, Antoine

    2011-10-01

    Today, waste recycling, (bottles, papers...), is a mechanical operation: the waste are crushed, fused and agglomerated in order to obtain new manufactured products (e.g. new bottles, clothes ...). The plastics recycling is the main application in the color sorting process. The colorless plastics recovered are more valuable than the colored plastics. Other emergent applications are in the paper sorting, where the main goal is to sort dyed paper from white papers. Up to now, Pellenc Selective Technologies has manufactured color sorting machines based on RGB cameras. Three dimensions (red, green and blue) are no longer sufficient to detect low quantities of dye in the considered waste. In order to increase the efficiency of the color detection, a new sorting machine, based on visible spectroscopy, has been developed. This paper presents the principles of the two approaches and their difference in terms of sorting performance, making visible spectroscopy a clear winner.

  18. Process for denitrating waste solutions containing nitrates and actinides with simultaneous separation of the actinides

    International Nuclear Information System (INIS)

    Gompper, K.

    1986-01-01

    The invention is intended to reduce the acid and nitrate content of nitrate waste solutions, to reduce the total salt content of the waste solution, to remove the actinides contained in it by precipitation, without any danger of violent reactions or an increase in the volume of the waste solution. The invention achieves this by mixing the waste solution with diethyl oxalate at room temperature and heating the mixture to at least 80 0 C. (orig./PW) [de

  19. Low-level burial grounds dangerous waste permit application design documents

    International Nuclear Information System (INIS)

    1990-08-01

    This document serves a supplement to the already existing ''Low-Level Burial Ground Dangerous Waste Permit Application Design Documents.'' This paper contains information regarding drawings, construction specifications, and liner/leachate compatibility test plans

  20. Stress corrosion cracking of candidate materials for nuclear waste containers

    International Nuclear Information System (INIS)

    Maiya, P.S.; Shack, W.J.; Kassner, T.F.

    1989-09-01

    Types 304L and 316L stainless steel (SS), Incoloy 825, Cu, Cu-30%Ni, and Cu-7%Al have been selected as candidate materials for the containment of high-level nuclear waste at the proposed Yucca Mountain Site in Nevada. The susceptibility of these materials to stress corrosion cracking has been investigated by slow-strain-rate tests (SSRTs) in water which simulates that from well J-13 (J-13 water) and is representative of the groundwater present at the Yucca Mountain site. The SSRTs were performed on specimens exposed to simulated J-13 water at 93 degree C and at a strain rate 10 -7 s -1 under crevice conditions and at a strain rate of 10 -8 s -1 under both crevice and noncrevice conditions. All the tests were interrupted after nominal elongation strains of 1--4%. Examination by scanning electron microscopy showed some crack initiation in virtually all specimens. Optical microscopy of metallographically prepared transverse sections of Type 304L SS suggests that the crack depths are small (<10 μm). Preliminary results suggest that a lower strain rate increases the severity of cracking of Types 304L and 316L SS, Incoloy 825, and Cu but has virtually no effect on Cu-30%Ni and Cu-7%Al. Differences in susceptibility to cracking were evaluated in terms of a stress ratio, which is defined as the ratio of the increase in stress after local yielding in the environment to the corresponding stress increase in an identical test in air, both computed at the same strain. On the basis of this stress ratio, the ranking of materials in order of increasing resistance to cracking is: Types 304L SS < 316L SS < Incoloy 825 congruent Cu-30%Ni < Cu congruent Cu-7%Al. 9 refs., 12 figs., 7 tabs

  1. Treatment of nanomaterial-containing waste in thermal waste treatment facilities; Behandlung nanomaterialhaltiger Abfaelle in thermischen Abfallbehandlungsanlagen

    Energy Technology Data Exchange (ETDEWEB)

    Vogel, Julia; Weiss, Volker [Umweltbundesamt, Dessau-Rosslau (Germany); Oischinger, Juergen; Meiller, Martin; Daschner, Robert [Fraunhofer Umsicht, Sulzbach-Rosenberg (Germany)

    2016-09-15

    There is already a multitude of products on the market, which contain synthetic nanomaterials (NM), and for the coming years an increase of such products can be expected. Consequently, it is predictable that more nanomaterial-containing waste will occur in the residual waste that is predominately disposed in thermal waste treatment plants. However, the knowledge about the behaviour and effects of nanomaterials from nanomaterial-containing waste in this disposal route is currently still low. A research project of the German Environment Agency on the ''Investigation of potential environmental impacts when disposing nanomaterial-containing waste in waste treatment plants'' will therefore dedicate itself to a detailed examination of emission pathways in the thermal waste treatment facilities. The tests carried out i.a. on an industrial waste incineration plant and a sludge incineration plant with controlled addition of titanium dioxide at the nanoscale, showed that no increase in the emissions of NM in the exhaust gas was detected. The majority of the NM was found in the combustion residues, particularly the slag.

  2. Design criteria burial containers for non-transuranic solid radioactive waste

    International Nuclear Information System (INIS)

    Hammond, J.E.

    1976-01-01

    The criteria, replace HW-83959 and apply to containers constructed specifically for the containment of beta-gamma radioactively contaminated waste removed from an area controlled by radiation work procedures, transported across an uncontrolled area where there is risk of a radiation release to the environs, and buried in an approved radioactive waste burial ground

  3. Precipitation and Deposition of Aluminum-Containing Phases in Tank Wastes

    International Nuclear Information System (INIS)

    Dabbs, Daniel M.; Aksay, Ilhan A.

    2005-01-01

    Aluminum-containing phases compose the bulk of solids precipitating during the processing of radioactive tank wastes. Processes designed to minimize the volume of high-level waste through conversion to glassy phases require transporting waste solutions near-saturated with aluminum-containing species from holding tank to processing center. The uncontrolled precipitation within transfer lines results in clogged pipes and lines and fouled ion exchangers, with the potential to shut down processing operations

  4. A container for storage and disposal of low-level waste

    International Nuclear Information System (INIS)

    Fish, R.L.; Butler, B.D.

    1989-01-01

    A unique concept for corrosion-resistant containers for storing and disposing of low-level radioactive, mixed and toxic wastes has been developed. The strength and low cost of carbon steel has been combined with the corrosion and abrasion resistance of a proprietary combination of polymers to provide an inexpensive alternative to currently available waste containers. The initial development effort has focused on a 55-gallon container, the B and W ECOSAFE-55 tm . However, Babcock and Wilcox (B and W) can develop a family of ECOSAFE waste containers using this technology to accommodate user-preferred configurations and volumes. The containers will be capable of accepting a wide range of low-level radioactive (LLRW) and industrial waste forms. Basic engineering design analyses and functional tests were performed to show compliance of the container with transportation functional requirements. These tests and analyses, along with chemical resistance tests, qualify the container for use in storing a wide range of radioactive and chemical wastes. For the container to be licensed for use as a high-integrity container in shallow land, low-level radioactive waste burial facilities, the Nuclear Regulatory Commission requires certain tests and analyses to demonstrate that container gross physical properties and identity can be maintained for 300 years. This paper describes the container concept in generic terms and provides information on the initial, ECOSAFE-55 container design, testing and engineering analysis efforts

  5. Properties of dune sand concrete containing coffee waste

    Directory of Open Access Journals (Sweden)

    Mohamed Guendouz

    2018-01-01

    Full Text Available In the last years, an increase of coffee beverages consumption has been observed all over the world; and its consumption increases the waste coffee grounds which will become an environmental problems. Recycling of this waste to produce new materials like sand concrete appears as one of the best solutions for reduces the problem of pollution. This work aims to study the possibility of recycling waste coffee grounds (Spent Coffee Grounds (SCG as a fine aggregate by replacing the sand in the manufacturing of dune sand concrete. For this; sand concrete mixes were prepared with substitution of sand with the spent coffee grounds waste at different percentage (0%, 5%, 10%, 15% and 20% by volume of the sand in order to study the influence of this wastes on physical (Workability, bulk density and porosity, mechanical (compressive and flexural strength and Thermal (Thermal conductivity and thermal diffusivity properties of dune sand concrete. The results showed that the use of spent coffee grounds waste as partial replacement of natural sand contributes to reduce workability, bulk density and mechanical strength of sand concrete mixes with an increase on its porosity. However, the thermal characteristics are improved and especially for a level of 15% and 20% of substitution. So, it is possible to obtain an insulating material which can be used in the various types of structural components. This study ensures that reusing of waste coffee grounds in dune sand concrete gives a positive approach to reduce the cost of materials and solve some environmental problems.

  6. Grout treatment facility dangerous waste permit application

    International Nuclear Information System (INIS)

    1992-07-01

    The Grout Treatment Facility (GTF) will provide permanent disposal for approximately 43 Mgal of low-level radioactive liquid waste currently being stored in underground tanks on the Hanford Site. The first step in permanent disposal is accomplished by solidifying the liquid waste with cementitious dry materials. The resulting grout is cast within underground vaults. This report on the GTF contains information on the following: Geologic data, hydrologic data, groundwater monitoring program, information, detection monitoring program, groundwater characterization drawings, building emergency plan--grout treatment facility, response action plan for grout treatment facility, Hanford Facility contingency plan, training course descriptions, overview of the Hanford Facility Grout Performance, assessment, bland use and zoning map, waste minimization plan, cover design engineering report, and clay liners (ADMIXTURES) in semiarid environments

  7. Characteristics of metal waste forms containing technetium and uranium

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  8. 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

  9. DESIGN ANALYSIS FOR THE DEFENSE HIGH-LEVEL WASTE DISPOSAL CONTAINER

    Energy Technology Data Exchange (ETDEWEB)

    G. Radulesscu; J.S. Tang

    2000-06-07

    The purpose of ''Design Analysis for the Defense High-Level Waste Disposal Container'' analysis is to technically define the defense high-level waste (DHLW) disposal container/waste package using the Waste Package Department's (WPD) design methods, as documented in ''Waste Package Design Methodology Report'' (CRWMS M&O [Civilian Radioactive Waste Management System Management and Operating Contractor] 2000a). The DHLW disposal container is intended for disposal of commercial high-level waste (HLW) and DHLW (including immobilized plutonium waste forms), placed within disposable canisters. The U.S. Department of Energy (DOE)-managed spent nuclear fuel (SNF) in disposable canisters may also be placed in a DHLW disposal container along with HLW forms. The objective of this analysis is to demonstrate that the DHLW disposal container/waste package satisfies the project requirements, as embodied in Defense High Level Waste Disposal Container System Description Document (SDD) (CRWMS M&O 1999a), and additional criteria, as identified in Waste Package Design Sensitivity Report (CRWMS M&Q 2000b, Table 4). The analysis briefly describes the analytical methods appropriate for the design of the DHLW disposal contained waste package, and summarizes the results of the calculations that illustrate the analytical methods. However, the analysis is limited to the calculations selected for the DHLW disposal container in support of the Site Recommendation (SR) (CRWMS M&O 2000b, Section 7). The scope of this analysis is restricted to the design of the codisposal waste package of the Savannah River Site (SRS) DHLW glass canisters and the Training, Research, Isotopes General Atomics (TRIGA) SNF loaded in a short 18-in.-outer diameter (OD) DOE standardized SNF canister. This waste package is representative of the waste packages that consist of the DHLW disposal container, the DHLW/HLW glass canisters, and the DOE-managed SNF in disposable

  10. DESIGN ANALYSIS FOR THE DEFENSE HIGH-LEVEL WASTE DISPOSAL CONTAINER

    International Nuclear Information System (INIS)

    Radulesscu, G.; Tang, J.S.

    2000-01-01

    The purpose of ''Design Analysis for the Defense High-Level Waste Disposal Container'' analysis is to technically define the defense high-level waste (DHLW) disposal container/waste package using the Waste Package Department's (WPD) design methods, as documented in ''Waste Package Design Methodology Report'' (CRWMS M andO [Civilian Radioactive Waste Management System Management and Operating Contractor] 2000a). The DHLW disposal container is intended for disposal of commercial high-level waste (HLW) and DHLW (including immobilized plutonium waste forms), placed within disposable canisters. The U.S. Department of Energy (DOE)-managed spent nuclear fuel (SNF) in disposable canisters may also be placed in a DHLW disposal container along with HLW forms. The objective of this analysis is to demonstrate that the DHLW disposal container/waste package satisfies the project requirements, as embodied in Defense High Level Waste Disposal Container System Description Document (SDD) (CRWMS M andO 1999a), and additional criteria, as identified in Waste Package Design Sensitivity Report (CRWMS M andQ 2000b, Table 4). The analysis briefly describes the analytical methods appropriate for the design of the DHLW disposal contained waste package, and summarizes the results of the calculations that illustrate the analytical methods. However, the analysis is limited to the calculations selected for the DHLW disposal container in support of the Site Recommendation (SR) (CRWMS M andO 2000b, Section 7). The scope of this analysis is restricted to the design of the codisposal waste package of the Savannah River Site (SRS) DHLW glass canisters and the Training, Research, Isotopes General Atomics (TRIGA) SNF loaded in a short 18-in.-outer diameter (OD) DOE standardized SNF canister. This waste package is representative of the waste packages that consist of the DHLW disposal container, the DHLW/HLW glass canisters, and the DOE-managed SNF in disposable canisters. The intended use of this

  11. Hanford facility dangerous waste permit application, PUREX storage tunnels

    International Nuclear Information System (INIS)

    Price, S.M.

    1997-01-01

    The Hanford Facility Dangerous Waste Permit Application is considered to be a single application organized into a General Information Portion (document number DOE/RL-91-28) and a Unit-Specific Portion. The scope of the Unit-Specific Portion is limited to Part B permit application documentation submitted for individual, operating treatment, storage, and/or disposal units, such as the PUREX Storage Tunnels (this document, DOE/RL-90-24). Both the General Information and Unit-Specific portions of the Hanford Facility Dangerous Waste Permit Application address the content of the Part B permit application guidance prepared by the Washington State Department of Ecology (Ecology 1996) and the US Environmental Protection Agency (40 Code of Federal Regulations 270), with additional information needs defined by the Hazardous and Solid Waste Amendments and revisions of Washington Administrative Code 173-303. For ease of reference, the Washington State Department of Ecology alpha-numeric section identifiers from the permit application guidance documentation (Ecology 1996) follow, in brackets, the chapter headings and subheadings. A checklist indicating where information is contained in the PUREX Storage Tunnels permit application documentation, in relation to the Washington State Department of Ecology guidance, is located in the Contents Section. Documentation contained in the General Information Portion is broader in nature and could be used by multiple treatment, storage, and/or disposal units (e.g., the glossary provided in the General Information Portion). Wherever appropriate, the PUREX Storage Tunnels permit application documentation makes cross-reference to the General Information Portion, rather than duplicating text. Information provided in this PUREX Storage Tunnels permit application documentation is current as of April 1997

  12. Simultaneous treatment of SO2 containing stack gases and waste water

    Science.gov (United States)

    Poradek, J. C.; Collins, D. D. (Inventor)

    1978-01-01

    A process for simultaneously removing sulfur dioxide from stack gases and the like and purifying waste water such as derived from domestic sewage is described. A portion of the gas stream and a portion of the waste water, the latter containing dissolved iron and having an acidic pH, are contacted in a closed loop gas-liquid scrubbing zone to effect absorption of the sulfur dioxide into the waste water. A second portion of the gas stream and a second portion of the waste water are controlled in an open loop gas-liquid scrubbing zone. The second portion of the waste water contains a lesser amount of iron than the first portion of the waste water. Contacting in the openloop scrubbing zone is sufficient to acidify the waste water which is then treated to remove solids originally present.

  13. Modeling of container failure and radionuclide release from a geologic nuclear waste repository

    International Nuclear Information System (INIS)

    Kim, Chang Lak; Kim, Jhin Wung; Choi, Kwang Sub; Cho, Chan Hee

    1989-02-01

    Generally, two processes are involved in leaching and dissolution; (1) chemical reactions and (2) mass transfer by diffusion. The chemical reaction controls the dissolution rates only during the early stage of exposure to groundwater. The exterior-field mass transfer may control the long-term dissolution rates from the waste solid in a geologic repository. Masstransfer analyses rely on detailed and careful application of the governing equations that describe the mechanistic processes of transport of material between and within phases. We develop analytical models to predict the radionuclide release rate into the groundwater with five different approaches: a measurement-based model, a diffusion model, a kinetics model, a diffusion-and-kinetics model, and a modified diffusion model. We also collected experimental leaching data for a partial validation of the radionuclide release model based on the mass transfer theory. Among various types of corrosions, pitting is the most significant because of its rapid growth. The failure time of the waste container, which also can be interpreted as the containment time, is a milestone of the performance of a repository. We develop analytical models to predict the pit growth rate on the container surface with three different approaches: an experimental method, a statistical method, and a mathematical method based on the diffusion theory. (Author)

  14. Waste heat recovery for offshore applications

    DEFF Research Database (Denmark)

    Pierobon, Leonardo; Kandepu, Rambabu; Haglind, Fredrik

    2012-01-01

    vary in the range 20-30%. There are several technologies available for onshore gas turbines (and low/medium heat sources) to convert the waste heat into electricity. For offshore applications it is not economical and practical to have a steam bottoming cycle to increase the efficiency of electricity...... production, due to low gas turbine outlet temperature, space and weight restrictions and the need for make-up water. A more promising option for use offshore is organic Rankine cycles (ORC). Moreover, several oil and gas platforms are equipped with waste heat recovery units to recover a part of the thermal...... energy in the gas turbine off-gas using heat exchangers, and the recovered thermal energy acts as heat source for some of the heat loads on the platform. The amount of the recovered thermal energy depends on the heat loads and thus the full potential of waste heat recovery units may not be utilized...

  15. Elaboration of new ceramic composites containing glass fibre production wastes

    Directory of Open Access Journals (Sweden)

    Rozenstrauha, I.

    2013-04-01

    Full Text Available Two main by-products or waste from the production of glass fibre are following: sewage sludge containing montmorillonite clay as sorbent material and ca 50% of organic matter as well as waste glass from aluminiumborosilicate glass fibre with relatively high softening temperature (> 600 ºC. In order to elaborate different new ceramic products (porous or dense composites the mentioned by-products and illitic clay from two different layers of Apriki deposit (Latvia with illite content in clay fraction up to 80-90% was used as a matrix. The raw materials were investigated by differential-thermal (DTA and XRD analysis. Ternary compositions were prepared from mixtures of 15–35 wt % of sludge, 20 wt % of waste glass and 45–65 wt % of clay and the pressed green bodies were thermally treated in sintering temperature range from 1080 to 1120 ºC in different treatment conditions. Materials produced in temperature range 1090–1100 ºC with the most optimal properties - porosity 38-52%, water absorption 39–47% and bulk density 1.35–1.67 g/cm3 were selected for production of porous ceramics and materials showing porosity 0.35–1.1%, water absorption 0.7–2.6 % and bulk density 2.1–2.3 g/cm3 - for dense ceramic composites. Obtained results indicated that incorporation up to 25 wt % of sewage sludge is beneficial for production of both ceramic products and glass-ceramic composites according to the technological properties. Structural analysis of elaborated composite materials was performed by scanning electron microscopy(SEM. By X-ray diffraction analysis (XRD the quartz, diopside and anorthite crystalline phases were detected.Durante la obtención de ciertas fibras de vidrio se generan dos subproductos o residuos principalmente: Lodo de arcilla montmorillonítica capaz de adsorber el 50 % de materia orgánica y un vidrio silicato alumínico con temperatura de reblandecimiento relativamente alta (> 600 ºC. Con el fin de elaborar nuevos

  16. Heat transfer effects in vertically emplaced high level nuclear waste container

    International Nuclear Information System (INIS)

    Moujaes, S.F.; Lei, Y.M.

    1994-01-01

    Modeling free convection heat transfer in an cylindrical annular enclosure is still an active area of research and an important problem to be addressed in the high level nuclear waste repository. For the vertically emplaced waste container, the air gap which is between the container shell and the rock borehole, have an important role of dissipating heat to surrounding rack. These waste containers are vertically emplaced in the borehole 300 meters below ground, and in a horizontal grid of 30 x 8 meters apart. The borehole will be capped after the container emplacement. The expected initial heat generated is between 3--4.74 kW per container depending on the type of waste. The goal of this study is to use a computer simulation model to find the borehole wall, air-gap and the container outer wall temperature distributions

  17. Application for approval to construct the Waste Receiving And Processing facility

    International Nuclear Information System (INIS)

    1993-02-01

    The following Application For Approval Of Construction is being submitted by the US Department of Energy, Richland Field Office pursuant to 40 CFR 61.07, ''Application for Approval of Construction or Modification,'' for the Waste Receiving and Processing (WRAP) Module 1 facility (also referred to as WRAP 1). The WRAP 1 facility will be a new source of radioactive emissions to the atmosphere. The WRAP 1 facility will be housed in the new 2336-W Building, which will be located in the 200 West Area south of 23rd Street and west of Dayton Avenue. The 200 West Area is located within the boundary of the Hanford Site. The mission of the WRAP 1 facility is to examine, assay, characterize, treat, and repackage solid radioactive and mixed waste to enable permanent disposal of the waste in accordance with all applicable regulations. The solid wastes to be handled in the WRAP 1 facility include low-level waste (LLW), Transuranic (TRU) waste, TRU mixed waste, and low-level mixed waste (LLMW). The WRAP 1 facility will only accept contact handled (CH) waste containers. CH waste is a waste category whose external surface dose rate does not exceed 200 mrem/h. These containers have a surface dose rate of less than 200 mrem/h

  18. The Welding Effect on Mechanical Strength of Low Level Radioactive Waste Drum Container

    International Nuclear Information System (INIS)

    Aisyah; Herlan Martono

    2007-01-01

    The treatment of compactable low level solid waste was started by compaction of 100 liter drum containing the waste using 600 kN hydraulic press in 200 liters drum. The 200 liter drum of waste container containing of compacted waste then immobilized with cement and stored in interm storage. The 200 liter drum of waste container made of carbon steel material to comply with a good mechanical strength request in order to be able to retain the waste content for long period. Welding is a one step in a waste drum container fabrication process that has an opportunity in decreasing these mechanical strength. The research is carried out by welding the waste drum container material sample by electric arc welding. Mechanical strength test carried out by measuring the tensile strength by using the tensile strength machine, hardness test by using Vickers hardness test and microstructure observation by using the optic microscope. The result shows that the welding cause the microstructure changes, its meaning of forming ferro oxide phase on welding area that leads to the brittle material, so that the mechanical strength has a decreasing slightly. Nevertheless the decreasing of mechanical strength is still in the range of safety limit. (author)

  19. Synthesis of studies on primary containers for MLA-VL wastes

    International Nuclear Information System (INIS)

    Bart, F.; Delassale, F.; Rey, F.; Helie, M.; Levoy, R.; Moitrier, C.; Sicardy, O.; Tiquet, P.

    2004-01-01

    The aim of this study is the presentation of studies realized on primary containers of medium activity long life level. These studies are realized in the framework of the axis 3 of the law of 1991 on the radioactive waste management. The specificity of this document is the presentation of container for ''random'' wastes chemically corrosive in order to complete the range of possible packages. Thus a special program has been developed to demonstrate a conditioning solution which offers to the waste producers a possibility of conditioning these wastes without a preliminary treatment. (A.L.B.)

  20. Process for denitrating waste solutions containing nitric acid actinides simultaneously separating the actinides

    International Nuclear Information System (INIS)

    Gompper, K.

    1984-01-01

    The invention should reduce the acid and nitrate content of waste solutions containing nitric acid as much as possible, should reduce the total salt content of the waste solution, remove the actinides contained in it by precipitation and reduce the α radio-activity in the remaining solution, without having to worry about strong reactions or an increase in the volume of the waste solution. The invention achieves this by mixing the waste solution with diethyl oxalate at room temperature and heating the mixture to at least 80 0 C. (orig.) [de

  1. Immobilization of INEL low-level radioactive wastes in ceramic containment materials

    International Nuclear Information System (INIS)

    Seymour, W.C.; Kelsey, P.V.

    1978-11-01

    INEL low-level radioactive wastes have an overall chemical composition that lends itself to self-containment in a ceramic-based material. Fewer chemical additives would be needed to process the wastes than to process high-level wastes or use a mixture containment method. The resulting forms of waste material could include a basalt-type glass or glass ceramic and a ceramic-type brick. Expected leach resistance is discussed in relationshp to data found in the literature for these materials and appears encouraging. An overview of possible processing steps for the ceramic materials is presented

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

    International Nuclear Information System (INIS)

    Luey, J.; Powell, T.D.

    1992-03-01

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

  3. User's manual for remote-handled transuranic waste container welding and inspection fixture

    International Nuclear Information System (INIS)

    Hauptmann, J.P.

    1985-09-01

    Rockwell Hanford Operations (Rockwell) has designed built, and tested a prototype remotely operated welding and inspection fixture to be used in making the closure weld on the remote-handled transuranic (RH-TRU) waste container. The RH-TRU waste container has an average TRU concentration in excess of 100 nCi/gm, and a surface radiation dose rate in excess of 200 mrem/h, but not exceeding 100 rem/h. The RH-TRU waste container is to be used by defense waste generator sites in the United States for final packaging of RH-TRU wastes and is compatible with the requirements of the Waste Isolation Pilot Plant (WIPP) and the WIPP handling system. Standard and stacked RH-TRU container designs are available. The standard container is 26 in. in dia. by 121 in. high; the stacked containers are 26 in. in dia. by 61.25 in. high. After loading, two stacked containers are fitted and welded together to form the identical measurements of the standard 121-in. container. The prototype RH-TRU waste container welding and inspection fixture was intended for test and evaluation only, and not for installation in an operating facility. The final RH-TRU waste container welding and inspection fixture drawings (see appendix) incorporate several changes made following operational testing of the original fixture. These modifications are identified in this manual. However, not all modifications have been functionally tested. The purpose of this manual is to aid waste generator sites in designing a remotely operated welding and inspection fixture that will conform to their own requirements. Modifications to the Rockwell design must be evaluated for structural and WIPP handling requirements. This manual also provides design philosophy, component vendor information, and cost estimates

  4. Polymer-Cement Composites Containing Waste Perlite Powder

    Directory of Open Access Journals (Sweden)

    Paweł Łukowski

    2016-10-01

    Full Text Available Polymer-cement composites (PCCs are materials in which the polymer and mineral binder create an interpenetrating network and co-operate, significantly improving the performance of the material. On the other hand, the need for the utilization of waste materials is a demand of sustainable construction. Various mineral powders, such as fly ash or blast-furnace slag, are successfully used for the production of cement and concrete. This paper deals with the use of perlite powder, which is a burdensome waste from the process of thermal expansion of the raw perlite, as a component of PCCs. The results of the testing of the mechanical properties of the composite and some microscopic observations are presented, indicating that there is a possibility to rationally and efficiently utilize waste perlite powder as a component of the PCC. This would lead to creating a new type of building material that successfully meets the requirements of sustainable construction.

  5. Substance Flow Analysis of Wastes Containing Polybrominated Diphenyl Ethers

    DEFF Research Database (Denmark)

    Vyzinkarova, Dana; Brunner, Paul H.

    2013-01-01

    materials. Therefore, end-of-life (EOL) plastic materials used for construction must be separated and properly treated, for example, in a state-of-the-art municipal solid waste (MSW) incinerator. In the case of cOctaBDE, the main flows are waste electrical and electronic equipment (WEEE) and, possibly......The present article examines flows and stocks of Stockholm Convention regulated pollutants, commercial penta- and octabrominated diphenyl ether (cPentaBDE, cOctaBDE), on a city level. The goals are to (1) identify sources, pathways, and sinks of these compounds in the city of Vienna, (2) determine...... the fractions that reach final sinks, and (3) develop recommendations for waste management to ensure their minimum recycling and maximum transfer to appropriate final sinks. By means of substance flow analysis (SFA) and scenario analysis, it was found that the key flows of cPentaBDE stem from construction...

  6. Hanford facility dangerous waste permit application, 616 Nonradioactive Dangerous Waste Storage Facility. Revision 2, Chapter 3.0, Waste characteristics supplemental information; Volume 1

    International Nuclear Information System (INIS)

    1993-01-01

    This report contains supplemental information concerning waste characteristics for numerous nonradioactive waste materials. Uniform hazardous waste manifests are included for routine as well as nonroutine waste streams. The manifests contain the following information: waste disposal analysis; general instructions; waste destination; and transportation representatives

  7. Hanford facility dangerous waste permit application, 616 Nonradioactive Dangerous Waste Storage Facility. Revision 2, Chapter 3.0, Waste characteristics supplemental information; Volume 2

    International Nuclear Information System (INIS)

    1993-01-01

    This report contains supplemental information concerning waste characteristics for numerous nonradioactive waste materials. Uniform hazardous waste manifests are included for routine as well as nonroutine waste streams. The manifests contain the following information: waste disposal analysis; general instructions; waste destination; and transportation representatives

  8. Hanford facility dangerous waste permit application, 616 Nonradioactive Dangerous Waste Storage Facility. Revision 2, Chapter 3.0, Waste characteristics supplemental information; Volume 3

    International Nuclear Information System (INIS)

    1993-01-01

    This report contains supplemental information concerning waste characteristics for numerous nonradioactive waste materials. Uniform hazardous waste manifests are included for routine as well as nonroutine waste streams. The manifests contain the following information: waste disposal analysis; general instructions; waste destination; and transportation representatives

  9. Mathematical models for diffusive mass transfer from waste package container with multiple perforations

    International Nuclear Information System (INIS)

    Lee, J.H.; Andrews, R.W.; Chambre, P.L.

    1996-01-01

    A robust engineered barrier system (EBS) is employed in the current design concept for the potential high-level nuclear waste repository at Yucca Mountain, Nevada, US. The primary component of the EBS is a multi-barrier waste package container. Simplifying the geometry of the cylindrical waste package container and the underlying invert into the equivalent spherical configuration, mathematical models are developed for steady-state and transient diffusive releases from the failed waste container with multiple perforations (or pit penetrations) at the boundary of the invert. Using the models the steady-state and transient diffusive release behaviors form the failed waste container are studied. The analyses show that the number of perforations, the size of perforation, the container wall thickness, the geometry of the waste container and invert, and the adsorption of radionuclide in the invert are the important parameters that control the diffusive release rate. It is emphasized that the failed (or perforated) waste package container can still perform as a potentially important barrier (or diffusion barrier) to radionuclide release

  10. STABILIZATION AND TESTING OF MERCURY CONTAINING WASTES: BORDEN SLUDGE

    Science.gov (United States)

    This report details the stability assessment of a mercury containing sulfide treatment sludge. Information contained in this report will consist of background data submitted by the geneerator, landfill data supplied by EPA and characterization and leaching studies conducted by UC...

  11. Remote mining for in-situ waste containment. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Martinelli, D.; Banta, L.; Peng, S. [and others

    1995-10-01

    This document presents the findings of a study conducted at West Virginia University to determine the feasibility of using a combination of longwall mining and standard landfill lining technologies to mitigate contamination of groundwater supplies by leachates from hazardous waste sites.

  12. Remote mining for in-situ waste containment. Final report

    International Nuclear Information System (INIS)

    Martinelli, D.; Banta, L.; Peng, S.

    1995-10-01

    This document presents the findings of a study conducted at West Virginia University to determine the feasibility of using a combination of longwall mining and standard landfill lining technologies to mitigate contamination of groundwater supplies by leachates from hazardous waste sites

  13. Chemical durability of glasses containing radioactive fission product waste

    International Nuclear Information System (INIS)

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

    1974-04-01

    Measurements made to determine the chemical durability of glasses for disposal of radioactive waste are discussed. The term glass covers materials varying from true glass with only minute quantities of crystallites, such as insoluble RuO 2 , to quasi glass-ceramics which are mostly crystalline. Chemical durability requirements and Soxhlet extractor leach tests are discussed

  14. Application of ion exchange in liquid radioactive waste management of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Pal, Puskar; Chopra, S K; Sharma, P D [Nuclear Power Corporation, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    The operation of nuclear power plants would necessarily result in generation of gaseous, liquid and solid radioactive wastes. The wastes are treated/conditioned to ensure that the permissible discharge limits laid down by Atomic Energy Regulatory Board of India are complied with. The wastes are segregated on activity levels, types of radioisotopes present and chemical nature of liquid streams. The basic philosophy of various treatment techniques is to concentrate and contain as much activity as possible. It is of utmost importance that the wastes are effectively treated by proven methods/processes. The radiochemical nature of waste generated is one of the parameters to select a treatment/conditioning method. The paper presents an outline of various processes adopted for treatment of liquid waste and ion exchange processes, their application in liquid waste management in detail. Projected quantities of liquid wastes for the current designs are included. (author). 2 tabs.

  15. Design compliance matrix waste sample container filling system for nested, fixed-depth sampling system

    International Nuclear Information System (INIS)

    BOGER, R.M.

    1999-01-01

    This design compliance matrix document provides specific design related functional characteristics, constraints, and requirements for the container filling system that is part of the nested, fixed-depth sampling system. This document addresses performance, external interfaces, ALARA, Authorization Basis, environmental and design code requirements for the container filling system. The container filling system will interface with the waste stream from the fluidic pumping channels of the nested, fixed-depth sampling system and will fill containers with waste that meet the Resource Conservation and Recovery Act (RCRA) criteria for waste that contains volatile and semi-volatile organic materials. The specifications for the nested, fixed-depth sampling system are described in a Level 2 Specification document (HNF-3483, Rev. 1). The basis for this design compliance matrix document is the Tank Waste Remediation System (TWRS) desk instructions for design Compliance matrix documents (PI-CP-008-00, Rev. 0)

  16. 40 CFR 264.316 - Disposal of small containers of hazardous waste in overpacked drums (lab packs).

    Science.gov (United States)

    2010-07-01

    ... HAZARDOUS WASTE TREATMENT, STORAGE, AND DISPOSAL FACILITIES Landfills § 264.316 Disposal of small containers of hazardous waste in overpacked drums (lab packs). Small containers of hazardous waste in overpacked... hazardous waste in overpacked drums (lab packs). 264.316 Section 264.316 Protection of Environment...

  17. 40 CFR 265.316 - Disposal of small containers of hazardous waste in overpacked drums (lab packs).

    Science.gov (United States)

    2010-07-01

    ... OPERATORS OF HAZARDOUS WASTE TREATMENT, STORAGE, AND DISPOSAL FACILITIES Landfills § 265.316 Disposal of small containers of hazardous waste in overpacked drums (lab packs). Small containers of hazardous waste... hazardous waste in overpacked drums (lab packs). 265.316 Section 265.316 Protection of Environment...

  18. Fabrication and closure development of nuclear waste containers for storage at the Yucca Mountain, Nevada repository

    International Nuclear Information System (INIS)

    Russell, E.W.; Nelson, T.A.; Domian, H.A.; LaCount, D.F.; Robitz, E.S.; Stein, K.O.

    1989-04-01

    US Congress and the President have determined that the Yucca Mountain site in Nevada is to be characterized to determine its suitability for construction of the first US high-level nuclear waste repository. Work in connection with this site is carried out within the Yucca Mountain Project (YMP). Lawrence Livermore National Laboratory (LLNL) has the responsibility for designing, developing, and projecting the performance of the waste package for the permanent storage of high-level nuclear waste. Babcock ampersand Wilcox (B ampersand W) is involved with the YMP as a subcontractor to LLNL. B ampersand W's role is to recommend and demonstrate a method for fabricating the metallic waste container and a method for performing the final closure of the container after it has been filled with waste. Various fabrication and closure methods are under consideration for the production of containers. This paper presents progress to date in identifying and evaluating the candidate manufacturing processes. 2 refs., 1 fig., 7 tabs

  19. Test procedures for polyester immobilized salt-containing surrogate mixed wastes

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  20. Incineration of urban solid waste containing radioactive sources

    Energy Technology Data Exchange (ETDEWEB)

    Ronchin, G.P., E-mail: giulio.ronchin@mail.polimi.i [Dipartimento di Energia (Sezione nucleare - Cesnef), Politecnico di Milano, Via Ponzio 34/3, 20133 Milano (Italy); Campi, F.; Porta, A.A. [Dipartimento di Energia (Sezione nucleare - Cesnef), Politecnico di Milano, Via Ponzio 34/3, 20133 Milano (Italy)

    2011-01-15

    Incineration of urban solid waste accidentally contaminated by orphan sources or radioactive material is a potential risk for environment and public health. Moreover, production and emission of radioactive fumes can cause a heavy contamination of the plant, leading to important economic detriment. In order to prevent such a hazard, in February 2004 a radiometric portal for detection of radioactive material in incoming waste has been installed at AMSA (Azienda Milanese per i Servizi Ambientali) 'Silla 2' urban solid waste incineration plant of Milan. Radioactive detections performed from installation time up to December 2006 consist entirely of low-activity material contaminated from radiopharmaceuticals (mainly {sup 131}I). In this work an estimate of the dose that would have been committed to population, due to incineration of the radioactive material detected by the radiometric portal, has been evaluated. Furthermore, public health and environmental effects due to incineration of a high-activity source have been estimated. Incineration of the contaminated material detected appears to have negligible effects at all; the evaluated annual collective dose, almost entirely conferred by {sup 131}I, is indeed 0.1 man mSv. Otherwise, incineration of a 3.7 x 10{sup 10} Bq (1 Ci) source of {sup 137}Cs, assumed as reference accident, could result in a light environmental contamination involving a large area. Although the maximum total dose, owing to inhalation and submersion, committed to a single individual appears to be negligible (less than 10{sup -8} Sv), the environmental contamination leads to a potential important exposure due to ingestion of contaminated foods. With respect to 'Silla 2' plant and to the worst meteorological conditions, the evaluated collective dose results in 0.34 man Sv. Performed analyses have confirmed that radiometric portals, which are today mainly used in foundries, represent a valid public health and environmental

  1. Quality control of radioactive waste disposal container for borehole project

    International Nuclear Information System (INIS)

    Mohamad Pauzi Ismail; Suhairy Sani; Azhar Azmi; Ilham Mukhriz Zainal Abidin

    2014-01-01

    This paper explained quality control of radioactive disposal container for the borehole project. Non-destructive Testing (NDT) is one of the quality tool used for evaluating the product. The disposal container is made of 316L stainless steel. The suitable NDT method for this object is radiography, ultrasonic, penetrant and eddy current testing. This container will be filled with radioactive capsules and cement mortar is grouted to fill the gap. The results of NDT measurements are explained and discussed. (author)

  2. Preliminary treatment of chlorinated waste streams containing fission products

    Energy Technology Data Exchange (ETDEWEB)

    Hudry, Damien; Bardez, Isabelle; Bart, Florence [CEA Marcoule DTCD/SECM/LM2C, BP 17171, 30207 Bagnols sur Ceze (France); Deniard, Philippe; Jobic, Stephane [Institut des Materiaux Jean Rouxel, Universite de Nantes, CNRS, BP 32229, 44322 Nantes cedex 3 (France); Rakhmatullin, Aydar [Conditions Extremes et Materiaux: Hautes Temperatures et Irradiations, CEMHTI-CNRS, 45071 Orleans cedex 2 (France); Bessada, Catherine [Conditions Extremes et Materiaux: Hautes Temperatures et Irradiations, CEMHTI-CNRS, 45071 Orleans cedex 2 (France); Universite d' Orleans, Faculte des Sciences, BP 6749, 45067 Orleans cedex 2 (France)

    2008-07-01

    Separating actinides from fission products (FP) by electrolytic techniques in a molten chloride medium produces high-level waste which, because of its high chlorine content, cannot be directly and quantitatively loaded in a glass matrix and therefore requires the development of new management methods. In this regard the strategy of submitting chlorinated waste streams to a preliminary treatment consists in separating the various types of FP from the solvent to minimize the ultimate high-level waste volume. Selective precipitation of the rare earth elements by NH{sub 4}H{sub 2}PO{sub 4} was investigated in a LiCl-KCl medium, and could constitute the first step in the purification process. Unlike the use of alkali orthophosphate, this method provides similar conversion factors with the simple addition of stoichiometric phosphorus (P:rare-earth = 1) and does not require excess phosphate (P:rare-earth = 5). This prevents the formation of a secondary Li{sub 3}PO{sub 4} phase. Moreover, NH{sub 4}H{sub 2}PO{sub 4} also allows chlorine bound to rare earth elements to be eliminated as NH{sub 4}Cl. The formation of HCl is highly probable.

  3. Process Description for the Retrieval of Earth Covered Transuranic (TRU) Waste Containers at the Hanford Site

    International Nuclear Information System (INIS)

    DEROSA, D.C.

    2000-01-01

    This document describes process and operational options for retrieval of the contact-handled suspect transuranic waste drums currently stored below grade in earth-covered trenches at the Hanford Site. Retrieval processes and options discussed include excavation, container retrieval, venting, non-destructive assay, criticality avoidance, incidental waste handling, site preparation, equipment, and shipping

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  5. Process Description for the Retrieval of Earth Covered Transuranic (TRU) Waste Containers at the Hanford Site

    Energy Technology Data Exchange (ETDEWEB)

    DEROSA, D.C.

    2000-01-13

    This document describes process and operational options for retrieval of the contact-handled suspect transuranic waste drums currently stored below grade in earth-covered trenches at the Hanford Site. Retrieval processes and options discussed include excavation, container retrieval, venting, non-destructive assay, criticality avoidance, incidental waste handling, site preparation, equipment, and shipping.

  6. Evaluation of potential mixed wastes containing lead, chromium, or used oil

    International Nuclear Information System (INIS)

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

    1987-01-01

    This paper presents the results of follow-on studies conducted by Brookhaven National Laboratory (BNL) for the Nuclear Regulatory Commission (NRC) on certain kinds of low-level waste (LLW) which could also be classified as hazardous waste subject to regulation by the Environmental Protection Agency (EPA). Such LLW is termed ''mixed waste.'' Additional data have been collected and evaluated on two categories of potential mixed waste, namely LLW containing metallic lead and LLW containing chromium. Additionally, LLW with organic liquids, especially liquid scintillation wastes, are reviewed. In light of a proposed EPA rule to list used oil as hazardous waste, the potential mixed waste hazard of used oil contaminated with radionuclides is discussed. It is concluded that the EPA test for determining whether a solid waste exhibits the hazardous characteristic of extraction procedure toxicity does not adequately simulate the burial environment at LLW disposal sites, and in particular, does not adequately assess the potential for dissolution and transport of buried metallic lead. Also, although chromates are, in general, not a normal or routine constitutent in commercial LLW (with the possible exception of chemical decontamination wastes), light water reactors which do use chromates might find it beneficial to consider alternative corrosion inhibitors. In addition, it is noted that if used oil is listed by the EPA as hazardous waste, LLW oil may be managed by a scheme including one or more of the following processes: incineration, immobilization, sorption, aqueous extraction and glass furnace processing

  7. The conceptual design of waste repository for radioactive waste from medical, industrial and research facilities containing comparatively high radioactivity

    International Nuclear Information System (INIS)

    Yamamoto, Masayuki; Hashimoto, Naro

    2002-02-01

    Advisory Committee on Nuclear Fuel Cycle Backend Policy reported the basic approach to the RI and Institute etc. wastes on March 2002. According to it, radioactive waste form medical, industrial and research facilities should be classified by their radioactivity properties and physical and chemical properties, and should be disposed in the appropriate types of repository with that classification. For the radioactive waste containing comparatively high radioactivity generated from reactors, NSC has established the Concentration limit for disposal. NSC is now discussing about the limit for the radioactive waste from medical, industrial and research facilities containing comparatively high radioactivity. Japan Nuclear Cycle Development Institute (JNC) preliminary studied about the repository for radioactive waste from medical, industrial and research facilities and discussed about the problems for design on H12. This study was started to consider those problems, and to develop the conceptual design of the repository for radioactive waste from medical, industrial and research facilities. Safety assessment for that repository is also performed. The result of this study showed that radioactive waste from medical, industrial and research facilities of high activity should be disposed in the repository that has higher performance of barrier system comparing with the vault type near surface facility. If the conditions of the natural barrier and the engineering barrier are clearer, optimization of the design will be possible. (author)

  8. Data for radioactive waste management and nuclear applications

    International Nuclear Information System (INIS)

    Stewart, D.C.

    1985-01-01

    This book is a specialized handbook on the management of radioactive waste, including information applicable to related applications. It consolidates information from many sources to develop techniques for dealing with radioactive waste management and features reports and other specialized data not available in libraries. It covers physical data, chemical data, types of radioactive wastes, and data for different operations

  9. Characterization and storage of liquid wastes containing 125Iodine in the laboratory for production of brachytherapy sources - IPEN

    International Nuclear Information System (INIS)

    Carvalho, Vitória S.; Souza, Daiane C.B. de; Barbosa, Nayane K.O.; Rodrigues, Bruna T.; Nogueira, Beatriz R.; Costa, Osvaldo L. da; Zeituni, Carlos A.; Vicente, Roberto; Rostelato, Maria E.C.M.

    2017-01-01

    Radioactive sources of Iodine-125 for medical applications have been developed at the Institute for Energy and Nuclear Research (IPEN) to meet the growing demand for medical applications such as brachytherapy. A dedicated laboratory is already being implemented at IPEN. Part of the processes involved in the production of sealed sources generate radioactive wastes that despite the short half-life (<100 days) have radioactive activity above the levels of exemption established by the Brazilian National Nuclear Energy Commission. Therefore, these wastes should receive appropriate treatment and storage until they reach the levels of release into the environment. This work aims to determine the volumes of the liquid wastes generated during the production stages of the sources, as well as to propose a temporary storage system for such wastes. The applied methodology consisted in determining the volumes of wastes generated in each production cell according to the manufacturing steps. After that, activities and activity concentrations were calculated for each container used for temporary storage inside the production laboratory. The total volume stored for one year in the temporary storage, as well as the rate of entry and exit of the liquid wastes were calculated according to the source production demand and the decay time of the radionuclide, respectively. The main results showed that the time required to reach sanitary sewage disposal values is within the period of operation of the facility. The total volume generated is also within the facility's temporary storage capacity

  10. TECHNICAL GUIDANCE DOCUMENT: CONSTRUCTION QUALITY MANAGEMENT FOR REMEDIAL ACTION AND REMEDIAL DESIGN WASTE CONTAINMENT SYSTEMS

    Science.gov (United States)

    This Technical Guidance Document is intended to augment the numerous construction quality control and construction quality assurance (CQC and CQA) documents that are available far materials associated with waste containment systems developed for Superfund site remediation. In ge...

  11. Monitoring the Durability Performance of Concrete in Nuclear Waste Containment. Technical Progress Report No. 3

    International Nuclear Information System (INIS)

    Ulm, Franz-Josef

    2000-01-01

    OAK-B135 Monitoring the Durability Performance of Concrete in Nuclear Waste Containment. Technical Progress Report No. 3(NOTE: Part II A item 1 indicates ''PAPER'', but a report is attached electronically)

  12. Assessing reliability and useful life of containers for disposal of irradiated fuel waste

    International Nuclear Information System (INIS)

    Doubt, G.

    1984-06-01

    Metal containers for fuel waste isolation are to be designed to last at least 500 years to provide a redundant barrier during the decay period of the high activity components of the waste. To meet the long-life requirement, containers must have a very low failure rate during the design mission, a low incidence of 'juvenile failures' due to undetected defects, and resistance to progressive deterioration from environmental processes. This paper summarizes studies to determine: (1) precedent for low failure rates and relevance to container longevity; (b) the likelihood of initial defects perforating the container before or shortly after emplacement, and estimates of material defect distribution; (c) the utility of reliability analysis techniques for estimating reliability and life of fuel waste containers; (d) other approaches to estimating container longevity and failure versus time distribution

  13. Method and apparatus for disposing a radioactive waste container to submarine bottom

    International Nuclear Information System (INIS)

    Shibata, Kiyoshi; Yoshida, Shoichi.

    1980-01-01

    Purpose: To completely eliminate a danger occurred by the rolling of a hull in the ocean in a method and apparatus for disposing radioactive waste container to submarine bottom by independently handling the radioactive waste containers when loading the container in a compartment carried on a barge and sinking the containers together with the compartment to the submarine bottom at its disposing time. Method: Radioactive waste containers are carried into a compartment loaded on a barge floating completely, and the barge is then applied with external force thereto by a ship or the like and sailed to the marine disposal area. Then, water is filled in the ballast tank of the barge to submerge the barge, the compartment is floated and separated from the containers, and water is charged into the compartment to sink the compartment. (Aizawa, K.)

  14. Hanford facility dangerous waste permit application, 616 Nonradioactive Dangerous Waste Storage Facility. Revision 2A

    International Nuclear Information System (INIS)

    Bowman, R.C.

    1994-04-01

    This permit application for the 616 Nonradioactive Dangerous Waste Storage Facility consists for 15 chapters. Topics of discussion include the following: facility description and general provisions; waste characteristics; process information; personnel training; reporting and record keeping; and certification

  15. Application to transfer radioactive waste to the Nevada Test Site

    International Nuclear Information System (INIS)

    1992-01-01

    All waste described in this application has been, and will be, generated by LANL in support of the nuclear weapons test program at the NTS. All waste originates on the NTS. DOE Order 5820.2A states that low-level radioactive waste shall be disposed of at the site where it is generated, when practical. Since the waste is produced at the NTS, it is cost effective for LANL to dispose of the waste at the NTS

  16. Corrosion issues in high-level nuclear waste containers

    Science.gov (United States)

    Asl, Samin Sharifi

    In this dissertation different aspects of corrosion and electrochemistry of copper, candidate canister material in Scandinavian high-level nuclear waste disposal program, including the thermodynamics and kinetics of the reactions that are predicted to occur in the practical system have been studied. A comprehensive thermodynamic study of copper in contact with granitic groundwater of the type and composition that is expected in the Forsmark repository in Sweden has been performed. Our primary objective was to ascertain whether copper would exist in the thermodynamically immune state in the repository, in which case corrosion could not occur and the issue of corrosion in the assessment of the storage technology would be moot. In spite of the fact that metallic copper has been found to exist for geological times in granitic geological formations, copper is well-known to be activated from the immune state to corrode by specific species that may exist in the environment. The principal activator of copper is known to be sulfur in its various forms, including sulfide (H2S, HS-, S2-), polysulfide (H2Sx, HSx -, Sx 2-), poly sulfur thiosulfate ( SxO3 2-), and polythionates (SxO6 2-). A comprehensive study of this aspect of copper chemistry has never been reported, and yet an understanding of this issue is vital for assessing whether copper is a suitable material for fabricating canisters for the disposal of HLNW. Our study identifies and explores those species that activate copper; these species include sulfur-containing entities as well as other, non-sulfur species that may be present in the repository. The effects of temperature, solution pH, and hydrogen pressure on the kinetics of the hydrogen electrode reaction (HER) on copper in borate buffer solution have been studied by means of steady-state polarization measurements, including electrochemical impedance spectroscopy (EIS). In order to obtain electrokinetic parameters, such as the exchange current density and the

  17. TECHNOLOGY FOR EFFICIENT USAGE OF HYDROCARBON-CONTAINING WASTE IN PRODUCTION OF MULTI-COMPONENT SOLID FUEL

    Directory of Open Access Journals (Sweden)

    B. M. Khroustalev

    2016-01-01

    Full Text Available The paper considers modern approaches to usage of hydrocarbon-containing waste as energy resources and presents description of investigations, statistic materials, analysis results on formation of hydrocarbon-containing waste in the Republic of Belarus. Main problems pertaining to usage of waste as a fuel and technologies for their application have been given in the paper. The paper describes main results of the investigations and a method for efficient application of viscous hydrocarbon-containing waste as an energy-packed component and a binding material while producing a solid fuel. A technological scheme, a prototype industrial unit which are necessary to realize a method for obtaining multi-component solid fuel are represented in the paper. A paper also provides a model of technological process with efficient sequence of technological operations and parameters of optimum component composition. Main factors exerting significant structure-formation influence in creation of structural composition of multi-component solid fuel have been presented in the paper. The paper gives a graphical representation of the principle for selection of mixture particles of various coarseness to form a solid fuel while using a briquetting method and comprising viscous hydrocarbon-containing waste. A dependence of dimensionless concentration g of emissions into atmosphere during burning of two-component solid fuel has been described in the paper. The paper analyzes an influence of the developed methodology for emission calculation of multi-component solid fuels and reveals a possibility to optimize the component composition in accordance with ecological function and individual peculiar features of fuel-burning equipment. Special features concerning storage and transportation, advantages and disadvantages, comparative characteristics, practical applicability of the developed multi-component solid fuel have been considered and presented in the paper. The paper

  18. High level waste containing granules coated and embedded in metal as an alternative to HLW glasses

    International Nuclear Information System (INIS)

    Neumann, W.

    1980-01-01

    Simulated high level waste containing granules were overcoated with pyrocarbon or nickel respectively. The coatings were performed by the use of chemical vapour deposition in a fluidized bed. The coated granules were embedded in an aluminium-silicon-alloy to improve the dissipation of radiation induced heat. The metal-granules-composites obtained were of improved product stability related to the high level waste containing glasses. (orig.) [de

  19. Precipitation of Aluminum Containing Species in Tank Wastes

    International Nuclear Information System (INIS)

    Mattigod, Shas V.; Hobbs, David; Parker, Kent E.; McCready, David E.

    2001-01-01

    Aluminisilicate deposit buildup experienced during the tank waste volume-reduction process at the Savannah River Site (SRS) required an evaporator to be shut down in October 1999. The Waste Processing Technology Section (WPTS) of Westinghouse Savannah River Company at SRS is now collaborating with team members from Pacific Northwest National Laboratory (PNNL) to verify the steady-state thermodynamic stability of aluminosilicate compounds under waste tank conditions in an attempt to eliminate the deposition and clogging problems. The data obtained at 40?C showed that formation and persistence of crystalline phases was dependent on the initial hydroxide concentrations. The formation and persistence of zeolite A occurred only at lower hydroxide concentrations, whereas increasing hydroxide concentrations appeared to promote the formation of sodalite and cancrinite. The data also showed that although zeolite A forms initially, it is a metastable phase that converts to more stable crystalline materials such as sodalite and cancrinite. Additionally, the rate of transformation of zeolite A appeared to increase with increasing hydroxide concentration. The data from tests conducted at 80?C revealed relatively rapid formation of sodalite and cancrinite. Although minor amounts of zeolite A were initially detected in some cases, the higher reaction temperatures seemed to promote very rapid transformation of this phase into more stable phases. Also, the higher temperature and hydroxide concentrations appeared to initiate kinetically fast crystallization of sodalite and cancrinite. More recent testing at SRS in support of the HLW evaporator plugging issue has shown similar trends in the formation of aluminosilicate phases. These tests were carried out under conditions more similar to those that occur in HLW tanks and evaporators. Comparison of our results with those reported above show very similar trends

  20. Process and closure system for a radioactive waste container

    International Nuclear Information System (INIS)

    Meyer, Andre.

    1974-01-01

    The closure process described is for a cylindrical radioactive waste drum. It makes use of a closure system for the drum comprising two lids separated by a twin flange seal. It consists essentially in placing a double flange 'O' ring inside the upper lip of the drum, and after filling has been completed, fitting the first lid on the twin flange 'O' ring and pushing down this lid whilst squashing the upper flange and then putting on the second lid in the usual prescribed manner. A description is also given of the drum sealing apparatus [fr

  1. Recovering of thorium contained in wastes from Thorium Purification Plant

    International Nuclear Information System (INIS)

    Brandao Filho, D.; Hespanhol, E.C.B.; Baba, S.; Miranda, L.E.T.; Araujo, J.A. de.

    1992-08-01

    A study has been developed in order to establish a chemical process for recovering thorium from wastes produced at the Thorium Purification Plant of the Instituto de Pesquisas Energeticas e Nucleares. The recovery of thorium in this process will be made by means of solvent extraction technique. Solutions of TBP/Varsol were employed as extracting agent during the runs. The influence of thorium concentration in the solution, aqueous phase acidity, volume ratio of the phases, percentage of TBP/Varsol and the contact time of the phases on the extraction of thorium and lanthanides was determined. (author)

  2. Inhibition of nuclear waste solutions containing multiple aggressive anions

    International Nuclear Information System (INIS)

    Congdon, J.W.

    1987-01-01

    The inhibition of localized corrosion of carbon steel in caustic, high-level radioactive waste solutions was studied using cyclic potentiodynamic polarization scans, supplemented by partially immersed coupon tests. The electrochemical tests provided a rapid and accurate means of determining the relationship between the minimum inhibitor requirements and the concentration of the aggressive anions in this system. Nitrate, sulfate, chloride, and fluoride were identified as aggressive anions, however, no synergistic effects were observed between these anions. This observation may have important theoretical implications because it tends to contradict the behavior of aggressive anions as predicted by existing theories for localized corrosion. 10 refs., 5 figs., 2 tabs

  3. Characterization and extraction of gold contained in foundry industrial wastes

    International Nuclear Information System (INIS)

    Vite T, J.; Vite T, M.; Diaz C, A.; Carreno de Leon, C.

    1999-01-01

    Gold was characterized and leached in foundry sands. These wastes are product among others of the automotive industry where they are used as molds material which are contaminated by diverse metals during the foundry. To fulfil the leaching process four coupled thermostat columns were used. To characterize the solid it was used the X-ray diffraction technique. For the qualitative analysis it was used the Activation analysis technique. Finally, for the study of liquors was used the Plasma diffraction spectroscopy (Icp-As) technique. The obtained results show that the process which was used the thermostat columns was more efficient, than the methods traditionally recommended. (Author)

  4. RCRA Part A Permit Application for Waste Management Activities at the Nevada Test Site, Part B Permit Application Hazardous Waste Storage Unit, Nevada Test Site, and Part B Permit Application - Explosives Ordnance Disposal Unit (EODU)

    International Nuclear Information System (INIS)

    2010-01-01

    The Area 5 Hazardous Waste Storage Unit (HWSU) was established to support testing, research, and remediation activities at the Nevada Test Site (NTS), a large-quantity generator of hazardous waste. The HWSU, located adjacent to the Area 5 Radioactive Waste Management Site (RWMS), is a prefabricated, rigid steel-framed, roofed shelter used to store hazardous nonradioactive waste generated on the NTS. No offsite generated wastes are managed at the HWSU. Waste managed at the HWSU includes the following categories: Flammables/Combustibles; Acid Corrosives; Alkali Corrosives; Oxidizers/Reactives; Toxics/Poisons; and Other Regulated Materials (ORMs). A list of the regulated waste codes accepted for storage at the HWSU is provided in Section B.2. Hazardous wastes stored at the HWSU are stored in U.S. Department of Transportation (DOT) compliant containers, compatible with the stored waste. Waste transfer (between containers) is not allowed at the HWSU and containers remain closed at all times. Containers are stored on secondary containment pallets and the unit is inspected monthly. Table 1 provides the metric conversion factors used in this application. Table 2 provides a list of existing permits. Table 3 lists operational Resource Conservation and Recovery Act (RCRA) units at the NTS and their respective regulatory status.

  5. RCRA Part A Permit Application for Waste Management Activities at the Nevada Test Site, Part B Permit Application Hazardous Waste Storage Unit, Nevada Test Site, and Part B Permit Application - Explosives Ordnance Disposal Unit (EODU)

    Energy Technology Data Exchange (ETDEWEB)

    NSTec Environmental Programs

    2010-06-17

    The Area 5 Hazardous Waste Storage Unit (HWSU) was established to support testing, research, and remediation activities at the Nevada Test Site (NTS), a large-quantity generator of hazardous waste. The HWSU, located adjacent to the Area 5 Radioactive Waste Management Site (RWMS), is a prefabricated, rigid steel-framed, roofed shelter used to store hazardous nonradioactive waste generated on the NTS. No offsite generated wastes are managed at the HWSU. Waste managed at the HWSU includes the following categories: Flammables/Combustibles; Acid Corrosives; Alkali Corrosives; Oxidizers/Reactives; Toxics/Poisons; and Other Regulated Materials (ORMs). A list of the regulated waste codes accepted for storage at the HWSU is provided in Section B.2. Hazardous wastes stored at the HWSU are stored in U.S. Department of Transportation (DOT) compliant containers, compatible with the stored waste. Waste transfer (between containers) is not allowed at the HWSU and containers remain closed at all times. Containers are stored on secondary containment pallets and the unit is inspected monthly. Table 1 provides the metric conversion factors used in this application. Table 2 provides a list of existing permits. Table 3 lists operational Resource Conservation and Recovery Act (RCRA) units at the NTS and their respective regulatory status.

  6. Recovery of Proteins and Chromium Complexes from Chromium – Containing Leather Waste (CCLW

    Directory of Open Access Journals (Sweden)

    B. Gutti

    2010-08-01

    Full Text Available Chromium – Containing Leather Waste (CCLW constitutes an environmental pollution problem to leather industries disposing the waste by landfill. The waste mainly consists of collagen and chromium III complexes. This work is a design of reactors to recover gelatin, polypeptides and chromium from CCLW. The results of the experiment shows that 68% of protein, based on dry weight of leather scraps, could be recovered. Three reactors with a total volume of 18 m3 was designed to handle 10,431 kg of waste generated from the tanning industries.

  7. Hanford Facility dangerous waste permit application, general information

    International Nuclear Information System (INIS)

    1993-05-01

    The current Hanford Facility Dangerous Waste Permit Application is considered to be a single application organized into a General Information Portion (this document, number DOE/RL-91-28) and a treatment, storage, and/or disposal Unit-Specific Portion, which includes documentation for individual TSD units (e.g., document numbers DOE/RL-89-03 and DOE/RL-90-01). Both portions consist of a Part A division and a Part B division. The Part B division consists of 15 chapters that address the content of the Part B checklists prepared by the Washington State Department of Ecology (Ecology 1987) and the US Environmental Protection Agency (40 Code of Federal Regulations 270), with additional information requirements mandated by the Hazardous and Solid Waste Amendments of 1984 and revisions of Washington Administrative Code 173-303. For ease of reference, the Washington State Department of Ecology checklist section numbers, in brackets, follow the chapter headings and subheadings. Documentation contained in the General Information Portion (i.e., this document, number DOE/RL-91-28) is broader in nature and applies to all treatment, storage, and/or disposal units for which final status is sought. Because of its broad nature, the Part A division of the General Information Portion references the Hanford Facility Dangerous Waste Part A Permit Application (document number DOE/RL-88-21), a compilation of all Part A documentation for the Hanford Facility

  8. Processing method and processing device for liquid waste containing surface active agent and radioactive material

    International Nuclear Information System (INIS)

    Nishi, Takashi; Matsuda, Masami; Baba, Tsutomu; Yoshikawa, Ryozo; Yukita, Atsushi.

    1998-01-01

    Washing liquid wastes containing surface active agents and radioactive materials are sent to a deaerating vessel. Ozone is blown into the deaerating vessel. The washing liquid wastes dissolved with ozone are introduced to a UV ray irradiation vessel. UV rays are irradiated to the washing liquid wastes, and hydroxy radicals generated by photodecomposition of dissolved ozone oxidatively decompose surface active agents contained in the washing liquid wastes. The washing liquid wastes discharged from the UV ray irradiation vessel are sent to an activated carbon mixing vessel and mixed with powdery activated carbon. The surface active agents not decomposed in the UV ray irradiation vessel are adsorbed to the activated carbon. Then, the activated carbon and washing liquid wastes are separated by an activated carbon separating/drying device. Radioactive materials (iron oxide and the like) contained in the washing liquid wastes are mostly granular, and they are separated and removed from the washing liquid wastes in the activated carbon separating/drying device. (I.N.)

  9. Method of processing cellulose filter sludge containing radioactive waste

    International Nuclear Information System (INIS)

    Shibata, Setsuo; Shibuya, Hidetoshi; Kusakabe, Takao; Kawakami, Hiroshi.

    1991-01-01

    To cellulose filter sludges deposited with radioactive wastes, 1 to 15% of cellulase based on the solid content of the filter sludges is caused to act in an aqueous medium with 4 to 8 pH at 10 to 50degC. If the pH value exceeds 8, hydrolyzing effect of cellulase is decreased, whereas a tank is corroded if the pH value is 4 or lower. If temperature is lower than 10degC, the rate of the hydrolysis reaction is too low to be practical. It is appropriate that the temperature is at the order of 40degC. If it exceeds 50degC, the cellulase itself becomes unstable. It is most effective that the amount of cellulase is about 8% and its addition by more than 15% is not effective. In this way, liquids in which most of filter sludges are hydrolyzed are processed as low level radioactive wastes. (T.M.)

  10. Characterization of Mechanical and Bactericidal Properties of Cement Mortars Containing Waste Glass Aggregate and Nanomaterials

    Directory of Open Access Journals (Sweden)

    Pawel Sikora

    2016-08-01

    Full Text Available The recycling of waste glass is a major problem for municipalities worldwide. The problem concerns especially colored waste glass which, due to its low recycling rate as result of high level of impurity, has mostly been dumped into landfills. In recent years, a new use was found for it: instead of creating waste, it can be recycled as an additive in building materials. The aim of the study was to evaluate the possibility of manufacturing sustainable and self-cleaning cement mortars with use of commercially available nanomaterials and brown soda-lime waste glass. Mechanical and bactericidal properties of cement mortars containing brown soda-lime waste glass and commercially available nanomaterials (amorphous nanosilica and cement containing nanocrystalline titanium dioxide were analyzed in terms of waste glass content and the effectiveness of nanomaterials. Quartz sand is replaced with brown waste glass at ratios of 25%, 50%, 75% and 100% by weight. Study has shown that waste glass can act as a successful replacement for sand (up to 100% to produce cement mortars while nanosilica is incorporated. Additionally, a positive effect of waste glass aggregate for bactericidal properties of cement mortars was observed.

  11. Characterization of Mechanical and Bactericidal Properties of Cement Mortars Containing Waste Glass Aggregate and Nanomaterials

    Science.gov (United States)

    Sikora, Pawel; Augustyniak, Adrian; Cendrowski, Krzysztof; Horszczaruk, Elzbieta; Rucinska, Teresa; Nawrotek, Pawel; Mijowska, Ewa

    2016-01-01

    The recycling of waste glass is a major problem for municipalities worldwide. The problem concerns especially colored waste glass which, due to its low recycling rate as result of high level of impurity, has mostly been dumped into landfills. In recent years, a new use was found for it: instead of creating waste, it can be recycled as an additive in building materials. The aim of the study was to evaluate the possibility of manufacturing sustainable and self-cleaning cement mortars with use of commercially available nanomaterials and brown soda-lime waste glass. Mechanical and bactericidal properties of cement mortars containing brown soda-lime waste glass and commercially available nanomaterials (amorphous nanosilica and cement containing nanocrystalline titanium dioxide) were analyzed in terms of waste glass content and the effectiveness of nanomaterials. Quartz sand is replaced with brown waste glass at ratios of 25%, 50%, 75% and 100% by weight. Study has shown that waste glass can act as a successful replacement for sand (up to 100%) to produce cement mortars while nanosilica is incorporated. Additionally, a positive effect of waste glass aggregate for bactericidal properties of cement mortars was observed. PMID:28773823

  12. Mechanical properties of concrete containing recycled concrete aggregate (RCA) and ceramic waste as coarse aggregate replacement

    Science.gov (United States)

    Khalid, Faisal Sheikh; Azmi, Nurul Bazilah; Sumandi, Khairul Azwa Syafiq Mohd; Mazenan, Puteri Natasya

    2017-10-01

    Many construction and development activities today consume large amounts of concrete. The amount of construction waste is also increasing because of the demolition process. Much of this waste can be recycled to produce new products and increase the sustainability of construction projects. As recyclable construction wastes, concrete and ceramic can replace the natural aggregate in concrete because of their hard and strong physical properties. This research used 25%, 35%, and 45% recycled concrete aggregate (RCA) and ceramic waste as coarse aggregate in producing concrete. Several tests, such as concrete cube compression and splitting tensile tests, were also performed to determine and compare the mechanical properties of the recycled concrete with those of the normal concrete that contains 100% natural aggregate. The concrete containing 35% RCA and 35% ceramic waste showed the best properties compared with the normal concrete.

  13. Review of LLNL Mixed Waste Streams for the Application of Potential Waste Reduction Controls

    International Nuclear Information System (INIS)

    Belue, A; Fischer, R P

    2007-01-01

    In July 2004, LLNL adopted the International Standard ISO 14001 as a Work Smart Standard in lieu of DOE Order 450.1. In support of this new requirement the Director issued a new environmental policy that was documented in Section 3.0 of Document 1.2, ''ES and H Policies of LLNL'', in the ES and H Manual. In recent years the Environmental Management System (EMS) process has become formalized as LLNL adopted ISO 14001 as part of the contract under which the laboratory is operated for the Department of Energy (DOE). On May 9, 2005, LLNL revised its Integrated Safety Management System Description to enhance existing environmental requirements to meet ISO 14001. Effective October 1, 2005, each new project or activity is required to be evaluated from an environmental aspect, particularly if a potential exists for significant environmental impacts. Authorizing organizations are required to consider the management of all environmental aspects, the applicable regulatory requirements, and reasonable actions that can be taken to reduce negative environmental impacts. During 2006, LLNL has worked to implement the corrective actions addressing the deficiencies identified in the DOE/LSO audit. LLNL has begun to update the present EMS to meet the requirements of ISO 14001:2004. The EMS commits LLNL--and each employee--to responsible stewardship of all the environmental resources in our care. The generation of mixed radioactive waste was identified as a significant environmental aspect. Mixed waste for the purposes of this report is defined as waste materials containing both hazardous chemical and radioactive constituents. Significant environmental aspects require that an Environmental Management Plan (EMP) be developed. The objective of the EMP developed for mixed waste (EMP-005) is to evaluate options for reducing the amount of mixed waste generated. This document presents the findings of the evaluation of mixed waste generated at LLNL and a proposed plan for reduction

  14. Thermoelectric cooling container for medical applications

    Energy Technology Data Exchange (ETDEWEB)

    Aivazov, A A; Shtern, Y I; Budaguan, B G; Makhrachev, K B; Pastor, M

    1997-07-01

    In this work the thermoelectric cooling container for storing and transportation of the medicine, particularly for insulin, is discussed. In the working volume the temperature is supported on the level of +4 C. The container can work in two operating conditions: with the power supply and without the power supply. Two removable blocks are used for this purpose. One block (thermoelectric) is used for the work with the power supply and another (passive)-for the work without power supply. The thermoelectric block has a 12V power supply, which is used in the automobiles, yachts and other kinds of transport. The temperature in the working volume is supported by the use of the Peltier effect. An electronic device is used in this block and stabilizes temperature on the level of +4 C and indicates information about working conditions. The thermoelectric container has a power supply block for work at 220(110)V. The working temperature in the container can be maintained in the absence of the power supply. In this case the necessary temperature conditions are supported by melting of the crystallized salt. For this purpose the container has a hermetic volume containing this salt and contacting with the working volume.

  15. A Prototype Scintillating Fibre Tracker for the Cosmic-Ray Muon Tomography of Legacy Nuclear Waste Containers

    OpenAIRE

    Kaiser, R.; Clarkson, A.; Hamilton, D. J.; Hoek, M.; Ireland, D.G.; Johnstone, J.R.; Keri, T.; Lumsden, S.; Mahon, D. F.; McKinnon, B.; Murray, M.; Nutbeam-Tuffs, S.; Shearer, C.; Staines, C.; Yang, G.

    2014-01-01

    Cosmic-ray muons are highly-penetrative charged particles observed at sea level with a flux of approximately 1 cm−2 min−1. They interact with matter primarily through Coulomb scattering which can be exploited in muon tomography to image objects within industrial nuclear waste containers. This paper presents the prototype scintillating-fibre detector developed for this application at the University of Glasgow. Experimental results taken with test objects are shown in comparison to results from...

  16. 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

  17. A process for treatment of mixed waste containing chemical plating wastes

    International Nuclear Information System (INIS)

    Anast, K.R.; Dziewinski, J.; Lussiez, G.

    1995-01-01

    The Waste Treatment and Minimization Group at Los Alamos National Laboratory has designed and will be constructing a transportable treatment system to treat low-level radioactive mixed waste generated during plating operations. The chemical and plating waste treatment system is composed of two modules with six submodules, which can be trucked to user sites to treat a wide variety of aqueous waste solutions. The process is designed to remove the hazardous components from the waste stream, generating chemically benign, disposable liquids and solids with low level radioactivity. The chemical and plating waste treatment system is designed as a multifunctional process capable of treating several different types of wastes. At this time, the unit has been the designated treatment process for these wastes: Destruction of free cyanide and metal-cyanide complexes from spent plating solutions; destruction of ammonia in solution from spent plating solutions; reduction of Cr VI to Cr III from spent plating solutions, precipitation, solids separation, and immobilization; heavy metal precipitation from spent plating solutions, solids separation, and immobilization, and acid or base neutralization from unspecified solutions

  18. Characterisation of concrete containers for radioactive waste in the engineering tranches system at the Yugoslav R.A waste storing center

    International Nuclear Information System (INIS)

    Plecas, I.; Peric, A.; Drljaca, J.; Kostadinovic, A.

    1987-10-01

    Low and intermediate level radioactive waste represents 90% of total R.A. waste. It is conditioned into special concrete containers. Since these concrete containers are to protect safely the radioactive waste for 300 years, the selection of materials and precise control of their physical and mechanical properties is very important. In this paper results obtained with some concrete compositions are described. (author)

  19. In situ testing of titanium and mild steel nuclear waste containers at the WIPP site

    International Nuclear Information System (INIS)

    Molecke, M.A.

    1990-01-01

    An overview of the Waste Isolation Pilot Plant (WIPP) in situ tests on the corrosion of titanium and mild steel for high level waste containers is presented. The tests at Sandia have moved out of the laboratory into a test underground facility in order to evaluate the performance of the waste package material. The tests are being performed under both near-reference and accelerated salt repository conditions. Some containers are filled with high level waste glass (non-radioactive); others contain electric heaters. Backfill material is either bentonite/sand or crushed salt. In other tests metals and glasses are exposed directly to brine. The tests are designed to study the corrosion and metallurgy of the canister and overpack materials; the feasibility and performance of backfill materials; and near-field effects such as brine migration

  20. Electrolytic treatment of liquid waste containing ammonium nitrate

    International Nuclear Information System (INIS)

    Komori, R.; Ogawa, N.; Ohtsuka, K.; Ohuchi, J.

    1981-01-01

    A study was made on the safe decomposition of ammonium nitrate, which is the main component of α-liquid waste from plutonium fuel facilities, by means of electrolytic reduction and thermal decomposition. In the first stage, ammonium nitrate is reduced to ammonium nitrite by electrolytic reduction using an electrolyser with a cation exchange membrane as a diaphragm. In the second stage, ammonium nitrite is decomposed to N 2 and H 2 O. The alkaline region and a low temperature are preferable for electrolytic reduction and the acidic region and high temperature for thermal decomposition. A basis was established for an ammonium nitrate treatment system in aqueous solution through the operation of a bench-scale unit, and the operating data obtained was applied to the basic design of a 10-m 3 /a facility. (author)

  1. Fracture during cooling of cast borosilicate glass containing nuclear wastes

    International Nuclear Information System (INIS)

    Smith, P.K.; Baxter, C.A.

    1981-09-01

    Procedures and techniques were evaluated to mitigate thermal stress fracture in waste glass as the glass cools after casting. The two principal causes of fracture identified in small-scale testing are internal thermal stresses arising from excessive thermal gradients when cooled too fast, and shear fracturing in the surface of the glass because the stainless steel canister shrinks faster than the glass on cooling. Acoustic emission and ceramographic techniques were used to outline an annealing schedule that requires at least three weeks of controlled cooling below 550 0 C to avoid excessive thermal gradients and corresponding stresses. Fracture arising from canister interactions cannot be relieved by slow cooling, but can be eliminated for stainless steel canisters by using ceramic paper, ceramic or graphite paste linings, or by choosing a canister material with a thermal expansion coefficient comparable to, or less than, that of the glass

  2. Viscosity of glasses containing simulated Savannah River Plant waste

    International Nuclear Information System (INIS)

    Plodinec, M.J.

    1978-08-01

    The viscosity of glass melts containing four simulated sludge types and two frit candidates (Frits 18 and 21) was measured over the temperature range 750 to 1200 0 C. The viscosity of melts made with either frit was reduced by the addition of high iron sludge, unchanged by average sludge, and increased by composite and high aluminum sludge. High aluminium sludge greatly increased the viscosity. Frit 21 (containing 4 wt % Li 2 O substituted for 4 wt % Na 2 O in Frit 18) was clearly better than Frit 18 in terms of its low viscosity. However, further reductions in viscosity are desirable, especially for glasses containing high aluminum sludge. Changing any frit component by 1 wt % did not significantly affect the viscosity of the glasses. Therefore, variability of 1 wt % in any frit component can be tolerated

  3. Management of wastes containing radioactivity from mining and milling of uranium ores in Northern Australia

    International Nuclear Information System (INIS)

    Costello, J.M.

    1977-01-01

    The procedures and controls to achieve safe management of wastes containing radioactivity during the mining and processing of uranium ores are mainly site specific depending on the nature, location and distribution of the ore and gangue material. Waste rock and below-ore-grade material containing low levels of radioactivity require disposal at the mine site. In open cut mining the material is generally stockpiled above ground, with revegetation and collection of run-off water. Some material may be used to backfill open cuts. Management of these wastes requires a thorough investigation of ground water hydrology and surface soil characteristics to control dissipation of radioactive material. Dust containing radon and radioactive particulate is produced during ore milling, and dusts of ore concentrate are generated during calcination and packaging of the yellowcake product. These dusts are managed by ventilation and filtration systems, working conditions, and discharges to atmosphere will be according to the Australian Code of Practice on Radiation Protection during Mining and Milling of Uranium Ores. The chemical waste stream from leaching and processing of the uranium ores contains the majority of the radioactivity resulting from radium and its decay products. Neutralised effluent is discharged into holding ponds for settling of solids. This paper describes the nature of wastes containing radioactivity resulting from the mining and milling of uranium, and illustrates modern engineering practices and monitoring procedures to manage the wastes, as described in the Environmental Impact statement produced by Ranger Uranium Mines Proprietary Limited for public hearings

  4. Management of wastes containing radioactivity from mining and milling uranium ores in Northern Australia

    International Nuclear Information System (INIS)

    Costello, J.M.

    1977-01-01

    The procedures and controls to achieve safe management of wastes containing radioactivity during the mining and processing of uranium ores are mainly site-specific depending on the nature, location and distribution of the ore and gangue material. Waste rock and below-ore-grade material containing low levels of radioactivity require disposal at the mine site. In open-cut mining the material is generally stockpiled above ground, with revegetation and collection of run-off water. Some material may be used to backfill open cuts. Management of these wastes requires a thorough investigation of groundwater hydrology and surface soil characteristics to control dissipation of radioactive material. Dust containing radon and radioactive particulate is produced during ore milling, and dusts of ore concentrate are generated during calcination and packaging of the yellowcake product. These dusts are managed by ventilation and filtration systems; working conditions and discharges to atmosphere will be according to the Australian Code of Practice on Radiation Protection during Mining and Milling of Uranium Ores. The chemical waste stream from leaching and processing of the uranium ores contains most of the radioactivity resulting from radium and its decay products. Neutralized effluent is discharged into holding ponds for settling solids. The paper describes the nature of wastes containing radioactivity resulting from the mining and milling of uranium, and illustrates modern engineering practices and monitoring procedures to manage the wastes, as described in the Environmental Impact Statement produced by Ranger Uranium Mines Pty Ltd (RUM) for public hearings. (author)

  5. Organic rankine cycle waste heat applications

    Science.gov (United States)

    Brasz, Joost J.; Biederman, Bruce P.

    2007-02-13

    A machine designed as a centrifugal compressor is applied as an organic rankine cycle turbine by operating the machine in reverse. In order to accommodate the higher pressures when operating as a turbine, a suitable refrigerant is chosen such that the pressures and temperatures are maintained within established limits. Such an adaptation of existing, relatively inexpensive equipment to an application that may be otherwise uneconomical, allows for the convenient and economical use of energy that would be otherwise lost by waste heat to the atmosphere.

  6. Zirconium phosphate waste forms for low-temperature stabilization of cesium-137-containing waste streams

    International Nuclear Information System (INIS)

    Singh, D.; Wagh, A.S.; Tlustochowicz.

    1996-04-01

    Novel chemically bonded phosphate ceramics are being developed and fabricated for low-temperature stabilization and solidification of waste streams that are not amenable to conventional high-temperature stabilization processes because volatiles are present in the wastes. A composite of zirconium-magnesium phosphate has been developed and shown to stabilize ash waste contaminated with a radioactive surrogate of 137 Cs. Excellent retainment of cesium in the phosphate matrix system was observed in Toxicity Characteristic Leaching Procedure tests. This was attributed to the capture of cesium in the layered zirconium phosphate structure by intercalation ion-exchange reaction. But because zirconium phosphate has low strength, a novel zirconium/magnesium phosphate composite waste form system was developed. The performance of these final waste forms, as indicated by compression strength and durability in aqueous environments, satisfy the regulatory criteria. Test results indicate that zirconium-magnesium-phosphate-based final waste forms present a viable technology for treatment and solidification of cesium-contaminated wastes

  7. A batch assay to measure microbial hydrogen sulfide production from sulfur-containing solid wastes

    International Nuclear Information System (INIS)

    Sun, Mei; Sun, Wenjie; Barlaz, Morton A.

    2016-01-01

    Large volumes of sulfur-containing wastes enter municipal solid waste landfills each year. Under the anaerobic conditions that prevail in landfills, oxidized forms of sulfur, primarily sulfate, are converted to sulfide. Hydrogen sulfide (H 2 S) is corrosive to landfill gas collection and treatment systems, and its presence in landfill gas often necessitates the installation of expensive removal systems. For landfill operators to understand the cost of managing sulfur-containing wastes, an estimate of the H 2 S production potential is needed. The objective of this study was to develop and demonstrate a biochemical sulfide potential (BSP) test to measure the amount of H 2 S produced by different types of sulfur-containing wastes in a relatively fast (30 days) and inexpensive (125 mL serum bottles) batch assay. This study confirmed the toxic effect of H 2 S on both sulfate reduction and methane production in batch systems, and demonstrated that removing accumulated H 2 S by base adsorption was effective for mitigating inhibition. H 2 S production potentials of coal combustion fly ash, flue gas desulfurization residual, municipal solid waste combustion ash, and construction and demolition waste were determined in BSP assays. After 30 days of incubation, most of the sulfate in the wastes was converted to gaseous or aqueous phase sulfide, with BSPs ranging from 0.8 to 58.8 mL H 2 S/g waste, depending on the chemical composition of the samples. Selected samples contained solid phase sulfide which contributed to the measured H 2 S yield. A 60 day incubation in selected samples resulted in 39–86% additional sulfide production. H 2 S production measured in BSP assays was compared with that measured in simulated landfill reactors and that calculated from chemical analyses. H 2 S production in BSP assays and in reactors was lower than the stoichiometric values calculated from chemical composition for all wastes tested, demonstrating the importance of assays to estimate the

  8. A batch assay to measure microbial hydrogen sulfide production from sulfur-containing solid wastes

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Mei, E-mail: msun8@uncc.edu [Department of Civil, Construction, and Environmental Engineering, North Carolina State University, Campus Box 7908, Raleigh, NC (United States); Sun, Wenjie, E-mail: wsun@smu.edu [Department of Civil, Construction, and Environmental Engineering, North Carolina State University, Campus Box 7908, Raleigh, NC (United States); Department of Civil and Environmental Engineering, Southern Methodist University, PO Box 750340, Dallas, TX (United States); Barlaz, Morton A., E-mail: barlaz@ncsu.edu [Department of Civil, Construction, and Environmental Engineering, North Carolina State University, Campus Box 7908, Raleigh, NC (United States)

    2016-05-01

    Large volumes of sulfur-containing wastes enter municipal solid waste landfills each year. Under the anaerobic conditions that prevail in landfills, oxidized forms of sulfur, primarily sulfate, are converted to sulfide. Hydrogen sulfide (H{sub 2}S) is corrosive to landfill gas collection and treatment systems, and its presence in landfill gas often necessitates the installation of expensive removal systems. For landfill operators to understand the cost of managing sulfur-containing wastes, an estimate of the H{sub 2}S production potential is needed. The objective of this study was to develop and demonstrate a biochemical sulfide potential (BSP) test to measure the amount of H{sub 2}S produced by different types of sulfur-containing wastes in a relatively fast (30 days) and inexpensive (125 mL serum bottles) batch assay. This study confirmed the toxic effect of H{sub 2}S on both sulfate reduction and methane production in batch systems, and demonstrated that removing accumulated H{sub 2}S by base adsorption was effective for mitigating inhibition. H{sub 2}S production potentials of coal combustion fly ash, flue gas desulfurization residual, municipal solid waste combustion ash, and construction and demolition waste were determined in BSP assays. After 30 days of incubation, most of the sulfate in the wastes was converted to gaseous or aqueous phase sulfide, with BSPs ranging from 0.8 to 58.8 mL H{sub 2}S/g waste, depending on the chemical composition of the samples. Selected samples contained solid phase sulfide which contributed to the measured H{sub 2}S yield. A 60 day incubation in selected samples resulted in 39–86% additional sulfide production. H{sub 2}S production measured in BSP assays was compared with that measured in simulated landfill reactors and that calculated from chemical analyses. H{sub 2}S production in BSP assays and in reactors was lower than the stoichiometric values calculated from chemical composition for all wastes tested, demonstrating

  9. Docker container analysis and application in microservice architecture

    OpenAIRE

    Benedik, Dejan

    2016-01-01

    Using container technology to run web applications and services is becoming increasingly common [9]. In this work we analyse the most popular container technology, Docker, and compare it with other Linux-based container technologies. Due to near-instant startup and minimal overhead, containers are facilitating a new way of designing software applications that is called microservice architecture. We analyse the approach and compare it with its alternatives. We examine some aspects of running s...

  10. Properties of concrete containing foamed concrete block waste as fine aggregate replacement

    Science.gov (United States)

    Muthusamy, K.; Budiea, A. M. A.; Zaidan, A. L. F.; Rasid, M. H.; Hazimmah, D. S.

    2017-11-01

    Environmental degradation due to excessive sand mining dumping at certain places and disposal of foamed concrete block waste from lightweight concrete producing industry are issues that should be resolved for a better and cleaner environment of the community. Thus, the main intention of this study is to investigate the potential of foamed concrete block waste as partial sand replacement in concrete production. The foamed concrete waste (FCW) used in this research that were supplied by a local lightweight concrete producing industry. The workability and compressive strength of concrete containing various percentage of foamed concrete waste as partial sand replacement has been investigated. Prior to the use, the foamed concrete waste were crushed to produce finer particles. Six concrete mixes containing various content of crushed foamed concrete waste that are 0%, 10%, 20%, 30%, 40% and 50% were used in this experimental work. Then the prepared specimens were placed in water curing until the testing age. Compressive strength test and flexural strength tests were conducted at 7, 14 and 28 days. The result shows that integration of crushed foamed concrete waste as partial sand replacement in concrete reduces the mix workability. It is interesting to note that both compressive strength and flexural strength of concrete improves when 30% crushed foamed concrete waste is added as partial sand replacement.

  11. Incineration of radioactive wastes containing only C-14 and H-3

    International Nuclear Information System (INIS)

    Garcia, Corazon M.

    1992-01-01

    C-14 and H-3 arc popularly used in chemical and biological research institutions in the Philippines. Most of the solid radioactive wastes generated by these institutions consist of combustible materials such as paper and accumulated environmental samples. Liquid wastes usually contain organic substances. The method proposed for managing C-14 and H-3 wastes is incineration which is expected to provide an acceptable means of disposal for C-14 and H-3 and their hazardous organic constituent. In the incineration process) the radioactively contaminated waste will be mixed with non-radioactive combustible wastes to lower the activity concentration and to improve the efficiency of combustion which will be carried out in a locally fabricated drum incinerator. The calculations presented determines the concentration limit for the incinerable wastes and the restriction on specific activity of the particles of the incinerable wastes containing C-14 or H-3 on the basis of the accepted air concentration and on the annual dose limit for an average radiation worker in the country. In the calculations for C-14, considerations were taken on the exposure received from the deposition of radioactive particles in the lungs containing unoxidized carbon. Calculations for H-3, however, is based on the assumption that the concentration of the radionuclide in the body water is the same as that in the environment. (author)

  12. Preliminary study for treatment methodology establishment of liquid waste containing uranium in refining facility lagoon

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung Jik; Lee, Kune Woo; Won, Hui Jun; Ahn, Byung Gil; Shim, Joon Bo

    1999-12-01

    The preliminary study which establishes the treatment methodology of the sludge waste containing uranium in the conversion facility lagoon was performed. The property of lagoon liquid waste such as the initial water content, the density including radiochemical analysis results were obtained using the samples taken from the lagoon. The objective of this study is to provide some basically needed materials for selection of the most proper lagoon waste treatment methodology by reviewing the effective processes and methods for minimizing the secondary waste resulting from the treatment and disposition of large amount of radioactive liquid waste according to the facility closing. The lagoon waste can be classified into two sorts, such as supernatant and precipitate. The supernatants contain uranium less than 5 ppm and their water content are about 35 percent. Therefore, supernatants are solutions composed of mainly salt components. However, the precipitates have lots of uranium compound contained in the coagulation matrix, and are formed as two kinds of crystalline structures. The most proper method minimizing the secondary waste would be direct drying and solidification of the supernatants and precipitates after separation of them by filtering. (author)

  13. Development of thermal conditioning technology for Alpha-containment wastes: Alpha-contaminated waste incineration technology

    International Nuclear Information System (INIS)

    Kim, Joon Hyung; Kim, Jeong Guk; Yang, Hee Chul; Choi, Byung Seon; Jeong, Myeong Soo

    1999-03-01

    As the first step of a 3-year project named 'development of alpha-contaminated waste incineration technology', the basic information and data were reviewed, while focusing on establishment of R and D direction to develop the final goal, self-supporting treatment of α- wastes that would be generated from domestic nuclear industries. The status on α waste incineration technology of advanced states was reviewed. A conceptual design for α waste incineration process was suggested. Besides, removal characteristics of volatile metals and radionuclides in a low-temperature dry off-gas system were investigated. Radiation dose assessments and some modification for the Demonstration-scale Incineration Plant (DSIP) at Korea Atomic Energy Research Institute (KAERI) were also done

  14. Development of thermal conditioning technology for Alpha-containment wastes: Alpha-contaminated waste incineration technology

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Joon Hyung; Kim, Jeong Guk; Yang, Hee Chul; Choi, Byung Seon; Jeong, Myeong Soo

    1999-03-01

    As the first step of a 3-year project named 'development of alpha-contaminated waste incineration technology', the basic information and data were reviewed, while focusing on establishment of R and D direction to develop the final goal, self-supporting treatment of {alpha}- wastes that would be generated from domestic nuclear industries. The status on {alpha} waste incineration technology of advanced states was reviewed. A conceptual design for {alpha} waste incineration process was suggested. Besides, removal characteristics of volatile metals and radionuclides in a low-temperature dry off-gas system were investigated. Radiation dose assessments and some modification for the Demonstration-scale Incineration Plant (DSIP) at Korea Atomic Energy Research Institute (KAERI) were also done.

  15. Far-Field Accumulation of Fissile Material From Waste Packages Containing Plutonium Disposition Waste Form

    International Nuclear Information System (INIS)

    J.P. Nicot

    2000-01-01

    The objective of this calculation is to estimate the quantity of fissile material that could accumulate in fractures in the rock beneath plutonium-ceramic (Pu-ceramic) and Mixed-Oxide (MOX) waste packages (WPs) as they degrade in the potential monitored geologic repository at Yucca Mountain. This calculation is to feed another calculation (Ref. 31) computing the probability of criticality in the systems described in Section 6 and then ultimately to a more general report on the impact of plutonium on the performance of the proposed repository (Ref. 32), both developed concurrently to this work. This calculation is done in accordance with the development plan TDP-DDC-MD-000001 (Ref. 9), item 5. The original document described in item 5 has been split into two documents: this calculation and Ref. 4. The scope of the calculation is limited to only very low flow rates because they lead to the most conservative cases for Pu accumulation and more generally are consistent with the way the effluent from the WP (called source term in this calculation) was calculated (Ref. 4). Ref. 4 (''In-Drift Accumulation of Fissile Material from WPs Containing Plutonium Disposition Waste Forms'') details the evolution through time (breach time is initial time) of the chemical composition of the solution inside the WP as degradation of the fuel and other materials proceed. It is the chemical solution used as a source term in this calculation. Ref. 4 takes that same source term and reacts it with the invert; this calculation reacts it with the rock. In addition to reactions with the rock minerals (that release Si and Ca), the basic mechanisms for actinide precipitation are dilution and mixing with resident water as explained in Section 2.1.4. No other potential mechanism such as flow through a reducing zone is investigated in this calculation. No attempt was made to use the effluent water from the bottom of the invert instead of using directly the effluent water from the WP. This

  16. Far-Field Accumulation of Fissile Material From Waste Packages Containing Plutonium Disposition Waste Form

    Energy Technology Data Exchange (ETDEWEB)

    J.P. Nicot

    2000-09-29

    The objective of this calculation is to estimate the quantity of fissile material that could accumulate in fractures in the rock beneath plutonium-ceramic (Pu-ceramic) and Mixed-Oxide (MOX) waste packages (WPs) as they degrade in the potential monitored geologic repository at Yucca Mountain. This calculation is to feed another calculation (Ref. 31) computing the probability of criticality in the systems described in Section 6 and then ultimately to a more general report on the impact of plutonium on the performance of the proposed repository (Ref. 32), both developed concurrently to this work. This calculation is done in accordance with the development plan TDP-DDC-MD-000001 (Ref. 9), item 5. The original document described in item 5 has been split into two documents: this calculation and Ref. 4. The scope of the calculation is limited to only very low flow rates because they lead to the most conservative cases for Pu accumulation and more generally are consistent with the way the effluent from the WP (called source term in this calculation) was calculated (Ref. 4). Ref. 4 (''In-Drift Accumulation of Fissile Material from WPs Containing Plutonium Disposition Waste Forms'') details the evolution through time (breach time is initial time) of the chemical composition of the solution inside the WP as degradation of the fuel and other materials proceed. It is the chemical solution used as a source term in this calculation. Ref. 4 takes that same source term and reacts it with the invert; this calculation reacts it with the rock. In addition to reactions with the rock minerals (that release Si and Ca), the basic mechanisms for actinide precipitation are dilution and mixing with resident water as explained in Section 2.1.4. No other potential mechanism such as flow through a reducing zone is investigated in this calculation. No attempt was made to use the effluent water from the bottom of the invert instead of using directly the effluent water from the

  17. An assessment of the feasibility of indefinite containment of Canadian nuclear fuel wastes

    International Nuclear Information System (INIS)

    Shoesmith, D.W.; King, F.; Ikeda, B.M.

    1995-05-01

    This report presents an analysis of the expected corrosion behaviour of nuclear fuel waste containers in a conceptual Canadian disposal vault. The container materials considered are dilute Ti alloys (Grades-2, -12 and -16) and oxygen-free copper. The corrosive conditions within the disposal vault change with time as the initially trapped oxygen is consumed and as the heat and γ-radiation produced by the waste decays. This evolution of the vault environment is broadly classified into an early, warm and oxidizing period followed by a period of long-term, stable, cool and non-oxidizing conditions. The corrosion behaviour of both types of material during these two periods is discussed, and various models that have been developed to predict the lifetimes of the containers are presented. The conclusion is that indefinite containment of the waste is feasible with both copper and titanium alloys under Canadian disposal conditions. (author). refs., tabs., figs

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  19. 40 CFR 61.155 - Standard for operations that convert asbestos-containing waste material into nonasbestos...

    Science.gov (United States)

    2010-07-01

    ... to the information requirements of § 61.07(b)(3), a (i) Description of waste feed handling and...) Disposed of as asbestos-containing waste material according to § 61.150, or (ii) Recycled as waste feed... waste feed to the process. (2) Collect and analyze monthly composite samples (one 200-gram (7-ounce...

  20. Effects of waste glass and waste foundry sand additions on reclaimed tiles containing sewage sludge ash.

    Science.gov (United States)

    Lin, Deng-Fong; Luo, Huan-Lin; Lin, Kuo-Liang; Liu, Zhe-Kun

    2017-07-01

    Applying sewage sludge ash (SSA) to produce reclaimed tiles is a promising recycling technology in resolving the increasing sludge wastes from wastewater treatment. However, performance of such reclaimed tiles is inferior to that of original ceramic tiles. Many researchers have therefore tried adding various industrial by-products to improve reclaimed tile properties. In this study, multiple materials including waste glass and waste foundry sand (WFS) were added in an attempt to improve physical and mechanical properties of reclaimed tiles with SSA. Samples with various combinations of clay, WFS, waste glass and SSA were made with three kiln temperatures of 1000°C, 1050°C, and 1100°C. A series of tests on the samples were next conducted. Test results showed that waste glass had positive effects on bending strength, water absorption and weight loss on ignition, while WFS contributed the most in reducing shrinkage, but could decrease the tile bending strength when large amount was added at a high kiln temperature. This study suggested that a combination of WFS from 10% to 15%, waste glass from 15% to 20%, SSA at 10% at a kiln temperature between 1000°C and 1050°C could result in quality reclaimed tiles with a balanced performance.

  1. SOLAR ENERGY APPLICATION IN WASTE TREATMENT- A REVIEW

    African Journals Online (AJOL)

    This review is an exposure on the various ways that solar energy can be harnessed for numerous waste treatment processes. Almost all forms of waste treatment require energy which is scarcely available considering the global energy crisis. The objective of this study is to enumerate the solar energy applications in waste ...

  2. Solid Waste Land Applications with Permits by the Iowa DNR

    Data.gov (United States)

    Iowa State University GIS Support and Research Facility — All types of facilities that handle solid waste, including: sanitary landfills, appliance demanufacturing facilities, transfer stations, land application sites,...

  3. The WIPP RCRA Part B permit application for TRU mixed waste disposal

    International Nuclear Information System (INIS)

    Johnson, J.E.

    1995-01-01

    In August 1993, the New Mexico Environment Department (NMED) issued a draft permit for the Waste Isolation Pilot Plant (WIPP) to begin experiments with transuranic (TRU) mixed waste. Subsequently, the Department of Energy (DOE) decided to cancel the on-site test program, opting instead for laboratory testing. The Secretary of the NMED withdrew the draft permit in 1994, ordering the State's Hazardous and Radioactive Waste Bureau to work with the DOE on submittal of a revised permit application. Revision 5 of the WIPP's Resource Conservation and Recovery Act (RCRA) Part B Permit Application was submitted to the NMED in May 1995, focusing on disposal of 175,600 m 3 of TRU mixed waste over a 25 year span plus ten years for closure. A key portion of the application, the Waste Analysis Plan, shifted from requirements to characterize a relatively small volume of TRU mixed waste for on-site experiments, to describing a complete program that would apply to all DOE TRU waste generating facilities and meet the appropriate RCRA regulations. Waste characterization will be conducted on a waste stream basis, fitting into three broad categories: (1) homogeneous solids, (2) soil/gravel, and (3) debris wastes. Techniques used include radiography, visually examining waste from opened containers, radioassay, headspace gas sampling, physical sampling and analysis of homogeneous wastes, and review of documented acceptable knowledge. Acceptable knowledge of the original organics and metals used, and the operations that generated these waste streams is sufficient in most cases to determine if the waste has toxicity characteristics, hazardous constituents, polychlorinated biphenyls (PBCs), or RCRA regulated metals

  4. Treatment alternatives of liquid radioactive waste containing uranium in phosphoric acid

    International Nuclear Information System (INIS)

    Bustamante Escobedo, Mauricio

    2003-01-01

    The UGDR, receives annually 100 [l] of liquid radioactive waste containing, highly acid (pH=0) uranium in phosphoric acid from the Laboratory of Chemical Analysis. This waste must be chemically and radiologically decontaminated before it can be discharged in accordance with local environmental standards. Chemical precipitation and evaporation test were carried out to define the operating conditions for the radiological decontamination of this radioactive waste and to obtain a solid waste that can be conditioned in a cement matrix. The evaporation process generates excellent rates of volume reduction, over 80%, but generates a pulp that is hard handle when submitted to a drying process. Chemical precipitation generates good results for decontaminating these solutions and reducing volume (above 50%) to obtain a uranium free effluent. The treatment with calcium carbonate generated an effluent with a low concentration of polluting agents. A preliminary test was carried out condition these solids in a cement matrix, using ratios of 0.45 waste/cement and 2 of water/cement. The mix prepared with waste from the sodium hydroxide treatment had low mechanical resistance resulting from the saline incrustations. The waste from the calcium carbonate treatment was very porous due to the water evaporation from the highly exothermic reaction between the waste and the cement. The mix of the calcium carbonate generated waste and the cement matrix needs to be optimized, since it generates favorable conditions for adhering with the cement matrix (au)

  5. Military Wastes-to-Energy Applications,

    Science.gov (United States)

    1980-11-01

    6 Residual) and Used Oil ( Automotive Crankcase Drainings) .......... ........................ 109 60 Properties of Coal: Bituminous, Subbituminous...facilities Physical/Chemical e Biomass removal from forest land may upset the Environment natural balance of terrestrial, aquatic, and atmospheric...ripout wastes (asbestos), sand- blasting wasts (organotin), and welding wastes (acetylene sludge); * Miscellaneous aircraft repair wastes -- brake

  6. Natural gas applications in waste management

    International Nuclear Information System (INIS)

    Tarman, P.B.

    1991-01-01

    The Institute of Gas Technology (IGT) is engaged in several projects related to the use of natural gas for waste management. These projects can be classified into four categories: cyclonic incineration of gaseous, liquid, and solid wastes; fluidized-bed reclamation of solid wastes; two-stage incineration of liquid and solid wastes; natural gas injection for emissions control. 5 refs., 8 figs

  7. Radiolysis effects on fuel corrosion within a failed nuclear waste container

    International Nuclear Information System (INIS)

    Sunder, S.; Shoeshmith, D.W.; Christensen, H.C.

    2003-01-01

    The concept of geological disposal of used nuclear fuel in corrosion resistant containers is being investigated in several countries. In the Canadian Nuclear Fuel Waste Management Program (CNFWMP), it is assumed that the used fuel will be disposed of in copper containers. Since the predicted lifetimes of these containers are very long (>106 years), only those containers emplaced with an undetected defect will fail within the period for which radionuclide release from the fuel must be considered. Early failure could lead to the entry of water into the container and subsequent release of radionuclides. The release rate of radionuclides from the used fuel will depend upon its dissolution rate. The primary mechanism for release will be the corrosion of the fuel driven by radiolytically-produced oxidants. The studies carried out to determine the effects of water radiolysis on fuel corrosion are reviewed, and some of the procedures used to predict corrosion rates of used fuel in failed nuclear waste containers described. (author)

  8. Heat transfer effects in vertically emplaced high level nuclear waste container

    International Nuclear Information System (INIS)

    Moujaes, S.F.; Lei, Y.M.

    1994-01-01

    Modeling free convection heat transfer in a cylindrical annular enclosure is still an active area of research and an important problem to be addressed in the high level nuclear waste repository. For the vertically emplaced waste container, the air gap which is between the container shell and the rock borehole, have an important role of dissipating heat to surrounding rock. These waste containers are vertically emplaced in the borehole 300 meters just below ground, and in a horizontal grid of 30 x 8 meters apart. The borehole will be capped after the container emplacement. The expected initial heat generated is between 3-4.74 kW per container depending on the type of waste. The goal of this study is to use a computer simulation model to find the borehole wall, air-gap and the container outer wall temperature distributions. The borehole wall temperature history has been found in the previous study, and was estimated to reach a maximum temperature of about 218 degrees C after 18 years from the emplacement. The temperature history of the rock surface is then used for the air-gap simulation. The problem includes convection and radiation heat transfer in a vertical enclosure. This paper will present the results of the convection in the air-gap over one thousand years after the containers' emplacement. During this long simulation period it was also observed that a multi-cellular air flow pattern can be generated in the air gap

  9. Essays of leaching in cemented products containing simulated waste from evaporator concentrated of PWR reactor

    International Nuclear Information System (INIS)

    Haucz, Maria Judite A.; Calabria, Jaqueline A. Almeida; Tello, Cledola Cassia O.; Candido, Francisco Donizete; Seles, Sandro Rogerio Novaes

    2011-01-01

    This paper evaluated the results from leaching resistance essays of cemented products, prepared from three distinct formulations, containing simulated waste of concentrated from the PWR reactor evaporator. The leaching rate is a parameter of qualification of solidified products containing radioactive waste and is determined in accordance with regulation ISO 6961. This procedure evaluates the capacity of transfer organic and inorganic substances presents in the waste through dissolution in the extractor medium. For the case of radioactive waste it is reached the more retention of contaminants in the cemented product, i.e.the lesser value of lixiviation rate. Therefore, this work evaluated among the proposed formulation that one which attend the criterion established in the regulation CNEN-NN-6.09

  10. Enviro-geotechnical considerations in waste containment system design and analysis

    International Nuclear Information System (INIS)

    Fang, H.Y.; Daniels, J.L.; Inyang, H.I.

    1997-01-01

    The effectiveness of waste control facilities hinges on careful evaluation of the overall planning, analysis and design of the entire system prior to construction. At present, most work is focused on the waste controlling system itself, with little attention given to the local environmental factors surrounding the facility sites. Containment materials including geomembranes, geotextiles and clay amended soils have received intense scrutiny. This paper, however, focuses on three relatively important issues relating to the characterization of the surrounding geomedia. Leakage through naturally occurring low-permeability soil layers, shrinkages swelling, cracking and effects of dynamic loads on system components are often responsible for a waste containment breach. In this paper, these mechanisms and their synergistic effects are explained in terms of the particle energy field theory. It is hoped that this additional information may assist the designer to be aware or take precaution to design safer future waste control facilities

  11. 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

  12. 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

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

    International Nuclear Information System (INIS)

    Luey, J.; Powell, T.D.

    1992-01-01

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

  14. LLNL/YMP Waste Container Fabrication and Closure Project; GFY technical activity summary

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1990-10-01

    The Department of Energy`s Office of Civilian Radioactive Waste Management (OCRWM) Program is studying Yucca Mountain, Nevada as a suitable site for the first US high-level nuclear waste repository. Lawrence Livermore National Laboratory (LLNL) has the responsibility for designing and developing the waste package for the permanent storage of high-level nuclear waste. This report is a summary of the technical activities for the LLNL/YMP Nuclear Waste Disposal Container Fabrication and Closure Development Project. Candidate welding closure processes were identified in the Phase 1 report. This report discusses Phase 2. Phase 2 of this effort involved laboratory studies to determine the optimum fabrication and closure processes. Because of budget limitations, LLNL narrowed the materials for evaluation in Phase 2 from the original six to four: Alloy 825, CDA 715, CDA 102 (or CDA 122) and CDA 952. Phase 2 studies focused on evaluation of candidate material in conjunction with fabrication and closure processes.

  15. The remote handling of canisters containing nuclear waste in glass at the Savannah River Plant

    International Nuclear Information System (INIS)

    Callan, J.E.

    1986-01-01

    The Defense Waste Processing Facility (DWPF) is a complete production area being constructed at the Savannah River Plant for the immobilization of nuclear waste in glass. The remote handling of canisters filled with nuclear waste in glass is an essential part of the process of the DWPF at the Savannah River Plant. The canisters are filled with nuclear waste containing up to 235,000 curies of radioactivity. Handling and movement of these canisters must be accomplished remotely since they radiate up to 5000 R/h. Within the Vitrification Building during filling, cleaning, and sealing, canisters are moved using standard cranes and trolleys and a specially designed grapple. During transportation to the Glass Waste Storage Building, a one-of-a-kind, specially designed Shielded Canister Transporter (SCT) is used. 8 figs

  16. Poisoning of bees with fluorine-containing industrial waste gases in Switzerland

    Energy Technology Data Exchange (ETDEWEB)

    Maurizio, A; Staub, M

    1956-01-01

    Mass poisoning of bees in the vicinity of Swiss Al factories was traced to high quantities of F in the waste gases. Plants and pollen in the vicinity contained 10-214 and 0.9-2.8 mg. % F resp., and collected rain water contained 0.1-10.4 mg. F/sg. dm. The dead bees each contained an av. of 15 ..gamma.. F.

  17. Conditions inside Water Pooled in a Failed Nuclear Waste Container and its Effect on Radionuclide Release

    Science.gov (United States)

    Hamdan, L. K.; Walton, J. C.; Woocay, A.

    2009-12-01

    Nuclear power use is expected to expand in the future, as part of the global clean energy initiative, to meet the world’s surging energy demand, and attenuate greenhouse gas emissions, which are mainly caused by fossil fuels. As a result, it is estimated that hundreds of thousands of metric tons of spent nuclear fuel (SNF) will accumulate. SNF disposal has major environmental (radiation exposure) and security (nuclear proliferation) concerns. Storage in unsaturated zone geological repositories is a reasonable solution for dealing with SNF. One of the key factors that determine the performance of the geological repository is the release of radionuclides from the engineered barrier system. Over time, the nuclear waste containers are expected to fail gradually due to general and localized corrosions and eventually infiltrating water will have access to the nuclear waste. Once radionuclides are released, they will be transported by water, and make their way to the accessible environment. Physical and chemical disturbances in the environment over the container will lead to different corrosion rates, causing different times and locations of penetration. One possible scenario for waste packages failure is the bathtub model, where penetrations occur on the top of the waste package and water pools inside it. In this paper the bathtub-type failed waste container is considered. We shed some light on chemical and physical processes that take place in the pooled water inside a partially failed waste container (bathtub category), and the effects of these processes on radionuclide release. Our study considers two possibilities: temperature stratification of the pooled water versus mixing process. Our calculations show that temperature stratification of the pooled water is expected when the waste package is half (or less) filled with water. On the other hand, when the waste package is fully filled (or above half) there will be mixing in the upper part of water. The effect of

  18. Studies on the recovery of 233U from phosphate containing aqueous waste using DBDECMP as extractant

    International Nuclear Information System (INIS)

    Sagar, V.B.; Oak, M.S.; Pawar, S.M.; Sivaramakrishnan, C.K.; Patil, S.K.

    1990-01-01

    A method for the recovery and purification of 233 U from phosphate containing analytical waste is developed. Extraction studies with Di-butyl N,N-diethylcarbamoylmethylphosphonate (DBDECMP) in xylene were carried out to explore the feasibility of separation and purification of 233 U from such wastes. Based on the data obtained, optimum conditions for the recovery of 233 U are suggested. (author) 11 refs.; 1 fig.; 3 tabs

  19. Multiple containment for LSA [low specific activity] and SCO [surface contaminated objects] wastes

    International Nuclear Information System (INIS)

    Burgess, M.H.

    1993-09-01

    Radioactive wastes are generally transported in the form of Low Specific Activity (LSA) materials or Surface Contaminated Objects (SCO). This report proposes that a method of acknowledging the beneficial effects of multiple containment for such wastes should be written into the 1996 Edition of the IAEA Transport Regulations. Experience used to assess risks from on-site movements of radioactive material in the UK can be applied to develop safety arguments justifying the alleviation of off-site transport risks. (UK)

  20. Flexible process options for the immobilisation of residues and wastes containing plutonium

    International Nuclear Information System (INIS)

    Stewart, M.W.A.; Moricca, S.A.; Day, R. A.; Begg, B. D.; Scales, C. R.; Maddrell, E. R.; Eilbeck, A. B.

    2007-01-01

    Residues and waste streams containing plutonium present unique technical, safety, regulatory, security, and socio-political challenges. In the UK these streams range from lightly plutonium contaminated materials (PCM) through to residue s resulting directly from Pu processing operations. In addition there are potentially stocks of Pu oxide powders whose future designation may be either a waste or an asset, due to their levels of contamination making their reuse uneconomic, or to changes in nuclear policy. While waste management routes exist for PCM, an immobilisation process is required for streams containing higher levels of Pu. Such a process is being developed by Nexia Solutions and ANSTO to treat and immobilise Pu waste and residues currently stored on the Sellafield site. The characteristics of these Pu waste streams are highly variable. The physical form of the Pu waste ranges from liquids, sludges, powders/granules, to solid components (e.g., test fuels), with the Pu present as an ion in solution, as a salt, metal, oxide or other compound. The chemistry of the Pu waste streams also varies considerably with a variety of impurities present in many waste streams. Furthermore, with fissile isotopes present, criticality is an issue during operations and in the store or repository. Safeguards and security concerns must be assessed and controlled. The process under development, by using a combination of tailored waste form chemistry combined with flexible process technology aims to develop a process line to handle a broad range of Pu waste streams. It aims to be capable of dealing with not only current arisings but those anticipated to arise as a result of future operations or policy changes. (authors)

  1. APPLICATION OF A GEOGRAPHIC INFORMATION SYSTEM FOR A CONTAINMENT SYSTEM LEAK DETECTION

    Science.gov (United States)

    The use of physical and hydraulic containment systems for the isolation of contaminated ground water associated with hazardous waste sites has increased during the last decade. Existing methodologies for monitoring and evaluating leakage from hazardous waste containment systems ...

  2. Position for determining gas-phase volatile organic compound concentrations in transuranic waste containers. Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Connolly, M.J.; Liekhus, K.J. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States). Idaho National Engineering and Environmental Lab.; Djordjevic, S.M.; Loehr, C.A.; Spangler, L.R. [Benchmark Environmental Corp. (United States)

    1998-06-01

    In the conditional no-migration determination (NMD) for the test phase of the Waste Isolation Pilot Plant (WIPP), the US Environmental Protection Agency (EPA) imposed certain conditions on the US Department of Energy (DOE) regarding gas phase volatile organic compound (VOC) concentrations in the void space of transuranic (TRU) waste containers. Specifically, the EPA required the DOE to ensure that each waste container has no layer of confinement that contains flammable mixtures of gases or mixtures of gases that could become flammable when mixed with air. The EPA also required that sampling of the headspace of waste containers outside inner layers of confinement be representative of the entire void space of the container. The EPA stated that all layers of confinement in a container would have to be sampled until DOE can demonstrate to the EPA that sampling of all layers is either unnecessary or can be safely reduced. A test program was conducted at the Idaho National Engineering and Environmental Laboratory (INEEL) to demonstrate that the gas phase VOC concentration in the void space of each layer of confinement in vented drums can be estimated from measured drum headspace using a theoretical transport model and that sampling of each layer of confinement is unnecessary. This report summarizes the studies performed in the INEEL test program and extends them for the purpose of developing a methodology for determining gas phase VOC concentrations in both vented and unvented TRU waste containers. The methodology specifies conditions under which waste drum headspace gases can be said to be representative of drum gases as a whole and describes a method for predicting drum concentrations in situations where the headspace concentration is not representative. The methodology addresses the approach for determining the drum VOC gas content for two purposes: operational period drum handling and operational period no-migration calculations.

  3. Position for determining gas phase volatile organic compound concentrations in transuranic waste containers. Revision 1

    International Nuclear Information System (INIS)

    Connolly, M.J.; Liekhus, K.J.; Djordjevic, S.M.; Loehr, C.A.; Spangler, L.R.

    1995-08-01

    In the conditional no-migration determination (NMD) for the test phase of the Waste Isolation Pilot Plant (WIPP), the US Environmental Protection Agency (EPA) imposed certain conditions on the US Department of Energy (DOE) regarding gas phase volatile organic compound (VOC) concentrations in the void space of transuranic (TRU) waste containers. Specifically, the EPA required the DOE to ensure that each waste container has no layer of confinement that contains flammable mixtures of gases or mixtures of gases that could become flammable when mixed with air. The EPA also required that sampling of the headspace of waste containers outside inner layers of confinement be representative of the entire void space of the container. The EPA stated that all layers of confinement in a container would have to be sampled until DOE can demonstrate to the EPA that sampling of all layers is either unnecessary or can be safely reduced. A test program was conducted at the Idaho National Engineering Laboratory (INEL) to demonstrate that the gas phase VOC concentration in the void space of each layer of confinement in vented drums can be estimated from measured drum headspace using a theoretical transport model and that sampling of each layer of confinement is unnecessary. This report summarizes the studies performed in the INEL test program and extends them for the purpose of developing a methodology for determining gas phase VOC concentrations in both vented and unvented TRU waste containers. The methodology specifies conditions under which waste drum headspace gases can be said to be representative of drum gases as a whole and describes a method for predicting drum concentrations in situations where the headspace concentration is not representative. The methodology addresses the approach for determining the drum VOC gas content for two purposes: operational period drum handling and operational period no-migration calculations

  4. Evaluation of doses during the handling and transport of radioactive wastes containers

    International Nuclear Information System (INIS)

    Kubik, I.; Kusovska, Z.; Hanusik, V.; Mrskova, A.; Kapisovsky, V.

    2000-01-01

    Radioactive waste products from the nuclear power plants (NPPs) must be isolated from contact with people for very long period of time. Low and intermediate-level waste will be disposed of in Slovakia in specially licensed Regional disposal facility which is located near the NPP Mochovce site. Radioactive waste accumulated in the Jaslovsk. Bohunice site, during the decommissioning process of the NPP A-1 and arising from the NPP V-1 and NPP V-2 operation, will be processed and shipped in standard concrete containers to the Mochovce Regional disposal facility. The treatment centre was build at the NPP Jaslovsk? Bohunice site which is in the trial operation now. It is supposed that radioactive waste containers will be transported by train from the treatment centre Jaslovsk? Bohunice to the site of Radioactive Waste Repository at Mochovce and by truck in the area of repository. To estimate the occupational radiation exposure during the transport the calculations of dose rates from the containers are necessary. The national regulations allow low level of radiation to emanate from the casks and containers. The maximum permissible volume radioactivity of wastes inside the container is limited in such a way that irradiation level should not exceed 2 mGy/h for the contact irradiation level and 0,1 mGy/h at 2-meter distance. MicroShield code was used to analyse shielding and assessing exposure from gamma radiation of containers to people. A radioactive source was conservatively modelled by homogenous mixture of radionuclides with concrete. Standard rectangular volume source and shield geometry is used in model calculations. The activities of the personnel during the transport and storage of containers are analysed and results of the evaluation of external dose rates and effective doses are described. (author)

  5. Position for determining gas-phase volatile organic compound concentrations in transuranic waste containers. Revision 2

    International Nuclear Information System (INIS)

    Connolly, M.J.; Liekhus, K.J.

    1998-06-01

    In the conditional no-migration determination (NMD) for the test phase of the Waste Isolation Pilot Plant (WIPP), the US Environmental Protection Agency (EPA) imposed certain conditions on the US Department of Energy (DOE) regarding gas phase volatile organic compound (VOC) concentrations in the void space of transuranic (TRU) waste containers. Specifically, the EPA required the DOE to ensure that each waste container has no layer of confinement that contains flammable mixtures of gases or mixtures of gases that could become flammable when mixed with air. The EPA also required that sampling of the headspace of waste containers outside inner layers of confinement be representative of the entire void space of the container. The EPA stated that all layers of confinement in a container would have to be sampled until DOE can demonstrate to the EPA that sampling of all layers is either unnecessary or can be safely reduced. A test program was conducted at the Idaho National Engineering and Environmental Laboratory (INEEL) to demonstrate that the gas phase VOC concentration in the void space of each layer of confinement in vented drums can be estimated from measured drum headspace using a theoretical transport model and that sampling of each layer of confinement is unnecessary. This report summarizes the studies performed in the INEEL test program and extends them for the purpose of developing a methodology for determining gas phase VOC concentrations in both vented and unvented TRU waste containers. The methodology specifies conditions under which waste drum headspace gases can be said to be representative of drum gases as a whole and describes a method for predicting drum concentrations in situations where the headspace concentration is not representative. The methodology addresses the approach for determining the drum VOC gas content for two purposes: operational period drum handling and operational period no-migration calculations

  6. Tunable lasers for waste management photochemistry applications

    International Nuclear Information System (INIS)

    Finch, F.T.

    1978-09-01

    A review of lasers with potential photochemical applications in waste management indicates that dye lasers, as a class, can provide tunable laser output through the visible and near-uv regions of the spectrum of most interest to photochemistry. Many variables can affect the performance of a specific dye laser, and the interactions of these variables, at the current state of the art, are complex. The recent literature on dye-laser characteristics has been reviewed and summarized, with emphasis on those parameters that most likely will affect the scaling of dye lasers in photochemical applications. Current costs are reviewed and correlated with output power. A new class of efficient uv lasers that appear to be scalable in both energy output and pulse rate, based on rare-gas halide excimers and similar molecules, is certain to find major applications in photochemistry. Because the most important developments are too recent to be adequately described in the literature or are the likely outcome of current experiments, the basic physics underlying the class of excimer lasers is described. Specific cost data are unavailable, but these new gas lasers should reflect costs similar to those of existing gas lasers, in particular, the pulsed CO 2 lasers. To complete the survey of tunable-laser characteristics, the technical characteristics of the various classes of lasers in the ir are summarized. Important developments in ir laser technology are being accelerated by isotope-separation research, but, initially at least, this portion of the spectrum is least likely to receive emphasis in waste-management-oriented photochemistry

  7. Ocean disposal option for bulk wastes containing naturally occurring radionuclides: an assessment case history

    International Nuclear Information System (INIS)

    Stull, E.A.; Merry-Libby, P.

    1985-01-01

    There are 180,000 m 3 of slightly contaminated radioactive wastes (36 pCi/g radium-226) currently stored at the US Department of Energy's Niagara Falls Storage Site (NFSS), near Lewiston, New York. These wastes resulted from the cleanup of soils that were contaminated above the guidelines for unrestricted use of property. An alternative to long-term management of these wastes on land is dispersal in the ocean. A scenario for ocean disposal is presented for excavation, transport, and emplacement of these wastes in an ocean disposal site. The potential fate of the wastes and impacts on the ocean environment are analyzed, and uncertainties in the development of two worst-case scenarios for dispersion and pathway analyses are discussed. Based on analysis of a worst-case pathway back to man, the incremental dose from ingesting fish containing naturally occurring radionuclides from ocean disposal of the NFSS wastes is insignificant. Ocean disposal of this type of waste appears to be a technically promising alternative to the long-term maintenance costs and eventual loss of containment associated with management in a near-surface land burial facility

  8. Report for slot cutter proof-of-principle test, Buried Waste Containment System project. Revision 1

    International Nuclear Information System (INIS)

    1998-01-01

    Several million cubic feet of hazardous and radioactive waste was buried in shallow pits and trenches within many US Department of Energy (US DOE) sites. The pits and trenches were constructed similarly to municipal landfills with both stacked and random dump waste forms such as barrels and boxes. Many of the hazardous materials in these waste sites are migrating into groundwater systems through plumes and leaching. On-site containment is one of the options being considered for prevention of waste migration. This report describes the results of a proof-of-principle test conducted to demonstrate technology for containing waste. This proof-of-principle test, conducted at the RAHCO International, Inc., facility in the summer of 1997, evaluated equipment techniques for cutting a horizontal slot beneath an existing waste site. The slot would theoretically be used by complementary equipment designed to place a cement barrier under the waste. The technology evaluated consisted of a slot cutting mechanism, muck handling system, thrust system, and instrumentation. Data were gathered and analyzed to evaluate the performance parameters

  9. Report for slot cutter proof-of-principle test, Buried Waste Containment System project. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-05-21

    Several million cubic feet of hazardous and radioactive waste was buried in shallow pits and trenches within many US Department of Energy (US DOE) sites. The pits and trenches were constructed similarly to municipal landfills with both stacked and random dump waste forms such as barrels and boxes. Many of the hazardous materials in these waste sites are migrating into groundwater systems through plumes and leaching. On-site containment is one of the options being considered for prevention of waste migration. This report describes the results of a proof-of-principle test conducted to demonstrate technology for containing waste. This proof-of-principle test, conducted at the RAHCO International, Inc., facility in the summer of 1997, evaluated equipment techniques for cutting a horizontal slot beneath an existing waste site. The slot would theoretically be used by complementary equipment designed to place a cement barrier under the waste. The technology evaluated consisted of a slot cutting mechanism, muck handling system, thrust system, and instrumentation. Data were gathered and analyzed to evaluate the performance parameters.

  10. Cost savings associated with landfilling wastes containing very low levels of uranium

    International Nuclear Information System (INIS)

    Boggs, C.J.; Shaddoan, W.T.

    1996-01-01

    The Paducah Gaseous Diffusion Plant (PGDP) has operated captive landfills (both residential and construction/demolition debris) in accordance with the Commonwealth of Kentucky regulations since the early 1980s. Typical waste streams allowed in these landfills include nonhazardous industrial and municipal solid waste (such as paper, plastic, cardboard, cafeteria waste, clothing, wood, asbestos, fly ash, metals, and construction debris). In July 1992, the U.S. Environmental Protection Agency issued new requirements for the disposal of sanitary wastes in a open-quotes contained landfill.close quotes These requirements were promulgated in the 401 Kentucky Administrative Record Chapters 47 and 48 that became effective 30 June 1995. The requirements for a new contained landfill include a synthetic liner made of high-density polyethylene in addition to the traditional 1-meter (3-foot) clay liner and a leachate collection system. A new landfill at Paducah would accept waste streams similar to those that have been accepted in the past. The permit for the previously existing landfills did not include radioactivity limits; instead, these levels were administratively controlled. Typically, if radioactivity was detected above background levels, the waste was classified as low-level waste (LLW), which would be sent off-site for disposal

  11. Selectivity of NF membrane for treatment of liquid waste containing uranium

    International Nuclear Information System (INIS)

    Oliveira, Elizabeth E.M.; Barbosa, Celina C.R.; Afonso, Julio C.

    2013-01-01

    The performance of two nanofiltration membranes were investigated for treatment of liquid waste containing uranium through two conditions permeation: permeation test and concentration test of the waste. In the permeation test solution permeated returned to the feed tank after collected samples each 3 hours. In the test of concentration the permeated was collected continuously until 90% reduction of the feed volume. The liquid waste ('carbonated water') was obtained during conversion of UF 6 to UO 2 in the cycle of nuclear fuel. This waste contains uranium concentration on average 7.0 mg L -1 , and not be eliminated to the environmental. The waste was permeated using a cross-flow membrane cell in the pressure of the 1.5 MPa. The selectivity of the membranes for separation of uranium was between 83% and 90% for both tests. In the concentration tests the waste was concentrated around for 5 times. The surface layer of the membranes was evaluated before and after the tests by infrared spectroscopy (ATR-FTIR), field emission microscopy (FESEM) and atomic force spectroscopy (AFM). The membrane separation process is a technique feasible to and very satisfactory for treatment the liquid waste. (author)

  12. Corrosion susceptibility of steel drums containing cemented intermediate level nuclear wastes

    Science.gov (United States)

    Duffó, Gustavo S.; Farina, Silvia B.; Schulz, Fátima M.; Marotta, Francesca

    2010-10-01

    Cementation processes are used as immobilization techniques for low or intermediate level radioactive waste for economical and safety reasons and for being a simple operation. In particular, ion-exchange resins commonly used for purification of radioactive liquid waste from nuclear reactors are immobilized before being stored to improve the leach resistance of the waste matrix and to maintain mechanical stability. Combustible solid radioactive waste can be incinerated and the resulting ashes can also be immobilized before storage. The immobilized resins and ashes are then contained in steel drums that may undergo corrosion depending on the presence of certain contaminants. The work described in this paper was aimed at evaluating the corrosion susceptibility of steel drums in contact with cemented ion-exchange resins and incineration ashes containing different concentrations of aggressive species (mostly chloride and sulphate ions). A special type of specimen was designed to simulate the cemented waste in the drum. The evolution of the corrosion potential and the corrosion current density of the steel, as well as the electrical resistivity of the matrix were monitored over a time period of 1 year. The results show the deleterious effect of chloride on the expected lifespan of the waste containers.

  13. Corrosion susceptibility of steel drums containing cemented intermediate level nuclear wastes

    International Nuclear Information System (INIS)

    Duffo, Gustavo S.; Farina, Silvia B.; Schulz, Fatima M.; Marotta, Francesca

    2010-01-01

    Cementation processes are used as immobilization techniques for low or intermediate level radioactive waste for economical and safety reasons and for being a simple operation. In particular, ion-exchange resins commonly used for purification of radioactive liquid waste from nuclear reactors are immobilized before being stored to improve the leach resistance of the waste matrix and to maintain mechanical stability. Combustible solid radioactive waste can be incinerated and the resulting ashes can also be immobilized before storage. The immobilized resins and ashes are then contained in steel drums that may undergo corrosion depending on the presence of certain contaminants. The work described in this paper was aimed at evaluating the corrosion susceptibility of steel drums in contact with cemented ion-exchange resins and incineration ashes containing different concentrations of aggressive species (mostly chloride and sulphate ions). A special type of specimen was designed to simulate the cemented waste in the drum. The evolution of the corrosion potential and the corrosion current density of the steel, as well as the electrical resistivity of the matrix were monitored over a time period of 1 year. The results show the deleterious effect of chloride on the expected lifespan of the waste containers.

  14. Mercury recovery from mercury-containing wastes using a vacuum thermal desorption system.

    Science.gov (United States)

    Lee, Woo Rim; Eom, Yujin; Lee, Tai Gyu

    2017-02-01

    Mercury (Hg)-containing waste from various industrial facilities is commonly treated by incineration or stabilization/solidification and retained in a landfill at a managed site. However, when highly concentrated Hg waste is treated using these methods, Hg is released into the atmosphere and soil environment. To eliminate these risks, Hg recovery technology using thermal treatment has been developed and commercialized to recover Hg from Hg-containing waste for safe disposal. Therefore, we developed Hg recovery equipment to treat Hg-containing waste under a vacuum of 6.67kPa (abs) at 400°C and recover the Hg. In addition, the dust generated from the waste was separated by controlling the temperature of the dust filtration unit to 230°C. Additionally, water and Hg vapors were condensed in a condensation unit. The Hg removal rate after waste treatment was 96.75%, and the Hg recovery rate as elemental Hg was 75.23%. Copyright © 2016 Elsevier Ltd. All rights reserved.

  15. IMPROVEMENTS IN CONTAINER MANAGEMENT OF TRANSURANIC (TRU) AND LOW LEVEL RADIOACTIVE WASTE STORED AT THE CENTRAL WASTE COMPLEX (CWC) AT HANFORD

    International Nuclear Information System (INIS)

    UYTIOCO EM

    2007-01-01

    The Central Waste Complex (CWC) is the interim storage facility for Resource Conservation and Recovery Act (RCRA) mixed waste, transuranic waste, transuranic mixed waste, low-level and low-level mixed radioactive waste at the Department of Energy's (DOE'S) Hanford Site. The majority of the waste stored at the facility is retrieved from the low-level burial grounds in the 200 West Area at the Site, with minor quantities of newly generated waste from on-site and off-site waste generators. The CWC comprises 18 storage buildings that house 13,000 containers. Each waste container within the facility is scanned into its location by building, module, tier and position and the information is stored in a site-wide database. As waste is retrieved from the burial grounds, a preliminary non-destructive assay is performed to determine if the waste is transuranic (TRU) or low-level waste (LLW) and subsequently shipped to the CWC. In general, the TRU and LLW waste containers are stored in separate locations within the CWC, but the final disposition of each waste container is not known upon receipt. The final disposition of each waste container is determined by the appropriate program as process knowledge is applied and characterization data becomes available. Waste containers are stored within the CWC based on their physical chemical and radiological hazards. Further segregation within each building is done by container size (55-gallon, 85-gallon, Standard Waste Box) and waste stream. Due to this waste storage scheme, assembling waste containers for shipment out of the CWC has been time consuming and labor intensive. Qualitatively, the ratio of containers moved to containers in the outgoing shipment has been excessively high, which correlates to additional worker exposure, shipment delays, and operational inefficiencies. These inefficiencies impacted the LLW Program's ability to meet commitments established by the Tri-Party Agreement, an agreement between the State of Washington

  16. Defense Waste Processing Facility (DWPF) Durability-Composition Models and the Applicability of the Associated Reduction of Constraints (ROC) Criteria for High TiO2 Containing Glasses

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Trivelpiece, C. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-08-30

    Radioactive high-level waste (HLW) at the Savannah River Site (SRS) has successfully been vitrified into borosilicate glass in the DWPF since 1996. Vitrification requires stringent product/process (P/P) constraints since the glass cannot be reworked once it has been poured into ten foot tall by two foot diameter canisters. A unique “feed forward” statistical process control (SPC) was developed for this control rather than relying on statistical quality control (SQC). In SPC, the feed composition to the DWPF melter is controlled prior to vitrification. In SQC, the glass product would be sampled after it is vitrified. Individual glass property-composition models form the basis for the “feed forward” SPC. The models transform constraints on the melt and glass properties into constraints on the feed composition going to the melter in order to determine, at the 95% confidence level, that the feed will be processable and that the durability of the resulting waste form will be acceptable to a geologic repository. The DWPF SPC system is known as the Product Composition Control System (PCCS). One of the process models within PCCS is known as the Thermodynamic Hydration Energy Reaction MOdel (THERMO™). The DWPF will soon be receiving increased concentrations of TiO2-, Na2O-, and Cs2O-enriched wastes from the Salt Waste Processing Facility (SWPF). The SWPF has been built to pretreat the high-curie fraction of the salt waste to be removed from the HLW tanks in the F- and H-Area Tank Farms at the SRS. In order to validate the existing TiO2 term in THERMO™ beyond 2.0 wt% in the DWPF, new durability data were developed over the target range of 2.00 to 6.00 wt% TiO2 and evaluated against the 1995 durability model. The durability was measured by the 7-day Product Consistency Test. This study documents the adequacy of the existing THERMO™ terms. It is recommended that the modified THERMO™ durability models and

  17. Investigations on cement/polymer Waste packages containing intermediate level waste and organic exchange resins

    Energy Technology Data Exchange (ETDEWEB)

    ELsourougy, M R; Zaki, A A; Aly, H F [Atomic energy authority, hot laboratory center, Cairo, (Egypt); Khalil, M Y [Nuclear engineering department, Alexandria university. Alexandria, (Egypt)

    1995-10-01

    Polymers can be added to cements to improve its nuclear waste immobilization properties. This trend in cementation processes is attracting attention and requiring through investigations. In this work, polymers of different kinds were added to ordinary portland cement for the purpose of solidifying intermediate level liquid wastes and organic ion exchange resins. Epoxy polymer such as Kemapoxy-150 reduced the leaching rate of cesium compared to cement alone. Latex to cement ratio less than 4% caused an increase in leaching rate of cesium. When cesium was absorbed to an organic resin its leachability was improved. 5 figs., 4 tabs.

  18. Investigations on cement/polymer Waste packages containing intermediate level waste and organic exchange resins

    International Nuclear Information System (INIS)

    ELsourougy, M.R.; Zaki, A.A.; Aly, H.F.; Khalil, M.Y.

    1995-01-01

    Polymers can be added to cements to improve its nuclear waste immobilization properties. This trend in cementation processes is attracting attention and requiring through investigations. In this work, polymers of different kinds were added to ordinary portland cement for the purpose of solidifying intermediate level liquid wastes and organic ion exchange resins. Epoxy polymer such as Kemapoxy-150 reduced the leaching rate of cesium compared to cement alone. Latex to cement ratio less than 4% caused an increase in leaching rate of cesium. When cesium was absorbed to an organic resin its leachability was improved. 5 figs., 4 tabs

  19. Fabrication development for high-level nuclear waste containers for the tuff repository

    International Nuclear Information System (INIS)

    Domian, H.A.; Holbrook, R.L.; LaCount, D.F.; Babcock and Wilcox Co., Alliance, OH

    1990-09-01

    This final report completes Phase 1 of an engineering study of potential manufacturing processes for the fabrication of containers for the long-term storage of nuclear waste. An extensive literature and industry review was conducted to identify and characterize various processes. A technical specification was prepared using the American Society of Mechanical Engineers Boiler ampersand Pressure Vessel Code (ASME BPVC) to develop the requirements. A complex weighting and evaluation system was devised as a preliminary method to assess the processes. The system takes into account the likelihood and severity of each possible failure mechanism in service and the effects of various processes on the microstructural features. It is concluded that an integral, seamless lower unit of the container made by back extrusion has potential performance advantages but is also very high in cost. A welded construction offers lower cost and may be adequate for the application. Recommendations are made for the processes to be further evaluated in the next phase when mock-up trials will be conducted to address key concerns with various processes and materials before selecting a primary manufacturing process. 43 refs., 26 figs., 34 tabs

  20. Selection criteria for container materials at the proposed Yucca Mountain high level nuclear waste repository

    International Nuclear Information System (INIS)

    Halsey, W.G.

    1989-11-01

    A geological repository has been proposed for the permanent disposal of the nation's high level nuclear waste at Yucca Mountain in the Nevada desert. The containers for this waste must remain intact for the unprecedented service lifetime of 1000 years. A combination of engineering, regulatory, and licensing requirements complicate the container material selection. In parallel to gathering information regarding the Yucca Mountain service environment and material performance data, a set of selection criteria have been established which compare candidate materials to the performance requirements, and allow a quantitative comparison of candidates. These criteria assign relative weighting to varied topic areas such as mechanical properties, corrosion resistance, fabricability, and cost. Considering the long service life of the waste containers, it is not surprising that the corrosion behavior of the material is a dominant factor. 7 refs

  1. Buckling design criteria for waste package disposal containers in mined salt repositories: Technical report

    International Nuclear Information System (INIS)

    Mallett, R.H.

    1986-12-01

    This report documents analytical and experimental results from a survey of the technical literature on buckling of thick-walled cylinders under external pressure. Based upon these results, a load factor is suggested for the design of waste package containers for disposal of high-level radioactive waste in repositories mined in salt formations. The load factor is defined as a ratio of buckling pressure to allowable pressure. Specifically, a load factor which ranges from 1.5 for plastic buckling to 3.0 for elastic buckling is included in a set of proposed buckling design criteria for waste disposal containers. Formulas are given for buckling design under axisymmetric conditions. Guidelines are given for detailed inelastic buckling analyses which are generally required for design of disposal containers

  2. Feasibility study using hypothesis testing to demonstrate containment of radionuclides within waste packages

    International Nuclear Information System (INIS)

    Thomas, R.E.

    1986-04-01

    The purpose of this report is to apply methods of statistical hypothesis testing to demonstrate the performance of containers of radioactive waste. The approach involves modeling the failure times of waste containers using Weibull distributions, making strong assumptions about the parameters. A specific objective is to apply methods of statistical hypothesis testing to determine the number of container tests that must be performed in order to control the probability of arriving at the wrong conclusions. An algorithm to determine the required number of containers to be tested with the acceptable number of failures is derived as a function of the distribution parameters, stated probabilities, and the desired waste containment life. Using a set of reference values for the input parameters, sample sizes of containers to be tested are calculated for demonstration purposes. These sample sizes are found to be excessively large, indicating that this hypothesis-testing framework does not provide a feasible approach for demonstrating satisfactory performance of waste packages for exceptionally long time periods

  3. PUREX Storage Tunnels dangerous waste permit application

    International Nuclear Information System (INIS)

    1991-12-01

    The PUREX Storage Tunnels are a mixed waste storage unit consisting of two underground railroad tunnels: Tunnel Number 1 designated 218-E-14 and Tunnel Number 2 designated 218-E-15. The two tunnels are connected by rail to the PUREX Plant and combine to provide storage space for 48 railroad cars (railcars). The PUREX Storage Tunnels provide a long-term storage location for equipment removed from the PUREX Plant. Transfers into the PUREX Storage Tunnels are made on an as-needed basis. Radioactively contaminated equipment is loaded on railcars and remotely transferred by rail into the PUREX Storage Tunnels. Railcars act as both a transport means and a storage platform for equipment placed into the tunnels. This report consists of part A and part B. Part A reports on amounts and locations of the mixed water. Part B permit application consists of the following: Facility Description and General Provisions; Waste Characteristics; Process Information; Groundwater Monitoring; Procedures to Prevent Hazards; Contingency Plan; Personnel Training; Exposure Information Report

  4. Effect of ingredients in waste water on property of ion exchange resin for uranium-contained waste water treatment

    International Nuclear Information System (INIS)

    Ren Junshu; Mu Tao; Zhang Wei; Yang Shengya

    2008-01-01

    The effect of ingredients in waste water on the property of ion exchange resin for uranium-contained waste water treatment was studied by the method of static ad- sorption combined with dynamic experiment. The experimental result shows that the efficiency or breackthrough volume of resin is reduced if there are other general anions, triethanolamine and oil in the solution. When the concentrations of CO 3 2- , HCO 3 - , SO 3 2- , Cl - in the solution are more than 0.24, 0.28, 0.23 and 0.09 mol/L, respectively, the concentrations of uranium in the outlet waste water will exceed 20 μg/L. The maximal allowable concentration of triethanolamine through the resin is no more than 250 mg/L. When the content of oil in the resin exceeds 1%(by quality), the breackthrough volume reduces by 16%, and when it exceeds 11%, the breackthrough volume almost loses at all. (authors)

  5. Pretreatment method for radioactive iodine-containing liquid wastes and pretreatment device

    International Nuclear Information System (INIS)

    Wakaida, Yasuo.

    1996-01-01

    Heretofore, radioactive iodine-containing liquid wastes have been discharged directly to a storing and decaying storage vessel to conduct a water draining treatment. In the present invention, the radioactive iodine-containing liquid wastes to be discharged are not discharged to the storage vessel directly but injected to a filling tank, as a pretreatment, to distinguish whether proteins are mixed in the liquid wastes or not. When proteins are mixed, miscellaneous materials such as proteins are recovered and removed by a protein processing system. When proteins are not mixed, radioactive iodine is recovered and removed directly by an iodine processing system. With such procedures, water draining treatment in the storing and decaying storage vessel is mitigated, and even when the amount of the radioactive iodine-containing liquid wastes is increased, the existent maintaining and decaying storage vessel can be used as it is. Accordingly, a safe water draining treatment with good efficiency can be conducted relative to radioactive iodine-containing liquid wastes at a reduced cost. (T.M.)

  6. Resource Conservation and Recovery Act (RCRA) Part B permit application for container storage units at the Oak Ridge Y-12 Plant

    Energy Technology Data Exchange (ETDEWEB)

    1994-08-01

    This document contains Part B of the Permit Application for Container Storage Units at the Oak Ridge Y-12 Plant. Sections cover the following areas: Facility description; Waste characteristics; Process information; Ground water monitoring; Procedures to prevent hazards; Contingency plan; Personnel training; Closure plan, post closure plan, and financial requirements; Recordkeeping; Other federal laws; Organic air emissions; Solid waste management units; and Certification.

  7. Resource Conservation and Recovery Act (RCRA) Part B permit application for container storage units at the Oak Ridge Y-12 Plant

    International Nuclear Information System (INIS)

    1994-08-01

    This document contains Part B of the Permit Application for Container Storage Units at the Oak Ridge Y-12 Plant. Sections cover the following areas: Facility description; Waste characteristics; Process information; Ground water monitoring; Procedures to prevent hazards; Contingency plan; Personnel training; Closure plan, post closure plan, and financial requirements; Recordkeeping; Other federal laws; Organic air emissions; Solid waste management units; and Certification

  8. Demonstration of close-coupled barriers for subsurface containment of buried waste. Conceptual test plan

    Energy Technology Data Exchange (ETDEWEB)

    Heiser, J. [Brookhaven National Laboratory, Upton, NY (United States); Dwyer, B. [Sandia National Laboratory, Albuquerque, NM (United States)

    1995-07-01

    Over the past five decades, the US Department of Energy (DOE) Complex sites have experienced numerous loss of confinement failures from underground storage tanks (USTs), piping systems, vaults, landfills, and other structures containing hazardous and mixed wastes. Consequently, efforts are being made to devise technologies that provide interim containment of waste sites while final remediation alternatives are developed. Barrier materials consisting of cement and polymer which will be emplaced beneath a 7500 liter tank. The stresses around the tank shall be evaluated during barrier construction.

  9. Fabrication and closure development of nuclear waste disposal containers for the Yucca Mountain Project: Status report

    International Nuclear Information System (INIS)

    Domian, H.A.; Robitz, E.S.; Conrardy, C.C.; LaCount, D.F.; McAninch, M.D.; Fish, R.L.; Russell, E.W.

    1991-09-01

    In GFY 89, a project was underway to determine and demonstrate a suitable method for fabricating thin-walled monolithic waste containers for service within the potential repository at Yucca Mountain. A concurrent project was underway to determine and demonstrate a suitable closure process for these containers after they have been filled with high-level nuclear waste. Phase 1 for both the fabrication and closure projects was a screening phase in which candidate processes were selected for further laboratory testing in Phase 2. This report describes the final results of the Phase 1 efforts. It also describes the preliminary results of Phase 2 efforts

  10. Precipitation and Deposition of Aluminum-Containing Species in Tank Wastes

    International Nuclear Information System (INIS)

    Mattigod, Shas V.; Hobbs, David T.; Wang, Li-Qiong; Dabbs, Daniel M.; Aksay, Ilhan A.

    2002-01-01

    Aluminum-containing phases represent the most prevalent solids that can appear or disappear during the processing of radioactive tank wastes. Processes such as sludge washing and leaching are designed to dissolve Al-containing phases and thereby minimize the volume of high-level waste glass required to encapsulate radioactive sludges. Unfortunately, waste-processing steps that include evaporation can involve solutions that are supersaturated with respect to cementitious aluminosilicates that result in unwanted precipitation and scale formation. Of all the constituents of tank waste, limited solubility cementitious aluminosilicates have the greatest potential for clogging pipes and transfer lines, fouling highly radioactive components such as ion exchangers, and completely shutting down processing operations. For instance, deposit buildup and clogged drain lines experienced during the tank waste volume-reduction process at the Savannah River Site (SRS) required an evaporator to be shut down in October 1999. The Waste Processing Technology Section of Westinghouse Savannah River Company at SRS now is collaborating with team members from Pacific Northwest National Laboratory (PNNL) to verify the thermodynamic stability of aluminosilicate compounds under waste tank conditions in an attempt to solve the deposition and clogging problems. The primary objectives of this study are (1) to understand the major factors controlling precipitation, heterogeneous nucleation, and growth phenomena of relatively insoluble aluminosilicates; (2) to determine the role of organics for inhibiting aluminosilicate formation, and (3) to develop a predictive tool to control precipitation, scale formation, and cementation under tank waste processing conditions. The results of this work will provide crucial information for (1) avoiding problematical sludge processing steps and (2) identifying and developing effective technologies to process retrieved sludges and supernatants before ultimate

  11. Decontamination flowsheet development for a waste oil containing mixed radioactive contaminants

    International Nuclear Information System (INIS)

    Vijayan, S.; Buckley, L.P.

    1993-01-01

    The majority of waste oils contaminated with both radioactive and hazardous components are generated in nuclear power plant, research lab. and uranium-refinery operations. The waste oils are complex, requiring a detailed examination of the waste management strategies and technology options. It may appear that incineration offers a total solution, but this may not be true in all cases. An alternative approach is to decontaminate the waste oils to very low contaminant levels, so that the treated oils can be reused, burned as fuel in boilers, or disposed of by commercial incineration. This paper presents selected experimental data and evaluation results gathered during the development of a decontamination flowsheet for a specific waste oil stores at Chalk River Labs. (CRL). The waste oil contains varying amounts of lube oils, grease, paint, water, particulates, sludge, light chloro- and fluoro-solvents, polychlorinated biphenyls (PCB), complexing chemicals, uranium, chromium, iron, arsenic and manganese. To achieve safe management of this radioactive and hazardous waste, several treatment and disposal methods were screened. Key experiments were performed at the laboratory-scale to confirm and select the most appropriate waste-management scheme based on technical, environmental and economic criteria. The waste-oil-decontamination flowsheet uses a combination of unit operations, including prefiltration, acid scrubbing, and aqueous-leachage treatment by precipitation, microfiltration, filter pressing and carbon adsorption. The decontaminated oil containing open-quotes de minimisclose quotes levels of contaminants will undergo chemical destruction of PCBs and final disposal by incineration. The recovered uranium will be recycled to a uranium milling process

  12. Chemically vapor deposited coatings for multibarrier containment of nuclear wastes

    International Nuclear Information System (INIS)

    Rusin, J.M.; Shade, J.W.; Kidd, R.W.; Browning, M.F.

    1981-01-01

    Chemical vapor deposition (CVD) was selected as a feasible method to coat ceramic cores, since the technology has previously been demonstrated for high-temperature gas-cooled reactor (HTGR) fuel particles. CVD coatings, including SiC, PyC (pyrolytic carbon), SiO 2 , and Al 2 O 3 were studied. This paper will discuss the development and characterization of PyC and Al 2 O 3 CVD coatings on supercalcine cores. Coatings were applied to 2 mm particles in either fluidized or vibrating beds. The PyC coating was deposited in a fluidized bed with ZrO 2 diluent from C 2 H 2 at temperatures between 1100 and 1200 0 C. The Al 2 O 3 coatings were deposited in a vibrated bed by a two-stage process to minimize loss of PyC during the overcoating operation. This process involved applying 10 μm of Al 2 O 3 using water vapor hydrolysis of AlCl 3 and then switching to the more surface-controlled hydrolysis via the H 2 + CO 2 reaction (3CO 2 + 3H 2 + 2AlCl 3 = Al 2 O 3 + 6HCl + 3CO). Typically, 50 to 80 μm Al 2 O 3 coatings were applied over 30 to 40 μm PyC coatings. The coatings were evaluated by metallographic examination, PyC oxidation tests, and leach resistance. After air oxidation for 100 hours at 750 0 C, the duplex PyC/Al 2 O 3 coated particles exhibited a weight loss of 0.01 percent. Leach resistance is being determined for temperatures from 50 to 150 0 C in various solutions. Typical results are given for selected ions. The leach resistance of supercalcine cores is significantly improved by the application of PyC and/or Al 2 O 3 coatings

  13. Clearance measurement for waste concerning contained radioactivity; Frei(gabe)messung von Abfall hinsichtlich enthaltener Radioaktivitaet

    Energy Technology Data Exchange (ETDEWEB)

    Sokcic-Kostic, Marina; Schultheis, Roland [NUKEM Technologies Engineering Services GmbH, Alzenau (Germany)

    2016-12-15

    Clearance measurements are always a compromise between requirements of the measurement technology and economic boundary conditions. Depending on the quantity and the type of waste, different solutions are obtained. For large volumes of more or less homogeneous waste, the conveyor belt method is the biggest favorite, which has already proved its suitability in practice. This is important, because numerous nuclear power stations are being decommissioned in Germany in the coming years and large quantities of waste will be arising. For some applications, e.g. Tritium or C-14, satisfying solutions either do not exist or are currently in the development stages. There is still great potential for the development of clearance methods.

  14. Design report for the interim waste containment facility at the Niagara Falls Storage Site

    International Nuclear Information System (INIS)

    1986-05-01

    Low-level radioactive residues from pitchblende processing and thorium- and radium-contaminated sand, soil, and building rubble are presently stored at the Niagara Falls Storage Site (NFSS) in Lewiston, New York. These residues and wastes derive from past NFSS operations and from similar operations at other sites in the United States conducted during the 1940s by the Manhattan Engineer District (MED) and subsequently by the Atomic Energy Commission (AEC). The US Department of Energy (DOE), successor to MED/AEC, is conducting remedial action at the NFSS under two programs: on-site work under the Surplus Facilities Managemnt Program and off-site cleanup of vicinity properties under the Formerly Utilized Sites Remedial Action Program. On-site remedial action consists of consolidating the residues and wastes within a designated waste containment area and constructing a waste containment facility to prevent contaminant migration. The service life of the system is 25 to 50 years. Near-term remedial action construction activities will not jeopardize or preclude implementation of any other remedial action alternative at a later date. Should DOE decide to extend the service life of the system, the waste containment area would be upgraded to provide a minimum service life of 200 years. This report describes the design for the containment system. Pertinent information on site geology and hydrology and on regional seismicity and meteorology is also provided. Engineering calculations and validated computer modeling studies based on site-specific and conservative parameters confirm the adequacy of the design for its intended purposes of waste containment and environmental protection

  15. Method of treating the waste liquid of a washing containing a radioactive substance

    International Nuclear Information System (INIS)

    Sawaguchi, Yusuke; Tsuyuki, Takashi; Kaneko, Masato; Sato, Yasuhiko; Yamaguchi, Takashi.

    1975-01-01

    Object: To separate waste liquid resulting from washing and which contains a radioactive substance and surface active agent into high purity water and a solid waste substance containing a small quantity of surface active agent. Structure: To waste liquid from a waste liquid tank is added a pH adjusting agent for adjusting the pH to 5.5, and the resultant liquid is sent to an agglomeration reaction tank, in which an inorganic agglomerating agent is added to the waste liquid to cause a major proportion of the radioactive substance and surface active agent to form flocks produced through agglomeration. Then, the waste liquid is sent from the agglomeration reaction tank to a froth separation tank, to which air is supplied through a perforated plate to cause frothing. The over-flowing liquid is de-frothed, and then the insoluble matter is separated as sludge, followed by hydroextraction and drying for solidification. The treated liquid extracted from a froth separation tank is sent to an agglomerating agent recovery tank for separation of the agglomeration agent, and then the residual surface active agent is removed by adsorption in an active carbon adsorption tower, followed by concentration by evaporation in an evaporating can. The concentrated liquid is extracted and then solidified with cement or asphalt. (Kamimura, M.)

  16. Waste water pilot plant research, development, and demonstration permit application

    International Nuclear Information System (INIS)

    1993-03-01

    This permit application has been prepared to obtain a research, development, and demonstration permit to perform pilot-scale treatability testing on the 242-A Evaporator process condensate waste water effluent stream. It provides the management framework, and controls all the testing conducted in the waste water pilot plant using dangerous waste. It also provides a waste acceptance envelope (upper limits for selected constituents) and details the safety and environmental protection requirements for waste water pilot plant testing. This permit application describes the overall approach to testing and the various components or requirements that are common to all tests. This permit application has been prepared at a sufficient level of detail to establish permit conditions for all waste water pilot plant tests to be conducted

  17. Repeated application of organic waste affects soil organic matter composition

    DEFF Research Database (Denmark)

    Peltre, Clément; Gregorich, Edward G.; Bruun, Sander

    2017-01-01

    Land application of organic waste is an important alternative to landfilling and incineration because it helps restore soil fertility and has environmental and agronomic benefits. These benefits may be related to the biochemical composition of the waste, which can result in the accumulation...... of different types of carbon compounds in soil. The objective of this study was to identify and characterise changes in soil organic matter (SOM) composition after repeated applications of organic waste. Soil from the CRUCIAL field experiment in Denmark was sampled after 12 years of annual application...... that there was accumulation in soil of different C compounds for the different types of applied organic waste, which appeared to be related to the degree to which microbial activity was stimulated and the type of microbial communities applied with the wastes or associated with the decomposition of applied wastes...

  18. Application bar-code system for solid radioactive waste management

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Y. H.; Kim, T. K.; Kang, I. S.; Cho, H. S.; Son, J. S. [KAERI, Taejon (Korea, Republic of)

    2004-07-01

    Solid radioactive wastes are generated from the post-irradiated fuel examination facility, the irradiated material examination facility, the research reactor, and the laboratories at KAERI. A bar-code system for a solid radioactive waste management of a research organization became necessary while developing the RAWMIS(Radioactive Waste Management Integration System) which it can generate personal history management for efficient management of a waste, documents, all kinds of statistics. This paper introduces an input and output application program design to do to database with data in the results and a stream process of a treatment that analyzed the waste occurrence present situation and data by bar-code system.

  19. Electrochemical probing of high-level radioactive waste tanks containing washed sludge and precipitates

    International Nuclear Information System (INIS)

    Bickford, D.F.; Congdon, J.W.; Oblath, S.B.

    1986-12-01

    At the US Department of Energy's Savannah River Plant, corrosion of carbon steel storage tanks containing alkaline, high-level radioactive waste is controlled by specification of limits on waste composition and temperature. Processes for the preparation of waste for final disposal will result in waste with low corrosion inhibitor concentrations and, in some cases, high aromatic organic concentrations, neither of which are characteristic of previous operations. Laboratory tests, conducted to determine minimum corrosion inhibitor levels indicated pitting of carbon steel near the waterline for proposed storage conditions. In situ electrochemical measurements of full-scale radioactive process demonstrations have been conducted to assess the validity of laboratory tests. Probes included pH, Eh (potential relative to a standard hydrogen electrode), tank potential, and alloy coupons. In situ results are compared to those of the laboratory tests, with particular regard given to simulated solution composition. Transition metal hydroxide sludge contains strong passivating species for carbon steel. Washed precipitate contains organic species that lower solution pH and tend to reduce passivating films, requiring higher inhibitor concentrations than the 0.01 molar nitrite required for reactor fuel reprocessing wastes. Periodic agitation, to keep the organic phase suspended, or cathodic protection are possible alternatives to higher nitrite inhibitor concentrations

  20. 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.

  1. Control of stress corrosion cracking in storage tanks containing radioactive waste

    International Nuclear Information System (INIS)

    Ondrejcin, R.S.; Rideout, S.P.; Donovan, J.A.

    1978-01-01

    Stress corrosion of carbon steel storage tanks containing alkaline nitrate radioactive waste, at the Savannah River Plant is controlled by specification of limits on waste composition and temperature. Cases of cracking have been observed in the primary steel shell of tanks designed and built before 1960 that were attributed to a combination of high residual stresses from fabrication welding and aggressiveness of fresh wastes from the reactor fuel reprocessing plants. The fresh wastes have the highest concentration of nitrate, which has been shown to be the cracking agent. Also as the waste solutions age and are reduced in volume by evaporation of water, nitrite and hydroxide ions become more concentrated and inhibit stress corrosion. Thus, by providing a heel of aged evaporated waste in tanks that receive fresh waste, concentrations of the inhibitor ions are maintained within specified ranges to protect against nitrate cracking. Tanks designed and built since 1960 have been made of steels with greater resistance to stress corrosion; these tanks have also been heat treated after fabrication to relieve residual stresses from construction operations. Temperature limits are also specified to protect against stress corrosion at elevated temperatures

  2. Electrochemical probing of high-level radioactive waste tanks containing washed sludge and precipitates

    International Nuclear Information System (INIS)

    Bickford, D.F.; Congdon, J.W.; Oblath, S.B.

    1987-01-01

    At the U.S. Department of Energy's Savannah River Plant, corrosion of carbon steel storage tanks containing alkaline, high-level radioactive waste is controlled by specification of limits on waste composition and temperature. Processes for the preparation of waste for final disposal will result in waste with low corrosion inhibitor concentrations and, in some cases, high aromatic organic concentrations, neither of which are characteristic of previous operations. Laboratory tests, conducted to determine minimum corrosion inhibitor levels indicated pitting of carbon steel near the waterline for proposed storage conditions. In situ electrochemical measurements of full-scale radioactive process demonstrations have been conducted to assess the validity of laboratory tests. Probes included pH, Eh (potential relative to a standard hydrogen electrode), tank potential, and alloy coupons. In situ results are compared to those of the laboratory tests, with particular regard given to simulated solution composition

  3. The TN-GEMINI: experience on a versatile alpha waste transport container

    International Nuclear Information System (INIS)

    Roland, V.; Chanzy, Y.

    2001-01-01

    The present paper discusses experience gained in moving alpha wastes and its teachings regarding transport aspects of D and D. Alpha wastes are generated in fuel cycle facilities such as those involved in reprocessing, in manufacture of mixed oxide fuel, and by research laboratories. If a significant amount of wastes has to be transported, then a Type B packaging is required. Developed by Transnucleaire and COGEMA, the TN GEMINI container enables nuclear facilities operators to optimise their alpha waste transport management, and more generally contribute to their D and D projects. After describing succinctly the design of the TN GEMINI, the paper will explain how the packaging is being operated. Teachings from experience will be shared. (orig.)

  4. Long time storage containers for spent fuels and vitrified wastes: synthesis of the studies

    International Nuclear Information System (INIS)

    Beziat, A.

    2004-01-01

    This report presents a synthesis of the studies relatives to the containers devoted to the long time spent fuels storage and vitrified wastes packages. These studies were realized in the framework of the axis 3 of the law of 1991 on the radioactive wastes management. The first part is devoted to the presentation of the studies. The container sizing studies which constitute the first containment barrier are then presented. The material choice and the closed system are also detailed. The studies were validate by the realization of containers models and an associated demonstration program is proposed. A synthesis of the technical and economical studies allowed to determine the components and operation costs. (A.L.B.)

  5. Application of the iron-enriched basalt waste form for immobilizing commercial transuranic waste

    International Nuclear Information System (INIS)

    Owen, D.E.

    1981-08-01

    The principal sources of commercial transuranic (TRU) waste in the United States are identified. The physical and chemical nature of the wastes from these sources are discussed. The fabrication technique and properties of iron-enriched basalt, a rock-like waste form developed for immobilizing defense TRU wastes, are discussed. The application of iron-enriched basalt to commercial TRU wastes is discussed. Review of commercial TRU wastes from mixed-oxide fuel fabrication, light water reactor fuel reprocessing, and miscellaneous medical, research, and industrial sources, indicates that iron-enriched basalt is suitable for most types of commercial TRU wastes. Noncombustible TRU wastes are dissolved in the high temperature, oxidizing iron-enriched basalt melt. Combustible TRU wastes are immobilized in iron-enriched basalt by incinerating the wastes and adding the TRU-bearing ash to the melt. Casting and controlled cooling of the melt produces a devitrified, rock-like iron-enriched basalt monolith. Recommendations are given for testing the applicability of iron-enriched basalt to commercial TRU wastes

  6. A process for containment removal and waste volume reduction to remediate groundwater containing certain radionuclides, toxic metals and organics

    International Nuclear Information System (INIS)

    Buckley, L.P.; Killey, D.R.W.; Vijayan, S.; Wong, P.C.F.

    1992-09-01

    A project to remove groundwater contaminants by an improved treatment process was performed during 1990 October--1992 March by Atomic Energy of Canada Limited for the United States Department of Energy, managed by Argonne National Laboratory. The goal was to generate high-quality effluent while minimizing secondary waste volume. Two effluent target levels, within an order of magnitude, or less than the US Drinking Water Limit, were set to judge the process effectiveness. The program employed mixed waste feeds containing cadmium, uranium, lead, iron, calcium, strontium-85-90, cesium-137, benzene and trichlorethylene in simulated and actual groundwater and soil leachate solutions. A combination of process steps consisting of sequential chemical conditioning, cross-flow microfiltration and dewatering by low temperature-evaporation, or filter pressing were effective for the treatment of mixed waste having diverse physico-chemical properties. A simplified single-stage version of the process was implemented to treat ground and surface waters contaminated with strontium-90 at the Chalk River Laboratories site. Effluent targets and project goals were met successfully

  7. Application of containment codes to LMFBRs in the United States

    International Nuclear Information System (INIS)

    Chang, Y.W.

    1977-01-01

    This paper describes the application of containment codes to predict the response of the fast reactor containment and the primary piping loops to HCDAs. Five sample problems are given to illustrate their applications. The first problem deals with the response of the primary containment to an HCDA. The second problem deals with the coolant flow in the reactor lower plenum. The third problem concerns sodium spillage and slug impact. The fourth problem deals with the response of a piping loop. The fifth problem analyzes the response of a reactor head closure. Application of codes in parametric studies and comparison of code predictions with experiments are also discussed. (Auth.)

  8. Application of containment codes to LMFBRs in the United States

    International Nuclear Information System (INIS)

    Chang, Y.W.

    1977-01-01

    The application of containment codes to predict the response of the fast reactor containment and the primary piping loops to HCDAs is described. Five sample problems are given to illustrate their applications. The first problem deals with the response of the primary containment to an HCDA. The second problem deals with the coolant flow in the reactor lower plenum. The third proem concerns sodium spillage and slug impact. The fourth problem deals with the response of a piping loop. The fifth problem analyzes the response of a reactor head closure. Application of codes in parametric studies and comparison of code predictions with experiments are also discussed

  9. Application of concrete to the treatment and disposal of radioactive waste in Japan

    International Nuclear Information System (INIS)

    Maki, Yasuro; Ohnuma, Hiroshi

    1992-01-01

    The paper presents the present state of application of concrete to treatment, storage and disposal of low level radioactive waste in Japan. In the 2nd section, the electric power supply and the kinds and volumes of radioactive waste from nuclear power plants in Japan are described. In the 3rd section, the applications of concrete to the treatment of radioactive waste are described. These are solidification with cement and containers made by various mortars and concretes. The application of concrete to disposal structures are presented in the 4th section; these are research on the durabity of concrete under disposal site condition, research on the filling the concrete pit with 200 l drum packed cement solidified wastes by prepacked concreting methods, and so on. And this section describes also the outlines of the low level radioactive disposal system at the Rokkasho site. (orig./DG)

  10. Container material and design considerations for storage of low-level radioactive waste

    International Nuclear Information System (INIS)

    Temus, C.J.

    1987-01-01

    With the threat of increased burial site restrictions and increased surcharges; the ease with which waste is sent to the burial site has been reduced. For many generators of waste the only alternative after maximizing volume reduction efforts is to store the waste. Even after working through the difficult decision of deciding what type of storage facility to have, the decision of what type of container to store the waste in has to still be made. This paper explores the many parameters that affect not only the material selection but also the design. The proper selection of materials affect the ability of the container to survive the storage period. The material selection also directly affects the design and utilization of the storage facility. The impacts to the facility include the functional aspects as well as its operational cost and liability as related to such things as fire insurance and active environmental control systems. The advantages and disadvantages of many of the common systems such as carbon steel, various coatings, polyethylene, stainless steel, composites and concrete will be discussed and evaluated. Recognizing that the waste is to be disposed of in the future differentiates it from waste that is shipped directly to the disposal site. The stored waste has to have the capability to be handled not only once like the disposal site waste but potentially several times before ultimate disposal. This handling may be by several different systems both at the storage facility and the burial site. Some of these systems due to ALARA considerations are usually remote requiring various interfaces, while not interfering with handling, transportation or disposal operations

  11. 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.

  12. Method of conversion of caustic liquid radioactive wastes containing sodium nitrates into solid insoluble products

    International Nuclear Information System (INIS)

    Barney, G.S.; Brownell, L.E.

    1975-01-01

    A proposal is made to convert caustic, liquid, radioactive wastes containing sodium nitrate into a solid product by reaction with powdered aluminium silicate at temperatures between 30 0 and 100 0 C, which is practically insoluble (10 -7 to 10 -10 g/cm 2 -day) and is thermally stable. A cancrinite is formed which binds the radioactive salts in the cage-like structure of its crystal lattice. The method is also suitable for liquid wastes from the Purex method as well as for wastes containing fission products of Cs 137 and Sr 90 in concentrations of 0.37 M to 0.01 M. Numerous detailed examples explain the invention. (UW/LH) [de

  13. Risks attached to container- and bunker-storage of nuclear waste

    International Nuclear Information System (INIS)

    Jager, D. de

    1987-12-01

    The results are presented of a literature study into the risks attached to the two dry-storage options selected by the Dutch Central Organization For Radioactive Waste (COVRA): the container- and the bunker-storage for irradiated nuclear-fuel elements and nuclear waste. Since the COVRA does not make it clear how these concepts should have to be realized, the experiences abroad with dry interim-storage are considered. In particular the Castor-container-storage and the bunker storage proposed in the committee MINSK (Possibilities of Interim-storage in the Netherlands of Irradiated nuclear-fuel elements and Nuclear waste) are studied further in depth. The committee MINSK has performed a study into the technical realizability of various interim-storage facilities, among which a storage in bunkers. (author). 75 refs.; 14 figs.; 16 tabs

  14. A novel technique for finding gas bubbles in the nuclear waste containers using Muon Scattering Tomography

    Science.gov (United States)

    Dobrowolska, M.; Velthuis, J.; Frazão, L.; Kikoła, D.

    2018-05-01

    Nuclear waste is deposited for many years in the concrete or bitumen-filled containers. With time hydrogen gas is produced, which can accumulate in bubbles. These pockets of gas may result in bitumen overflowing out of the waste containers and could result in spread of radioactivity. Muon Scattering Tomography is a non-invasive scanning method developed to examine the unknown content of nuclear waste drums. Here we present a method which allows us to successfully detect bubbles larger than 2 litres and determine their size with a relative uncertainty resolution of 1.55 ± 0.77%. Furthermore, the method allows to make a distinction between a conglomeration of bubbles and a few smaller gas volumes in different locations.

  15. Light Duty Utility Arm System applications for tank waste remediation

    International Nuclear Information System (INIS)

    Carteret, B.A.

    1994-10-01

    The Light Duty Utility Arm (LDUA) System is being developed by the US Department of Energy's (DOE's) Office of Technology Development (OTD, EM-50) to obtain information about the conditions and contents of the DOE's underground storage tanks. Many of these tanks are deteriorating and contain hazardous, radioactive waste generated over the past 50 years as a result of defense materials production at a member of DOE sites. Stabilization and remediation of these waste tanks is a high priority for the DOE's environmental restoration program. The LDUA System will provide the capability to obtain vital data needed to develop safe and cost-effective tank remediation plans, to respond to ongoing questions about tank integrity and leakage, and to quickly investigate tank events that raise safety concerns. In-tank demonstrations of the LDUA System are planned for three DOE sites in 1996 and 1997: Hanford, Idaho National Engineering Laboratory (INEL), and Oak Ridge National Laboratory (ORNL). This paper provides a general description of the system design and discusses a number of planned applications of this technology to support the DOE's environmental restoration program, as well as potential applications in other areas. Supporting papers by other authors provide additional in-depth technical information on specific areas of the system design

  16. Detection of tiny amounts of fissile materials in large-sized containers with radioactive waste

    Science.gov (United States)

    Batyaev, V. F.; Skliarov, S. V.

    2018-01-01

    The paper is devoted to non-destructive control of tiny amounts of fissile materials in large-sized containers filled with radioactive waste (RAW). The aim of this work is to model an active neutron interrogation facility for detection of fissile ma-terials inside NZK type containers with RAW and determine the minimal detectable mass of U-235 as a function of various param-eters: matrix type, nonuniformity of container filling, neutron gen-erator parameters (flux, pulse frequency, pulse duration), meas-urement time. As a result the dependence of minimal detectable mass on fissile materials location inside container is shown. Nonu-niformity of the thermal neutron flux inside a container is the main reason of the space-heterogeneity of minimal detectable mass in-side a large-sized container. Our experiments with tiny amounts of uranium-235 (<1 g) confirm the detection of fissile materials in NZK containers by using active neutron interrogation technique.

  17. Optimisation by mathematical modeling of physicochemical characteristics of concrete containers in radioactive waste management

    Directory of Open Access Journals (Sweden)

    Plećaš Ilija

    2013-01-01

    Full Text Available A method for obtaining an optimal concrete container composition used for storing radioactive waste from nuclear power plants is developed. It is applied to the radionuclides 60Co, 137Cs, 85Sr, and 54Mn. A set of recipes for concrete composition leading to an optimal solution is given.

  18. ENVIRONMENTAL IMPACT OF THE STORED DUST-LIKE ZINC AND IRON CONTAINING WASTES

    Directory of Open Access Journals (Sweden)

    Tatyana A. Lytaeva

    2017-05-01

    On the basis of laboratory research and field observations of the environmental components in the impact area of the storage of dust-like zinc and iron containing wastes, the article describes regularities of formation of hydrogeochemical halos of contamination by heavy metals and iron. Results include also the description of changes in physico-chemical groundwater composition under the storage area.

  19. The market-incentive recycling system for waste packaging containers in Taiwan

    International Nuclear Information System (INIS)

    Bor Yunchang, Jeffrey; Chien, Y.-L.; Hsu, Esher

    2004-01-01

    This paper presents a new market-incentive (MI) system to recycle waste-packaging containers in Taiwan. Since most used packaging containers have no or insufficient market value, the government imposes a combined product charge and subsidy policy to provide enough economic incentive for recycling various kinds of packaging containers, such as iron, aluminum, paper, glass and plastic. Empirical results show that the new MI approach has stimulated and established the recycling market for waste-packaging containers. The new recycling system has provided 18,356 employment opportunities and generated NT$ 6.97 billion in real-production value and NT$ 3.18 billion in real GDP during the 1998 survey year. Cost-effectiveness analysis constitutes the theoretical foundation of the new scheme, whereas data used to compute empirical product charge are from two sources: marketing surveys of internal conventional costs of solid-waste collection, disposal and recycling in Taiwan, and benefit transfer of external environmental costs in the United States. The new recycling policy designed by the authors provides a reasonable solution for solid-waste management in a country with limited land resources such as Taiwan

  20. STABILIZATION OF A MIXED WASTE SLUDGE SURROGATE CONTAINING MORE THAN 260 PPM MERCURY

    International Nuclear Information System (INIS)

    Smith, W. J.; Feizollahi, F.; Brimley, R.

    2002-01-01

    In an earlier demonstration of an innovative mercury stabilization technology for the Department of Energy, ATG's full-scale process stabilized mercury in soils that initially contained more than 260 ppm of mercury of unknown speciation. The treated waste satisfied the leaching standards for mercury that qualify wastes containing less than 260 ppm for land disposal. This paper describes the extension of that work to demonstrate a full-scale process for the stabilization of a representative sludge that contained more than 260 ppm of Hg of several mercury species. RCRA (Resource Conservation and Recovery Act) regulations now require the recovery of mercury from any waste containing more than 260 ppm of mercury, usually with thermal retorts. The results of this work with a surrogate sludge, and of the previous work with an actual soil, support a proposal now before the U.S. EPA (Environmental Protection Agency) to allow such wastes to be stabilized without retorting. The full-scale demonstration with a sulfide reagent reduced the mercury concentrations in extracts of treated sludge below the relevant leaching standard, a Universal Treatment Standard (UTS) limit of 0.025 mg mercury per liter of leachate generated by the Toxicity Characteristic Leaching Procedure (TCLP). The sulfide formulation reduced the concentration to about onehalf the UTS limit

  1. Preliminary sensitivity analyses of corrosion models for BWIP [Basalt Waste Isolation Project] container materials

    International Nuclear Information System (INIS)

    Anantatmula, R.P.

    1984-01-01

    A preliminary sensitivity analysis was performed for the corrosion models developed for Basalt Waste Isolation Project container materials. The models describe corrosion behavior of the candidate container materials (low carbon steel and Fe9Cr1Mo), in various environments that are expected in the vicinity of the waste package, by separate equations. The present sensitivity analysis yields an uncertainty in total uniform corrosion on the basis of assumed uncertainties in the parameters comprising the corrosion equations. Based on the sample scenario and the preliminary corrosion models, the uncertainty in total uniform corrosion of low carbon steel and Fe9Cr1Mo for the 1000 yr containment period are 20% and 15%, respectively. For containment periods ≥ 1000 yr, the uncertainty in corrosion during the post-closure aqueous periods controls the uncertainty in total uniform corrosion for both low carbon steel and Fe9Cr1Mo. The key parameters controlling the corrosion behavior of candidate container materials are temperature, radiation, groundwater species, etc. Tests are planned in the Basalt Waste Isolation Project containment materials test program to determine in detail the sensitivity of corrosion to these parameters. We also plan to expand the sensitivity analysis to include sensitivity coefficients and other parameters in future studies. 6 refs., 3 figs., 9 tabs

  2. Implementation of radon measurements to evaluate the suitability of using cement containers for storing radioactive waste containing Ra-226

    International Nuclear Information System (INIS)

    Shweikani, R.; Kheituo, M.; Hushari, M.; Ali, A. F.

    2003-12-01

    This work aimed at studying radon diffusion through walls of cubic cement containers containing inside radioactive waste rich in Radium-226. In addition, the effect of the wall thickness on radon exhalation and external gamma exposure were also studded. Cubic cement molds were prepared with different dimensions ranged from 5 to 11 cm containing central cubic holes to contain the radioactive materials with dimensions ranged from 2 to 7 cm. The thicknesses of the walls were varied from 1 to 4 cm. Radon exhalation was studied by placing each pre-prepared cement specimen in a tightly closed glass container (desiccators, volume 7 liters) provided with input and output gas system circulation for one week. Active method (Lucas cell) was used to measure the concentration of radon in the container. It was noticed that radon concentration increased with the increase of the radioactive materials inside the specimens. This was simply explained as it is due to the increase of the amount of radium-226 in the specimen with will definitely lead to the increase of radon production. In addition, it was noticed that radon concentration were increased by increasing the thickness of the specimen wall for fixed amount of the radioactive materials inside. This result was unexpected. Therefore, many attempts were performed to explain it. For that, the mechanism of cement solidifications and structure of cement after solidification were studied. The conditions which affect the size and number of the formed pores in the specimens were also studied assuming that increasing the wall thickness will increase porosity and lead to the increase diffusion paths. It was concluded that it is possible to use the cubic cement containers to stop gamma radiation from the radioactive materials, but it is not possible to use them to stop radon unless special arrangements are performed. (author)

  3. Chemical stability of seven years aged cement-PET composite waste form containing radioactive borate waste simulates

    Energy Technology Data Exchange (ETDEWEB)

    Saleh, H.M., E-mail: hosamsaleh70@yahoo.com [Radioisotope Department, Atomic Energy Authority, Dokki (Egypt); Tawfik, M.E. [Department of Polymers and Pigments, National Research Center, Dokki (Egypt); Bayoumi, T.A. [Radioisotope Department, Atomic Energy Authority, Dokki (Egypt)

    2011-04-15

    Different samples of radioactive borate waste simulate [originating from pressurized water reactors (PWR)] have been prepared and solidified after mixing with cement-water extended polyester composite (CPC). The polymer-cement composite samples were prepared from recycled poly (ethylene terephthalate) (PET) waste and cement paste (water/cement ratio of 40%). The prepared samples were left to set at room temperature (25 deg. C {+-} 5) under humid conditions. After 28 days curing time the obtained specimens were kept in their molds to age for 7 years under ambient conditions. Cement-polymer composite waste form specimens (CPCW) have been subjected to leach tests for both {sup 137}Cs and {sup 60}Co radionuclides according to the method proposed by the International Atomic Energy Agency (IAEA). Leaching tests were justified under various factors that may exist within the disposal site (e.g. type of leachant, surrounding temperature, leachant behavior, the leachant volume to CPCW surface area...). The obtained data after 260 days of leaching revealed that after 7 years of aging the candidate cement-polymer composite (CPC) containing radioactive borate waste samples are characterized by adequate chemical stability required for the long-term disposal process.

  4. Testing, verification and application of CONTAIN for severe accident analysis of LMFBR-containments

    International Nuclear Information System (INIS)

    Langhans, J.

    1991-01-01

    Severe accident analysis for LMFBR-containments has to consider various phenomena influencing the development of containment loads as pressure and temperatures as well as generation, transport, depletion and release of aerosols and radioactive materials. As most of the different phenomena are linked together their feedback has to be taken into account within the calculation of severe accident consequences. Otherwise no best-estimate results can be assured. Under the sponsorship of the German BMFT the US code CONTAIN is being developed, verified and applied in GRS for future fast breeder reactor concepts. In the first step of verification, the basic calculation models of a containment code have been proven: (i) flow calculation for different flow situations, (ii) heat transfer from and to structures, (iii) coolant evaporation, boiling and condensation, (iv) material properties. In the second step the proof of the interaction of coupled phenomena has been checked. The calculation of integrated containment experiments relating natural convection flow, structure heating and coolant condensation as well as parallel calculation of results obtained with an other code give detailed information on the applicability of CONTAIN. The actual verification status allows the following conclusion: a caucious analyst experienced in containment accident modelling using the proven parts of CONTAIN will obtain results which have the same accuracy as other well optimized and detailed lumped parameter containment codes can achieve. Further code development, additional verification and international exchange of experience and results will assure an adequate code for the application in safety analyses for LMFBRs. (orig.)

  5. Liquid metal reactor applications of the CONTAIN code

    International Nuclear Information System (INIS)

    Carroll, D.E.; Bergeron, K.D.; Gido, R.; Valdez, G.D.; Scholtyssek, W.

    1988-01-01

    The CONTAIN code is the NRC's best-estimate code for the evaluation of the conditions that may exist inside a reactor containment building during a severe accident. Included in the phenomena modeled are thermal-hydraulics, radiant and convective heat transfer, aerosol loading and transient response, fission product transport and heating effects, and interactions of sodium and corium with the containment atmosphere and structures. CONTAIN has been used by groups in Japan and West Germany to assess its ability to analyze accident consequences for liquid metal reactor (LMR) plants. In conjunction with this use, collaborative efforts to improve the modeling have been pursued. This paper summarizes the current state of the version of CONTAIN that has been enhanced with extra capabilities for LMR applications. A description of physical models is presented, followed by a review of validation exercises performed with CONTAIN. Some demonstration calculations of an integrated LMR application are presented

  6. 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.

  7. The use of repassivation potential in predicting the performance of high-level nuclear waste container materials

    International Nuclear Information System (INIS)

    Sridhar, N.; Dunn, D.; Cragnolino, G.

    1995-01-01

    Localized corrosion in aqueous environments forms an important bounding condition for the performance assessment of high-level waste (HLW) container materials. A predictive methodology using repassivation potential is examined in this paper. It is shown, based on long-term (continuing for over 11 months) testing of alloy 825, that repassivation potential of deep pits or crevices is a conservative and robust parameter for the prediction of localized corrosion. In contrast, initiation potentials measured by short-term tests are non-conservative and highly sensitive to several surface and environmental factors. Corrosion data from various field tests and plant equipment performance are analyzed in terms of the applicability of repassivation potential. The applicability of repassivation potential for predicting the occurrence of stress corrosion cracking (SCC) and intergranular corrosion in chloride containing environments is also examined

  8. Optimization of Waste Collection System Using Underground Containers with Source Separation Plan (Case Study: Zone 3 of Yazd Municipality, Iran

    Directory of Open Access Journals (Sweden)

    Maryam Morakabatchian

    2017-12-01

    Conclusion: Optimization of urban waste collection system using underground containers for wet waste and the use of temporary stations of dry wastes, considering the significant economic, environmental and aesthetic advantages can be considered as an appropriate option in Iranian cities especially in areas with hot and humid weather such as Yazd.

  9. Improvement of the cold flow characteristics of biodiesel containing dissolved polymer wastes using acetone

    Directory of Open Access Journals (Sweden)

    Pouya Mohammadi

    2014-03-01

    Full Text Available Due to the fast fossil fuel depletion and at the same time global warming phenomenon anticipated for the next coming years, the necessity of developing alternative fuels e.g. biofuels (i.e. bioethanol, biodiesel, biogas and etc. has turned into an important concern. Recently, the application of the bio-solvency properties of biodiesel for recycling waste polymers has been highlighted. However, the impact of polymer dissolution on cold flow characteristics of biodiesel was never investigated. The present study was set to explore the impact of different solvents in stabilizing biodiesel-polymer solution. Among them, acetone was proved to be the best fuel stabilizer. Subsequently, cold flow characteristic i.e. cloud point, of the biodiesel-polymer-acetone fuel was found to have improved (decreased due to the inclusion of acetone. Finally, flash point analysis of the fuel blends containing acetone was done to ensured high safety of the fuel blend by dramatically increasing the flash point values of biodiesel-polymer fuel blends.

  10. Sulphate solubility and sulphate diffusion in oxide glasses: implications for the containment of sulphate-bearing nuclear wastes; Solubilite et cinetiques de diffusion des sulfates dans differents verres d'oxydes: application au conditionnement des dechets nucleaires sulfates

    Energy Technology Data Exchange (ETDEWEB)

    Lenoir, M.

    2009-09-15

    The thesis deals with sulphate solubility and sulphate diffusion in oxide glasses, in order to control sulphate incorporation and sulphate volatilization in nuclear waste glasses. It was conducted on simplified compositions, in the SiO{sub 2}-B{sub 2}O{sub 3}-R{sub 2}O (R = Li, Na, K, Cs), SiO{sub 2}-B{sub 2}O{sub 3}-BaO and V{sub 2}O{sub 5}-B{sub 2}O{sub 3}-BaO systems. These compositions allowed us to study the influence of the nature of network-modifying ions (Li{sup +}, Na{sup +}, K{sup +}, Cs{sup +} or Ba{sup 2+}) and also of former elements (Si, B, V), on structure and properties of glasses. Sulphate volatility is studied in sodium borosilicate melts using an innovative technique of sulphate quantitation with Raman spectroscopy. This technique is useful to obtain kinetic curves of sulphate volatilization. The establishment of a model to fit these curves leads to the determination of diffusion coefficients of sulphate. These diffusion coefficients can thus be compared to diffusion coefficients of other species, determined by other techniques and presented in the literature. They are also linked to diffusion coefficients in relation with the viscosity of the melts. Concerning sulphate solubility in glasses, it depends on glass composition and on the nature of sulphate incorporated. Sulphate incorporation in alkali borosilicate glasses leads to the formation of a sulphate layer floating on top of the melt. Sulphate incorporation in barium borosilicate and boro-vanadate glasses leads to the crystallization of sulphate species inside the vitreous matrix. Moreover, sulphate solubility is higher in these glasses than in alkali borosilicates. Finally, exchanges between cations present in glasses and cations present in the sulphate phase are also studied. (author)

  11. The application of PLC in 60Co container inspection system

    International Nuclear Information System (INIS)

    Huang Yibin; Xiang Xincheng

    2001-01-01

    The author discusses the interlock technique of 60 Co container inspection system, and introduces the hardware structure and program of interlock control system using PLC. Due to adopting PLC distributed control, the system works stably and reliably. The successful application of PLC in 60 Co container inspection system has some use for reference in nuclear technology field

  12. Experimental investigation of the fatigue behaviour of asphalt concrete mixtures containing waste iron powder

    International Nuclear Information System (INIS)

    Arabani, M.; Mirabdolazimi, S.M.

    2011-01-01

    Research highlights: → This paper presents the first model of the fatigue behaviour of iron-asphalt mixtures in the world. → This model is able to describe the fatigue behaviour of iron-asphalt under dynamic loading. → Coarse surface, high stiffness and angularity of iron powder lead to enhanced fatigue performance. → The model illustrates that the use of iron powder has a considerable effect on tensile strain of HMA. → The use of this type of waste material could be a helpful solution for less polluted environment. - Abstract: The use of additives and admixtures in the construction of asphalt concrete pavements to strengthen them against dynamic loads has increased considerably in recent years. Recent research has shown that employing desirable waste materials in hot mix asphalts (HMAs) improves their dynamic properties noticeably. The study of some special cases, such as the addition of blast furnace slag and metallic materials of waste electronic instruments to HMA, has led to a considerable increase in the ability of HMAs to tolerate fatigue phenomena and repeated loading. Based on experimental studies, a model is proposed to describe the fatigue behaviour of asphalt mixtures containing waste iron powder. The results of this research show an important increase in the strength of asphalt mixtures containing waste iron powder against fatigue phenomena in comparison to conventional HMAs.

  13. Virtual environmental applications for buried waste characterization technology evaluation report

    International Nuclear Information System (INIS)

    1995-05-01

    The project, Virtual Environment Applications for Buried Waste Characterization, was initiated in the Buried Waste Integrated Demonstration Program in fiscal year 1994. This project is a research and development effort that supports the remediation of buried waste by identifying and examining the issues, needs, and feasibility of creating virtual environments using available characterization and other data. This document describes the progress and results from this project during the past year

  14. Virtual environmental applications for buried waste characterization technology evaluation report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-05-01

    The project, Virtual Environment Applications for Buried Waste Characterization, was initiated in the Buried Waste Integrated Demonstration Program in fiscal year 1994. This project is a research and development effort that supports the remediation of buried waste by identifying and examining the issues, needs, and feasibility of creating virtual environments using available characterization and other data. This document describes the progress and results from this project during the past year.

  15. Separation of uranium and common impurities from solid analytical waste containing plutonium

    International Nuclear Information System (INIS)

    Pathak, Nimai; Kumar, Mithlesh; Thulasidas, S.K.; Hon, N.S.; Kulkarni, M.J.; Mhatre, Amol; Natarajan, V.

    2014-07-01

    The report describes separation of uranium (U) and common impurities from solid analytical waste containing plutonium (Pu). This will be useful in recovery of Pu from nuclear waste. This is an important activity of any nuclear program in view of the strategic importance of Pu. In Radiochemistry Division, the trace metal analysis of Pu bearing fuel materials such as PuO 2 , (U,Pu)O 2 and (U,Pu)C are being carried out using the DC arc-Carrier Distillation technique. During these analyses, solid analytical waste containing Pu and 241 Am is generated. This comprises of left-over of samples and prepared charges. The main constituents of this waste are uranium oxide, plutonium oxide and silver chloride used as carrier. This report describes the entire work carried out to separate gram quantities of Pu from large amounts of U and mg quantities of 241 Am and the effect of leaching of the waste with nitric acid as a function of batch size. The effect of leaching the solid analytical waste of (U,Pu)O 2 and AgCl with concentrated nitric acid for different time intervals was also studied. Later keeping the time constant, the effect of nitric acid molarity on the leaching of U and Pu was investigated. Four different lots of the waste having different amounts were subjected to multiple leaching with 8 M nitric acid, each for 15 minutes duration. In all the experiments the amount of Uranium, Plutonium and other impurities leached were determined using ICP as an excitation source. The results are discussed in this report. (author)

  16. Development of electrochemical ion exchange electrodes for the treatment of wastes containing chromium or cesium ions

    International Nuclear Information System (INIS)

    Manosso, Helena Cristina

    2006-01-01

    Nowadays, environmental preservation using technologies that do not attack it, generating non-toxic residues and reduced volumes, has been discussed. Hazardous effluents, containing metals, as chromium, have been poured in the soils and rivers, degrading the water. Not different are the problems originated from some nuclear activities that generate wastes, as in chemical research laboratories. Although those wastes are not poured in the environment, sometimes they are inadequately stored, what can cause serious accidents. With the purpose of solving this problem, there are some techniques to waste treatment, between them there is the electrochemical ion exchange (EIX). EIX is an advanced process that has advantages over traditional ion exchange and the fact of using the electron as the only reagent, reduces the volume of the solution to be treated. This technique consists of development of an electrode, where an ion exchanger is physically incorporated in an electrode structure with a binder. In this study, cationic resin Amberlite CG-50 and zirconium phosphate have been chosen for the separation of chromium and cesium from waste, respectively. They were chosen because they present high chemical stability in oxidizing media and at ionizing radiation. The quantity of charcoal, graphite and binder used in formulation of electrode have been studied either. Before choosing the best electrode, it was verified sorption percentage of 99,3% for chromium and 99,8% for cesium. The greater advantage of this process is the total elution of chromium as much as cesium, without reagents addition, being possible to reuse the electrode without losing its capacity. Beside on the results, a continuous process for the wastes containing Cr and Cs, using a flux electrolytic cell (CELFLUX) of high retention capacity, was presented. The high efficiency of this cell for both retention and elution, leading to an important reduction of waste volume, and, every more, making possible the

  17. Biochemical process of low level radioactive liquid simulation waste containing detergent

    International Nuclear Information System (INIS)

    Kundari, Noor Anis; Putra, Sugili; Mukaromah, Umi

    2015-01-01

    Research of biochemical process of low level radioactive liquid waste containing detergent has been done. Thse organic liquid wastes are generated in nuclear facilities such as from laundry. The wastes that are cotegorized as hazard and poison materials are also radioactive. It must be treated properly by detoxification of the hazard and decontamination of the radionuclides to ensure that the disposal of the waste meets the requirement of standard quality of water. This research was intended to determine decontamination factor and separation efficiensies, its kinetics law, and to produce a supernatant that ensured the environmental quality standard. The radioactive element in the waste was thorium with activity of 5.10 −5 Ci/m 3 . The radioactive liquid waste which were generated in simulation plant contains detergents that was further processed by aerobic biochemical process using SGB 103 bacteria in a batch reactor equipped with aerators. Two different concentration of samples were processed and analyzed for 212 hours and 183 hours respectively at a room temperature. The product of this process is a liquid phase called as supernatant and solid phase material called sludge. The chemical oxygen demand (COD), biological oxygen demand (BOD), suspended solid (SS), and its alpha activity were analyzed. The results show that the decontamination factor and the separation efficiency of the lower concentration samples are higher compared to the samples with high concentration. Regarding the decontamination factor, the result for 212 hours processing of waste with detergent concentration of 1.496 g/L was 3.496 times, whereas at the detergent concentration of 0.748 g/L was 15.305 times for 183 hours processing. In case of the separation efficiency, the results for both samples were 71.396% and 93.465% respectively. The Bacterial growth kinetics equation follow Monod’s model and the decreasing of COD and BOD were first order with the rate constant of 0.01 hour −1

  18. Biochemical process of low level radioactive liquid simulation waste containing detergent

    Energy Technology Data Exchange (ETDEWEB)

    Kundari, Noor Anis, E-mail: nooranis@batan.go.id; Putra, Sugili; Mukaromah, Umi [Sekolah Tinggi Teknologi Nuklir – Badan Tenaga Nuklir Nasional Jl. Babarsari P.O. BOX 6101 YKBB Yogyakarta 55281 Telp : (0274) 48085, 489716, Fax : (0274) 489715 (Indonesia)

    2015-12-29

    Research of biochemical process of low level radioactive liquid waste containing detergent has been done. Thse organic liquid wastes are generated in nuclear facilities such as from laundry. The wastes that are cotegorized as hazard and poison materials are also radioactive. It must be treated properly by detoxification of the hazard and decontamination of the radionuclides to ensure that the disposal of the waste meets the requirement of standard quality of water. This research was intended to determine decontamination factor and separation efficiensies, its kinetics law, and to produce a supernatant that ensured the environmental quality standard. The radioactive element in the waste was thorium with activity of 5.10{sup −5} Ci/m{sup 3}. The radioactive liquid waste which were generated in simulation plant contains detergents that was further processed by aerobic biochemical process using SGB 103 bacteria in a batch reactor equipped with aerators. Two different concentration of samples were processed and analyzed for 212 hours and 183 hours respectively at a room temperature. The product of this process is a liquid phase called as supernatant and solid phase material called sludge. The chemical oxygen demand (COD), biological oxygen demand (BOD), suspended solid (SS), and its alpha activity were analyzed. The results show that the decontamination factor and the separation efficiency of the lower concentration samples are higher compared to the samples with high concentration. Regarding the decontamination factor, the result for 212 hours processing of waste with detergent concentration of 1.496 g/L was 3.496 times, whereas at the detergent concentration of 0.748 g/L was 15.305 times for 183 hours processing. In case of the separation efficiency, the results for both samples were 71.396% and 93.465% respectively. The Bacterial growth kinetics equation follow Monod’s model and the decreasing of COD and BOD were first order with the rate constant of 0

  19. Biochemical process of low level radioactive liquid simulation waste containing detergent

    Science.gov (United States)

    Kundari, Noor Anis; Putra, Sugili; Mukaromah, Umi

    2015-12-01

    Research of biochemical process of low level radioactive liquid waste containing detergent has been done. Thse organic liquid wastes are generated in nuclear facilities such as from laundry. The wastes that are cotegorized as hazard and poison materials are also radioactive. It must be treated properly by detoxification of the hazard and decontamination of the radionuclides to ensure that the disposal of the waste meets the requirement of standard quality of water. This research was intended to determine decontamination factor and separation efficiensies, its kinetics law, and to produce a supernatant that ensured the environmental quality standard. The radioactive element in the waste was thorium with activity of 5.10-5 Ci/m3. The radioactive liquid waste which were generated in simulation plant contains detergents that was further processed by aerobic biochemical process using SGB 103 bacteria in a batch reactor equipped with aerators. Two different concentration of samples were processed and analyzed for 212 hours and 183 hours respectively at a room temperature. The product of this process is a liquid phase called as supernatant and solid phase material called sludge. The chemical oxygen demand (COD), biological oxygen demand (BOD), suspended solid (SS), and its alpha activity were analyzed. The results show that the decontamination factor and the separation efficiency of the lower concentration samples are higher compared to the samples with high concentration. Regarding the decontamination factor, the result for 212 hours processing of waste with detergent concentration of 1.496 g/L was 3.496 times, whereas at the detergent concentration of 0.748 g/L was 15.305 times for 183 hours processing. In case of the separation efficiency, the results for both samples were 71.396% and 93.465% respectively. The Bacterial growth kinetics equation follow Monod's model and the decreasing of COD and BOD were first order with the rate constant of 0.01 hour-1.

  20. A Prototype Scintillating-Fibre Tracker for the Cosmic-ray Muon Tomography of Legacy Nuclear Waste Containers

    Science.gov (United States)

    Kaiser, R.; Clarkson, A.; Hamilton, D. J.; Hoek, M.; Ireland, D. G.; Johnston, J. R.; Keri, T.; Lumsden, S.; Mahon, D. F.; McKinnon, B.; Murray, M.; Nutbeam-Tuffs, S.; Shearer, C.; Staines, C.; Yang, G.; Zimmerman, C.

    2014-03-01

    Cosmic-ray muons are highly-penetrative charged particles observed at sea level with a flux of approximately 1 cm-2 min-1. They interact with matter primarily through Coulomb scattering which can be exploited in muon tomography to image objects within industrial nuclear waste containers. This paper presents the prototype scintillating-fibre detector developed for this application at the University of Glasgow. Experimental results taken with test objects are shown in comparison to results from GEANT4 simulations. These results verify the simulation and show discrimination between the low, medium and high-Z materials imaged.

  1. A Prototype Scintillating-Fibre Tracker for the Cosmic-ray Muon Tomography of Legacy Nuclear Waste Containers

    Directory of Open Access Journals (Sweden)

    Kaiser R.

    2014-03-01

    Full Text Available Cosmic-ray muons are highly-penetrative charged particles observed at sea level with a flux of approximately 1 cm−2 min−1. They interact with matter primarily through Coulomb scattering which can be exploited in muon tomography to image objects within industrial nuclear waste containers. This paper presents the prototype scintillating-fibre detector developed for this application at the University of Glasgow. Experimental results taken with test objects are shown in comparison to results from GEANT4 simulations. These results verify the simulation and show discrimination between the low, medium and high-Z materials imaged.

  2. Self-consolidating concretes containing waste PET bottles as sand replacement

    Science.gov (United States)

    Khalid, Faisal Sheikh; Azmi, Nurul Bazilah; Mazenan, Puteri Natasya; Shahidan, Shahiron; Othman, Nor hazurina; Guntor, Nickholas Anting Anak

    2018-02-01

    This study evaluates the effect of self-consolidating concrete (SCC) containing waste polyethylene terephthalate (PET) granules on the fresh, mechanical and water absorption properties. Fine aggregates were replaced from 0% to 8% by PET granules. The fresh properties of SCC containing PET granules were determined using slump flow and V-funnel flow time tests. The compressive and splitting tensile strength were evaluated. The results indicated that utilization of waste PET granules in production of SCC could be an effective way for recycling purpose. The maximum amount of PET replacement should be limited to 5%. Exceeding 5% of PET content may result in an increase of V-funnel flow time to overpass the limiting value, decrease in strength. The production of high performance SCC containing 5% PET granules satisfies all the requirements for SCC with satisfactory outputs.

  3. Studies of corrosion in metallic container for storage of high level radioactive wastes

    International Nuclear Information System (INIS)

    Azkarate, I.; Madina, V.; Insausti, M.

    1999-01-01

    The metallic container is one of the most important barriers that, along with engineered and natural barriers, will isolate high level nuclear waste in saline and granite geological formations from the geosphere. However, general and localized corrosion modes such as stress corrosion cracking (SCC), pitting, crevice corrosion and hydrogen damage can be active under disposal conditions, so the corrosion behaviour of the metal container material must be carefully studied. Several metals and their alloys have been proposed for the fabrication of nuclear waste containers including carbon steels, stainless steels, titanium and titanium alloys and copper and copper-base alloys. Carbon steels and copper alloys are considered for the two rock formations, titanium is considered for salt environments and the stainless steel only in the case of a granite formation. (Author)

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

    International Nuclear Information System (INIS)

    Vinson, D.W.; Bullen, D.B.

    1995-01-01

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

  5. Passive 3D imaging of nuclear waste containers with Muon Scattering Tomography

    Science.gov (United States)

    Thomay, C.; Velthuis, J.; Poffley, T.; Baesso, P.; Cussans, D.; Frazão, L.

    2016-03-01

    The non-invasive imaging of dense objects is of particular interest in the context of nuclear waste management, where it is important to know the contents of waste containers without opening them. Using Muon Scattering Tomography (MST), it is possible to obtain a detailed 3D image of the contents of a waste container on reasonable timescales, showing both the high and low density materials inside. We show the performance of such a method on a Monte Carlo simulation of a dummy waste drum object containing objects of different shapes and materials. The simulation has been tuned with our MST prototype detector performance. In particular, we show that both a tungsten penny of 2 cm radius and 1 cm thickness, and a uranium sheet of 0.5 cm thickness can be clearly identified. We also show the performance of a novel edge finding technique, by which the edges of embedded objects can be identified more precisely than by solely using the imaging method.

  6. Development and assessment of a fiber reinforced HPC container for radioactive waste

    International Nuclear Information System (INIS)

    Roulet, A.; Pineau, F.; Chanut, S.; Thibaux, Th.

    2007-01-01

    As part of its research into solutions for concrete disposal containers for long-lived radioactive waste, Andra defined requirements for high-performance concretes with enhanced porosity, diffusion, and permeability characteristics. It is the starting point for further research into severe conditions of containment and durability. To meet these objectives, Eiffage TP consequently developed a highly fibered High Performance Concrete (HPC) design mix using CEM V cement and silica fume. Then, mockups were produced to characterize the performance various concepts of containers with this new concrete mix. These mockups helped to identify possible manufacturing problems, and particularly the risk of cracking due to restrained shrinkage. (authors)

  7. In-Drift Accumulation of Fissile Material From Waste Packages Containing Plutonium Disposition Waste Forms

    International Nuclear Information System (INIS)

    H.W. Stockman; S. LeStrange

    2000-01-01

    The objective of this calculation is to provide estimates of the amount of fissile material flowing out of the waste package (source term) and the accumulation of fissile elements (U and Pu) in a crushed-tuff invert. These calculations provide input for the analysis of repository impacts of the Pu-ceramic waste forms. In particular, the source term results are used as input to the far-field accumulation calculation reported in Ref. 51, and the in-drift accumulation results are used as inputs for the criticality calculations reported in Ref. 2. The results are also summarized and interpreted in Ref. 52. The scope of this calculation is the waste package (WP) Viability Assessment (VA) design, which consists of an outer corrosion-allowance material (CAM) and an inner corrosion-resistant material (CRM). This design is used in this calculation in order to be consistent with earlier Pu-ceramic degradation calculations (Ref. 15). The impact of the new Enhanced Design Alternative-I1 (EDA-11) design on the results will be addressed in a subsequent report. The design of the invert (a leveling foundation, which creates a level surface of the drift floor and supports the WP mounting structure) is consistent with the EDA-I1 design. The invert will be composed of crushed stone and a steel support structure (Ref. 17). The scope of this calculation is also defined by the nominal degradation scenario, which involves the breach of the WP (Section 10.5.1.2, Ref. 48), followed by the influx of water. Water in the WP may, in time, gradually leach the fissile components and neutron absorbers out of the ceramic waste forms. Thus, the water in the WP may become laden with dissolved actinides (e.g., Pu and U), and may eventually overflow or leak from the WP. Once the water leaves the WP, it may encounter the invert, in which the actinides may reprecipitate. Several factors could induce reprecipitation; these factors include: the high surface area of the crushed stone, and the presence of

  8. In-Drift Accumulation of Fissile Material From Waste Packages Containing Plutonium Disposition Waste Form

    Energy Technology Data Exchange (ETDEWEB)

    H.W> Stockman; S. LeStrange

    2000-09-28

    The objective of this calculation is to provide estimates of the amount of fissile material flowing out of the waste package (source term) and the accumulation of fissile elements (U and Pu) in a crushed-tuff invert. These calculations provide input for the analysis of repository impacts of the Pu-ceramic waste forms. In particular, the source term results are used as input to the far-field accumulation calculation reported in Ref. 51, and the in-drift accumulation results are used as inputs for the criticality calculations reported in Ref. 2. The results are also summarized and interpreted in Ref. 52. The scope of this calculation is the waste package (WP) Viability Assessment (VA) design, which consists of an outer corrosion-allowance material (CAM) and an inner corrosion-resistant material (CRM). This design is used in this calculation in order to be consistent with earlier Pu-ceramic degradation calculations (Ref. 15). The impact of the new Enhanced Design Alternative-I1 (EDA-11) design on the results will be addressed in a subsequent report. The design of the invert (a leveling foundation, which creates a level surface of the drift floor and supports the WP mounting structure) is consistent with the EDA-I1 design. The invert will be composed of crushed stone and a steel support structure (Ref. 17). The scope of this calculation is also defined by the nominal degradation scenario, which involves the breach of the WP (Section 10.5.1.2, Ref. 48), followed by the influx of water. Water in the WP may, in time, gradually leach the fissile components and neutron absorbers out of the ceramic waste forms. Thus, the water in the WP may become laden with dissolved actinides (e.g., Pu and U), and may eventually overflow or leak from the WP. Once the water leaves the WP, it may encounter the invert, in which the actinides may reprecipitate. Several factors could induce reprecipitation; these factors include: the high surface area of the crushed stone, and the presence of

  9. Grout formulation for disposal of low-level and hazardous waste streams containing fluoride

    Science.gov (United States)

    McDaniel, E.W.; Sams, T.L.; Tallent, O.K.

    1987-06-02

    A composition and related process for disposal of hazardous waste streams containing fluoride in cement-based materials is disclosed. the presence of fluoride in cement-based materials is disclosed. The presence of fluoride in waste materials acts as a set retarder and as a result, prevents cement-based grouts from setting. This problem is overcome by the present invention wherein calcium hydroxide is incorporated into the dry-solid portion of the grout mix. The calcium hydroxide renders the fluoride insoluble, allowing the grout to set up and immobilize all hazardous constituents of concern. 4 tabs.

  10. Effect of chloride concentration and pH on pitting corrosion of waste package container materials

    International Nuclear Information System (INIS)

    Roy, A.K.; Fleming, D.L.; Gordon, S.R.

    1996-12-01

    Electrochemical cyclic potentiodynamic polarization experiments were performed on several candidate waste package container materials to evaluate their susceptibility to pitting corrosion at 90 degrees C in aqueous environments relevant to the potential underground high-level nuclear waste repository. Results indicate that of all the materials tested, Alloy C-22 and Ti Grade-12 exhibited the maximum corrosion resistance, showing no pitting or observable corrosion in any environment tested. Efforts were also made to study the effect of chloride ion concentration and pH on the measured corrosion potential (Ecorr), critical pitting and protection potential values

  11. Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

    International Nuclear Information System (INIS)

    Bullen, D.B.; Gdowski, G.E.

    1988-08-01

    Three copper-based alloys and three iron- to nickel-based austenitic alloys are being considered as possible materials for fabrication of high-level radioactive-waste disposal containers. The waste will include spent fuel assemblies from reactors as well as high-level waste in borosilicate glass and will be sent to the prospective site at Yucca Mountain, Nevada, for disposal. The copper-based alloy materials are CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni). The austenitic materials are Types 304L and 316L stainless steels and Alloy 825. The waste-package containers must maintain substantially complete containment for at least 300 yr and perhaps as long as 1000 yr, and they must be retrievable from the disposal site during the first 50 yr after emplacement. The containers will be exposed to high temperatures and high gamma radiation fields from the decay of high-level waste. This volume surveys the available data on the phase stability of both groups of candidate alloys. The austenitic alloys are reviewed in terms of the physical metallurgy of the iron-chromium-nickel system, martensite transformations, carbide formation, and intermetallic-phase precipitation. The copper-based alloys are reviewed in terms of their phase equilibria and the possibility of precipitation of the minor alloying constituents. For the austenitic materials, the ranking based on phase stability is: Alloy 825 (best), Type 316L stainless steel, and then Type 304L stainless steel (worst). For the copper-based materials, the ranking is: CDA 102 (oxygen-free copper) (best), and then both CDA 715 and CDA 613. 75 refs., 24 figs., 6 tabs

  12. State Waste Discharge Permit application, 100-N Sewage Lagoon

    International Nuclear Information System (INIS)

    1994-06-01

    As part of the Hanford Federal Facility Agreement and Consent Order negotiations (Ecology et al. 1994), the US Department of Energy, Richland Operations Office, the US Environmental Protection Agency, and the Washington State Department of Ecology agreed that liquid effluent discharges to the ground on the Hanford Site which affect groundwater or have the potential to affect groundwater would be subject to permitting under the structure of Chapter 173--216 (or 173--218 where applicable) of the Washington Administrative Code, the State Waste Discharge Permit Program. As a result of this decision, the Washington State Department of Ecology and the US Department of Energy, Richland Operations Office entered into Consent Order No. DE 91NM-177, (Ecology and DOE-RL 1991). This document constitutes the State Waste Discharge Permit application for the 100-N Sewage Lagoon. Since the influent to the sewer lagoon is domestic waste water, the State Waste Discharge Permit application for Public Owned Treatment Works Discharges to Land was used. Although the 100-N Sewage Lagoon is not a Public Owned Treatment Works, the Public Owned Treatment Works application is more applicable than the application for industrial waste water. The 100-N Sewage Lagoon serves the 100-N Area and other Hanford Site areas by receiving domestic waste from two sources. A network of sanitary sewer piping and lift stations transfers domestic waste water from the 100-N Area buildings directly to the 100-N Sewage Lagoon. Waste is also received by trucks that transport domestic waste pumped from on site septic tanks and holding tanks. Three ponds comprise the 100-N Sewage Lagoon treatment system. These include a lined aeration pond and stabilization pond, as well as an unlined infiltration pond. Both piped-in and trucked-in domestic waste is discharged directly into the aeration pond

  13. Cost estimate of high-level radioactive waste containers for the Yucca Mountain Site Characterization Project

    Energy Technology Data Exchange (ETDEWEB)

    Russell, E.W.; Clarke, W. [Lawrence Livermore National Lab., CA (United States); Domian, H.A. [Babcock and Wilcox Co., Lynchburg, VA (United States); Madson, A.A. [Kaiser Engineers California Corp., Oakland, CA (United States)

    1991-08-01

    This report summarizes the bottoms-up cost estimates for fabrication of high-level radioactive waste disposal containers based on the Site Characterization Plan Conceptual Design (SCP-CD). These estimates were acquired by Babcock and Wilcox (B&S) under sub-contract to Lawrence Livermore National Laboratory (LLNL) for the Yucca Mountain Site Characterization Project (YMP). The estimates were obtained for two leading container candidate materials (Alloy 825 and CDA 715), and from other three vendors who were selected from a list of twenty solicited. Three types of container designs were analyzed that represent containers for spent fuel, and for vitrified high-level waste (HLW). The container internal structures were assumed to be AISI-304 stainless steel in all cases, with an annual production rate of 750 containers. Subjective techniques were used for estimating QA/QC costs based on vendor experience and the specifications derived for the LLNL-YMP Quality Assurance program. In addition, an independent QA/QC analysis is reported which was prepared by Kasier Engineering. Based on the cost estimates developed, LLNL recommends that values of $825K and $62K be used for the 1991 TSLCC for the spent fuel and HLW containers, respectively. These numbers represent the most conservative among the three vendors, and are for the high-nickel anstenitic steel (Alloy 825). 6 refs., 7 figs.

  14. Cost estimate of high-level radioactive waste containers for the Yucca Mountain Site Characterization Project

    International Nuclear Information System (INIS)

    Russell, E.W.; Clarke, W.; Domian, H.A.; Madson, A.A.

    1991-08-01

    This report summarizes the bottoms-up cost estimates for fabrication of high-level radioactive waste disposal containers based on the Site Characterization Plan Conceptual Design (SCP-CD). These estimates were acquired by Babcock and Wilcox (B ampersand S) under sub-contract to Lawrence Livermore National Laboratory (LLNL) for the Yucca Mountain Site Characterization Project (YMP). The estimates were obtained for two leading container candidate materials (Alloy 825 and CDA 715), and from other three vendors who were selected from a list of twenty solicited. Three types of container designs were analyzed that represent containers for spent fuel, and for vitrified high-level waste (HLW). The container internal structures were assumed to be AISI-304 stainless steel in all cases, with an annual production rate of 750 containers. Subjective techniques were used for estimating QA/QC costs based on vendor experience and the specifications derived for the LLNL-YMP Quality Assurance program. In addition, an independent QA/QC analysis is reported which was prepared by Kasier Engineering. Based on the cost estimates developed, LLNL recommends that values of $825K and $62K be used for the 1991 TSLCC for the spent fuel and HLW containers, respectively. These numbers represent the most conservative among the three vendors, and are for the high-nickel anstenitic steel (Alloy 825). 6 refs., 7 figs

  15. Preliminary analysis of the creep behaviour of nuclear fuel-waste container materials

    International Nuclear Information System (INIS)

    Dutton, R.; Leitch, B.W.; Crosthwaite, J.L.; Kasprick, G.R.

    1996-12-01

    In the Canadian Nuclear Fuel Waste Management Program, it is proposed that nuclear fuel waste be placed in a durable container and disposed of in a deep underground vault. Consideration of various disposal-container designs has identified either titanium or copper as the material suitable for constructing the container shell. As part of the R and D program to examine the structural integrity of the container, creep tests are being conducted on commercially pure titanium and oxygen-free copper. This report presents the preliminary data obtained. It also describes the evaluation of various constitutive equations to represent the creep curves, thus providing the basis for extrapolation of the creep behaviour over the design lifetime of the container. In this regard, a specific focus is placed on equations derived from the 0-Projection Concept. Recognizing that the container lifetime will be determined by the onset of tertiary creep leading to creep rupture, we present the results of the metallographic examination of creep damage. This shows that the tertiary stage in titanium is associated with the formation of transgranular cavities within the region of localized necking of the creep specimens. In contrast, creep damage in copper is in the form of intergranular cavities uniformly distributed throughout the gauge length. These results are analyzed within the context of the extant literature, and their implications for future container design are discussed. (author)

  16. Applicability of slags as waste forms for hazardous waste

    International Nuclear Information System (INIS)

    Bates, J.K.; Buck, E.C.; Dietz, N.L.; Wronkiewicz, D.J.; Feng, X.; Whitworth, C.; Filius, K.; Battleson, D.

    1994-01-01

    Slags, which are a combination of glassy and ceramic phases, were produced by the Component Development and Integration Facility, using a combination of soil and metal feeds. The slags were tested for durability using accelerated test methods in both water vapor and liquid water for time periods up to 179 days. The results indicated that under both conditions there was little reaction of the slag, in terms of material released to solution, or the reaction of the slag to form secondary mineral phases. The durability of the slags tested exceeded that of current high-level nuclear glass formulations and are viable materials, for waste disposal

  17. Safety analysis of geologic containment of long life radioactive wastes. Critical assessment of existing methods and proposition of prospective approach

    International Nuclear Information System (INIS)

    Masure, P.; Gedefroy, P.; Imauven, C.

    1983-01-01

    Existing methods of risk analysis applied to disposal of long-lived radioactive waste in geologic formations are rewieved. A prospective analysis method for containment performances is proposed, deduced in the burial system from the combination of interaction between wastes, repository, host rock, surrounding geosphere, of natural evolution of each component of the system, sudden or chance events that could break waste containment. The method is based on the elaboration of four basic schemes graded in difficulties to facilitate comparisons

  18. Chemistry of application of calcination/dissolution to the Hanford tank waste inventory

    International Nuclear Information System (INIS)

    Delegard, C.H.; Elcan, T.D.; Hey, B.E.

    1994-05-01

    Approximately 330,000 metric tons of sodium-rich radioactive waste originating from separation of plutonium from irradiated uranium fuel are stored in underground tanks at the Hanford Site in Washington State. Fractionation of the waste into low-level waste (LLW) and high-level waste (HLW) streams is envisioned via partial water dissolution and limited radionuclide extraction operations. Under optimum conditions, LLW would contain most of the chemical bulk while HLW would contain virtually all of the transuranic and fission product activity. Calcination at around 850 C, followed by water dissolution, has been proposed as an alternative initial treatment of Hanford Site waste to improve waste dissolution and the envisioned LLW/HLW split. Results of literature and laboratory studies are reported on the application of calcination/dissolution (C/D) to the fractionation of the Hanford Site tank waste inventory. Both simulated and genuine Hanford Site waste materials were used in the lab tests. To evaluation confirmed that C/D processing reduced the amount of several components from the waste. The C/D dissolutions of aluminum and chromium allow redistribution of these waste components from the HLW to the LLW fraction. Comparisons of simple water-washing with C/D processing of genuine Hanford Site waste are also reported based on material (radionuclide and chemical) distributions to solution and solid residue phases. The lab results show that C/D processing yielded superior dissolution of aluminum and chromium sludges compared to simple water dissolution. 57 refs., 26 figs., 18 tabs

  19. Chemistry of application of calcination/dissolution to the Hanford tank waste inventory

    Energy Technology Data Exchange (ETDEWEB)

    Delegard, C.H.; Elcan, T.D.; Hey, B.E.

    1994-05-01

    Approximately 330,000 metric tons of sodium-rich radioactive waste originating from separation of plutonium from irradiated uranium fuel are stored in underground tanks at the Hanford Site in Washington State. Fractionation of the waste into low-level waste (LLW) and high-level waste (HLW) streams is envisioned via partial water dissolution and limited radionuclide extraction operations. Under optimum conditions, LLW would contain most of the chemical bulk while HLW would contain virtually all of the transuranic and fission product activity. Calcination at around 850 C, followed by water dissolution, has been proposed as an alternative initial treatment of Hanford Site waste to improve waste dissolution and the envisioned LLW/HLW split. Results of literature and laboratory studies are reported on the application of calcination/dissolution (C/D) to the fractionation of the Hanford Site tank waste inventory. Both simulated and genuine Hanford Site waste materials were used in the lab tests. To evaluation confirmed that C/D processing reduced the amount of several components from the waste. The C/D dissolutions of aluminum and chromium allow redistribution of these waste components from the HLW to the LLW fraction. Comparisons of simple water-washing with C/D processing of genuine Hanford Site waste are also reported based on material (radionuclide and chemical) distributions to solution and solid residue phases. The lab results show that C/D processing yielded superior dissolution of aluminum and chromium sludges compared to simple water dissolution. 57 refs., 26 figs., 18 tabs.

  20. Application and research of special waste plasma disposal technology

    International Nuclear Information System (INIS)

    Lan Wei

    2007-12-01

    The basic concept of plasma and the principle of waste hot plasma disposal technology are simply introduced. Several sides of application and research of solid waste plasma disposal technology are sumed up. Compared to the common technology, the advantages of waste hot plasma disposal technology manifest further. It becomes one of the most prospective and the most attended high tech disposal technology in particular kind of waste disposal field. The article also simply introduces some experiment results in Southwest Institute of Physics and some work on the side of importation, absorption, digestion, development of foreign plasma torch technology and researching new power sources for plasma torch. (authors)