WorldWideScience

Sample records for clay-based waste containment

  1. Modification of clay-based waste containment materials

    Energy Technology Data Exchange (ETDEWEB)

    Adu-Wusu, K. [DuPont Central Research and Development, Newark, DE (United States); Whang, J.M. [DuPont Specialty Chemicals, Deepwater, NJ (United States); McDevitt, M.F. [DuPont Central Research and Development, Wilmington, DE (United States)

    1997-12-31

    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.

  2. ALKALI-ACTIVATED CEMENT MORTARS CONTAINING RECYCLED CLAY-BASED CONSTRUCTION AND DEMOLITION WASTE

    Directory of Open Access Journals (Sweden)

    F. Puertas

    2015-09-01

    Full Text Available The use of clay-based waste as an aggregate for concrete production is an amply studied procedure. Nonetheless, research on the use of this recycled aggregate to prepare alkaline cement mortars and concretes has yet to be forthcoming. The present study aimed to determine: the behaviour of this waste as a pozzolan in OPC systems, the mechanical strength in OPC, alkali-activated slag (AAS and fly ash (AAFA mortars and the effect of partial replacement of the slag and ash themselves with ground fractions of the waste. The pozzolanic behaviour of clay-based waste was confirmed. Replacing up to 20 % of siliceous aggregate with waste aggregate in OPC mortars induced a decline in 7 day strength (around 23 wt. %. The behaviour of waste aggregate in AAMs mortars, in turn, was observed to depend on the nature of the aluminosilicate and the replacement ratio used. When 20 % of siliceous aggregate was replaced by waste aggregate in AAS mortars, the 7 day strength values remained the same (40 MPa. In AAFA mortars, waste was found to effectively replace both the fly ash and the aggregate. The highest strength for AAFA mortars was observed when they were prepared with both a 50 % replacement ratio for the ash and a 20 % ratio for the aggregate.

  3. Using mixture design of experiments to assess the environmental impact of clay-based structural ceramics containing foundry wastes.

    Science.gov (United States)

    Coronado, M; Segadães, A M; Andrés, A

    2015-12-15

    This work describes the leaching behavior of potentially hazardous metals from three different clay-based industrial ceramic products (wall bricks, roof tiles, and face bricks) containing foundry sand dust and Waelz slag as alternative raw materials. For each product, ten mixtures were defined by mixture design of experiments and the leaching of As, Ba, Cd, Cr, Cu, Mo, Ni, Pb, and Zn was evaluated in pressed specimens fired simulating the three industrial ceramic processes. The results showed that, despite the chemical, mineralogical and processing differences, only chrome and molybdenum were not fully immobilized during ceramic processing. Their leaching was modeled as polynomial equations, functions of the raw materials contents, and plotted as response surfaces. This brought to evidence that Cr and Mo leaching from the fired products is not only dependent on the corresponding contents and the basicity of the initial mixtures, but is also clearly related with the mineralogical composition of the fired products, namely the amount of the glassy phase, which depends on both the major oxides contents and the firing temperature. PMID:26252997

  4. Screening of waste for use in clay-based bricks in the Arctic

    DEFF Research Database (Denmark)

    Belmonte, Louise Josefine; Ottosen, Lisbeth M.; Kirkelund, Gunvor Marie;

    2014-01-01

    of hazardous waste, municipal solid waste incineration (MSWI) ashes and minetailings from Greenland, were investigated in order to determine their potential suitability for incorporationin the production of clay-based bricks. Furthermore, the MSWI fly ash was subjected to two remediation techniques......Clay-based ceramics, such as bricks, are heterogeneous materials, which can incorporate raw materials ofwide ranging compositions, without impairing their technical properties (Dondi et al., 1997a,b). Due to thisability, bricks have become a popular material in waste management research worldwide...... and several studies have demonstrated that clay-based bricks and tiles can successfully accommodate waste types,such as incineration ashes, mine tailings and dredged harbour sediments (Zhang et al., 2011; Roy et al.,2007; Mezencevova et al., 2012). In the vulnerable Arctic environment, the impact...

  5. Radioactive waste conditioning by way of their introduction into clay base ceramic matrices

    International Nuclear Information System (INIS)

    Conditions for fixation of ash from radioactive wastes burnup, hydroxide pulps formed during precipitation-purification works in radiochemical technology, bottoms from NPPs liquid radioactive wastes evaporation are worked out primarily on simulators. It is shown that ceramics including 30-40% by wastes mass, roasted at the temperature of 1000-1050 deg C gas an apparent density of 2.1-2.5 g/cm3, compression endurance limit of 40-70 MPa and radionuclide leaching rate of 10-6-10-8 g(cm2xday). 9 refs.; 2 figs.; 6 tabs

  6. 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...

  7. Kaolinitic clay-based grouting demonstration

    International Nuclear Information System (INIS)

    An innovative Kaolinitic Clay-Based Grouting Demonstration was performed under the Mine Waste Technology Program (MWTP), funded by the U.S. Environmental Protection Agency (EPA) and jointly administered by the EPA and the U.S. Department of Energy (DOE). The objective of the technology was to demonstrate the effectiveness of kaolinitic clay-based grouting in reducing/eliminating infiltration of surface and shallow groundwater through fractured bedrock into underground mine workings. In 1993, the Mike Horse Mine was selected as a demonstration site for the field implementation and evaluation of the grouting technology. The mine portal discharge ranged between 114 to 454 liters per minute (30 to 120 gpm) of water containing iron, zinc, manganese, and cadmium at levels exceeding the National Drinking Water Maximum Contaminant Levels. The grout formulation was designed by the developer Morrison Knudsen Corporation/Spetstamponazhgeologia (MK/STG), in May 1994. Grout injection was performed by Hayward Baker, Inc. under the directive of MSE Technology Applications, Inc. (MSE-TA) during fall of 1994. The grout was injected into directionally-drilled grout holes to form a grout curtain at the project site. Post grout observations suggest the grout was successful in reducing the infiltration of the surface and shallow groundwater from entering the underground mine workings. The proceeding paper describes the demonstration and technology used to form the subsurface barrier in the fracture system

  8. VEGETATIVE COVERS FOR WASTE CONTAINMENT

    Science.gov (United States)

    Disposal of municipal ahd hazardous waste in the United States is primarily accomplished by containment in lined and capped landfills. Evapotranspiration cover systems offer an alternative to conventional landfill cap systems. These covers work on completely different principles ...

  9. 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.

  10. 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.

  11. Informative document halogenated hydrocarbon-containing waste

    NARCIS (Netherlands)

    Verhagen H

    1992-01-01

    This "Informative document halogenated hydrocarbon-containing waste" forms part of a series of "Informative documents waste materials". These documents are conducted by RIVM on the instructions of the Directorate General for the Environment, Waste Materials Directorate, in behal

  12. General strategy / clay based concepts

    International Nuclear Information System (INIS)

    This session gathers 4 articles dealing with: the Ontario Power Generation's proposed deep geologic repository for low and intermediate level radioactive waste, Bruce site, Ontario, Canada (M.R. Jensen, F.K. King); the design and realisation of the PRACLAY experimental gallery at the Hades URF (W. Bastiaens, F. Bernier); the extension of the Mont Terri Rock Laboratory and experiment programme (P. Bossart, H.J. Alheid, J. Delay, E. Frank, M. Hugi, K. Kiho, T. Rothfuchs); and the European bentonites as alternatives to MX-80 (D. Koch)

  13. 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...

  14. Ground freezing for containment of hazardous waste

    Energy Technology Data Exchange (ETDEWEB)

    Sayles, F.N.; Iskandar, I.K.

    1998-07-01

    The freezing of ground for the containment of subsurface hazardous waste is a promising method that is environmentally friendly and offers a safe alternative to other methods of waste retention in many cases. The frozen soil method offers two concepts for retaining waste. One concept is to freeze the entire waste area into a solid block of frozen soil thus locking the waste in situ. For small areas where the contaminated soil does not include vessels that would rupture from frost action, this concept may be simpler to install. A second concept, of course, is to create a frozen soil barrier to confine the waste within prescribed unfrozen soil boundaries; initial research in this area was funded by EPA, Cincinnati, OH, and the Army Corps of Engineers. The paper discusses advantages and limitations, a case study from Oak Ridge, TN, and a mesh generation program that simulates the cryogenic technology.

  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. Radioactive waste containment - a literature study

    International Nuclear Information System (INIS)

    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)

  17. 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. PMID:26764134

  18. Direct conversion of halogen-containing wastes to borosilicate glass

    International Nuclear Information System (INIS)

    Glass has become a preferred waste form worldwide for radioactive wastes: however, there are limitations. Halogen-containing wastes can not be converted to glass because halogens form poor-quality waste glasses. Furthermore, halides in glass melters often form second phases that create operating problems. A new waste vitrification process, the Glass Material Oxidation and dissolution System (GMODS), removes these limitations by converting halogen-containing wastes into borosilicate glass and a secondary, clean, sodium-halide stream

  19. Buried waste containment system materials. Final Report

    Energy Technology Data Exchange (ETDEWEB)

    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.

  20. 黏土基氮·磷·钾缓释肥的制备及其释放特征%Preparation of Clay based Slow-release Fertilizers Containing Nitrogen,Phosphorus and Potassium and its Release Characteristics

    Institute of Scientific and Technical Information of China (English)

    罗阳坡; 赵赛锋; 潘国祥; 黄丽芬; 黄登丰; 陈超楠; 顾丽君

    2012-01-01

    [目的]利用黏土作为载体制备黏土基氮、磷、钾缓释肥,是提高化肥利用率的有效途径.[方法]以膨润土和高岭土为载体,采用研磨法制备了多种黏土基氮、磷、钾肥及复合肥,并用淋溶试验评价了肥料的释放性能,得到了氮、磷、钾的释放特征曲线.[结果]相比于传统化学肥料,使用高岭土和膨润土复合的氮、磷、钾肥对氮、磷、钾元素都起到了较好的缓释效果.膨润土对氮、磷、钾的缓释性能要优于高岭土,其原因在于膨润土单元层内电荷不平衡及边缘大量断键的存在,并且其水合膨胀后具有较大的层空间,能够将氮、磷、钾锁定在膨润土层间,不会轻易流失.制得的膨润土基复合肥CLAY-N-P-K中氮、磷、钾都具有较好的缓释性能,就释放速率而言,氮的释放速率最快,钾次之,磷最慢.[结论]研究膨润土基氮、磷、钾肥的缓释特征对缓释肥的生产和使用有重要价值.%[Objective] It was an effective way to improve the utilization efficiency of release fertilizer using clay as a carrier for preparation of nitrogen,phosphorus,and potassium fertilizer. [Method] A variety of clay-based nitrogen,phosphorus,potassium fertilizer and their complex fertilizers were prepared by grinding methods using bentonite and kaolin as supporters. Release characteristic curves of nitrogen,phosphorus and potassium were obtained by the leaching experiments. [Result] Compared with the traditional fertilizers, the release performances of nitrogen, phosphorus and potassium using bentonite and kaolin as supporters were better. Moreover,bentonite was superior to kaolin. The reason was that the charge disbalance within layer sheets of bentonite,and there was a large number of broken bonds in the edge of bentonite,then a large expansion space after hydration, which made nitrogen, phosphorus and potassium were locked in the interlayer of bentonite, and wouldn't lose easily. Nitrogen

  1. Predicting the effects of microbial activity on the corrosion of copper nuclear fuel waste disposal containers

    International Nuclear Information System (INIS)

    Microbially influenced corrosion (MIC) of copper nuclear fuel waste containers may occur in a disposal vault located 500-1000 m underground in the granitic rock of the Canadian Shield. The extent and diversity of microbial activity in the vault is expected to be limited initially because of the aggressive conditions produced by γ-radiation, elevated temperatures and desiccation of the clay-based buffer in which the containers will be embedded. Experimental results on the heat- and radiation-sensitivity of the natural microbiota in buffer material are presented. The data suggest that the low water activity in the buffer material will severely limit the growth of microbes near the container. The most likely form of MIC involves sulphate-reducing bacteria (SRB). Electrochemical experiments using a clay-covered copper electrode have shown that sulphide ions produced by SRB could diffuse through buffer material and induce corrosion of the container. A method to predict the long-term corrosion behaviour is presented. (author)

  2. 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.

  3. Waste container fabrication from recycled DOE metal

    Energy Technology Data Exchange (ETDEWEB)

    Motl, G.P.; Burns, D.D.

    1994-02-15

    The Department of Energy (DOE) has more than 2.5 million tons of radioactive scrap metal (RSM) that is either in inventory or expected to be generated over the next 25 years as major facilities within the weapons complex are decommissioned. Much of this material cannot be surface decontaminated. In an attempt to conserve natural resources and to avoid burial of this material at DOE disposal sites, options are now being explored to {open_quotes}beneficially reuse{close_quotes} this material in applications where small amounts of radioactivity are not a detriment. One example is where RSM is currently being beneficially used to fabricate shield blocks for use in DOE medium energy physics programs. This paper describes other initiatives now underway within DOE to utilize RSM to fabricate other products, such as radioactive waste shipping, storage and disposal containers.

  4. 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.

  5. General strategy, clay based disposal concepts and integration (GSI)

    International Nuclear Information System (INIS)

    This session gathers 20 articles (posters) dealing with: the assessment of backfill materials and methods for deposition tunnels; HTV-1: a semi technical scale testing of a multi-layer hydraulic shaft sealing system; the development of water content adjust method by mixing powdered-ice and chilled bentonite: application to the construction of bentonite engineered barriers by shot-clay method; repository design issues related to the thermal impact induced by heat emitting radioactive waste; pillared clays, using Romanian montmorillonite; the simulation of differential settlements of clay based engineered barrier systems in a geo-centrifuge; the critical issues regarding clay behaviour in the KBS-3H repository design; an alternative buffer material experiment; assessing the performance of a swelling clay tunnel seal and issues identified in the course of its operation; the activation of a Ca-bentonite as buffer material; a large diameter borehole type repository in the clays for radioactive waste long term storage; the erosion of backfill materials during the installation phase; the behaviour of the clay cover of a site for very low level nuclear waste: field flexion tests; the laboratory tests made on three different backfill candidates for the Swedish KBS- 3V concept; the engineering geological clay research for radioactive waste repository in Slovakia; the ESDRED project, module 1 - Design, fabrication, assembly, handling and packaging of buffer rings; the laboratory experiments on the sealing ability of bentonite pellets; the screening of bentonite resources for use as an engineered barrier component in deep geologic repositories; the assessment of the radionuclide release from the near-field environment of a spent nuclear fuel geological repository; and the emplacement tests with granular bentonite

  6. The stress corrosion cracking of copper containers for the disposal of high-level nuclear waste

    International Nuclear Information System (INIS)

    Stress corrosion cracking (SCC) is a possible failure mode for Cu containers in an underground nuclear waste disposal vault. It is difficult to guarantee that SCC will never initiate on Cu containers based only on the results of relatively short-term experiments. Therefore, the extent of SCC is being predicted based on the argument that the rate of crack propagation will be limited. Several environmental factors will limit the rate of cracking, including: the general absence of aggressive SCC agents in the vault, the limited time of rapid strain of the container shell and the limited supply rate of oxidants (principally dissolved O2) to the container surface through the compacted clay-based material in which the containers will be placed. In the first part of this study, the effect of oxidant flux on the crack velocity is being determined. The SCC behavior of two high-copper alloys has been determined in nitrite-containing environments over a range of oxidant fluxes. In NO2- solutions, transgranular SCC is observed. There is evidence for discontinuous crack advance, including crack arrest markings on fracture surfaces and correlated noise events in the electrochemical potential and applied load signals. Crack velocities of 4--8 nm/s are observed in constant extension rate tests in 0.1 mol·dm-3 sodium nitrite (NaNO2) with applied current densities of 0.1--1.0 micro·cm-2. The maximum crack length for a Cu container has been estimated based on the observed dependence of the crack velocity on the oxidant flux and the predicted time dependence of the oxidant flux to the containers in a disposal vault

  7. Development of polymer concrete radioactive waste management containers

    International Nuclear Information System (INIS)

    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)

  8. 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)

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

    OpenAIRE

    I. K. Romanovich; M. I. Balonov; Barkovsky, A.N.

    2016-01-01

    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.

  10. 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.

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

    International Nuclear Information System (INIS)

    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

  12. Predicting the effects of microbial activity on the corrosion of copper nuclear fuel waste disposal containers. AECL research No. AECL-11598

    Energy Technology Data Exchange (ETDEWEB)

    King, F.; Stroes-Gascoyne, S.

    1996-12-31

    Microbially influenced corrosion of copper nuclear fuel waste containers may occur in a disposal vault buried in granitic rock. The extent and diversity of microbial activity in the vault is expected to be limited initially because of the aggressive conditions produced by gamma radiation, elevated temperatures, and desiccation of the clay-based buffer in which the containers will be emplaced. This paper presents new evidence regarding the claim that a virtually sterile zone will be created around the container, and describes experiments studying the effects of remote sulphate-reducing bacteria activity on the long-term corrosion of the container. A method for predicting the consequences for the container lifetime is also presented.

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

    International Nuclear Information System (INIS)

    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

  14. Method for treating waste containing stainless steel

    International Nuclear Information System (INIS)

    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 Fe3O4. Time of exposure to oxygen is controlled so that the iron does not oxidize into Fe2O3. 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

  15. Erosion of clay-based grouts in simulated rock fractures

    International Nuclear Information System (INIS)

    The paper presents a laboratory study on the erosion of clay-based grouts in a simulated rock fracture and in a simulated rock fracture network. The apparatus specially constructed for these experiments and the testing procedure are described. The testing results have shown that a partially eroded clay-based grout may still be effective in sealing rock fractures and that the addition of cement in a clay grout can minimize erosion

  16. New materials for the containment of radioactive wastes

    International Nuclear Information System (INIS)

    Asbestos-cement is a new material that can be used in the containment or storage of radioactive waste, because it can act as intermediate storage for high activity waste dispersed in this material or else be used in the shape of definitive storage containers

  17. Defense High Level Waste Disposal Container System Description

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-10-12

    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

  18. Corrosion of 316L stainless steels MAVL wastes containers

    International Nuclear Information System (INIS)

    The long lived and medium activity wastes are conditioned or could be re-conditioned in primary drums of 316L stainless steels. In the framework of wastes storage, these drums will be placed in concrete containers; each containers would contain one or more drums. This document recalls global information on the corrosion of stainless steels, analyzes specific conditions bond to the drums conditioning in concrete containers and the nature of the wastes, and details the consequences on the possible risks of external and internal corrosion of the drums. (A.L.B.)

  19. Clays in natural and engineered barriers for radioactive waste confinement

    International Nuclear Information System (INIS)

    Andra organised an International Symposium on the use of Natural and Engineered Clay-based Barriers for the Containment of Radioactive Waste hold at the Congress Centre of Tours, France, in March 2005. The symposium provided an opportunity to take stock of the potential properties of the clay-based materials present in engineered or natural barriers in order to meet the containment specifications of a deep geological repository for radioactive waste. It was intended for specialists working in the various disciplines involved with clays and clay based minerals, as well as scientists from agencies and organisations dealing with investigations on the disposal of high-level and long-lived radioactive waste. The themes of the Symposium included geology, geochemistry, transfers of materials, alteration processes, geomechanics, as well as the recent developments regarding the characterisation of clays, as well as experiments in surface and underground laboratories. The symposium consisted of plenary sessions, parallel specialized sessions and poster sessions. (author)

  20. Clays in natural and engineered barriers for radioactive waste confinement

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-07-01

    Andra organised an International Symposium on the use of Natural and Engineered Clay-based Barriers for the Containment of Radioactive Waste hold at the Congress Centre of Tours, France, in March 2005. The symposium provided an opportunity to take stock of the potential properties of the clay-based materials present in engineered or natural barriers in order to meet the containment specifications of a deep geological repository for radioactive waste. It was intended for specialists working in the various disciplines involved with clays and clay based minerals, as well as scientists from agencies and organisations dealing with investigations on the disposal of high-level and long-lived radioactive waste. The themes of the Symposium included geology, geochemistry, transfers of materials, alteration processes, geomechanics, as well as the recent developments regarding the characterisation of clays, as well as experiments in surface and underground laboratories. The symposium consisted of plenary sessions, parallel specialized sessions and poster sessions. (author)

  1. The Robertsfors waste container. Historic and technical documentation

    International Nuclear Information System (INIS)

    This report concerns the so called Robertsfors waste container and its history. The purpose of the report is to contribute to the knowledge about the design of the container and about its radioactive content in order to facilitate the final disposal of the radioactive material. After the general elections in Sweden in 1976 the new government made the start-up of new power reactors conditional on that the owner of the plants could prove that the spent fuel could be disposed of in a safe way. By the mid seventies, the possibility to use ceramic containers for final disposal of high level radioactive waste was identified within the Swedish company ASEA. The ASEA high pressure technology was to be used for the manufacturing and sealing of the containers through hot isostatic pressing. The waste container project was given very high priority by the ASEA management. Due to the political situation, ASEA wanted to do a practical experiment comprising encapsulation of an irradiated fuel rod to prove that ceramic waste containers constituted a viable solution to the waste problem. An experimental fuel rod, length approximately 0.5 m, irradiated for about four years in the Swedish BWR Oskarshamn 1, was chosen for the experiment. The ceramic container was manufactured and sealed at the ASEA high pressure laboratory at Robertsfors in northern Sweden. The Robertsfors container is now temporarily stored in an intermediate storage used for radioactive waste at Studsvik

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

    International Nuclear Information System (INIS)

    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

  3. Controlled Containment, Radioactive Waste Management in the Netherlands

    Energy Technology Data Exchange (ETDEWEB)

    Codee, H.

    2002-02-26

    All radioactive waste produced in The Netherlands is managed by COVRA, the central organization for radioactive waste. The Netherlands forms a good example of a country with a small nuclear power program which will end in the near future. However, radioisotope production, nuclear research and other industrial activities will continue to produce radioactive waste. For the small volume, but broad spectrum of radioactive waste, including TENORM, The Netherlands has developed a management system based on the principles to isolate, to control and to monitor the waste. Long term storage is an essential element of the management system and forms a necessary step in the strategy of controlled containment that will ultimately result in final removal of the waste. Since the waste will remain retrievable for long time new technologies and new disposal options can be applied when available and feasible.

  4. Treatment of actinide-containing organic waste

    International Nuclear Information System (INIS)

    A method has been developed for reducing the volume of organic wastes and recovering the actinide elements. The waste, together with gaseous oxygen (air) is introduced into a molten salt, preferably an alkali metal carbonate such as sodium carbonate. The bath is kept at 7500 - 10000C and 0.5 - 10 atm to thermally decompose and partially oxidize the waste, while substantially reducing its volume. The gaseous effluent, mainly carbon dioxide and water vapour, is vented to the atmosphere through a series of filters to remove trace amounts of actinide elements or particulate alkali metal salts. The remaining combustion products are entrained in the molten salt. Part of the molten salt-combustion product mixture is withdrawn and mixed with an aqueous medium. Insoluble combustion products are then removed from the aqueous medium and are leached with a mixture of hydrofluoric and nitric acids to solubilize the actinide elements. The actinide elements are easily recovered from the acid solution using conventional techniques. (DN)

  5. Plasma processing of carbon-containing technical aggregations and wastes

    Science.gov (United States)

    Cherednichenko, V. S.; An'shakov, A. S.; Faleev, V. A.; Danilenko, A. A.

    2008-12-01

    The plasma gasification of technical aggregations is experimentally studied using the utilization of solid domestic wastes as an example. A shaft electric furnace is described, and the experimental and calculated data are analyzed and compared. The high-temperature gasification of carbon-containing wastes is shown to be a promising process.

  6. Research into Smell Emitted by Containers for Public Waste

    OpenAIRE

    Tadas Lukauskas; Eglė Zuokaitė

    2012-01-01

    Waste is generally accepted as any materials and objects that a holder discards, wants to discard or is required to be discarded. The article deals with the smell of prefabricated containers for household waste produced under normal domestic activities. The paper discusses the advantages and disadvantages of open, shallow and underground Molok containers, installation options and geometric parameters. Research has been conducted referring to air samples taken from three open, three shallow an...

  7. Die Design for Running System of Waste Containers

    OpenAIRE

    Osmel Pérez Acosta; Reinaldo Pérez Sierra; Tania Rodríguez Moliner; Miguel Pérez Sosa

    2014-01-01

    Product deterioration possessing waste containers and their involvement in the collection of solid waste in Cuban cities, the present research is developed in order to make the design of the dies necessary for obtaining system components running of the containers themselves. These systems allow shooting baskets countless repair and revitalization of manufacturing a basket 100 % Cuban. For the design of these dies are taken in account the availability of technology. In this paper, specifically...

  8. Biodegradable containers from green waste materials

    Science.gov (United States)

    Sartore, Luciana; Schettini, Evelia; Pandini, Stefano; Bignotti, Fabio; Vox, Giuliano; D'Amore, Alberto

    2016-05-01

    Novel biodegradable polymeric materials based on protein hydrolysate (PH), derived from waste products of the leather industry, and poly(ethylene glycol) diglycidyl ether (PEG) or epoxidized soybean oil (ESO) were obtained and their physico-chemical properties and mechanical behaviour were evaluated. Different processing conditions and the introduction of fillers of natural origin, as saw dust and wood flour, were used to tailor the mechanical properties and the environmental durability of the product. The biodegradable products, which are almost completely manufactured from renewable-based raw materials, look promising for several applications, particularly in agriculture for the additional fertilizing action of PH or in packaging.

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

    OpenAIRE

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

    2015-01-01

    Little research has been done to determine emissions of engineered nanomaterials (ENM) from currently available nano-enabled consumer products. While ENM release is expected to occur throughout the life cycle of the products, this study focuses on the product end-of-life (EOL) phase. We used 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 individualproducts to an ...

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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 60Co gamma rays and 244Cm 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 137Cs and 90Sr), 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 238Pu and 239Pu will predominate; the hydrogen and oxygen pressures will increase to >200 psi

  14. Assessment of gas flammability in transuranic waste container

    International Nuclear Information System (INIS)

    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

  15. Method of processing radioactive liquid waste containing soduium nitrate

    International Nuclear Information System (INIS)

    Sulfuric acid is added to radioactive liquid wastes containing sodium nitrate and heated to convert sodium nitrate into sodium sulfate and remove nitric acid as fumes. Then, calcium oxide or calcium hydroxide is added to the resultant liquid wastes containing sodium sulfate into a solution of calcium sulfate and sodium hydroxide. Then, solid-liquid separation is applied to take out, as a solid, calcium sulfate containing most portion of radioactive materials. Since no burnable materials such as asphalt are not used as in the prior art method, it is possible, according to the present invention, to reduce the fire hazard and remarkably decrease the formation of solidification products. (S.T.)

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

    International Nuclear Information System (INIS)

    Nuclear industry produces a wide range of low and intermediate level liquid radioactive wastes which can include different radionuclides such as 90Sr. 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 Pu4+ to Pu3+ 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 vitrified. A test was

  17. The change in bioavailability of organic matter associated with clay-based buffer materials as a result of heat and radiation treatment

    International Nuclear Information System (INIS)

    Compacted clay-based buffer surrounds corrosion-resistant waste containers in the Canadian concept for nuclear fuel waste disposal. Clays naturally contain small quantities of organic matter that may be resistant to bacterial degradation. The containers with highly radioactive material would subject the surrounding buffer to both heat and radiation. Both could potentially break down complex organic material to smaller, more bioavailable compounds. This could stimulate microbial growth and possibly affect gas production, microbially-influenced corrosion or radionuclide migration. Experiments were carried out in which buffer was heated at 60 and 90 C for periods of 2, 4 and 6 weeks, in some cases followed by irradiation to 25 kGy. Unheated buffer was also irradiated to 25 and 50 kGy at different moisture contents. The treated materials were subsequently suspended in distilled water, shaken for 24 h and centrifuged to remove the solids. The 0.22 microm filter-sterilized leachates were inoculated with equal volumes of fresh groundwater and incubated at room temperature for 10 d to determine the increase in total and viable bacteria compared to a groundwater control. Results indicated that leachates from buffer subjected to heat, radiation or combinations of these, had a stimulating effect on both total and viable cell counts in groundwater, compared to unamended groundwater controls. This stimulating effect was generally most pronounced for viable counts and could be larger than two orders of magnitude. Leachates from untreated buffer material also stimulated the growth of groundwater bacteria, but to a lesser extent than leachates from heat- and radiation-treated buffer material. The effects of heat and radiation on nutrient availability in clay-based sealing materials should, therefore, be taken into account when attempting to quantify the effects of microbial activity on vault performance

  18. 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.

  19. Radioactive waste containing vessel for underground disposal

    International Nuclear Information System (INIS)

    A canister and an over packing vessel for containing the canister are assembled. Preceding to underground disposal, a sealing medium under a pressurized state is sealed to the space in the overpacking vessel. As the sealing medium, non-compressible medium of a liquid or a fluid such as a mineral oil, vegetable oil, cement mortar is used in an ordinary cases. Alternatively, gases such as of nitrogen or argon are applied for the purpose of increasing pressure of the space portion to more than the pressure of underground water and the like. In the state of the underground disposal, difference of pressure between the external pressure such as of underground water and a pressure of the sealed medium is applied to the wall of the over packing vessel. With such a constitution, since the wall of the over packing vessel can be reduced, the weight and the cost can be reduced. (I.N.)

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

    International Nuclear Information System (INIS)

    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

  1. 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.

  2. Containment of Solid Wastes in some Large Scandinavian Cities

    DEFF Research Database (Denmark)

    Du-Thinh, Kien

    1998-01-01

    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...... 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...

  3. 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.

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

    International Nuclear Information System (INIS)

    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. 14C released from the hot fuel-cladding surface can leak out of penetrations, and air inleakage can mobilize additional 14C 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

  5. Die Design for Running System of Waste Containers

    Directory of Open Access Journals (Sweden)

    Osmel Pérez Acosta

    2014-11-01

    Full Text Available Product deterioration possessing waste containers and their involvement in the collection of solid waste in Cuban cities, the present research is developed in order to make the design of the dies necessary for obtaining system components running of the containers themselves. These systems allow shooting baskets countless repair and revitalization of manufacturing a basket 100 % Cuban. For the design of these dies are taken in account the availability of technology. In this paper, specifically, describes the production of the piece called saucer, emphasizing the design of the die cutting thereof. These are also given the materials used in each of the components.

  6. Russian Containers for Transportation of Solid Radioactive Waste

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-02-28

    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.

  7. In situ containment and stabilization of buried waste

    Energy Technology Data Exchange (ETDEWEB)

    Allan, M.L.; Kukacka, L.E.; Heiser, J.H.

    1992-11-01

    The objective of the project was to develop, demonstrate and implement advanced grouting materials for the in-situ installation of impermeable, durable subsurface barriers and caps around waste sites and for the in-situ stabilization of contaminated soils. Specifically, the work was aimed at remediation of the Chemical Waste (CWL) and Mixed Waste Landfills (MWL) at Sandia National Laboratories (SNL) as part of the Mixed Waste Landfill Integrated Demonstration (MWLID). This report documents this project, which was conducted in two subtasks. These were (1) Capping and Barrier Grouts, and (2) In-situ Stabilization of Contaminated Soils. Subtask 1 examined materials and placement methods for in-situ containment of contaminated sites by subsurface barriers and surface caps. In Subtask 2 materials and techniques were evaluated for in-situ chemical stabilization of chromium in soil.

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

    International Nuclear Information System (INIS)

    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.)

  9. Substance Flow Analysis of Wastes Containing Polybrominated Diphenyl Ethers

    DEFF Research Database (Denmark)

    Vyzinkarova, Dana; Brunner, Paul H.

    2013-01-01

    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...... 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......, vehicles. Most EOL vehicles are exported from Vienna and pose a continental, rather than a local, problem. According to the modeling, approximately 73% of cOctaBDE reached the final sink MSW incinerator, and 17% returned back to consumption by recycling. Secondary plastics, made from WEEE, may thus contain...

  10. Processing method for salt containing radioactive liquid waste

    International Nuclear Information System (INIS)

    A mixed solution of ferrocyanate and copper sulfate is added to salt-containing radioactive liquid wastes, then pH is controlled to 9 to 11, and they are stood still to coprecipitate and separate radioactive nuclides. The precipitated sludges are condensed by evaporation and the resultant condensed liquid wastes are solidified, if necessary, by using asphalts. Further, the coprecipitated and separated supernatants are passed through a filter of activated carbon or a hollow thread membrane for removing remaining radioactive materials. With such procedures, the amount of liquid condensates generated during the evaporation and condensation step is reduced greatly, and the amount of generated solids is reduced also in a case of applying solidification. Further, since iron cruds are precipirated and separated simultaneously with coprecipitation, loads applied to the filter is reduced upon subsequent filtration of the supernatants, thereby enabling to use the filter for a long period of time, and the accompanying generation of wastes is also reduced. (T.M.)

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

    International Nuclear Information System (INIS)

    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 105 and 4.4 x 104 R/hr showed that methane-saturated groundwater was more reducing than identical experiments where Ar was substituted for CH4. Dissolved methane, therefore, may be beneficial to the waste package in limiting the solubility of redox sensitive radionuclides such a 99Tc. 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

  12. 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

  13. WASTE CONTAINER AND WASTE PACKAGE PERFORMANCE MODELING TO SUPPORT SAFETY ASSESSMENT OF LOW AND INTERMEDIATE-LEVEL RADIOACTIVE WASTE DISPOSAL.

    Energy Technology Data Exchange (ETDEWEB)

    SULLIVAN, T.

    2004-06-30

    Prior to subsurface burial of low- and intermediate-level radioactive wastes, a demonstration that disposal of the wastes can be accomplished while protecting the health and safety of the general population is required. The long-time frames over which public safety must be insured necessitates that this demonstration relies, in part, on computer simulations of events and processes that will occur in the future. This demonstration, known as a Safety Assessment, requires understanding the performance of the disposal facility, waste containers, waste forms, and contaminant transport to locations accessible to humans. The objective of the coordinated research program is to examine the state-of-the-art in testing and evaluation short-lived low- and intermediate-level waste packages (container and waste form) in near surface repository conditions. The link between data collection and long-term predictions is modeling. The objective of this study is to review state-of-the-art modeling approaches for waste package performance. This is accomplished by reviewing the fundamental concepts behind safety assessment and demonstrating how waste package models can be used to support safety assessment. Safety assessment for low- and intermediate-level wastes is a complicated process involving assumptions about the appropriate conceptual model to use and the data required to support these models. Typically due to the lack of long-term data and the uncertainties from lack of understanding and natural variability, the models used in safety assessment are simplistic. However, even though the models are simplistic, waste container and waste form performance are often central to the case for making a safety assessment. An overview of waste container and waste form performance and typical models used in a safety assessment is supplied. As illustrative examples of the role of waste container and waste package performance, three sample test cases are provided. An example of the impacts of

  14. Treatment technology analysis for mixed waste containers and debris

    International Nuclear Information System (INIS)

    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

  15. 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.

  16. Mixed Waste Encapsulation in Polyester Resins. Treatment for Mixed Wastes Containing Salts. Mixed Waste Focus Area. OST Reference #1685

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    1999-09-01

    Throughout the Department of Energy (DOE) complex there are large inventories of homogeneous solid mixed wastes, such as treatment residues, fly ashes, and sludges that contain relatively high concentrations (greater than 15% by weight) of salts. The inherent solubility of nitrate, sulfate, and chloride salts makes traditional cement stabilization of these waste streams difficult, expensive, and challenging. Salts can effect the setting rate of cements and can react with cement hydration products to form expansive and cement damaging compounds. Many of these salt wastes are in a dry granular form and are the by-product of treating spent acidic and metal solutions used to recover and reformulate nuclear weapons materials over the past 50 years. At the Idaho National Engineering and Environmental Laboratory (INEEL) alone, there is approximately 8,000 cubic meters of nitrate salts (potassium and sodium nitrate) stored above ground with an earthen cover. Current estimates indicate that over 200 million kg of contaminated salt wastes exist at various DOE sites. Continued primary treatment of waste water coupled with the use of mixed waste incinerators may generate an additional 5 million kg of salt-containing, mixed waste residues each year. One of the obvious treatment solutions for these salt-containing wastes is to immobilize the hazardous components to meet Environmental Protection Agency/Resource Conservation and Recovery Act (EPA/RCRA) Land Disposal Restrictions (LDR), thus rendering the mixed waste to a radioactive waste only classification. One proposed solution is to use thermal treatment via vitrification to immobilize the hazardous component and thereby substantially reduce the volume, as well as provide exceptional durability. However, these melter systems involve expensive capital apparatus with complicated off-gas systems. In addition, the vitrification of high salt waste may cause foaming and usually requires extensive development to specify glass

  17. Clay-based polymer nanocomposites: research and commercial development.

    Science.gov (United States)

    Zeng, Q H; Yu, A B; Lu, G Q; Paul, D R

    2005-10-01

    This paper reviews the recent research and development of clay-based polymer nanocomposites. Clay minerals, due to their unique layered structure, rich intercalation chemistry and availability at low cost, are promising nanoparticle reinforcements for polymers to manufacture low-cost, lightweight and high performance nanocomposites. We introduce briefly the structure, properties and surface modification of clay minerals, followed by the processing and characterization techniques of polymer nanocomposites. The enhanced and novel properties of such nanocomposites are then discussed, including mechanical, thermal, barrier, electrical conductivity, biodegradability among others. In addition, their available commercial and potential applications in automotive, packaging, coating and pigment, electrical materials, and in particular biomedical fields are highlighted. Finally, the challenges for the future are discussed in terms of processing, characterization and the mechanisms governing the behaviour of these advanced materials. PMID:16245517

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

    International Nuclear Information System (INIS)

    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

  19. Corrosion of carbon steel nuclear waste containers in marine sediment

    International Nuclear Information System (INIS)

    The report describes a study of the corrosion of carbon steel nuclear waste containers in deep ocean sediments, which had the objective of estimating the metal allowance needed to ensure that the containers were not breached by corrosion for 1000 years. It was concluded that under such disposal conditions carbon steel would not be subject to localised corrosion or hydrogen embrittlement, and therefore the study concentrated on evaluating the rate of general attack. This was carried out by developing a mechanistically based mathematical model which was formulated on the conservative assumption that the corrosion would be under activation control, and would not be impeded by the formation of corrosion product layers. This model predicted that an allowance of 33 mm would be required for a 1000 year life. (author)

  20. Treatment and recycling of asbestos-cement containing waste

    Energy Technology Data Exchange (ETDEWEB)

    Colangelo, F. [Department of Technology, University Parthenope, Naples (Italy); Cioffi, R., E-mail: raffaele.cioffi@uniparthenope.it [Department of Technology, University Parthenope, Naples (Italy); Lavorgna, M.; Verdolotti, L. [Institute for Biomedical and Composite Materials - CNR, Naples (Italy); De Stefano, L. [Institute for Microelectronics and Microsystems - CNR, Naples (Italy)

    2011-11-15

    Highlights: {yields} Asbestos-cement wastes are hazardous. {yields} High energy milling treatment at room temperature allows mineralogical and morphological transformation of asbestos phases. {yields} The obtained milled powders are not-hazardous. {yields} The inert powders can be recycled as pozzolanic materials. {yields} The hydraulic mortars containing the milled inert powders are good building materials. - Abstract: The remediation of industrial buildings covered with asbestos-cement roofs is one of the most important issues in asbestos risk management. The relevant Italian Directives call for the above waste to be treated prior to disposal on landfill. Processes able to eliminate the hazard of these wastes are very attractive because the treated products can be recycled as mineral components in building materials. In this work, asbestos-cement waste is milled by means of a high energy ring mill for up to 4 h. The very fine powders obtained at all milling times are characterized to check the mineralogical and morphological transformation of the asbestos phases. Specifically, after 120 min of milling, the disappearance of the chrysotile OH stretching modes at 3690 cm{sup -1}, of the main crystalline chrysotile peaks and of the fibrous phase are detected by means of infrared spectroscopy and X-ray diffraction and scanning electron microscopy analyses, respectively. The hydraulic behavior of the milled powders in presence of lime is also tested at different times. The results of thermal analyses show that the endothermic effects associated to the neo-formed binding phases significantly increase with curing time. Furthermore, the technological efficacy of the recycling process is evaluated by preparing and testing hydraulic lime and milled powder-based mortars. The complete test set gives good results in terms of the hydration kinetics and mechanical properties of the building materials studied. In fact, values of reacted lime around 40% and values of compressive

  1. Treatment and recycling of asbestos-cement containing waste

    International Nuclear Information System (INIS)

    Highlights: → Asbestos-cement wastes are hazardous. → High energy milling treatment at room temperature allows mineralogical and morphological transformation of asbestos phases. → The obtained milled powders are not-hazardous. → The inert powders can be recycled as pozzolanic materials. → The hydraulic mortars containing the milled inert powders are good building materials. - Abstract: The remediation of industrial buildings covered with asbestos-cement roofs is one of the most important issues in asbestos risk management. The relevant Italian Directives call for the above waste to be treated prior to disposal on landfill. Processes able to eliminate the hazard of these wastes are very attractive because the treated products can be recycled as mineral components in building materials. In this work, asbestos-cement waste is milled by means of a high energy ring mill for up to 4 h. The very fine powders obtained at all milling times are characterized to check the mineralogical and morphological transformation of the asbestos phases. Specifically, after 120 min of milling, the disappearance of the chrysotile OH stretching modes at 3690 cm-1, of the main crystalline chrysotile peaks and of the fibrous phase are detected by means of infrared spectroscopy and X-ray diffraction and scanning electron microscopy analyses, respectively. The hydraulic behavior of the milled powders in presence of lime is also tested at different times. The results of thermal analyses show that the endothermic effects associated to the neo-formed binding phases significantly increase with curing time. Furthermore, the technological efficacy of the recycling process is evaluated by preparing and testing hydraulic lime and milled powder-based mortars. The complete test set gives good results in terms of the hydration kinetics and mechanical properties of the building materials studied. In fact, values of reacted lime around 40% and values of compressive strength in the range of 2

  2. Activity release from waste packages containing LL and IL waste forms under mechanical and thermal stresses

    International Nuclear Information System (INIS)

    For transport and handling of radioactive waste packages in an underground repository safety assessments are being performed to keep any unacceptable radiation hazards from the operational staff and the population in the site neighborhood. Therefore experiments were carried out to determine source terms for activity release from waste packages containing cemented waste forms in case of heavy mechanical and thermal impacts. Mechanical impact was applied by drop test with a maximum energy input of 3.105 Nm. A special cage construction around the target (reinforced concrete covered by a 80 mm steel plate) allows the collection of the airborne fines with a particle size of < 10 μm by using micro filters in a defined geometry. In addition, in two experiments the particle fraction with an aerodynamic diameter between 1 μm and 20 μm was determined using a cascade impactor. Additional laboratory experiments were performed to determine comparative values for different waste forms. In case of thermal impact, the temperature profiles in the waste forms were measured and the release of added indicators (Cs, Sr, Eu) was determined. Further laboratory experiments were performed with inactive samples to determine the temperature dependence of water release (Thermogravimetric-Analysis)

  3. 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.

  4. Immobilization of calcium sulfate contained in demolition waste

    International Nuclear Information System (INIS)

    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

  5. Stress corrosion cracking of candidate waste container materials

    International Nuclear Information System (INIS)

    Six alloys have been selected as candidate container materials for the storage of high-level nuclear waste at the proposed Yucca Mountain site in Nevada. These materials are Type 304L stainless steel (SS), Type 316L SS, Incology 825, P-deoxidized Cu, Cu-30%Ni, and Cu-7% Al. The present program has been initiated to determine whether any of these materials can survive for 300 years in the site environment without developing through-wall stress corrosion cracks, and to assess the relative resistance of these materials to stress corrosion cracking (SCC). A series of slow-strain-rate tests (SSRTs) in simulated Well J-13 water which is representative of the groundwater present at the Yucca Mountain site has been completed, and crack-growth-rate (CGR) tests are also being conducted under the same environmental conditions. 13 refs., 60 figs., 22 tabs

  6. Waste injection risk identification: keys to control waste containment and procure a safe waste injection operation

    Energy Technology Data Exchange (ETDEWEB)

    Ovalle, Adriana P.; Ronderos, Julio R. [M-I SWACO, Houston, TX (United States); Francisco, Francisco F.

    2008-07-01

    As the world faces new challenges to protect the environment from all human-generated wastes, self-imposed industry standards as well as governmental regulations support new green politics to prevent environmental damage due to spillage during the course of operations. As such, the oil industry produces wastes from the drilling and production phases which ultimately are required to be disposed of in a safe manner. Waste Injection (WI) has been selected as the sound engineering and cost-effective final disposal methodology by many operators and legislators based on the capability to achieve zero discharge in a safe and efficient manner when compared to other existing proven technologies. This is particularly true for large-scale projects where WI has been strategically implemented as an integral component in field developments because of the commitment to the environment and the acceptance of subsurface engineering by local legislation. With the view of an assured process, the project development and implementation of WI technology is carefully designed using risk-based analysis that comprehends fracturing studies of the area of injection, logistics, equipment specification, and operation monitoring with the objective to perform a seamless and risk-free job. This paper addresses WI planning and implementation methodology and cites real examples to demonstrate the value of proper preparation of the injection operation to attain maximum efficiency under QHSE standards. (author)

  7. Long term chemical durability studies of vitrified waste products containing sulphate bearing high level radioactive waste

    International Nuclear Information System (INIS)

    Evaluation of the term durability of the vitrified waste product (VWP) is of paramount importance for ascertaining safe containment of radionuclide immobilized in the matrix, because leaching is the principle mechanism through which radionuclide can migrate to human environment. Sodium released out from the glass was taken as the index element to examine the leach rate as a function of time. Average leach rate of VWPs based on barium borosilicate glass matrix immobilizing sulphate bearing HLW is 2.32'10-6 g.cm-2, day-1 after a period of 710 days at 373 deg K using demineralised water as leachant indicating adequate leach resistance of the conditioned product. The paper presented here describes the outcome of the work carried out for studying long term chemical durability of the vitrified waste period. (author)

  8. 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.

  9. Disposal of phosphogypsum waste containing enhanced levels of radioactivity

    International Nuclear Information System (INIS)

    Full text: From production of phosphoric acid based on the reaction of phosphorite with sulphuric acid, manufacturers either have released the phosphogypsum containing uranium decay products into the aquatic environment or have stockpiled the phosphogypsum on land. In Portugal two factories have produced phosphoric acid by this wet chemistry method during several decades, from the 30s till the late 80s. The radioactivity remaining in the phosphogypsum depends upon the composition of the raw material used and upon the efficiency of the chemical reaction method. In one factory, using mainly phosphorites from Syria and Tunisia, 226Ra concentrations in the gypsum were at about 600 Bq kg-1 and at about the same level for 210Pb and 210Po. In another factory, using mainly phosphorites imported from Morocco, radionuclide concentrations in gypsum were higher, at about 1000 Bq kg-1 for the same radionuclides. Phosphogypsum waste stockpiled on land and uncovered may undergo weathering, including the slow dissolution of calcium sulphate by rain water. This process may be accompanied with partial dissolution of 226Ra, which leaches from the stockpiles, whereas the less soluble 210Pb and 210Po nuclides may remain in the gypsum. In one place, the stockpiles of phosphogypsum have been exposed in the open air for years until recent coverage with soil and vegetation. This remedial action to confine the phosphogypsum have reduced surface runoff, radium leaching and waste disposal. It may have contributed also to reduce radon emanation. In another site, the gypsum stockpiles are still uncovered in the open air. The disposal site was a former salt evaporation basin with compact, highly impermeable, fine grained grounds. The gypsum stockpiles are surrounded by ditches to retain rain water drainage. In the water accumulated in the ditches high concentrations of 226Ra were measured as well as relatively high concentrations of 210Pb and 210Po, although these ones associated mainly to

  10. Biofilm treatment of soil for waste containment and remediation

    Energy Technology Data Exchange (ETDEWEB)

    Turner, J.P.; Dennis, M.L.; Osman, Y.A.; Chase, J.; Bulla, L.A. [Univ. of Wyoming, Laramie, WY (United States)

    1997-12-31

    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{sub sat} using a flexible-wall permeameter. Saturated hydraulic conductivity (k{sub sat}) was reduced from 1 x 10{sup -5} cm/sec to 2 x 10{sup -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{sub 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.

  11. [Pyrolysis characteristics of medical waste compositions containing PVC (polyvinyl chloride)].

    Science.gov (United States)

    Deng, Na; Zhang, Yu-Feng; Zhao, Wei; Ma, Hong-Ting; Wei, Li-Li

    2008-03-01

    To obtain pyrolysis characteristics of medical waste compositions containing PVC (polyvinyl chloride), thermogravimetric study of tube for transfusion (TFT) and sample collector for urine (SCFU) was carried out using the thermogravimetric analyser (TGA) with N2. The heat change in pyrolysis process was analyzed and the properties of pyrolysis residues are reported. The mathematics model with two-step and four-reaction was established to simulate the pyrolysis process. The results show that: 1) The pyrolysis mechanism of the two samples is in agreement with that of PVC. The decomposition process appears two stages in 200 - 390 degrees C and 390 - 550 degrees C, which are clearly expressed with two prominent peaks with maximum rate of weight loss at about 315 degrees C and 470 degrees C. 2) Complex ingredients in samples result in irregular and uneven shape of DTG peaks, in which plasticizer lowers the antichloration temperature and enhances the weight loss rate. 3) The model could satisfactorily describe the weight loss and differential process of TFT and SCFU.

  12. Method for calcining nuclear waste solutions containing zirconium and halides

    Science.gov (United States)

    Newby, Billie J.

    1979-01-01

    A reduction in the quantity of gelatinous solids which are formed in aqueous zirconium-fluoride nuclear reprocessing waste solutions by calcium nitrate added to suppress halide volatility during calcination of the solution while further suppressing chloride volatility is achieved by increasing the aluminum to fluoride mole ratio in the waste solution prior to adding the calcium nitrate.

  13. Remote automated material handling of radioactive waste containers

    International Nuclear Information System (INIS)

    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

  14. Application of solid waste containing lead for gamma ray shielding material

    OpenAIRE

    SARAEE, Rezaee Ebrahim; POURAJAM BAFERANI, S.; TAHMASEBI, O.

    2015-01-01

    Abstract. The basic strategies to decrease solid waste disposal problems have focused on the reduction of waste production and recovery of usable materials using waste and making raw materials. Generally, various materials have been used for radiation shielding in different areas and situations. In this study, a novel shielding material produced by a metallurgical solid waste containing lead has been analyzed in order to make a shielding material against gamma radiation. The photon total mass...

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Daniel M. Dabbs; Ilhan A. Aksay

    2005-01-12

    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.

  17. Considerations in estimating corrosion of metallic containers in nuclear waste repositories

    International Nuclear Information System (INIS)

    Metallic containers for high-level nuclear waste are expected to isolate waste from the repository environment for at least 1000 years. Forms of corrosive attack that could lead to premature failure of the containers and some of the difficulties in predicting corrosion behavior in repositories over long periods of time are discussed

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

    International Nuclear Information System (INIS)

    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.

  19. FY 1996 solid waste integrated life-cycle forecast container summary volume 1 and 2

    International Nuclear Information System (INIS)

    For the past six years, a waste volume forecast has been collected annually from onsite and offsite generators that currently ship or are planning to ship solid waste to the Westinghouse Hanford Company's Central Waste Complex (CWC). This document provides a description of the containers expected to be used for these waste shipments from 1996 through the remaining life cycle of the Hanford Site. In previous years, forecast data have been reported for a 30-year time period; however, the life-cycle approach was adopted this year to maintain consistency with FY 1996 Multi-Year Program Plans. This document is a companion report to the more detailed report on waste volumes: WHC-EP0900, FY 1996 Solid Waste Integrated Life-Cycle Forecast Volume Summary. Both of these documents are based on data gathered during the FY 1995 data call and verified as of January, 1996. These documents are intended to be used in conjunction with other solid waste planning documents as references for short and long-term planning of the WHC Solid Waste Disposal Division's treatment, storage, and disposal activities over the next several decades. This document focuses on the types of containers that will be used for packaging low-level mixed waste (LLMW) and transuranic waste (both non-mixed and mixed) (TRU(M)). The major waste generators for each waste category and container type are also discussed. Containers used for low-level waste (LLW) are described in Appendix A, since LLW requires minimal treatment and storage prior to onsite disposal in the LLW burial grounds. The FY 1996 forecast data indicate that about 100,900 cubic meters of LLMW and TRU(M) waste are expected to be received at the CWC over the remaining life cycle of the site. Based on ranges provided by the waste generators, this baseline volume could fluctuate between a minimum of about 59,720 cubic meters and a maximum of about 152,170 cubic meters

  20. 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

  1. Iron Phosphate Glass-Containing Hanford Waste Simulant

    Energy Technology Data Exchange (ETDEWEB)

    Sevigny, Gary J.; Kimura, Marcia L.; Fischer, Christopher M.; Schweiger, M. J.; Rodriguez, Carmen P.; Kim, Dong-Sang; Riley, Brian J.

    2012-01-18

    Resolution of the nation's high-level tank waste legacy requires the design, construction, and operation of large and technically complex one-of-a-kind processing waste treatment and vitrification facilities. While the ultimate limits for waste loading and melter efficiency have yet to be defined or realized, significant reductions in glass volumes for disposal and mission life may be possible with advancements in melter technologies and/or glass formulations. This test report describes the experimental results from a small-scale test using the research-scale melter (RSM) at Pacific Northwest National Laboratory (PNNL) to demonstrate the viability of iron-phosphate-based glass with a selected waste composition that is high in sulfate (4.37 wt% SO3). The primary objective of the test was to develop data to support a cost-benefit analysis related to the implementation of phosphate-based glasses for Hanford low-activity waste (LAW) and/or other high-level waste streams within the U.S. Department of Energy complex. The testing was performed by PNNL and supported by Idaho National Laboratory, Savannah River National Laboratory, Missouri University of Science and Technology, and Mo-Sci Corporation.

  2. Iron Phosphate Glass-Containing Hanford Waste Simulant

    Energy Technology Data Exchange (ETDEWEB)

    Sevigny, Gary J.; Kimura, Marcia L.; Fischer, Christopher M.; Schweiger, Michael J.; Kim, Dong-Sang

    2011-08-01

    Resolution of the nation’s high level tank waste legacy requires the design, construction, and operation of large and technically complex one-of-a-kind processing waste treatment and vitrification facilities. While the ultimate limits for waste loading and melter efficiency have yet to be defined or realized, significant reductions in glass volumes for disposal and mission life may be possible with advancements in melter technologies and/or glass formulations. This test report describes the experimental results from a small-scale test using the research scale melter (RSM) at Pacific Northwest National Laboratory (PNNL) to demonstrate the viability of iron phosphate-based glass with a selected waste composition that is high in sulfates (4.37 wt% SO3). The primary objective of the test was to develop data to support a cost-benefit analysis as related to the implementation of phosphate-based glasses for Hanford low activity waste (LAW) and/or other high-level waste streams within the U.S. Department of Energy complex. The testing was performed by PNNL and supported by Idaho National Laboratory, Savannah River National Laboratory, and Mo-Sci Corporation.

  3. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-09-30

    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).

  5. Treatment of Pu-containing waste by acid digestion (wet combustion)

    International Nuclear Information System (INIS)

    Acid digestion as a process of treatment of plutonium-containing solid waste was developed and demonstrated under conditions of an active operation with respect to the recovery of plutonium. The process composes the following main steps: waste shredding, waste carbonisation, waste oxidation and conversion of plutonium oxide to plutonium sulphate, off-gas treatment, acid recovery and plutonium separation. The technical, safety and operational details of the plant will be presented. Furthermore, methods of the purification of separate plutonium and solidification of secondary waste for final disposal will be described. (orig./RW)

  6. 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.

  7. Properties of lightweight cement-based composites containing waste polypropylene

    Science.gov (United States)

    Záleská, Martina; Pavlíková, Milena; Pavlík, Zbyšek

    2016-07-01

    Improvement of buildings thermal stability represents an increasingly important trend of the construction industry. This work aims to study the possible use of two types of waste polypropylene (PP) for the development of lightweight cement-based composites with enhanced thermal insulation function. Crushed PP waste originating from the PP tubes production is used for the partial replacement of silica sand by 10, 20, 30, 40 and 50 mass%, whereas a reference mixture without plastic waste is studied as well. First, basic physical and thermal properties of granular PP random copolymer (PPR) and glass fiber reinforced PP (PPGF) aggregate are studied. For the developed composite mixtures, basic physical, mechanical, heat transport and storage properties are accessed. The obtained results show that the composites with incorporated PP aggregate exhibit an improved thermal insulation properties and acceptable mechanical resistivity. This new composite materials with enhanced thermal insulation function are found to be promising materials for buildings subsoil or floor structures.

  8. Waste treatment of fission product solutions containing aluminium nitrate

    International Nuclear Information System (INIS)

    In the Rossendorf molybdenum-99 production facility AMOR short-term irradiated aluminium clad fuel elements from the Rossendorf Research Reactor are reprocessed. Following extractive recovery of the enriched uranium the facility system has to be disposed of the fission product-Al(NO3)3 solution. Investigations on waste conditioning of such solutions are presented. (author)

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

    International Nuclear Information System (INIS)

    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

  10. 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.

  11. Design and testing of Type A containers for packaging radioactive waste. Revision 1

    International Nuclear Information System (INIS)

    The Toxic Waste Control Group at the Lawrence Livermore National Laboratory tested numerous Type A containers for use in the shipping of retrievable and disposable radioactive waste, specifically Transuranic waste, to identify and adopt a container that meets test criteria established by the Department of Transportation (49 CFR 173.398). This report summarizes the test results. Several containers passed DOT tests, but were unacceptable for use because of cost, maneuverability, size or shape, weight, or potential fire hazard during closure. The TX-4 passed all DOT tests and met LLNL requirements for handling, safety, and cost

  12. The Robertsfors waste container. Historic and technical documentation; Robertsforsbehaallaren. Historisk och teknisk dokumentation

    Energy Technology Data Exchange (ETDEWEB)

    Vaernild, Ola (OV Konsult, Vaesteraas (Sweden))

    2008-07-01

    This report concerns the so called Robertsfors waste container and its history. The purpose of the report is to contribute to the knowledge about the design of the container and about its radioactive content in order to facilitate the final disposal of the radioactive material. After the general elections in Sweden in 1976 the new government made the start-up of new power reactors conditional on that the owner of the plants could prove that the spent fuel could be disposed of in a safe way. By the mid seventies, the possibility to use ceramic containers for final disposal of high level radioactive waste was identified within the Swedish company ASEA. The ASEA high pressure technology was to be used for the manufacturing and sealing of the containers through hot isostatic pressing. The waste container project was given very high priority by the ASEA management. Due to the political situation, ASEA wanted to do a practical experiment comprising encapsulation of an irradiated fuel rod to prove that ceramic waste containers constituted a viable solution to the waste problem. An experimental fuel rod, length approximately 0.5 m, irradiated for about four years in the Swedish BWR Oskarshamn 1, was chosen for the experiment. The ceramic container was manufactured and sealed at the ASEA high pressure laboratory at Robertsfors in northern Sweden. The Robertsfors container is now temporarily stored in an intermediate storage used for radioactive waste at Studsvik

  13. 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...

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-18

    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.

  16. Early detection and evaluation of waste through sensorized containers for a collection monitoring application

    International Nuclear Information System (INIS)

    The present study describes a novel application for use in the monitoring of municipal solid waste, based on distributed sensor technology and geographical information systems. Original field testing and evaluation of the application were carried out in Pudong, Shanghai (PR China). The local waste management system in Pudong features particular requirements related to the rapidly increasing rate of waste production. In view of the fact that collected waste is currently deployed to landfills or to incineration plants within the context investigated, the key aspects to be taken into account in waste collection procedures include monitoring of the overall amount of waste produced, quantitative measurement of the waste present at each collection point and identification of classes of material present in the collected waste. The case study described herein focuses particularly on the above mentioned aspects, proposing the implementation of a network of sensorized waste containers linked to a data management system. Containers used were equipped with a set of sensors mounted onto standard waste bins. The design, implementation and validation procedures applied are subsequently described. The main aim to be achieved by data collection and evaluation was to provide for feasibility analysis of the final device. Data pertaining to the content of waste containers, sampled and processed by means of devices validated on two purpose-designed prototypes, were therefore uploaded to a central monitoring server using GPRS connection. The data monitoring and management modules are integrated into an existing application used by local municipal authorities. A field test campaign was performed in the Pudong area. The system was evaluated in terms of real data flow from the network nodes (containers) as well as in terms of optimization functions, such as collection vehicle routing and scheduling. The most important outcomes obtained were related to calculations of waste weight and

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

    International Nuclear Information System (INIS)

    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)

  18. WESTERN RESEARCH INSTITUTE CONTAINED RECOVERY OF OILY WASTES (CROW) PROCESS - ITER

    Science.gov (United States)

    This report summarizes the findings of an evaluation of the Contained Recovery of Oily Wastes (CROW) technology developed by the Western Research Institute. The process involves the injection of heated water into the subsurface to mobilize oily wastes, which are removed from the ...

  19. 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.

  20. The Al-containing wastes technology of recycling for alumina, coagulants and building materials production

    Institute of Scientific and Technical Information of China (English)

    2005-01-01

    @@ The Al-containing wastes are generated by a row of industrial plants as hydroalumocarbonate residuum, underwastes water, foundry slag, mud, catalysts, mineral part of coals and others. These wastes is cycling in technological processes that cause to extra energy costs, processes stages difficulties and negatively affecting to environment.

  1. The Al-containing wastes technology of recycling for alumina, coagulants and building materials production

    Institute of Scientific and Technical Information of China (English)

    Lainer; U.; A.; Tuzhilin; A.; S.; Perekhoda; S.; P.; Vetchinkina; T.; N.; Samoilov; E.; N.

    2005-01-01

    The Al-containing wastes are generated by a row of industrial plants as hydroalumocarbonate residuum, underwastes water, foundry slag, mud, catalysts, mineral part of coals and others. These wastes is cycling in technological processes that cause to extra energy costs, processes stages difficulties and negatively affecting to environment.……

  2. 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...

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

    Energy Technology Data Exchange (ETDEWEB)

    Ulm, Franz-Josef

    2000-03-31

    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)

  4. Treatment and recycling of asbestos-cement containing waste.

    Science.gov (United States)

    Colangelo, F; Cioffi, R; Lavorgna, M; Verdolotti, L; De Stefano, L

    2011-11-15

    The remediation of industrial buildings covered with asbestos-cement roofs is one of the most important issues in asbestos risk management. The relevant Italian Directives call for the above waste to be treated prior to disposal on landfill. Processes able to eliminate the hazard of these wastes are very attractive because the treated products can be recycled as mineral components in building materials. In this work, asbestos-cement waste is milled by means of a high energy ring mill for up to 4h. The very fine powders obtained at all milling times are characterized to check the mineralogical and morphological transformation of the asbestos phases. Specifically, after 120 min of milling, the disappearance of the chrysotile OH stretching modes at 3690 cm(-1), of the main crystalline chrysotile peaks and of the fibrous phase are detected by means of infrared spectroscopy and X-ray diffraction and scanning electron microscopy analyses, respectively. The hydraulic behavior of the milled powders in presence of lime is also tested at different times. The results of thermal analyses show that the endothermic effects associated to the neo-formed binding phases significantly increase with curing time. Furthermore, the technological efficacy of the recycling process is evaluated by preparing and testing hydraulic lime and milled powder-based mortars. The complete test set gives good results in terms of the hydration kinetics and mechanical properties of the building materials studied. In fact, values of reacted lime around 40% and values of compressive strength in the range of 2.17 and 2.29 MPa, are measured. PMID:21924550

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

    International Nuclear Information System (INIS)

    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)

  6. Effect of Heat Treatment Temperature on Properties of Chinese Calcined Flint Clay Based Plastic Refractories

    Institute of Scientific and Technical Information of China (English)

    ZHANG Wei; DAI Wenyong; YU Xinfeng; LI Liang

    2009-01-01

    Effects of different heat treatment temperatures on properties of Chinese calcined flint clay based plastic refractories were investigated using Chinese calcined flint clay as starting material,aluminum sulfate and fireclay as binding system.The results showed that with temperature rising,Chinese calcined flint clay based plastic refractories shrinked firstly and then expanded.The modulus of rupture (MOR) and the cold crushing strength (CCS) increased firstly and then decreased from 110 ℃ to 600 ℃,then increased obviously.Thermal expansion coefficient increased from 110 ℃ to 760 ℃,decreased from 760 ℃ to 1 300 ℃,and increased from 1 300 ℃ to 1 500 ℃.

  7. Applicability of insoluble tannin to treatment of waste containing americium

    International Nuclear Information System (INIS)

    The applicability of insoluble tannin adsorbent to the treatment of aqueous waste contaminated with americium has been investigated. Insoluble tannin is considered highly applicable because it consists of only carbon, hydrogen and oxygen and so its volume can be easily reduced by incineration. This report describes measurements of the americium distribution coefficient in low concentration nitric acid. The americium distribution coefficients were found to decrease with increasing concentration of nitric acid and sodium nitrate, and with increasing temperature. At 25 C in 2.0 x 10-3 M HNO3, the distribution coefficient was found to be 2000 ml g-1. The adsorption capacity was determined by column experiments using europium as a simulant of americium, and found to be 7 x 10-3 mmol g-1-dried tannin in 0.01 M HNO3 at 25 C, which corresponds to approximately 1.7 mg-241Am/g-adsorbent(dried). The prospect of applying the adsorbent to the treatment of aqueous waste contaminated with americium appears promising. (orig.)

  8. 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.)

  9. 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.

  10. Method of processing waste water containing actinide element by fixed tannin

    International Nuclear Information System (INIS)

    Since the waste water from a nuclear power plant is generally in an alkaline range above pH 8, in the case of processing waste water by using fixed tannin, fixed tannin is partially leached to unstable, and the adsorbing elimination rate of actinide elements contained in waste water is decreased. Accordingly, the fixed tannin is immersed and brought into contact with an aqueous solution of ammonia for pre-treatment. it is necessary that the pH value of the aqueous solution of ammonia used is higher than that of the waste water containing the actinide elements, and preferably, within a range of 10 to 12. The time of contact for ensuring the effect of the pre-treatment is at least 30 minutes for the lower limit and 60 minutes for the upper limit. In this way, the adsorbing performance itself can be improved and the processing performance for radioactive waste water can be improved. (T.M.)

  11. Clay-based grout injection in crystalline rock

    International Nuclear Information System (INIS)

    In the sealing of an underground disposal facilities for the high-level radioactive waste, a concept of the clay grouting in the sealing of the underground facilities applied to the hard rock is summarized, based on the results of clay grouting experiments Japan Nuclear Cycle Development Institute (JNC) has performed. JNC performed the clay grouting experiments in-situ of the hard rock. In the experiments, clay grout slurry was injected to the fractures on the floor of the test tunnel and to the excavated damage zone around the key cut off the excavated damage zone along the tunnel. Through the results of these experiments, the injected grout slurry to the target excavated damage zone area improved the hydraulic conductivity of the target area using the injection boreholes opened from the wall of the tunnel. Regarding the adequate design of the clay grouting in the hard rock, information of the fracture characterization (scale and distribution), distribution of the excavated damage zone (hydraulic characteristics), selection of the clay material, injection technique, target area of the injection of the grout (position and region) and so on is required. (author)

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

    Science.gov (United States)

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

    2014-09-01

    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 Fe3+/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.

  13. Container and waste pile standards for owners and operators of hazardous waste facilities: consolidated permit regulations--Environmental Protection Agency. Amendments to interim final rule.

    Science.gov (United States)

    1981-11-01

    The Environmental Protection Agency (EPA) is today promulgating amendments to the hazardous waste management regulations regarding the management of hazardous waste in containers and piles and associated permit regulations (40 CFR Part 264, Subparts I and L, and Part 122, Subpart B). These amendments better tailor the standards to the particular type of hazard posed by specific situations. The standards for containers are amended to waive the containment system requirements for wastes that do not contain free liquids, provided that the wastes are protected from contact with accumulated liquid. The standards for waste piles are amended to waive the containment system requirements for wastes that do not contain free liquids, provided that the pile is protected from precipitation by a structure and from surface water run-on and wind dispersal of the waste by the structure or some other means. The Agency believes these amendments believes these amendments will not reduce the level of protection of human health and the environment.

  14. Radionuclide transport through penetrations in nuclear waste containers

    International Nuclear Information System (INIS)

    Penetrations may result from corrosion or cracking and may be through the container material or through deposits of corrosion products. The analysis deals with the resultant radionuclide transport, but not with how these penetrations occur. We provide numerical illustrations for diffusive nuclide flux through these apertures from mathematical expressions. 2 refs., 2 figs

  15. 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.

  16. 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)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

  19. 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

  20. Poly-urea spray elastomer for waste containment applications

    International Nuclear Information System (INIS)

    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

  1. Poly-urea spray elastomer for waste containment applications

    Energy Technology Data Exchange (ETDEWEB)

    Miller, C.J. [Wayne State Univ., Detroit, MI (United States); Cheng, S.C.J. [Drexel Univ., Philadelphia, PA (United States); Tanis, R. [Foamseal, Lapeer, MI (United States)

    1997-12-31

    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. Application of service examinations to transuranic waste container integrity at the Hanford Site

    International Nuclear Information System (INIS)

    Transuranic waste containers in retrievable storage trenches at the Hanford Site and their storage environment are described. The containers are of various types, predominantly steel 0.21-m3 (55-gal) drums and boxes of many different sizes and materials. The storage environment is direct soil burial and aboveground storage under plastic tarps with earth on top of the tarps. Available data from several transuranic waste storage sites are summarized and degradation rates are projected for containers in storage at the Hanford Site

  3. Application of service examinations to transuranic waste container integrity at the Hanford Site. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Duncan, D.R.; Burbank, D.A. Jr.; Anderson, B.C.; Demiter, J.A.

    1993-09-01

    Transuranic waste containers in retrievable storage trenches at the Hanford Site and their storage environment are described. The containers are of various types, predominantly steel 0.21-m{sup 3} (55-gal) drums and boxes of many different sizes and materials. The storage environment is direct soil burial and aboveground storage under plastic tarps with earth on top of the tarps. Available data from several transuranic waste storage sites are summarized and degradation rates are projected for containers in storage at the Hanford Site.

  4. Synthesis of studies on common ELD storage containers for the MA-VL wastes

    International Nuclear Information System (INIS)

    The aim of this document is to present the results of different studies realized on the common container during long time storage and during the deep underground disposal. These studies are realized in the framework of the 2 and 3 axis of the law on the radioactive wastes management of 1991. The common container is the external envelop collecting many primary wastes packages. The results are presented in eight chapters: the initial data, the main functions of the container, the based options for the sizing, the constitutive concrete, the functional demonstrations, the technological demonstrations, the old packages examinations, the transport analysis. (A.L.B.)

  5. A novel clay-based catalytic material: Preparation and properties

    Energy Technology Data Exchange (ETDEWEB)

    Lussier, R.J. (W.R. Grace Co.-Conn., Baltimore, Md. (USA))

    1991-05-01

    A novel acid-leached calcined laolin has been prepared by careful control of the calcination and acid leach conditions. A narrow calcination window gives an extremely acid-reactive calcined kaolin, which develops high surface areas at a rate much faster than that of samples calcined outside this range. This more acid active calcined kaolin also allows the use of extremely low levels of acid, which results in most of the alumina being in the solid phase during the entire leach step. Al{sup 27} NMR results indicate that most acid-reactive calcined clay has the lowest level of octahedral and the highest level of five-coordinate Al. Acids containing anions that do not complex with aluminum such as hydrochloric, nitric, or aluminum chloride work in this process, while acids containg anions that complex with aluminum such a sulfuric or phosphoric do not lead to the same high surface area, catalytically active products. Properly calcined and leached materials show a broad distribution of pores centered at about 40 (angstrom).

  6. 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... CFR parts 173, 178, and 179), if those regulations specify a particular inside container for the waste... hazardous waste in overpacked drums (lab packs). 264.316 Section 264.316 Protection of...

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Bullen, D.B.; Gdowski, G.E. (Science and Engineering Associates, Inc., Pleasanton, CA (USA)); Weiss, H. (Lawrence Livermore National Lab., CA (USA))

    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.

  9. Methods for treating and conditioning of 14C containing health care waste

    International Nuclear Information System (INIS)

    Health care radioactive waste was previously accepted at Necsa and disposed of on this site in near-surface trenches. This practice was terminated by the regulator during 1997 and since then waste drums have been stored and have now become a Necsa liability. These waste drums containing unknown quantities of 14C. About 2500 drums have been accumulated over the years at the Necsa site. The 14C and 3H contents could not be determined with non-destructive assay methods. A study to minimize the further accumulation of 14C containing health care waste was undertaken and some new regulations implemented to prevent further increase of the liability.The bio-hazardous nature of the waste proved to be the main complication in the development of appropriate characterization and conditioning methods. Possible methods to sterilize the waste as a first step were consequently investigated, and this regards two interesting options received attention. The first was the so-called Stericycle ETD process, during which the waste is shredded in an enclosed environment and then sterilized by means of a technique known as Electro Thermal De-activation, and the second was sterilization with Gamma rays. The latter method had the advantage that shredding and repacking were not required.Once the waste was sterilized the waste could be characterized. The most practical method to do this was to compact the drum in a supercompactor and to analyze the liquid released from the drum during compaction in a laboratory.Reasonably accurate estimates of the 14C contents of the waste packages were obtained in this way and at the same time the waste volume to be disposed of was reduced by at least a factor of four. The option to dispose of the waste without doing any quantification of the 14C was also investigated. This option does not require the waste drums to be opened and therefore no sterilization is required. Characterization is in this case limited to assaying the drums for nuclides that can be

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

    Energy Technology Data Exchange (ETDEWEB)

    Fang, H.Y.; Daniels, J.L.; Inyang, H.I. [Univ. of Massachusetts, Lowell, MA (United States)

    1997-12-31

    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. Flexible process options for the immobilisation of residues and wastes containing plutonium

    International Nuclear Information System (INIS)

    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)

  12. Influence of Nitrogen Containing Wastes Addition on Natural Aerobic Composting of Rice Straw

    OpenAIRE

    Thaniya Kaosol; Suchinun Kiepukdee; Prawit Towatana

    2012-01-01

    Problem statement: Rice straw is an agricultural residue. Typically, the rice straw can be burn in the rice field after the harvesting process. The burning can cause air pollution. Another alternative rice straw management method is animal feed. The amount of rice straw is enormus in Thailand. Another sustainable way to manage rice straw is required. Rice straw is used as main waste to compost with nitrogen containing wastes such as golden apple snail, cattle dung and urea in natural aerobic ...

  13. Process for the storage of borate containing radioactive wastes by vitrification

    International Nuclear Information System (INIS)

    For storage of radioactive waste by vitrification the radioactive waste concentrates from borate-containing liquids are mixed with glass-forming aggregates. The borates make up a major part of the glass product. A glass product with good chemical and physical properties for storage is produced by heating to produce a glass-forming melt. Lead oxides and silicates in particular are considered suitable aggregate materials. (orig.)

  14. 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.

  15. Position for determining gas phase volatile organic compound concentrations in transuranic waste containers. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Connolly, M.J.; Liekhus, K.J. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States); Djordjevic, S.M.; Loehr, C.A.; Spangler, L.R. [Benchmark Environmental Corp., Albuquerque, NM (United States)

    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.

  16. Report for slot cutter proof-of-principle test, Buried Waste Containment System project. Revision 1

    International Nuclear Information System (INIS)

    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

  17. 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.

  18. 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.

  19. Calculational technique to predict combustible gas generation in sealed radioactive waste containers

    International Nuclear Information System (INIS)

    Certain forms of nuclear waste, when subjected to ionizing radiation, produce combustible mixtures of gases. The production of these gases in sealed radioactive waste containers represents a significant safety concern for the handling, shipment and storage of waste. The US Nuclear Regulatory Commission (NRC) acted on this safety concern in September 1984 by publishing an information notice requiring waste generators to demonstrate, by tests or measurements, that combustible mixtures of gases are not present in radioactive waste shipments; otherwise the waste must be vented within 10 days of shipping. A task force, formed by the Edison Electric Institute to evaluate these NRC requirements, developed a calculational method to quantify hydrogen gas generation in sealed containers. This report presents the calculational method along with comparisons to actual measured hydrogen concentrations from EPICOR II liners, vented during their preparation for shipment. As a result of this, the NRC recently altered certain waste shipment Certificates-Of-Compliance to allow calculations, as well as tests and measurements, as acceptable means of determining combustible gas concentration. This modification was due in part to work described herein

  20. Selectivity of NF membrane for treatment of liquid waste containing uranium

    International Nuclear Information System (INIS)

    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 UF6 to UO2 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)

  1. Selectivity of NF membrane for treatment of liquid waste containing uranium

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Elizabeth E.M.; Barbosa, Celina C.R., E-mail: eemo@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Afonso, Julio C., E-mail: julio@iq.ufrj.br [Universidade Federal do Rio de Janeiro(UFRJ), Rio de Janeiro, RJ (Brazil). Inst. de Quimica. Dept. de Quimica

    2013-07-01

    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{sub 6} to UO{sub 2} in the cycle of nuclear fuel. This waste contains uranium concentration on average 7.0 mg L{sup -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)

  2. IMPROVEMENTS IN CONTAINER MANAGEMENT OF TRANSURANIC (TRU) AND LOW LEVEL RADIOACTIVE WASTE STORED AT THE CENTRAL WASTE COMPLEX (CWC) AT HANFORD

    Energy Technology Data Exchange (ETDEWEB)

    UYTIOCO EM

    2007-11-14

    The Central Waste Complex (CWC) is the interim storage facility for Resource Conservation & 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

  3. Corrosion studies on containment materials for vitrified heat generating waste

    International Nuclear Information System (INIS)

    Mean corrosion rates of carbon steels, monitored by Rsub(p) measurements on specimens in on-going long term immersion tests, are presented. True corrosion rates measured on specimens from two dismantled tests after > 2 years exposure were about 25 μm yr-1 for both cast and forged steel buried in granite at 90 C but only approx. 3 and 7 μm yr-1 for the same materials, respectively, in bentonite. Extreme value statistical analysis of maximum pit penetrations observed in experimental studies, to compensate for the small area of test specimens compared with a container, indicates that after 1000 years the maximum pit depth could be 200 mm. Overall, tests with γ-radiation on carbon steel specimens immersed in deaerated seawater at 90 C show that there is an acceleration of corrosion rate with continued exposure at the three radiation dose rates used. However in deaerated groundwater at 90 C the general corrosion rate of forged 0.2% carbon steel is -1 at a dose rate of 105 Rads h-1. Threshold stresses for the initiation of stress corrosion cracking in carbon steel parent and weld metal have been estimated. Preliminary experiments have been initiated to investigate the effect of sulphate reducing bacteria on the corrosion of carbon steel buried in bentonite. (author)

  4. Buckling design criteria for waste package disposal containers in mined salt repositories: Technical report

    International Nuclear Information System (INIS)

    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

  5. Feasibility study using hypothesis testing to demonstrate containment of radionuclides within waste packages

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    KAMBERG, L.D.

    2000-04-01

    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.

  7. 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)

    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

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

    Science.gov (United States)

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

    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 (H2S) 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 H2S 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 H2S produced by different types of sulfur-containing wastes in a relatively fast (30days) and inexpensive (125mL serum bottles) batch assay. This study confirmed the toxic effect of H2S on both sulfate reduction and methane production in batch systems, and demonstrated that removing accumulated H2S by base adsorption was effective for mitigating inhibition. H2S 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 30days of incubation, most of the sulfate in the wastes was converted to gaseous or aqueous phase sulfide, with BSPs ranging from 0.8 to 58.8mLH2S/g waste, depending on the chemical composition of the samples. Selected samples contained solid phase sulfide which contributed to the measured H2S yield. A 60day incubation in selected samples resulted in 39-86% additional sulfide production. H2S production measured in BSP assays was compared with that measured in simulated landfill reactors and that calculated from chemical analyses. H2S 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 microbial sulfide production

  9. 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.

  10. Application of Waste Liquids Containing Lignin from Pulp-producing Industry to CWM Preparation

    Institute of Scientific and Technical Information of China (English)

    HUANG Ding-guo; TADAHIRO Murakata; TAKESHI Higuchi; SHIMIO Sato

    2004-01-01

    Three kinds of craft waste liquids, which are by-products in the pulp industry and contain much lignin,were used as dispersing additives for preparing Horonai coal CWM (coal water mixture). The experiments showed that the CWM exhibited the lowest viscosity when it was diluted with an appropriate amount of water with the waste eiquids added. The experiments also indicated that the maximum coal concentration in the 62.5% (mass fraction), and 56.5% is the maximum coal mass fraction of the CWM prepared without additives. These data show the effectiveness of the waste liquids as the additives for preparing CWMs. The zeta potential of coal particles in the CWMs changed with the addition of lignin. From the change, the steric repulsion effect of the lignin adsorbed on the coal particles is concluded to be mainly responsible for the CWM dispersion. The waste liquids contain less sulfur than PSSNa(polystyrene sulfonate sodium salt), a typical dispersant which is currently used for preparing the commercial CWM, when the sulfur content in the unit mass of the solid matters within the waste liquids is compared with that in unit mass of PSSNa. This fact suggests that the waste liquids are more advantageous than PSSNa as far as air pollutants are concerned.

  11. 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

  12. Control of stress corrosion cracking in storage tanks containing radioactive waste

    International Nuclear Information System (INIS)

    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

  13. In situ containment and stabilization of buried waste. Annual report FY 1992

    Energy Technology Data Exchange (ETDEWEB)

    Allan, M.L.; Kukacka, L.E.; Heiser, J.H.

    1992-11-01

    The objective of the project was to develop, demonstrate and implement advanced grouting materials for the in-situ installation of impermeable, durable subsurface barriers and caps around waste sites and for the in-situ stabilization of contaminated soils. Specifically, the work was aimed at remediation of the Chemical Waste (CWL) and Mixed Waste Landfills (MWL) at Sandia National Laboratories (SNL) as part of the Mixed Waste Landfill Integrated Demonstration (MWLID). This report documents this project, which was conducted in two subtasks. These were (1) Capping and Barrier Grouts, and (2) In-situ Stabilization of Contaminated Soils. Subtask 1 examined materials and placement methods for in-situ containment of contaminated sites by subsurface barriers and surface caps. In Subtask 2 materials and techniques were evaluated for in-situ chemical stabilization of chromium in soil.

  14. Long time storage containers for spent fuels and vitrified wastes: synthesis of the studies

    International Nuclear Information System (INIS)

    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.)

  15. Method of encapsulating radioactive or other dangerous waste and a container for this waste

    International Nuclear Information System (INIS)

    The matter is made insoluble for water, placed in a gasstight container and isostatically compacted to a solid body. The container has a bellow-formed outer wall and an inner capsule which is gas-permeable. The top and the bottom are plain and gas-tight. (G.B.)

  16. A process for containment removal and waste volume reduction to remediate groundwater containing certain radionuclides, toxic metals and organics

    International Nuclear Information System (INIS)

    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

  17. Description of the solid waste container corrosion program at the Hanford Site

    International Nuclear Information System (INIS)

    Waste management and environmental restoration are the Prime missions of the Hanford site, owned by the Department of Energy and operated by a management and operations contractor. The Site is located in southeast Washington State; its focus since World War II was the production of nuclear material to be used in atomic weapons but now is environmental cleanup. The cleanup of the site presents formidable challenges. The degradation of containers used to store radioactive and hazardous waste presents one of these challenges. Such containers, primarily 55 gallon (208 liter) drums, have been stored for eventual retrieval and re-packing for final disposal, some since 1970, in various types of environments. The expected degradation during storage must be estimated, verified, and predicted to allow prudent waste storage. several programs have been put into place at the Hanford Site to facilitate corrosion measurement and prediction

  18. A Study of Self-Burial of a Radioactive Waste Container by Deep Rock Melting

    Directory of Open Access Journals (Sweden)

    Wenzhen Chen

    2013-01-01

    Full Text Available Aiming at the problem of radioactive waste disposal, the concept and mechanism of self-burial by deep rock melting are presented. The rationality and feasibility of self-burial by deep rock melting are analyzed by comparing with deep geological burial. The heat threshold during the process of contact melting around a spherical heat source is defined. The descent velocities and burial depths of spherical waste containers with varying radius are calculated. The calculated depth is much smaller than that obtained in the related literature. The scheme is compared with the deep geological burial that is currently carried out by the main nuclear countries. It is found that, at the end of melting, a radioactive waste container can reach deep strata that are isolated from groundwater.

  19. The market-incentive recycling system for waste packaging containers in Taiwan

    International Nuclear Information System (INIS)

    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. 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.

  1. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Charles W. Solbrig; Kenneth J. Bateman

    2010-11-01

    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.

  3. Experimental investigation of the fatigue behaviour of asphalt concrete mixtures containing waste iron powder

    International Nuclear Information System (INIS)

    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.

  4. MICROBE-METAL-INTERACTIONS FOR THE BIOTECHNOLOGICAL TREATMENT OF METAL-CONTAINING SOLID WASTE

    Institute of Scientific and Technical Information of China (English)

    Helmut Brandl; Mohammad A. Faramarzi

    2006-01-01

    In nature, microbes are involved in weathering of rocks, in mobilization of metals from minerals, and in metal precipitation and deposition. These microbiological principles and processes can be adapted to treat particulate solid wastes. Especially the microbiological solubilization of metals from solid minerals (termed bioleaching) to obtain metal values is a well-known technique in the mining industry. We focus here on non-mining mineral wastes to demonstrate the applicability of mining-based technologies for the treatment of metal-containing solid wastes. In the case study presented, microbial metal mobilization from particulate fly ash (originating from municipal solid waste incineration) by Acidithiobacilli resulted in cadmium, copper, and zinc mobilization of >80%, whereas lead, chromium, and nickel were mobilized by 2, 11 and 32%, respectively. In addition, the potential of HCN-forming bacteria (Chromobacterium violaceum,Pseudomonas fluorescens) was investigated to mobilize metals when grown in the presence of solid materials (e.g.,copper-containing ores, electronic scrap, spent automobile catalytic converters). C. violaceum was found capable of mobilizing nickel as tetracyanonickelate from fine-grained nickel powder. Gold was microbially solubilized as dicyanoaurate from electronic waste. Additionally, cyanide-complexed copper was detected during biological treatment of shredded printed circuit-board scraps. Water-soluble copper and platinum cyanide were also detected during the treatment of spent automobile catalytic converters.

  5. 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

  6. 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.

  7. 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

  8. Transient and steady-state radionuclide transport through penetrations in nuclear waste containers

    International Nuclear Information System (INIS)

    In this paper we analyze the transport of radionuclides through penetrations in nuclear waste containers. Penetrations may result from corrosion or cracks and may occur in the original container material, in degraded or corroded material, or in deposits of corrosion products. We do not consider how these penetrations occur or the characteristics of expected penetrations in waste containers. We are concerned only with the analytical formulation and solutions of equations to predict rates of mass transfer through penetrations of specified size and geometry. Expressions for the diffusive mass transfer rates through apertures are presented. We present numerical illustrations for steady-state mass-transfer rates through a circular hole, including concentration isopleths. The results are extended to multiple holes, including a criterion for hole spacing wherein superposition of single-hole solutions can be used. Results illustrated for holes in thin-walled containers show that significant mass transfer can occur even if a small fraction of the container area is perforated. We also illustrate the case of holes facing a water gap, instead of being in intimate contact with porous rock. In this case the radionuclide flux from many small holes approaches that from a bare waste cylinder

  9. Studies of corrosion in metallic container for storage of high level radioactive wastes

    International Nuclear Information System (INIS)

    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)

  10. Passive 3D imaging of nuclear waste containers with Muon Scattering Tomography

    International Nuclear Information System (INIS)

    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

  11. 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.

  12. Hazards Associated with Legacy Nitrate Salt Waste Drums Managed under the Container Isolation Plan

    Energy Technology Data Exchange (ETDEWEB)

    Funk, David John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Clark, David Lewis [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-01-07

    At present, there are 29 drums of nitrate waste salts (oxidizers with potentially acidic liquid bearing RCRA characteristics D001 and D002) that are awaiting processing, specifically to eliminate these characteristics and to allow for ultimate disposition at WIPP. As a result of the Feb. 14th, 2014 drum breach at WIPP, and the subsequent identification of the breached drum as a product ofLANL TRU waste disposition on May 15th, 2014, these 29 containers were moved into the Perrnacon in Dome 231 at TA-54 Area G, as part of the New Mexico Environment Department (NMED) approved container isolation plan. The plan is designed to mitigate hazards associated with the nitrate salt bearing waste stream. The purpose of this document is to articulate the hazards associated with un-remediated nitrate salts while in storage at LANL. These hazards are distinctly different from the Swheat-remediated nitrate salt bearing drums, and this document is intended to support the request to remove the un-remediated drums from management under the container isolation plan. Plans to remediate and/or treat both of these waste types are being developed separately, and are beyond the scope of this document.

  13. Implementation of Exhaust Gas Recirculation for Double Stage Waste Heat Recovery System on Large Container Vessel

    DEFF Research Database (Denmark)

    Andreasen, Morten; Marissal, Matthieu; Sørensen, Kim;

    2014-01-01

    of recovering some of the waste heat from the exhaust gas. This heat is converted into electrical energy used on-board instead of using auxiliary engines. Exhaust Gas Recirculation (EGR) systems, are recirculating a part of the exhaust gas through the engine combustion chamber to reduce emissions. WHRS combined......Concerned to push ships to have a lower impact on the environment, the International Maritime Organization are implementing stricter regulation of NOx and SOx emissions, called Tier III, within emission control areas (ECAs). Waste Heat Recovery Systems (WHRS) on container ships consist...

  14. Treatment of Zn-Containing Acidic Waste Water by Emulsion Liquid Membrane Process

    Institute of Scientific and Technical Information of China (English)

    王士柱; 何培炯; 郝东萍; 朱永贝睿

    2002-01-01

    Zn-containing waste water from a viscose staple fiber plant has been treated using the emulsion liquid membrane (ELM) process since 1995. The flow sheet and operating parameters of the ELM process are introduced. After adjusting the membrane composition, changing the emulsion phase ratio, and adding a scrubbing step, the ELM process operated normally without trouble for emulsion splitting and mass transport throughput. The splitter voltage was decreased to 3.55 kV. The zinc concentration of treated waste water was lowered to less than 10 mgL-1. More than 95% of the zinc was recovered and reused.

  15. Effect of chloride concentration and pH on pitting corrosion of waste package container materials

    International Nuclear Information System (INIS)

    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

  16. 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.

  17. Cost estimate of high-level radioactive waste containers for the Yucca Mountain Site Characterization Project

    International Nuclear Information System (INIS)

    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

  18. Influence of Nitrogen Containing Wastes Addition on Natural Aerobic Composting of Rice Straw

    Directory of Open Access Journals (Sweden)

    Thaniya Kaosol

    2012-01-01

    Full Text Available Problem statement: Rice straw is an agricultural residue. Typically, the rice straw can be burn in the rice field after the harvesting process. The burning can cause air pollution. Another alternative rice straw management method is animal feed. The amount of rice straw is enormus in Thailand. Another sustainable way to manage rice straw is required. Rice straw is used as main waste to compost with nitrogen containing wastes such as golden apple snail, cattle dung and urea in natural aerobic composting reactors. The golden apple snail is a pesticide and cattle dung is an animal waste. Both materials are all waste of low values. The main purpose of this study was to determine the influence of nitrogen containing wastes addition to rice straw on the performance of natural aerobic composting process in terms of the following parameters: pH, temperature, organic matter, C/N ratio, electrical conductivity and GI. The impact of this study is to reuse agriculture residue by composting. Approach: The experiments was consisted of three reactors. The reactor 1 contains the rice straws and golden apple snails while the reactor 2 contains the rice straws, golden apple snails and urea. The reactor 3 contains the rice straws, cattle dung and urea. The experiments were carried out in designed natural aerobic reactors (60 L under controlled laboratory conditions over 60 days. The analysis was done every 5 days however the temperature was measured daily. Results: The experimental results showed that the initial C/N ratio was 30.7, 30.3 and 31.8 in the reactor 1, 2 and 3, respectively. After the 60-day period, the final C/N ratio was reduced to 17.9, 16.9 and 18.4 in the reactor 1, 2 and 3, respectively. The main nutrients (N: P: K from all reactors achieved the standard level for Thai compost standard. The rice straw as agricultural residue was suitable for co-composting with golden apple snails and cattle dung as the nitrogen containing wastes. Conclusion: The

  19. 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)

    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

  20. Processing device for plutonium-containing liquid wastes by using tannin

    International Nuclear Information System (INIS)

    Insoluble tannin adsorbs uranium and uranium elements extremely efficiently. Accordingly, liquid wastes containing Pu are passed through at least two adsorbing towers filled with the insoluble tannin. Pu in the liquid wastes is adsorbed to the insoluble tannin, so that a draining standard can be satisfied by a single process without using combination with other methods. Since tannin has a high Pu adsorbing performance, the adsorbing towers can be reduced in the size. In addition, since insoluble tannin comprises carbon, hydrogen and oxygen, even if spent adsorbent having no more Pu adsorbing performance is burnt, it can be released without contaminating environment. On the other hand, an extremely slight amount of U and Pu adsorbed to the insoluble tannin are oxidized, and these oxides can be formed into an MOX powder, thereby enabling to minimize the residues after processing liquid wastes. (T.M.)

  1. Pore size distribution, strength, and microstructure of portland cement paste containing metal hydroxide waste

    Energy Technology Data Exchange (ETDEWEB)

    Majid, Z.A.; Mahmud, H.; Shaaban, M.G.

    1996-12-31

    Stabilization/solidification of hazardous wastes is used to convert hazardous metal hydroxide waste sludge into a solid mass with better handling properties. This study investigated the pore size development of ordinary portland cement pastes containing metal hydroxide waste sludge and rice husk ash using mercury intrusion porosimetry. The effects of acre and the addition of rice husk ash on pore size development and strength were studied. It was found that the pore structures of mixes changed significantly with curing acre. The pore size shifted from 1,204 to 324 {angstrom} for 3-day old cement paste, and from 956 to 263 {angstrom} for a 7-day old sample. A reduction in pore size distribution for different curing ages was also observed in the other mixtures. From this limited study, no conclusion could be made as to any correlation between strength development and porosity. 10 refs., 6 figs., 3 tabs.

  2. A new type B ISO container for transportation of alpha waste

    International Nuclear Information System (INIS)

    The French Atomic Energy Commission (CEA) operates several facilities which produce various transuranic wastes. These wastes are generally stored in metallic drums. There is a need for transportation of more than 1000 drums per year to intermediate storage sites and in the future to final storage sites which will be managed by ANDRA the French nuclear waste Agency. To answer this need, TRANSNUCLEAIRE has developed the TN-GEMINI II, a large dimension 'type B' container for use in France and in Europe. This packaging fits on ISO 20 ft standard truck trailers using conventional tractors, and is also compatible with french railcar dimensions. The main improvements brought by this design are: 1) high payload: 40 drums, 200 liter type, 2) versatility for transport of large size contaminated parts 3) simple operational features. (J.P.N.)

  3. The effect of devitrification on leaching rate of glass containing simulated high level liquid waste (HLLW)

    International Nuclear Information System (INIS)

    Effect of devitrification on leaching rate of glass named G1 and G2 each contains 20 wt% and 30wt% of waste has been studied. devitrification of waste - glass has been carried out by heating those specimens at 850oC for 10, 18, 26, 34, 42 and 50 hours respectively. The weight percentage of crystal in waste glass was determined by X-ray diffractometer and leaching rate was determined by soxhlet apparatus at 100oC for 24 hours. The longer heating time, the more weight percentage of crystal is formed. The results show that leaching rate of G2 specimens are higher than those of G1. For G1 the leaching rate at 850oC in 20 times than without heating, and for G2 leaching rate is 15.7 times than without heating. (author)

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

    International Nuclear Information System (INIS)

    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. DEMONSTRATiON OF A SUBSURFACE CONTAINMENT SYSTEM FOR INSTALLATION AT DOE WASTE SITES

    Energy Technology Data Exchange (ETDEWEB)

    Thomas J. Crocker; Verna M. Carpenter

    2003-05-21

    Between 1952 and 1970, DOE buried mixed waste in pits and trenches that now have special cleanup needs. The disposal practices used decades ago left these landfills and other trenches, pits, and disposal sites filled with three million cubic meters of buried waste. This waste is becoming harmful to human safety and health. Today's cleanup and waste removal is time-consuming and expensive with some sites scheduled to complete cleanup by 2006 or later. An interim solution to the DOE buried waste problem is to encapsulate and hydraulically isolate the waste with a geomembrane barrier and monitor the performance of the barrier over its 50-yr lifetime. The installed containment barriers would isolate the buried waste and protect groundwater from pollutants until final remediations are completed. The DOE has awarded a contract to RAHCO International, Inc.; of Spokane, Washington; to design, develop, and test a novel subsurface barrier installation system, referred to as a Subsurface Containment System (SCS). The installed containment barrier consists of commercially available geomembrane materials that isolates the underground waste, similar to the way a swimming pools hold water, without disrupting hazardous material that was buried decades ago. The barrier protects soil and groundwater from contamination and effectively meets environmental cleanup standards while reducing risks, schedules, and costs. Constructing the subsurface containment barrier uses a combination of conventional and specialized equipment and a unique continuous construction process. This innovative equipment and construction method can construct a 1000-ft-long X 34-ft-wide X 30-ft-deep barrier at construction rates to 12 Wday (8 hr/day operation). Life cycle costs including RCRA cover and long-term monitoring range from approximately $380 to $590/cu yd of waste contained or $100 to $160/sq ft of placed barrier based upon the subsurface geology surrounding the waste. Project objectives for Phase

  6. Mechanical and toxicological evaluation of concrete artifacts containing waste foundry sand.

    Science.gov (United States)

    Mastella, Miguel Angelo; Gislon, Edivelton Soratto; Pelisser, Fernando; Ricken, Cláudio; da Silva, Luciano; Angioletto, Elídio; Montedo, Oscar Rubem Klegues

    2014-08-01

    The creation of metal parts via casting uses molds that are generally made from sand and phenolic resin. The waste generated after the casting process is called waste foundry sand (WFS). Depending on the mold composition and the casting process, WFS can contain substances that prevent its direct emission to the environment. In Brazil, this waste is classified according to the Standard ABNT NBR 10004:2004 as a waste Class II (Non-Inert). The recycling of this waste is limited because its characteristics change significantly after use. Although the use (or reuse) of this byproduct in civil construction is a technically feasible alternative, its effects must be evaluated, especially from mechanical and environmental points of view. Thus, the objective of this study is to investigate the effect of the use of WFS in the manufacture of cement artifacts, such as masonry blocks for walls, structural masonry blocks, and paving blocks. Blocks containing different concentrations of WFS (up to 75% by weight) were produced and evaluated using compressive strength tests (35 MPa at 28 days) and toxicity tests on Daphnia magna, Allium cepa (onion root), and Eisenia foetida (earthworm). The results showed that there was not a considerable reduction in the compressive strength, with values of 35 ± 2 MPa at 28 days. The toxicity study with the material obtained from leaching did not significantly interfere with the development of D. magna and E. foetida, but the growth of the A. cepa species was reduced. The study showed that the use of this waste in the production of concrete blocks is feasible from both mechanical and environmental points of view. PMID:24582355

  7. Hydrogen Concentration in the Inner-Most Container within a Pencil Tank Overpack Packaged in a Standard Waste Box Package

    Energy Technology Data Exchange (ETDEWEB)

    Marusich, Robert M.

    2012-01-25

    A set of steady state diffusion flow equations, for the hydrogen diffusion from one bag to the next bag (or one plastic waste container to another), within a set of nested waste bags (or nested waste containers), are developed and presented. The input data is then presented and justified. Inputting the data for each volume and solving these equations yields the steady state hydrogen concentration in each volume. The input data (permeability of the bag surface and closure, dimensions and hydrogen generation rate) and equations are analyzed to obtain the hydrogen concentrations in the innermost container for a set of containers which are analyzed for the TRUCON code for the general waste containers and the TRUCON code for the Pencil Tank Overpacks (PTO) in a Standard Waste Box (SWB).

  8. The treatment of liquid radioactive waste containing Americium by using a cation exchange method

    International Nuclear Information System (INIS)

    A research in the treatment of a liquid radioactive waste containing americium has been done. The liquid radioactive waste used in this research was standard solution of U dan Ce with the initial activity of 100 ppm. The experimental investigation is aimed at a study of the effects of the waste pH, the column dimension of IR-120 cation exchanger which is expressed as L/D, the flow rate of a liquid waste and the influence of thiocyanate as a complex agent against the efficiency of a decontamination for uranium and cerium element. The experiment was done by passing downward the feed of uranium and cerium solution into an IR-120 type of cation exchanger with the L/D of 11.37. From the experimental parameters done in this research where the influence of waste pH was varied from 3 - 8, the geometric column (L/D) 11.37, the liquid flow rate was from 2.5 - 10 ml/m and the thiocyanate concentration was between 100 ppm-500 ppm can be concluded that the optimum operational condition for the ion exchange achieved were the waste pH for uranium = 4 and the waste pH for cerium = 6, the flow rate = 2.5 ml/men. From the given maximum value of DF for uranium = 24 (DE = 95.83%) and of DF for cerium = 40 (DE = 97.5%), it can also be concluded that this investigation is to be continued in order that the greater value of DF/DE can be achieved

  9. LONG-TERM CORROSION TESTING OF CANDIDATE MATERIALS FOR HIGH-LEVEL RADIOACTIVE WASTE CONTAINMENT

    International Nuclear Information System (INIS)

    Preliminary results are presented from the long-term corrosion test program of candidate materials for the high-level radioactive waste packages that would be emplaced in the potential repository at Yucca Mountain, Nevada. The present waste package design is based on a multi-barrier concept having an inner container of a corrosion resistant material and an outer container of a corrosion allowance material. Test specimens have been exposed to simulated bounding environments that may credibly develop in the vicinity of the waste packages. Corrosion rates have been calculated for weight loss and crevice specimens, and U-bend specimens have been examined for evidence of stress corrosion cracking (SCC). Galvanic testing has been started recently and initial results are forthcoming. Pitting characterization of test specimens will be conducted in the coming year. This test program is expected to continue for a minimum of five years so that long-term corrosion data can be determined to support corrosion model development, performance assessment, and waste package design

  10. Northwest Hazardous Waste Research, Development, and Demonstration Center: Program Plan. [Contains glossary

    Energy Technology Data Exchange (ETDEWEB)

    1988-02-01

    The Northwest Hazardous Waste Research, Development, and Demonstration Center was created as part of an ongoing federal effort to provide technologies and methods that protect human health and welfare and environment from hazardous wastes. The Center was established by the Superfund Amendments and Reauthorization Act (SARA) to develop and adapt innovative technologies and methods for assessing the impacts of and remediating inactive hazardous and radioactive mixed-waste sites. The Superfund legislation authorized $10 million for Pacific Northwest Laboratory to establish and operate the Center over a 5-year period. Under this legislation, Congress authorized $10 million each to support research, development, and demonstration (RD and D) on hazardous and radioactive mixed-waste problems in Idaho, Montana, Oregon, and Washington, including the Hanford Site. In 1987, the Center initiated its RD and D activities and prepared this Program Plan that presents the framework within which the Center will carry out its mission. Section 1.0 describes the Center, its mission, objectives, organization, and relationship to other programs. Section 2.0 describes the Center's RD and D strategy and contains the RD and D objectives, priorities, and process to be used to select specific projects. Section 3.0 contains the Center's FY 1988 operating plan and describes the specific RD and D projects to be carried out and their budgets and schedules. 9 refs., 18 figs., 5 tabs.

  11. Autoclave inactivation of infectious radioactive laboratory waste contained within a charcoal filtration system

    International Nuclear Information System (INIS)

    A model system was developed previously for disposal of solid laboratory waste that is both radioactive and heat sensitive, e.g., HIV. A double polypropylene bag with charcoal vent filter and absorbent was designed to meet requirements for both steam sterilization and disposal as solid radioactive waste. Earlier work demonstrated the effective containment of radioactive gases by the filter and inactivation of organisms as heat sensitive as HIV. The authors sought to broaden the application of this model to ensure inactivation of microorganisms that are more heat resistant than HIV. The efficacy of steam sterilization using water or solutions of iodophor, hypochlorite, or hydrogen peroxide was studied under constant temperature and time conditions. The systems were monitored with internal probes, physical, chemical, and biological indicators. Biological indicators documented inactivation when bags containing hydrogen peroxide (3%) were autoclaved for 60 min at 121C. Synergistic activity between hydrogen peroxide and autoclave conditions significantly reduced processing time

  12. Hydraulic containment of low-level radioactive waste disposal sites: [Final technical report

    International Nuclear Information System (INIS)

    This document describes the use of impermeable barriers for the containment of liquid radioactive wastes at low-level radioactive waste disposal sites. Included are a review of existing barrier systems, assessments of laboratory and field data, and simulations of system performance under humid and arid conditions. Alternatives are identified as the most promising of the existing systems based on retention of irradiated water, field installation feasibility, and response to aggressive permeation. In decreasing order of preference, the favored systems are asphalt slurry, high density polyethylene synthetic liner, polyvinyl chloride synthetic liner, lean portland cement concrete, and compacted bentonite liner. It should be stressed that all five of these alternatives effectively retain irradiated water in the humid and arid simulations. Recommendations on the design and operation of the hydraulic containment system and suggestions on avenues for future research are included. 102 refs., 27 figs., 23 tabs

  13. A novel shielding material prepared from solid waste containing lead for gamma ray

    Science.gov (United States)

    Erdem, Mehmet; Baykara, Oktay; Doğru, Mahmut; Kuluöztürk, Fatih

    2010-09-01

    Human beings are continuously exposed to cosmogenic radiation and its products in the atmosphere from naturally occurring radioactive materials (NORM) within Earth, their bodies, houses and foods. Especially, for the radiation protection environments where high ionizing radiation levels appear should be shielded. Generally, different materials are used for the radiation shielding in different areas and for different situations. In this study, a novel shielding material produced by a metallurgical solid waste containing lead was analyzed as shielding material for gamma radiation. The photon total mass attenuation coefficients ( μ/ ρ) were measured and calculated using WinXCom computer code for the novel shielding material, concrete and lead. Theoretical and experimental values of total mass attenuation coefficient of the each studied sample were compared. Consequently, a new shielding material prepared from the solid waste containing lead could be preferred for buildings as shielding materials against gamma radiation.

  14. Criticality study of the storage of radioactive waste containing 235U

    International Nuclear Information System (INIS)

    The purpose of this study is to define the conditions of storage of nuclear waste drums containing 350 g of 235U (per drum). This study is valid for a square pitch stacking of cylindrical drums whose height/diameter ratio does not exceed 3. The reflector effect of concrete is taken into account. This study defines a conservative case that can be used under any hypothesis of moderation, of radiation coupling between drums and of fissile material density. (A.C.)

  15. The Use of Bioleaching Methods for the Recovery of Metals Contained in Sulfidic Mining Wastes

    OpenAIRE

    Guezennec, Anne-Gwénaëlle; Delclaud, Marie; Savreux, Frederic; Jacob, Jérome; d'Hugues, Patrick

    2014-01-01

    Mining wastes can contain base and precious metals, but also metalloids and rare earth elements that are nowadays considered as highly critical for the industrial development of the European Union. The development of alternative routes to conventional processing is still required in order to decrease the cost associated with the treatment of these resources, which are more complex in composition and with lower grades. An ecologically acceptable and yet economic alternative for processing of l...

  16. Assessment of two thermally treated drill mud wastes for landfill containment applications.

    Science.gov (United States)

    Carignan, Marie-Pierre; Lake, Craig B; Menzies, Todd

    2007-10-01

    Offshore oil and gas drilling operations generate significant amounts of drill mud waste, some of which is transported onshore for subsequent thermal treatment (i.e. via thermal remediation). This treatment process results in a mineral waste by-product (referred to as thermally treated drill mud waste; TTDMW). Bentonites are originally present in many of the drill mud products and it is hypothesized that TTDMW can be utilized in landfill containment applications (i.e. cover or base liner). The objective of this paper is to examine the feasibility of this application by performing various physical and chemical tests on two TTDMW samples. It is shown that the two TTDMW samples contained relatively small amounts of clay-sized minerals although hydraulic conductivity values are found to be less than 10(-8) m/s. Organic carbon contents of the samples were approximately 2%. Mineralogy characterization of the samples confirmed varying amounts of smectite, however, peak friction angles for a TTDMW sample was greater than 36 degrees. Chemical characterization of the TTDMW samples show potential leaching of barium and small amounts of other heavy metals. Discussion is provided in the paper on suggestions to assist in overcoming regulatory issues associated with utilization of TTDMW in landfill containment applications. PMID:17985664

  17. Greater-than-Class C low-level radioactive waste shipping package/container identification and requirements study. National Low-Level Waste Management Program

    Energy Technology Data Exchange (ETDEWEB)

    Tyacke, M.

    1993-08-01

    This report identifies a variety of shipping packages (also referred to as casks) and waste containers currently available or being developed that could be used for greater-than-Class C (GTCC) low-level waste (LLW). Since GTCC LLW varies greatly in size, shape, and activity levels, the casks and waste containers that could be used range in size from small, to accommodate a single sealed radiation source, to very large-capacity casks/canisters used to transport or dry-store highly radioactive spent fuel. In some cases, the waste containers may serve directly as shipping packages, while in other cases, the containers would need to be placed in a transport cask. For the purpose of this report, it is assumed that the generator is responsible for transporting the waste to a Department of Energy (DOE) storage, treatment, or disposal facility. Unless DOE establishes specific acceptance criteria, the receiving facility would need the capability to accept any of the casks and waste containers identified in this report. In identifying potential casks and waste containers, no consideration was given to their adequacy relative to handling, storage, treatment, and disposal. Those considerations must be addressed separately as the capabilities of the receiving facility and the handling requirements and operations are better understood.

  18. Design study on containers for geological disposal of high-level radioactive waste

    International Nuclear Information System (INIS)

    A study has been made of the requirements and design features for containers to isolate vitrified high-level radioactive waste from the environment for a period of 500 to 1000 years. The requirements for handling, storing and transporting containers have been identified following a study of disposal operations, and the pressures and temperatures which may possibly be experienced in clay, granite and salt formations have been estimated. A range of possible container designs have been proposed to satisfy the requirements of each of the disposal environments. Alternative design concepts in corrosion resistant or corrosion allowance material have been suggested. Some resist pressure by using a structural shell leaving the contents unstressed whereas others transmit loads to their contents. Potentially suitable container shell materials have been selected following a review of corrosion studies and although metals have not been specified in detail, titanium alloys and low carbon steels are thought to be appropriate for corrosion resistant and corrosion allowance designs respectively. Performance requirements for container filler materials have been identified and candidate materials assessed. However, no entirely suitable materials have been found and further research is required in this area. A preliminary container stress analysis has shown the importance of thermal modelling and that if lead is used as a filler it dominates the stress response of the container. Possible methods of manufacturing disposal containers have been assessed and found to be generally feasible although filling operations and container closure could be difficult

  19. The development of special ISO freight containers for the transport of low level radioactive waste

    International Nuclear Information System (INIS)

    During the operation and maintenance of nuclear power stations, and other nuclear facilities, solid waste materials such as paper, plastics, filters, clothing, wood and metallic items are produced which are lightly or potentially radioactively contaminated. These items of trash are generally classified as low level waste (LLW) which, in the UK, is defined as having a radioactivity content of not more than 12 GBq/ton beta/gamma (about 300 mCi/t) and 4 GBq/ton alpha (about 100 mCi/t). LLW does not normally require to be shielded during normal handling and transport. LLW in the UK is routinely disposed at a special site at Drigg in Cumbria and until recently the disposal method used has been simple tumble-tipping into shallow trenches excavated in clay. Large re-usable tipping containers were used to transport the waste by road to the disposal site. Although various studies had confirmed the continued technical, safety and environmental acceptability of the simple disposal practices at Drigg, it was recognized that improvements would have to be made, mainly for presentational purposes. The new disposal concept adopted at Drigg was to construct concrete lined engineered vaults in which the containerized waste would be stacked uniformly. It was therefore necessary to develop a new method of waste packaging that was compatible with the new disposal concept. A number of proposals were considered. The authors proposed a system that would use ISO freight containers as both transport and disposal packages. This system was adopted and has been in service since mid 1988

  20. A methodology to analyze the creep behaviour of nuclear fuel waste containers

    International Nuclear Information System (INIS)

    The concept for the disposal of used-fuel waste from CANDU reeactors operating in Canada comprises a system of natural and engineered barriers surrounding the waste in a mined vault situated at a depth of 500 - 1000 m in plutonic rock of the Canadian Shield. The fuel would be packaged in a highly durable metal container, within a matrix of compacted particulate. The design of the container takes into account that it would be subjected to an external hydrostatic pressure. Consideration of the rate of radioactive decay of the radionuclides contained in the fuel, suggests that the lifetime of the container should be at least 500 years. Consequently, the role of creep deformation, and the possibility of creep rupture of the container shell, must be included in the assessment of time-dependent mechanical integrity. This report describes an analytical approach that can be used to quantify the long-term creep properties of the container material and facilitate the engineering design. The overall objective is to formulate a constitutive creep equation that provides the required input for a finite element computer model being developed to analyze the elastic-plastic behaviour of the container. Alternative forms of such equations are reviewed. It is shown that the capability of many of these equations to extrapolate over long time scales is limited by their empirical nature. Thus, the recommended equation is based on current mechanistic understanding of creep deformation and creep rupture. A criterion for determining the onset of material failure by creep rupture, that could be used in the design of containers with extended structural integrity, is proposed. Interpretation and extrapolation will be supported by the complementary Deformation and Fracture Mechanism Maps. (author) 103 refs., 2 tabs., 54 figs

  1. Design study on containers for geological disposal of high-level radioactive waste. Phase 2

    International Nuclear Information System (INIS)

    This study has considered the feasibility of three designs for containers which would isolate the waste from the environment for a minimum period of 500 to 1000 years. The candidate container designs were taken from the results of a previous study by Ove Arup and Partners (1985) and were developed as the study progressed. Their major features can be summarized as follows: Type A: A thin-walled corrosion-resistant metal shell filled with lead or cement grout. Type B: An unfilled thick-walled carbon steel shell. Type C: an unfilled carbon steel shell plated externally with corrosion-resistant metal. Reference repository conditions in clay, granite and salt, reference disposal operations and metals corrosion data have been taken from various European Community radioactive waste management research and engineering projects. The study concludes that design types A and B are feasible in manufacturing terms but design Type C is not. Furthermore, a titanium-palladium alloy is considered the most suitable metal for Type A container shells and lead is the preferred filler. The analysis shows that design Types A and B both have adequate resistance to pressure and temperature loadings and both would resist accidental impact damage when upright. A reduction in waste heat output at disposal would lower the stress levels in Type A containers but would have virtually no effect on Type B. There is insufficient data to compare the relative costs and benefits of design Types A and B. In conclusion design Types A and B are both considered feasible but Type A would require more development than Type B. In both cases further research is needed to confirm the long-term corrosion performance of the candidate materials. It is recommended that model containers should be produced to demonstrate the proposed methods of manufacture and that they should be tested to validate the analytical techniques used

  2. Effects of resource activities upon repository siting and waste containment with reference to bedded salt

    International Nuclear Information System (INIS)

    The primary consideration for the suitability of a nuclear waste repository site is the overall ability of the repository to safely contain radioactive waste. This report is a discussion of the past, present, and future effects of resource activities on waste containment. Past and present resource activities which provide release pathways (i.e., leaky boreholes, adjacent mines) will receive initial evaluation during the early stages of any repository site study. However, other resource activities which may have subtle effects on containment (e.g., long-term pumping causing increased groundwater gradients, invasion of saline water causing lower retardation) and all potential future resource activities must also be considered during the site evaluation process. Resource activities will affect both the siting and the designing of repositories. Ideally, sites should be located in areas of low resource activity and low potential for future activity, and repository design should seek to eliminate or minimize the adverse effects of any resource activity. Buffer zones should be created to provide areas in which resource activities that might adversely affect containment can be restricted or curtailed. This could mean removing large areas of land from resource development. The impact of these frozen assets should be assessed in terms of their economic value and of their effect upon resource reserves. This step could require a major effort in data acquisition and analysis followed by extensive numerical modeling of regional fluid flow and mass transport. Numerical models should be used to assess the effects of resource activity upon containment and should include the cumulative effects of different resource activities. Analysis by other methods is probably not possible except for relatively simple cases

  3. Chloride ions promoted the catalytic wet peroxide oxidation of phenol over clay-based catalysts.

    Science.gov (United States)

    Zhou, Shiwei; Zhang, Changbo; Xu, Rui; Gu, Chuantao; Song, Zhengguo; Xu, Minggang

    2016-01-01

    Catalytic wet peroxide oxidation (CWPO) of phenol over clay-based catalysts in the presence and absence of NaCl was investigated. Changes in the H2O2, Cl(-), and dissolved metal ion concentration, as well as solution pH during phenol oxidation, were also studied. Additionally, the intermediates formed during phenol oxidation were detected by liquid chromatography-mass spectroscopy and the chemical bonding information of the catalyst surfaces was analyzed by X-ray photoelectron spectroscopy (XPS). The results showed that the presence of Cl(-) increased the oxidation rate of phenol to 155%, and this phenomenon was ubiquitous during the oxidation of phenolic compounds by H2O2 over clay-based catalysts. Cl(-)-assisted oxidation of phenol was evidenced by several analytical techniques such as mass spectroscopy (MS) and XPS, and it was hypothesized that the rate-limiting step was accelerated in the presence of Cl(-). Based on the results of this study, the CWPO technology appears to be promising for applications in actual saline phenolic wastewater treatment. PMID:26942523

  4. Testing of low-temperature stabilization alternatives for salt containing mixed wastes - Approach and results to date

    International Nuclear Information System (INIS)

    Through its annual process of identifying technology deficiencies associated with waste treatment, the Department of Energy's (DOE) Mixed Waste Focus Area (MWFA) determined that the former DOE weapons complex lacks efficient mixed waste stabilization technologies for salt containing wastes. These wastes were generated as sludge and solid effluents from various primary nuclear processes involving acids and metal finishing; and well over 10,000 cubic meters exist at 6 sites. In addition, future volumes of these problematic wastes will be produced as other mixed waste treatment methods such as incineration and melting are deployed. The current method used to stabilize salt waste for compliant disposal is grouting with Portland cement. This method is inefficient since the highly soluble and reactive chloride, nitrate, and sulfate salts interfere with the hydration and setting processes associated with grouting. The inefficiency results from having to use low waste loadings to ensure a durable and leach resistant final waste form. The following five alternatives were selected for MWFA development funding in FY97 and FY98: phosphate bonded ceramics; sol-gel process; polysiloxane; polyester resin; and enhanced concrete. Comparable evaluations were planned for the stabilization development efforts. Under these evaluations each technology stabilized the same type of salt waste surrogates. Final waste form performance data such as compressive strength, waste loading, and leachability could then be equally compared. Selected preliminary test results are provided in this paper

  5. Potential use of densified polymer-pastefill mixture as waste containment barrier materials.

    Science.gov (United States)

    Fall, M; Célestin, J; Sen, H F

    2010-12-01

    Mining activities generate a large amount of solid waste, such as waste rock and tailings. The surface disposal of such waste can create several environmental and geotechnical problems. Public perception and strict government regulations with regards to the disposal of such waste compel the mining industry to develop new strategies which are environmentally sound and cost effective. In this scenario, recycling of such waste into mining or civil engineering construction materials have become a great challenge for the mining and civil engineering community. Hence, in this study, taking advantage of the inherent low hydraulic conductivity of paste tailings (pastefill), small amounts (0.05, 0.1, 0.2, 0.5%) of a super absorbent polymer (SAP) are added to the latter after moisturizing the tailings. The resulting densified polymer-pastefill (PP) materials are compacted and submitted to permeability tests at room temperature and performance tests under cyclic freeze-thaw and wet-dry conditions to evaluate their suitability as a barrier for waste containment facilities. Valuable results are obtained. It is found that the hydraulic conductivity of the proposed barrier material (PP) decreases as the amount of SAP increases. Hydraulic conductivity values as low as 1 × 10(-7) and 6 × 10(-9)cm/s are obtained for PPs which contain 0.1-0.5% SAP, respectively. The PP material also shows relatively good resistance to cyclic freeze-thaw and wet-dry stresses. The results show that negligible to acceptable changes in hydraulic conductivity occur after five freeze-thaw and six wet-dry cycles. None of the changes reach one order of magnitude. As a final step, a cost analysis is undertaken to evaluate the economical benefits that could be drawn from such a proposed barrier material. When compared to a conventional compacted sand-bentonite barrier with 12% bentonite concentration, it is found that the benefit realized could be estimated to 98, 96 and 90% when using PP material that

  6. Evaluation of dry-solids-blend material source for grouts containing 106-AN waste: Final report

    International Nuclear Information System (INIS)

    Stabilization/solidification technology is one of the most widely used techniques for the treatment and ultimate disposal of both radioactive and chemically hazardous wastes. Cement-based products, commonly referred to as grouts, are the predominant materials of choice because of their low associated processing costs, compatibility with a wide variety of disposal scenarios, and ability to meet stringent processing and performance requirements. Such technology is being utilized in a Grout Treatment Facility (GTF) by the Westinghouse Hanford Company (WHC) for the disposal of various wastes, including 106-AN wastes, located on the Hanford Reservation. The WHC personnel have developed a grout formula for 106-AN disposal that is designed to meet stringent performance requirements. This formula consists of a dry-solids blend containing 40 wt % limestone, 28 wt % granulated blast furnace slag (BFS), 28 wt % American Society for Testing and Materials (ASTM) Class F fly ash, and 4 wt % Type I-II-LA Portland cement. This blend is mixed with 106-AN at a mix ratio of 9 lb of dry-solids blend per gallon of waste. This report documents the final results of efforts at Oak Ridge National Laboratory in support of WHC's Grout Technology Program to assess the effects of the source of the dry-solids-blend materials on the resulting grout formula

  7. In-situ containment and stabilization of buried waste: Annual report FY 1994

    International Nuclear Information System (INIS)

    The two landfills of specific interest are the Chemical Waste Landfill (CWL) and the Mixed Waste Landfill (MWL), both located at Sandia National Laboratory. The work is comprised of two subtasks: (1) In-Situ Barriers and (2) In-Situ Stabilization of Contaminated Soils. The main environmental concern at the CWL is a chromium plume resulting from disposal of chromic acid and chromic sulfuric acid into unlined pits. This program has investigated means of in-situ stabilization of chromium contaminated soils and placement of containment barriers around the CWL. The MWL contains a plume of tritiated water. In-situ immobilization of tritiated water with cementitious grouts was not considered to be a method with a high probability of success and was not pursued. This is discussed further in Section 5.0. Containment barriers for the tritium plume were investigated. FY 94 work focused on stabilization of chromium contaminated soil with blast furnace slag modified grouts to bypass the stage of pre-reduction of Cr(6), barriers for tritiated water containment at the MWL, continued study of barriers for the CWL, and jet grouting field trials for CWL barriers at an uncontaminated site at SNL. Cores from the FY 93 permeation grouting field trails were also tested in FY 94

  8. Selection of candidate canister materials for high-level nuclear waste containment in a tuff repository

    International Nuclear Information System (INIS)

    A repository located at Yucca Mountain at the Nevada Test Site is a potential site for permanent geological disposal of high-level nuclear waste. The repository can be located in a horizon in welded tuff, a volcanic rock, which is above the static water level at this site. The environmental conditions in this unsaturated zone are expected to be air and water vapor dominated for much of the containment period. Type 304L stainless steel is the reference material for fabricating canisters to contain the solid high-level wastes. Alternative stainless alloys are considered because of possible susceptibility of 304L to localized and stress forms of corrosion. For the reprocessed glass wastes, the canisters serve as the recipient for pouring the glass with the result that a sensitized microstructure may develop because of the times at elevated temperatures. Corrosion testing of the reference and alternative materials has begun in tuff-conditioned water and steam environments. 21 references, 8 figures, 8 tables

  9. Process for converting sodium nitrate-containing, caustic liquid radioactive wastes to solid insoluble products

    Science.gov (United States)

    Barney, Gary S.; Brownell, Lloyd E.

    1977-01-01

    A method for converting sodium nitrate-containing, caustic, radioactive wastes to a solid, relatively insoluble, thermally stable form is provided and comprises the steps of reacting powdered aluminum silicate clay, e.g., kaolin, bentonite, dickite, halloysite, pyrophyllite, etc., with the sodium nitrate-containing radioactive wastes which have a caustic concentration of about 3 to 7 M at a temperature of 30.degree. C to 100.degree. C to thereby entrap the dissolved radioactive salts in the aluminosilicate matrix. In one embodiment the sodium nitrate-containing, caustic, radioactive liquid waste, such as neutralized Purex-type waste, or salts or oxide produced by evaporation or calcination of these liquid wastes (e.g., anhydrous salt cake) is converted at a temperature within the range of 30.degree. C to 100.degree. C to the solid mineral form-cancrinite having an approximate chemical formula 2(NaAlSiO.sub.4) .sup.. xSalt.sup.. y H.sub.2 O with x = 0.52 and y = 0.68 when the entrapped salt is NaNO.sub.3. In another embodiment the sodium nitrate-containing, caustic, radioactive liquid is reacted with the powdered aluminum silicate clay at a temperature within the range of 30.degree. C to 100.degree. C, the resulting reaction product is air dried eitheras loose powder or molded shapes (e.g., bricks) and then fired at a temperature of at least 600.degree. C to form the solid mineral form-nepheline which has the approximate chemical formula of NaAlSiO.sub.4. The leach rate of the entrapped radioactive salts with distilled water is reduced essentially to that of the aluminosilicate lattice which is very low, e.g., in the range of 10.sup.-.sup.2 to 10.sup.-.sup.4 g/cm.sup.2 -- day for cancrinite and 10.sup.-.sup.3 to 10.sup.-.sup.5 g/cm.sup.2 -- day for nepheline.

  10. Corrosion behaviour of container materials for geological disposal of high level radioactive waste

    International Nuclear Information System (INIS)

    The disposal of high level radioactive waste in geological formations, based on the multibarrier concept, may include the use of a container as one of the engineered barriers. In this report the requirements imposed on this container and the possible degradation processes are reviewed. Further on an overview is given of the research being carried out by various research centres in the European Community on the assessment of the corrosion behaviour of candidate container materials. The results obtained on a number of materials under various testing conditions are summarized and evaluated. As a result, three promising materials have been selected for a detailed joint testing programme. It concerns two highly corrosion resistant alloys, resp. Ti-Pd (0.2 Pd%) and Hastelloy C4 and one consumable material namely a low carbon steel. Finally the possibilities of modelling the corrosion phenomena are discussed

  11. Progress in welding studies for Canadian nuclear fuel waste disposal containers

    International Nuclear Information System (INIS)

    This report describes the progress in the development of closure-welding technology for Canadian nuclear fuel waste disposal containers. Titanium, copper and Inconel 625 are being investigated as candidate materials for fabrication of these containers. Gas-tungsten-arc welding, gas metal-arc-welding, resistance-heated diffusion bonding and electron beam welding have been evaluated as candidate closure welding processes. Characteristic weldment properties, relative merits of welding techniques, suitable weld joint configurations and fit-up tolerances, and welding parameter control ranges have been identified for various container designs. Furthermore, the automation requirements for candidate welding processes have been assessed. Progress in the development of a computer-controlled remote gas-shielded arc welding system is described

  12. Greater-than-Class C low-level radioactive waste shipping package/container identification and requirements study

    International Nuclear Information System (INIS)

    This report identifies a variety of shipping packages (also referred to as casks) and waste containers currently available or being developed that could be used for greater-than-Class C (GTCC) low-level waste (LLW). Since GTCC LLW varies greatly in size, shape, and activity levels, the casks and waste containers that could be used range in size from small, to accommodate a single sealed radiation source, to very large-capacity casks/canisters used to transport or dry-store highly radioactive spent fuel. In some cases, the waste containers may serve directly as shipping packages, while in other cases, the containers would need to be placed in a transport cask. For the purpose of this report, it is assumed that the generator is responsible for transporting the waste to a Department of Energy (DOE) storage, treatment, or disposal facility. Unless DOE establishes specific acceptance criteria, the receiving facility would need the capability to accept any of the casks and waste containers identified in this report. In identifying potential casks and waste containers, no consideration was given to their adequacy relative to handling, storage, treatment, and disposal. Those considerations must be addressed separately as the capabilities of the receiving facility and the handling requirements and operations are better understood

  13. Alkaline degradation of organic materials contained in TRU wastes under repository conditions

    International Nuclear Information System (INIS)

    Alkaline degradation tests for 9 organic materials were conducted under the conditions of TRU waste disposal: anaerobic alkaline conditions. The tests were carried out at 90degC for 91 days. The sample materials for the tests were selected from the standpoint of constituent organic materials of TRU wastes. It has been found that cellulose and plastic solidified products are degraded relatively easily and that rubbers are difficult to degrade. It could be presumed that the alkaline degradation of organic materials occurs starting from the functional group in the material. Therefore, the degree of degradation difficulty is expected to be dependent on the kinds of functional group contained in the organic material. (author)

  14. Determination of mass attenuation coefficients of concretes containing ulexite and ulexite concentrator waste

    International Nuclear Information System (INIS)

    Highlights: • Concretes containing ulexite and ulexite concentrator waste were produced. • Mass attenuation coefficients were determined for 59.54 and 80.99 keV. • Mass attenuation coefficients depend on the rate of these materials in concrete. • Shielding capacity of concrete can be enhanced by using these materials. - Abstract: Purposes of this study are to examine photon attenuation properties of concretes including ulexite and ulexite concentrator waste and to present an alternative shielding material in order to decrease the intensity of gamma radiation. In order to investigate the radiation transmission of these concretes, mass attenuation coefficients at 59.54 and 80.99 keV photons energies were measured by executing a transmission geometry with NaI(Tl) scintillation detector and calculated by WinXCom computer program

  15. Standard practices for dissolving glass containing radioactive and mixed waste for chemical and radiochemical analysis

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2000-01-01

    1.1 These practices cover techniques suitable for dissolving glass samples that may contain nuclear wastes. These techniques used together or independently will produce solutions that can be analyzed by inductively coupled plasma atomic emission spectroscopy (ICP-AES), inductively coupled plasma mass spectrometry (ICP-MS), atomic absorption spectrometry (AAS), radiochemical methods and wet chemical techniques for major components, minor components and radionuclides. 1.2 One of the fusion practices and the microwave practice can be used in hot cells and shielded hoods after modification to meet local operational requirements. 1.3 The user of these practices must follow radiation protection guidelines in place for their specific laboratories. 1.4 Additional information relating to safety is included in the text. 1.5 The dissolution techniques described in these practices can be used for quality control of the feed materials and the product of plants vitrifying nuclear waste materials in glass. 1.6 These pr...

  16. Extraction and separation of uranium from simulated uranium-containing liquid wastes of Ningyo-toge Environmental Engineering Center

    International Nuclear Information System (INIS)

    An effective mass processing equipment using solvent extraction method, named 'emulsion flow extractor', is the most promising apparatus for removal and recovery of uranium from uranium-containing liquid wastes originated from decontamination of uranium-contaminated fluoride waste in the uranium conversion test facility and of used gas centrifuges in the uranium enrichment facility at Ningyo-toge environmental engineering center of Japan Atomic Energy Agency. Prior to application of the emulsion flow extractor for actual uranium-containing liquid wastes of Ningyo-toge environmental engineering center, properties of some phosphorous extractants for extraction and separation of uranium and constituents from simulated liquid wastes were examined through batch tests. These preliminary tests revealed that D2EHPA would be a promising candidate for extractant used for treatment of the actual uranium-containing liquid wastes, and that the extractants with a surfactant like AOT would not be useful. (author)

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

    International Nuclear Information System (INIS)

    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

  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. Part 1: Participatory Ergonomics Approach to Waste Container Handling Utilizing a Multidisciplinary Team

    Energy Technology Data Exchange (ETDEWEB)

    Zalk, D.M.; Tittiranonda, P.; Burastero, S.; Biggs, T.W.; Perry, C.M.; Tageson, R.; Barsnick, L.

    2000-02-07

    This multidisciplinary team approach to waste container handling, developed within the Grassroots Ergonomics process, presents participatory ergonomic interpretations of quantitative and qualitative aspects of this process resulting in a peer developed training. The lower back, shoulders, and wrists were identified as frequently injured areas, so these working postures were a primary focus for the creation of the workers' training. Handling procedures were analyzed by the team to identify common cycles involving one 5 gallon (60 pounds), two 5 gallons (60 and 54 pounds), 30 gallon (216 pounds), and 55 gallon (482 pounds) containers: lowering from transporting to/from transport vehicles, loading/unloading on transport vehicles, and loading onto pallet. Eleven experienced waste container handlers participated in this field analysis. Ergonomic exposure assessment tools measuring these field activities included posture analysis, posture targeting, Lumbar Motion Monitor{trademark} (LMM), and surface electromyography (sEMG) for the erector spinae, infraspinatus, and upper trapezius muscles. Posture analysis indicates that waste container handlers maintained non-neutral lower back postures (flexion, lateral bending, and rotation) for a mean of 51.7% of the time across all activities. The right wrist was in non-neutral postures (radial, ulnar, extension, and flexion) a mean of 30.5% of the time and the left wrist 31.4%. Non-neutral shoulder postures (elevation) were the least common, occurring 17.6% and 14.0% of the time in the right and left shoulders respectively. For training applications, each cycle had its own synchronized posture analysis and posture target diagram. Visual interpretations relating to the peak force modifications of the posture target diagrams proved to be invaluable for the workers' understanding of LMM and sEMG results (refer to Part II). Results were reviewed by the team's field technicians and their interpretations were developed

  20. DURABILITY OF GREEN CONCRETE WITH TERNARY CEMENTITIOUS SYSTEM CONTAINING RECYCLED AGGREGATE CONCRETE AND TIRE RUBBER WASTES

    Directory of Open Access Journals (Sweden)

    MAJID MATOUQ ASSAS

    2016-06-01

    Full Text Available All over the world billions of tires are being discarded and buried representing a serious ecological threat. Up to now a small part is recycled and millions of tires are just stockpiled, landfilled or buried. This paper presents results about the properties and the durability of green concrete contains recycled concrete as a coarse aggregate with partial replacement of sand by tire rubber wastes for pavement use. Ternary cementious system, Silica fume, Fly ash and Cement Kiln Dust are used as partial replacement of cement by weight. Each one replaced 10% of cement weight to give a total replacement of 30%. The durability performance was assessed by means of water absorption, chloride ion permeability at 28 and 90 days, and resistance to sulphuric acid attack at 1, 7, 14 and 28 days. Also to the compression behaviors for the tested specimens at 7, 14, 28 and 90 days were detected. The results show the existence of ternary cementitious system, silica fly ash and Cement Kiln Dust minimizes the strength loss associated to the use of rubber waste. In this way, up to 10% rubber content and 30% ternary cementious system an adequate strength class value (30 MPa, as required for a wide range of common structural uses, can be reached both through natural aggregate concrete and recycled aggregate concrete. Results also show that, it is possible to use rubber waste up to 15% and still maintain a high resistance to acid attack. The mixes with 10%silica fume, 10% fly ash and 10% Cement Kiln Dust show a higher resistance to sulphuric acid attack than the reference mix independently of the rubber waste content. The mixes with rubber waste and ternary cementious system was a lower resistance to sulphuric acid attack than the reference mix.

  1. The Treatment of Low Level Radioactive Liquid Waste Containing Detergent by Biological Activated Sludge Process

    International Nuclear Information System (INIS)

    The treatment of low level radioactive liquid waste containing persil detergent from laundry operation of contaminated clothes by activated sludge process has been done, for alternative process replacing the existing treatment by evaporation. The detergent concentration in water solution from laundry operation is 14.96 g/l. After rinsing operation of clothes and mixing of laundry water solution with another liquid waste, the waste water solution contains about ≤ 1.496 g/l of detergent and 10-3 Ci/m3 of Cs-137 activity. The simulation waste having equivalent activity of Cs-137 10-3 Ci/m3, detergent content (X) 1.496, 0.748, 0.374, 0.187, 0.1496 and 0.094 g/l on BOD value respectively 186, 115, 71, 48, 19, and 16 ppm was processed by activated sludge in reactor of 18.6 l capacity on ambient temperature. It is used Super Growth Bacteria (SGB) 102 and SGB 104, nitrogen and phosphor nutrition, and aeration. The result show that bacteria of SGB 102 and SGB 104 were able to degrade the persil detergent for attaining standard quality of water release category B in which BOD values 6 ppm. It was need 30 hours for X ≤ 0.187 g/l, 50 hours for 0.187 < X ≤ 0.374 g/l, 75 hours for 0.374 < X ≤ 0.748, and 100 hours for 0.748 < X ≤ 1.496 g/l. On the initial period the bacteria of SGB 104 interact most quickly to degrade the detergent comparing SGB 102. Biochemical oxidation process decontaminate the solution on the decontamination factor of 350, Cs-137 be concentrate in sludge by complexing with the bacteria wall until the activity of solution be become very low. (author)

  2. Precipitation and Deposition of Aluminum-Containing Phases in Tank Wastes. Final Report

    International Nuclear Information System (INIS)

    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. The principal focus of our research was to maintain the fluidity of aluminum- or silicon-containing suspensions and solutions during transport, whether by preventing particle formation, stabilizing colloidal particles in suspension, or by combining partial dissolution with particle stabilization. We have found that all of these can be effected in aluminum-containing solutions using the simple organic, citric acid. Silicon-containing solutions were found to be less tractable, but we have strong indications that chemistries similar to the citric acid/aluminum suspensions can be effective in maintaining silicon suspensions at high alkalinities. In the first phase of our study, we focused on the use of simple organics to raise the solubility of aluminum oxyhydroxides in high alkaline aqueous solvents. In a limited survey of common organic acids, we determined that citric acid had the highest potential to achieve our goal. However, our subsequent investigation revealed that the citric acid appeared to play two roles in the solutions: first, raising the concentration of aluminum in highly alkaline solutions by breaking up or inhibiting 'seed' polycations and thereby delaying the nucleation and growth of particles; and second, stabilizing nanometer-sized particles in suspension when nucleation did occur. The second phase of our work involved the solvation of silicon, again in solutions of high alkalinity. Here, the use of polyols was determined to be effective in

  3. Study of the sulphate expansion phenomenon in concrete: behaviour of the cemented radioactive wastes containing sulphate

    International Nuclear Information System (INIS)

    Sulphate attack is one of the major degradation processes of concrete. It is especially important in storing cemented radioactive wastes containing sulphate. In this thesis, we have thoroughly investigated the degradation mechanisms of cemented radioactive wastes by sulphate. The CaO-Al2O3-SO3-H2O systems with and without alkalis are studied. For the system without alkalis, experimental results show that it is the formation of a secondary ettringite under external water supply by steric effect that causes the expansion. For the system with alkalis, the ettringite does not appear while a new mineral called 'U', a sodium-substituted AFm phase is detected. This phase is shown to be responsible for the expansion and destruction of the specimens. The conditions for the formation, the product of solubility and many means of its synthesis are discussed, and a complete list of the inter-reticular distances file is given. The behaviour of the different types of cemented wastes containing sulphate are then studied with a special focus on the U phase on entity which was heretofore very little understood. The following three hypothetical mechanisms of sulphate expansion are proposed: the formation of the secondary U phase, the transformation of the U phase to the ettringite and the topochemical hydration of thenardite into mirabilite. Experiments on a simplified system have demonstrated clearly that the formation of the secondary U phase can induce enormous expansion by steric effect, this justifying the first assumption. Simulation by the mass and volume balances is carried out thereafter and enables us to estimate the expansion induced by the formation of the secondary U phase in the cemented wastes. The second assumption is also well verified by a series of leaching tests in different solutions on mixtures containing the U phase. On the basis of the analysis of the specimens under leaching, it has been assumed that the expansion is associated with the inter

  4. Towards zero discharge of chromium-containing leather waste through improved alkali hydrolysis.

    Science.gov (United States)

    Mu, Changdao; Lin, Wei; Zhang, Mingrang; Zhu, Qingshi

    2003-01-01

    The treatment of chromium-containing leather waste (CCLW), the major solid waste generated at the post-tanning operations of leather processing, has the potential to generate value-added leather chemicals. Various alkali and enzymatic hydrolysis were compared, and calcium oxide was found to be important for effective (but still incomplete) hydrolysis. Three possible reasons are given for the incomplete hydrolysis under alkaline conditions. Data for 19 amino acids are presented for four different treatment products. On the basis of the results, a novel three-step CCLW treatment process is proposed. The gelatin extracted in the first step is chemically modified to produce leather finishing agents. The collagen hydrolysates isolated in the second step are used as proteinic retanning agents by chemical modification. The remaining chrome cake is further hydrolyzed with acids in the third step, and the obtained chromium-containing protein hydrolysates could be used for the preparation of chromium-containing retanning agents for leather industry. The proposed three-step process provides a feasible zero discharge process for the treatment of CCLW. PMID:14583246

  5. Corrosion studies on containment materials for vitrified high level nuclear waste

    International Nuclear Information System (INIS)

    Progress is reported on a research programme designed to identify containment materials that could be used to isolate nuclear waste for 500 to 1000 years after disposal. The main emphasis in this reporting period has been on the general corrosion of carbon steels selected as candidates for corrosion allowance containers. Carbon steel coupons embedded in crushed granite under aerated synthetic granite ground water at 90 deg C for six months exhibited a general corrosion rate of about 20 μm yr-1. Eighteen additional long term tank immersion tests are in progress to investigate the corrosion behaviour of plain and welded carbon steel samples under a range of experimental conditions. Experiments with carbon steel exposed for 7500 hours to deaerated seawater at 90 deg C demonstrated that the maximum general corrosion rate was -1 in the absence of oxygen. Preliminary results from identical experiments conducted in a low dose rate cobalt-60 radiation cell indicate that this rate of corrosion was unaffected by a radiation dose of 285 Rh-1. A mathematical model has been formulated to describe the general corrosion behaviour of carbon steel containers buried in an environment typical of a waste repository. This has indicated that the long term general corrosion rate will settle at approx. 3.5 μm yr-1. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, J.C.; Van Konynenburg, R.A.; McCright, R.D. (Lawrence Livermore National Lab., CA (USA)); Bullen, D.B. (Science and Engineering Associates, Inc., Pleasanton, CA (USA))

    1988-04-01

    Three iron- to nickel-based austenitic alloys (Types 304L and 316L stainless steels and Alloy 825) are being considered as candidate materials for the fabrication of high-level radioactive-waste containers. Waste will include fuel assemblies from reactors as well as high-level waste in borosilicate glass forms, and will be sent to the prospective repository at Yucca Mountain, Nevada. The decay of radionuclides in the repository will result in the generation of substantial heat and in fluences of gamma radiation. Container materials may undergo any of several modes of degradation in this environment, including atmospheric oxidation; uniform aqueous phase corrosion; pitting; crevice corrosion; sensitization and intergranular stress corrosion cracking (IGSCC); and transgranular stress corrosion cracking (TGSCC). This report is an analysis of data relevant to the pitting, crevice corrosion, and stress corrosion cracking (SCC) of the three austenitic candidate alloys. The candidates are compared in terms of their susceptibilities to these forms of corrosion. Although all three candidates have demonstrated pitting and crevice corrosion in chloride-containing environments, Alloy 825 has the greatest resistance to these types of localized corrosion (LC); such resistance is important because pits can penetrate the metal and serve as crack initiation sites. Both Types 304L and 316L stainless steels are susceptible to SCC in acidic chloride media. In contrast, SCC has not been documented in Alloy 825 under comparable conditions. Gamma radiation has been found to enhance SCC in Types 304 and 304L stainless steels, but it has no detectable effect on the resistance of Alloy 825 to SCC. Furthermore, while the effects of microbiologically induced corrosion have been observed for 300-series stainless steels, nickel-based alloys such as Alloy 825 seem to be immune to such problems. 211 refs., 49 figs., 10 tabs.

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

    International Nuclear Information System (INIS)

    Three iron- to nickel-based austenitic alloys (Types 304L and 316L stainless steels and Alloy 825) are being considered as candidate materials for the fabrication of high-level radioactive-waste containers. Waste will include fuel assemblies from reactors as well as high-level waste in borosilicate glass forms, and will be sent to the prospective repository at Yucca Mountain, Nevada. The decay of radionuclides in the repository will result in the generation of substantial heat and in fluences of gamma radiation. Container materials may undergo any of several modes of degradation in this environment, including atmospheric oxidation; uniform aqueous phase corrosion; pitting; crevice corrosion; sensitization and intergranular stress corrosion cracking (IGSCC); and transgranular stress corrosion cracking (TGSCC). This report is an analysis of data relevant to the pitting, crevice corrosion, and stress corrosion cracking (SCC) of the three austenitic candidate alloys. The candidates are compared in terms of their susceptibilities to these forms of corrosion. Although all three candidates have demonstrated pitting and crevice corrosion in chloride-containing environments, Alloy 825 has the greatest resistance to these types of localized corrosion (LC); such resistance is important because pits can penetrate the metal and serve as crack initiation sites. Both Types 304L and 316L stainless steels are susceptible to SCC in acidic chloride media. In contrast, SCC has not been documented in Alloy 825 under comparable conditions. Gamma radiation has been found to enhance SCC in Types 304 and 304L stainless steels, but it has no detectable effect on the resistance of Alloy 825 to SCC. Furthermore, while the effects of microbiologically induced corrosion have been observed for 300-series stainless steels, nickel-based alloys such as Alloy 825 seem to be immune to such problems. 211 refs., 49 figs., 10 tabs

  8. System of large transport containers for waste from dismantling light water and gas-cooled nuclear reactors. Volume 2

    International Nuclear Information System (INIS)

    The purpose of this volume is to assess the means of transportation of decommissioning wastes, costs of transport, radiological detriment attributable to transport and develops conceptual designs of large transport containers. The document ends with Conclusions and Recommendations

  9. An optimized process for tritium-containing waste water collection of High-Temperature Gas-cooled Reactor

    International Nuclear Information System (INIS)

    Highlights: • An optimized process for tritium-containing waste water collection of High-Temperature Gas-cooled Reactor was developed. • The optimized process and verification experiment using the HTR-10 were presented in detail. • A large quantity of high-dose tritium-containing waste water was successfully collected in commissioning experiment of the improved HTR-10. • The optimized process was proved to be reliable to avoid the large emission of radioactive waste water to the environment. - Abstract: An optimized process for tritium-containing waste water collection of High-Temperature Gas-cooled Reactor (HTGR) was developed and experimentally verified using the 10 MW High-Temperature Gas-cooled Reactor-test module (HTR-10). Compared with the previous process, an auxiliary molecular sieve bed was added in helium purification regeneration system and new operation process was proposed to collect tritium-containing waste water. In this paper, the optimized process and verification experiment were presented in detail. In commissioning experiment of the improved HTR-10, a large quantity of high-dose tritium-containing waste water was successfully collected in the water separator of helium purification regeneration system, with the specific activity being 6.1 × 109 Bq/L. The verification experiment confirms that the optimized process is effective and reliable for the demonstration plant design of High Temperature Gas-cooled Reactor-Pebble bed module (HTR-PM) to avoid the large emission of detrimentally radioactive waste water to the environment

  10. The Precipitation Process of Liquid Wastes Containing Contaminant Am withBarium Sulfate

    International Nuclear Information System (INIS)

    The investigated of the reduction volume liquid wastes containing ofAmericium nuclide contaminant has been done. The reduction volume was done byadding barium sulfate coagulant. The experimental procedure that has beendone by adding regent of barium nitrate and natrium sulfate to the wasteswith its preadjusted pH, then by utilizing the jar test equipment was carriedout the fast stirring speed for 5 minutes and the gentle agitation for 30minutes, therefor its floc and supernatant will be formed. The resultedbarium sulfate floc will trap radionuclide in the wastes. The Variableinvestigated were: the concentration of barium sulfate, pH of the wastes, theflash mixing rate, the gentle agitation rate. The investigated barium sulfateconcentration variable was started from 100 ppm up to 800 ppm. Theinvestigated pH variable was started from pH 7 up to pH 13. The investigatedflash mixing rate were 75, 100, 125, 150, 175, 200, 225, 250 rpm. Theinvestigated gentle agitation variable were 20, 30, 40, 50 rpm. The bestresult which was represented by decontaminating factor (DF) was found frombarium sulfate concentration of 300 ppm and pH 11, and the flash mixing rateof 200 rpm and the gentle agitation rate of 20 rpm, with the separationefficiency = 97.2 %. (author)

  11. Removal of organic dyes using Cr-containing activated carbon prepared from leather waste.

    Science.gov (United States)

    Oliveira, Luiz C A; Coura, Camila Van Zanten; Guimarães, Iara R; Gonçalves, Maraisa

    2011-09-15

    In this work, hydrogen peroxide decomposition and oxidation of organics in aqueous medium were studied in the presence of activated carbon prepared from wet blue leather waste. The wet blue leather waste, after controlled pyrolysis under CO(2) flow, was transformed into chromium-containing activated carbons. The carbon with Cr showed high microporous surface area (up to 889 m(2)g(-1)). Moreover, the obtained carbon was impregnated with nanoparticles of chromium oxide from the wet blue leather. The chromium oxide was nanodispersed on the activated carbon, and the particle size increased with the activation time. It is proposed that these chromium species on the carbon can activate H(2)O(2) to generate HO radicals, which can lead to two competitive reactions, i.e. the hydrogen peroxide decomposition or the oxidation of organics in water. In fact, in this work we observed that activated carbon obtained from leather waste presented high removal of methylene blue dye combining the adsorption and oxidation processes.

  12. Corrosion of steel drums containing simulated radioactive waste of low and intermediate level

    International Nuclear Information System (INIS)

    Ion-exchange resins are frequently used during the operation of nuclear power plants and constitute radioactive waste of low and intermediate level. For the final disposal inside the repository the resins are immobilized by cementation and placed inside steel drums. The eventful contamination of the resins with aggressive species may cause corrosion problems to the drums. In order to assess the incidence of this phenomenon and to estimate the lifespan of the steel drums, in the present work, the corrosion susceptibility of steel drums in contact with cemented ion-exchange resins contaminated with different aggressive species was studied. The aggressive species studied were chloride ions (main ionic species of concern) and sulphate ions (produced during radiolysis of the cationic exchange-resins after cementation). The corrosion rate of the steel was monitored over a time period of 900 days and a chemical and morphological analysis of the corrosion products formed on the steel in each condition was performed. When applying the results obtained in the present work to estimate the corrosion depth of the drums containing the cemented radioactive waste after a period of 300 years (foreseen durability of the Low and Intermediate Level Radioactive Waste facility in Argentina), it was found that in the most unfavourable case (high chloride contamination), the corrosion penetration will be considerably lower than the thickness of the wall of the steel drums. (author)

  13. Green route for the utilization of chrome shavings (chromium-containing solid waste) in tanning industry.

    Science.gov (United States)

    Rao, Jonnalagadda Raghava; Thanikaivelan, Palanisamy; Sreeram, Kalarical Janardhanan; Nair, Balachandran Unni

    2002-03-15

    Chromium-containing wastes from various industrial sectors are under critical review. Leather processing is one such industrial activity that generates chromium-bearing wastes in different forms. One of them is chrome shavings, and this contributes to an extent of 10% of the quantum of raw skins/hides processed, amounting to 0.8 million ton globally. In this study, the high protein content of chrome shavings has been utilized for reduction of chromium(VI) in the preparation of chrome tanning agent. This approach has been exploited for the development of two products: one with chrome shavings alone as reducing agent and the other with equal proportion of chrome shavings and molasses. The developed products exhibit more masking due to the formation of intermediate organic oligopeptides. This has been corroborated through the spectral, hydrolysis, and species-wise distribution studies. The formation of these organic masking agents helps in chrome tanning by shifting the precipitation point of chromium to relatively higher pH levels. Hence, the developed products find use as chrome tanning agents for leather processing, thus providing a means for better utilization of chrome shaving wastes. PMID:11944695

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

    International Nuclear Information System (INIS)

    This volume surveys the effects of welding on the degradation modes of three austenitic alloys: Types 304L and 316L stainless steels and Alloy 825. These materials are candidates for the fabrication of containers for the long-term storage of high-level nuclear waste. The metallurgical characteristics of fusion welds are reviewed here and related to potential degradation modes of the containers. Three specific areas are discussed in depth: (1) decreased resistance to corrosion in the forms of preferential corrosion, sensitization, and susceptibility to stress corrosion cracking, (2) hot cracking in the heat-affected zone and the weld zone, and (3) formation of intermetallic phases. The austenitic alloys are ranked as follows in terms of overall weldability: Alloy 825 (best) > Type 316L stainless steel > Type 304L stainless steel (worst). 108 refs., 31 figs., 7 tabs

  15. Production of technical-grade sodium citrate from glycerol-containing biodiesel waste by Yarrowia lipolytica.

    Science.gov (United States)

    Kamzolova, Svetlana V; Vinokurova, Natalia G; Lunina, Julia N; Zelenkova, Nina F; Morgunov, Igor G

    2015-10-01

    The production of technical-grade sodium citrate from the glycerol-containing biodiesel waste by Yarrowia lipolytica was studied. Batch experiments showed that citrate was actively produced within 144 h, then citrate formation decreased presumably due to inhibition of enzymes involved in this process. In contrast, when the method of repeated batch cultivation was used, the formation of citrate continued for more than 500 h. In this case, the final concentration of citrate in the culture liquid reached 79-82 g/L. Trisodium citrate was isolated from the culture liquid filtrate by the addition of a small amount of NaOH, so that the pH of the filtrate increased to 7-8. This simple and economic isolation procedure gave the yield of crude preparation containing trisodium citrate 5.5-hydrate up to 82-86%.

  16. Estimation of the atmospheric corrosion on metal containers in industrial waste disposal.

    Science.gov (United States)

    Baklouti, M; Midoux, N; Mazaudier, F; Feron, D

    2001-08-17

    Solid industrial waste are often stored in metal containers filled with concrete, and placed in well-aerated warehouses. Depending on meteorological conditions, atmospheric corrosion can induce severe material damages to the metal casing, and this damage has to be predicted to achieve safe storage. This work provides a first estimation of the corrosivity of the local atmosphere adjacent to the walls of the container through a realistic modeling of heat transfer phenomena which was developed for this purpose. Subsequent simulations of condensation/evaporation of the water vapor in the atmosphere were carried out. Atmospheric corrosion rates and material losses are easily deduced. For handling realistic data and comparison, two different meteorological contexts were chosen: (1) an oceanic and damp atmosphere and (2) a drier storage location. Some conclusions were also made for the storage configuration in order to reduce the extent of corrosion phenomena. PMID:11489528

  17. Estimating Time Loss Effects On Municipal Solid Waste Collection Using Haul Container System In Uyo Nigeria

    Directory of Open Access Journals (Sweden)

    Obot E. Essien

    2013-10-01

    Full Text Available - Time loss in time and motion study of the collection of municipal solid waste in Uyo metropolis was observed to affect the round-trip time, the solid waste generation rate and the collection efficiency of the haul container system of solid waste management, and hence needed information to drive control or reduction in the service. The result showed that its effects depended on the truck, route zone and operators skill in maneuvering the routes to reduce the dead ends and waste hours. Seven components of time losses with values ranging from 7 to 40 minutes per trip were measured, giving valuable total times loss per service truck per day as 2.0 hr for zones 2, 3 and 6, and 1.95hr for zone 4. The time loss for collection efficiency showed significant difference (P = 0.05 between zones and trucks, and varied as 19%, 20%, 7% and 30% for trucks 046, 053, 060 and 072 used in zones 03, 02, 04 and 06 respectively. Trucks for zones 05 and 01 were invalid. The available time was thus reduced. With average cycle time of 17.30 min to 24.21 min per trip, such loss time, in turn, reduced collection efficiency by 20 to 25% per truck thereby reducing the total trips and daily turnover. Recommendations include micro-routing principles, operators’ motivation with team spirit and avoidance of observed start-up delays. Also route re-design of more dense zones and sparsely populated zones are recommended in order to bring trip time to near equality.

  18. Cement Solidification Method For Intermediate-Level Liquid Waste Containing Sodium Sulphate (Na2SO4)

    International Nuclear Information System (INIS)

    A new cement solidification method for intermediate-level liquid waste containing large amounts of sodium sulphate (Na2SO4) has been developed. This method involves two safety concepts for disposal sites: reduction in the amount of sulphate ion (SO42-) released from solidified wastes and reduction in the amount of hydrogen gas generated due to radiolysis of the water present in the solidified waste. In order to eliminate SO42- release from solidified wastes, two chemical reactions were important in our solidification method: (1) Barium-compounds (Ba(OH)2.8H2O, etc) were reacted with SO42- to form BaSO4, and (2) using alumina cement material, SO42- was mineralized as ettringite, 3CaO.Al2O3.3CaSO3.2H2O. Based on leaching tests, the amount of SO42- released from the solidified forms into ion exchange water under anaerobic conditions was less than 1 x 10-3 mol/L. Thus, this method should be effective in preventing engineered concrete barrier layers from cracking. In order to evaluate the amount of hydrogen gas generated from cement solids due to radiolysis of hydrated and non-hydrated water in the solid, gamma-ray irradiation experiments on solidified alumina cement (ALC), solidified ordinary portland cement (OPC), solidified ordinary portland cement blended with blast-furnace slag (OPC-BFS), and synthetic ettringite were performed. As a result, the generation rate of hydrogen gas from ALC was less than those from OPC and OPC-BFS and approximately equal to that from ettringite. (authors)

  19. Solid waste containing persistent organic pollutants in Serbia: From precautionary measures to the final treatment (case study).

    Science.gov (United States)

    Stevanovic-Carapina, Hristina; Milic, Jelena; Curcic, Marijana; Randjelovic, Jasminka; Krinulovic, Katarina; Jovovic, Aleksandar; Brnjas, Zvonko

    2016-07-01

    Sustainable solid waste management needs more dedicated attention in respect of environmental and human health protection. Solid waste containing persistent organic pollutants is of special concern, since persistent organic pollutants are persistent, toxic and of high risk to human health and the environment. The objective of this investigation was to identify critical points in the Serbian system of solid waste and persistent organic pollutants management, to assure the life cycle management of persistent organic pollutants and products containing these chemicals, including prevention and final destruction. Data were collected from the Serbian competent authorities, and led us to identify preventive actions for solid waste management that should reduce or minimise release of persistent organic pollutants into the environment, and to propose actions necessary for persistent organic pollutants solid waste. The adverse impact of persistent organic pollutants is multidimensional. Owing to the lack of treatment or disposal plants for hazardous waste in Serbia, the only option at the moment to manage persistent organic pollutants waste is to keep it in temporary storage and when conditions are created (primarily financial), such waste should be exported for destruction in hazardous waste incinerators. Meanwhile, it needs to be assured that any persistent organic pollutants management activity does not negatively impact recycling flows or disturb progress towards a more circular economy in Serbia.

  20. Solid waste containing persistent organic pollutants in Serbia: From precautionary measures to the final treatment (case study).

    Science.gov (United States)

    Stevanovic-Carapina, Hristina; Milic, Jelena; Curcic, Marijana; Randjelovic, Jasminka; Krinulovic, Katarina; Jovovic, Aleksandar; Brnjas, Zvonko

    2016-07-01

    Sustainable solid waste management needs more dedicated attention in respect of environmental and human health protection. Solid waste containing persistent organic pollutants is of special concern, since persistent organic pollutants are persistent, toxic and of high risk to human health and the environment. The objective of this investigation was to identify critical points in the Serbian system of solid waste and persistent organic pollutants management, to assure the life cycle management of persistent organic pollutants and products containing these chemicals, including prevention and final destruction. Data were collected from the Serbian competent authorities, and led us to identify preventive actions for solid waste management that should reduce or minimise release of persistent organic pollutants into the environment, and to propose actions necessary for persistent organic pollutants solid waste. The adverse impact of persistent organic pollutants is multidimensional. Owing to the lack of treatment or disposal plants for hazardous waste in Serbia, the only option at the moment to manage persistent organic pollutants waste is to keep it in temporary storage and when conditions are created (primarily financial), such waste should be exported for destruction in hazardous waste incinerators. Meanwhile, it needs to be assured that any persistent organic pollutants management activity does not negatively impact recycling flows or disturb progress towards a more circular economy in Serbia. PMID:27281225

  1. Status and use of the Rocky Flats Environmental Technology Site Pipe Overpack Container for TRU waste storage and shipments

    International Nuclear Information System (INIS)

    The Pipe Overpack Container was designed to optimize shipments of high plutonium content transuranic waste from Rocky Flats Environmental Technology Site (RFETS) to Waste Isolation Pilot Plant (WIPP). The container was approved for use in the TRUPACT-II shipping container by the Nuclear Regulatory Commission in February 1997. The container optimizes shipments to WIPP by increasing the TRUPACT-II criticality limit from 325 fissile grams equivalent (FGE) to 2,800 FGE and provides additional shielding for handling wastes with high americium-241 (Am-241) content. The container was subsequently evaluated and approved for storage of highly dispersible TRU wastes and residues at RFETS. Thermal evaluation of the container shows that the container will mitigate the impact of a worst case thermal event from reactive or potentially pyrophoric materials. These materials contain hazards postulated by the Defense Nuclear Facilities Safety Board for interim storage. Packaging these reactive or potentially pyrophoric residues in the container without stabilizing the materials is under consideration at RFETS. The design, testing, and evaluations used in the approvals, and the current status of the container usage, will be discussed

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

    International Nuclear Information System (INIS)

    Three copper-based alloys, CDA 102 (oxygen-free, high-purity copper), CDA 613 (aluminum bronze), and CDA 715 (Cu-30Ni), are candidates for the fabrication of high-level radioactive-waste disposal containers. Waste will include spent fuel assemblies from reactors as well as borosilicate glass, and will be sent to the prospective repository site at Yucca Mountain in Nye County, Nevada. The decay of radionuclides will result in the generation of substantial heat and in fluxes of gamma radiation outside the containers. In this environment, container materials might degrade by atmospheric oxidation, general aqueous phase corrosion, localized corrosion (LC), and stress corrosion cracking (SCC). This volume is a critical survey of available data on pitting and crevice corrosion of the copper-based candidates. Pitting and crevice corrosion are two of the most common forms of LC of these materials. Data on the SCC of these alloys is surveyed in Volume 4. Pitting usually occurs in water that contains low concentrations of bicarbonate and chloride anions, such as water from Well J-13 at the Nevada Test Site. Consequently, this mode of degradation might occur in the repository environment. Though few quantitative data on LC were found, a tentative ranking based on pitting corrosion, local dealloying, crevice corrosion, and biofouling is presented. CDA 102 performs well in the categories of pitting corrosion, local dealloying, and biofouling, but susceptibility to crevice corrosion diminishes its attractiveness as a candidate. The cupronickel alloy, CDA 715, probably has the best overall resistance to such localized forms of attack. 123 refs., 11 figs., 3 tabs

  3. Properties and solubility of chrome in iron alumina phosphate glasses containing high level nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    Huang, W. [School of Materials Science and Engineering, Tongji Univ., Shanghai, SH (China); Day, D.E.; Ray, C.S.; Kim, C.W.; Reis, S.T.D. [Univ. of Missouri-Rolla (United States). Graduate Center for Materials Research

    2004-10-01

    Chemical durability, glass formation tendency, and other properties of iron alumina phosphate glasses containing 70 wt% of a simulated high level nuclear waste (HLW), doped with different amounts of Cr{sub 2}O{sub 3}, have been investigated. All of the iron alumina phosphate glasses had an outstanding chemical durability as measured by their small dissolution rate (1 . 10{sup -9} g/(cm{sup 2} . min)) in deionized water at 90 C for 128 d, their low normalized mass release as determined by the product consistency test (PCT) and a barely measurable corrosion rate of <0.1 g/(m{sup 2} . d) after 7 d at 200 C by the vapor hydration test (VHT). The solubility limit for Cr{sub 2}O{sub 3} in the iron phosphate melts was estimated at 4.1 wt%, but all of the as-annealed melts contained a few percent of crystalline Cr{sub 2}O{sub 3} that had no apparent effect on the chemical durability. The chemical durability was unchanged after deliberate crystallization, 48 h at 650 C. These iron phosphate waste forms, with a waste loading of at least 70 wt%, can be readily melted in commercial refractory crucibles at 1250 C for 2 to 4 h, are resistant to crystallization, meet all current US Department of Energy requirements for chemical durability, and have a solubility limit for Cr{sub 2}O{sub 3} which is at least three times larger than that for borosilicate glasses. (orig.)

  4. Misinterpretation on the risk of radioactive cesium contained in the disaster wastes

    International Nuclear Information System (INIS)

    Osaka Prefectural Government accepted the disaster wastes contained radioactive cesium after investigation them during one year. I explained the process and discussed about the risk management by people and the self-government body. The environmental pollution by radioactive cesium and Act on Special Measures concerning the Handling of Pollution by Radioactive Materials, the progress of treatment of debris, the concentration of radioactive cesium in debris, the acceptance conditions of debris contained small amount of radioactive cesium, evaluation of effects of radioactive materials in debris on the environment, and citizen's opinion of Osaka prefecture are described. The important investigation area of radioactive contamination on the basis of Act on Special Measures concerning the Handling of Pollution by Radioactive Materials, total amount of waste from Fukushima nuclear accident and debris in Miyagi, Iwate and Fukushima prefecture, the concentration of radioactive cesium in debris in Rikuzentakata and Miyako city as of September, 2011, and cumulative number of citizen's opinion to Osaka are illustrated. (S.Y.)

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

    International Nuclear Information System (INIS)

    Three iron- to nickel-based austenitic alloys and three copper-based alloys are being considered as candidate materials for the fabrication of high-level radioactive-waste disposal containers. The austenitic alloys are Types 304L and 316L stainless steels and the high-nickel material Alloy 825. The copper-based alloys are CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni). Waste in the forms of both spent fuel assemblies from reactors and borosilicate glass will be sent to the prospective repository at Yucca Mountain, Nevada. The decay of radionuclides will result in the generation of substantial heat and gamma radiation. Container materials may undergo any of several modes of degradation in this environment, including undesirable phase transformations due to a lack of phase stability; atmospheric oxidation; general aqueous corrosion; pitting; crevice corrosion; intergranular stress corrosion cracking; and transgranular stress corrosion cracking. Problems specific to welds, such as hot cracking, may also occur. A survey of the literature has been prepared as part of the process of selecting, from among the candidates, a material that is adequate for repository conditions. The modes of degradation are discussed in detail in the survey to determine which apply to the candidate alloys and the extent to which they may actually occur. The eight volumes of the survey are summarized in Sections 1 through 8 of this overview. The conclusions drawn from the survey are also given in this overview

  6. Microbially influenced corrosion of copper nuclear fuel waste containers in a Canadian disposal vault

    International Nuclear Information System (INIS)

    An assessment of the potential for microbially influenced corrosion (MIC) of copper nuclear fuel waste containers in a Canadian disposal vault is presented. The assessment is based on a consideration of the microbial activity within a disposal vault, the reported cases of MIC of Cu alloys in the literature and the known corrosion behaviour of Cu. Because of the critical role of biofilms in the reported cases of MIC, their formation and properties are discussed in detail. Next, the literature on the MIC of Cu alloys is briefly reviewed. The various MIC mechanisms proposed are critically discussed and the implications for the corrosion of Cu containers considered. In the majority of literature cases, MIC depends on alternating aerated and deaerated environments, with accelerated corrosion being observed when fresh aerated water replaces stagnant water, e.g., the MIC of Cu-Ni heat exchangers in polluted seawater and the microbially influenced pitting of Cu water pipes. Finally, because of the predominance of corrosion by sulphate-reducing bacteria (SRB) in the MIC literature, the abiotic behaviour of Cu alloys in sulphide solutions is also reviewed. The effect of the evolving environment in a disposal vault on the extent and location of microbial activity is discussed. Biofilm formation on the container surface is considered unlikely throughout the container lifetime, but especially initially when the environmental conditions will be particularly aggressive. Microbial activity in areas of the vault away from the container is possible, however. Corrosion of the container could then occur if microbial metabolic by-products diffuse to the container surface. Sulphide, produced by the action of SRB are considered to be the most likely cause of container corrosion. It is concluded that the only likely form of MIC of Cu containers will result from sulphide produced by SRB diffusing to the container surface. A modelling procedure for predicting the extent of corrosion is

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

    Energy Technology Data Exchange (ETDEWEB)

    Vinson, D.W.; Nutt, W.M.; Bullen, D.B. [Iowa State Univ. of Science and Technology, Ames, IA (United States)

    1995-06-01

    Oxidation and atmospheric corrosion data suggest that addition of Cr provides the greatest improvement in oxidation resistance. Cr-bearing cast irons are resistant to chloride environments and solutions containing strongly oxidizing constituents. Weathering steels, including high content and at least 0.04% Cu, appear to provide adequate resistance to oxidation under temperate conditions. However, data from long-term, high-temperature oxidation studies on weathering steels were not available. From the literature, it appears that the low alloy steels, plain carbon steels, cast steels, and cast irons con-ode at similar rates in an aqueous environment. Alloys containing more than 12% Cr or 36% Ni corrode at a lower rate than plain carbon steels, but pitting may be worse. Short term tests indicate that an alloy of 9Cr-1Mo may result in increased corrosion resistance, however long term data are not available. Austenitic cast irons show the best corrosion resistance. A ranking of total corrosion performance of the materials from most corrosion resistant to least corrosion resistant is: Austenitic Cast Iron; 12% Cr = 36% Ni = 9Cr-1Mo; Carbon Steel = Low Alloy Steels; and Cast Iron. Since the materials to be employed in the Advanced Conceptual Design (ACD) waste package are considered to be corrosion allowance materials, the austenitic cast irons, high Cr steels, high Ni steels and the high Cr-Mo steels should not be considered as candidates for the outer containment barrier. Based upon the oxidation and corrosion data available for carbon steels, low alloy steels, and cast irons, a suitable list of candidate materials for a corrosion allowance outer barrier for an ACD waste package could include, A516, 2.25%Cr -- 1%Mo Steel, and A27.

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

    International Nuclear Information System (INIS)

    Three copper-based alloys --- CDA 102 (OFHC copper), CDA 613 (aluminum bronze), and CDA 715 (Cu-30Ni) --- are being considered as possible materials for the fabrication of high-level radioactive-waste disposal containers. Waste will include fuel assemblies from reactors as well as borosilicate glass forms, and will be sent to the prospective repository at Yucca Mountain, Nevada, for emplacement. The three copper-based alloys discussed here are being considered in addition to the iron- to nickel-based austenitic materials discussed in Volume 3. The decay of radionuclides will result in substantial heat generation and in fluxes of gamma radiation. In this environment, container materials may degrade by atmospheric oxidation, uniform aqueous phase corrosion, pitting, crevice corrosion, transgranular stress corrosion cracking (TGSCC) in tarnishing environments, or intergranular stress corrosion cracking (IGSCC) in nontarnishing environments. This report is a critical survey of available data on the stress corrosion cracking (SCC) of the three copper-based alloys. The requisite conditions for TGSCC and IGSCC include combinations of stress, oxygen, ammonia or nitrite, and water. Note that nitrite is generated by gamma radiolysis of moisture films in air but that ammonia is not. TGSCC has been observed in CDA 102 and CDA 613 exposed to moist ammonia-containing environments whereas SCC has not been documented for CDA 715 under similar conditions. SCC is also promoted in copper by nitrite ions. Furthermore, phosphorus-deoxidized copper is unusually susceptible to embrittlement in such environments. The presence of tin in CDA 613 prevents IGSCC. It is believed that tin segregates to grain boundaries, where it oxidizes very slowly, thereby inhibiting the oxidation of aluminum. 117 refs., 27 figs., 9 tabs

  9. Polymer-based composite materials for the fabrication of containers for the disposal of radioactive waste

    International Nuclear Information System (INIS)

    The use of carbon fibre reinforced PEEK for the fabrication of a spent nuclear fuel storage container was investigated with the irradiation of samples in the mixed radiation field of the SLOWPOKE-2 nuclear reactor at various temperatures (20oC to 75oC) and doses (up to 1.0 MGy). Mechanical testing showed that the irradiated sample properties rarely deviated from the un-irradiated samples. Chemical testing showed that the irradiated samples exhibited a greater degree of crosslinking and improved mechanical strength. Polypropylene, nylon 6,6, polycarbonate, and polyurethane, all with and without glass fibre reinforcement were also irradiated using the SLOWPOKE-2 reactor at doses from 0.5 MGy to 6.0 MGy, followed by chemical and mechanical testing to determine their suitability for low level waste storage containers. Results indicated that the major effect of irradiation was an increase in crosslinking. Simulated groundwater conditions combined with irradiation for glass fibre reinforced polycarbonate and polyurethane included immersion in a 1 M NaOH (pH 1) or a 1 M HC1 (pH 13) solution for a one month period followed by irradiation at doses of 0.5 kGy to 3.0 kGy in the SLOWPOKE-2 reactor. Flexural testing showed that the combination of chemical exposure and irradiation on these systems resulted in decrease of approximately 10% in flexural yield stress for all pH conditions. Work is ongoing to determine the combined effects of irradiation, immersion, and temperature on Nylon 6,6, polyurethane, and epoxy based composite materials. Mechanical testing results combined with mathematical modeling will lead to the establishment of a system for the determination of a polymer composite's long term performance as a nuclear waste storage container. (author)

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

    International Nuclear Information System (INIS)

    Oxidation and atmospheric corrosion data suggest that addition of Cr provides the greatest improvement in oxidation resistance. Cr-bearing cast irons are resistant to chloride environments and solutions containing strongly oxidizing constituents. Weathering steels, including high content and at least 0.04% Cu, appear to provide adequate resistance to oxidation under temperate conditions. However, data from long-term, high-temperature oxidation studies on weathering steels were not available. From the literature, it appears that the low alloy steels, plain carbon steels, cast steels, and cast irons con-ode at similar rates in an aqueous environment. Alloys containing more than 12% Cr or 36% Ni corrode at a lower rate than plain carbon steels, but pitting may be worse. Short term tests indicate that an alloy of 9Cr-1Mo may result in increased corrosion resistance, however long term data are not available. Austenitic cast irons show the best corrosion resistance. A ranking of total corrosion performance of the materials from most corrosion resistant to least corrosion resistant is: Austenitic Cast Iron; 12% Cr = 36% Ni = 9Cr-1Mo; Carbon Steel = Low Alloy Steels; and Cast Iron. Since the materials to be employed in the Advanced Conceptual Design (ACD) waste package are considered to be corrosion allowance materials, the austenitic cast irons, high Cr steels, high Ni steels and the high Cr-Mo steels should not be considered as candidates for the outer containment barrier. Based upon the oxidation and corrosion data available for carbon steels, low alloy steels, and cast irons, a suitable list of candidate materials for a corrosion allowance outer barrier for an ACD waste package could include, A516, 2.25%Cr -- 1%Mo Steel, and A27

  11. Laboratory performance testing of an extruded bitumen containing a surrogate, sodium nitrate-based, low-level aqueous waste

    International Nuclear Information System (INIS)

    Laboratory results of a comprehensive regulatory performance test program, using an extruded bitumen and a surrogate, sodium nitrate-based waste, have been compiled at the Oak Ridge National Laboratory (ORNL). The testing has shown that the relatively viscous form of oxidized bitumen that was used has been able to meet all performance requirements. Using a 53-mm Werner and Pfleiderer extruder, operated by personnel of WasteChem Corporation of Paramus, New Jersey, laboratory-scale, molded samples of ASTM D312, type III, air-blown bitumen were prepared for laboratory performance testing. A surrogate, low-level, mixed liquid waste, formulated to represent an actual on-site waste at ORNL, was used. The mixed liquid waste contained approximately 30 wt % sodium nitrate, in addition to eight heavy metals, cold cesium, and strontium. Samples tested contained three levels of waste loading: that is, 40, 50, and 60 wt % salt. Performance test results include the 90-day American Nuclear Society (ANS) 16.1 leach test, with leach indices reported for all cations and anions, in addition to the EP toxicity test, at all levels of waste loading. Additionally, test results presented include the unconfined compressive strength and surface morphology utilizing scanning electron microscopy (SEM). Data presented include correlations between waste form loading and test results, in addition to their relationship to regulatory performance requirements

  12. Corrosion behaviour of container materials for geological disposal of high-level waste

    International Nuclear Information System (INIS)

    Within the framework of the Community R and D programme on management and storage of radioactive waste (shared cost action), a research activity is aiming at the assessment of the corrosion behaviour of potential container materials for the geological disposal of vitrified high-level waste. In a joint programme, three promising reference materials are being tested in environments representative of the three considered geological formations, clay, salt and granite. Samples of the three reference materials, Ti-0.2% Pd, Hastelloy C4 and a low carbon steel were provided by the Commission to the participating laboratories respectively: Studiecentrum voor Kernenergie (SCK/CEN) at Mol (Belgium), Kernforschungszentrum (KfK) at Karlsruhe (Federal Republic of Germany), Commissariat a l'Energie Atomique (CEA) at Fontenay-aux-Roses (France), the Atomic Energy Research Establishment (AERE) at Harwell (United Kingdom) and the Centre National de la Recherche Scientifique (CNRS) at Vitry (France). In this report, the results obtained during the year 1984 are described

  13. Hydration of blended cement pastes containing waste ceramic powder as a function of age

    Science.gov (United States)

    Scheinherrová, Lenka; Trník, Anton; Kulovaná, Tereza; Pavlík, Zbyšek; Rahhal, Viviana; Irassar, Edgardo F.; Černý, Robert

    2016-07-01

    The production of a cement binder generates a high amount of CO2 and has high energy consumption, resulting in a very adverse impact on the environment. Therefore, use of pozzolana active materials in the concrete production leads to a decrease of the consumption of cement binder and costs, especially when some type of industrial waste is used. In this paper, the hydration of blended cement pastes containing waste ceramic powder from the Czech Republic and Portland cement produced in Argentina is studied. A cement binder is partially replaced by 8 and 40 mass% of a ceramic powder. These materials are compared with an ordinary cement paste. All mixtures are prepared with a water/cement ratio of 0.5. Thermal characterization of the hydrated blended pastes is carried out in the time period from 2 to 360 days. Simultaneous DSC/TG analysis is performed in the temperature range from 25 °C to 1000 °C in an argon atmosphere. Using this thermal analysis, we identify the temperature, enthalpy and mass changes related to the liberation of physically bound water, calcium-silicate-hydrates gels dehydration, portlandite, vaterite and calcite decomposition and their changes during the curing time. Based on thermogravimetry results, we found out that the portlandite content slightly decreases with time for all blended cement pastes.

  14. Biological technologies for the removal of sulfur containing compounds from waste streams: bioreactors and microbial characteristics.

    Science.gov (United States)

    Li, Lin; Zhang, Jingying; Lin, Jian; Liu, Junxin

    2015-10-01

    Waste gases containing sulfur compounds, such as hydrogen sulfide, sulfur dioxide, thioethers, and mercaptan, produced and emitted from industrial processes, wastewater treatment, and landfill waste may cause undesirable issues in adjacent areas and contribute to atmospheric pollution. Their control has been an area of concern and research for many years. As alternative to conventional physicochemical air pollution control technologies, biological treatment processes which can transform sulfur compounds to harmless products by microbial activity, have gained in popularity due to their efficiency, cost-effectiveness and environmental acceptability. This paper provides an overview of the current biological techniques used for the treatment of air streams contaminated with sulfur compounds as well as the advances made in the past year. The discussion focuses on bioreactor configuration and design, mechanism of operation, insights into the overall biological treatment process, and the characterization of the microbial species present in bioreactors, their populations and their interactions with the environment. Some bioreactor case studies are also introduced. Finally, the perspectives on future research and development needs in this research area were also highlighted. PMID:26250546

  15. Biological technologies for the removal of sulfur containing compounds from waste streams: bioreactors and microbial characteristics.

    Science.gov (United States)

    Li, Lin; Zhang, Jingying; Lin, Jian; Liu, Junxin

    2015-10-01

    Waste gases containing sulfur compounds, such as hydrogen sulfide, sulfur dioxide, thioethers, and mercaptan, produced and emitted from industrial processes, wastewater treatment, and landfill waste may cause undesirable issues in adjacent areas and contribute to atmospheric pollution. Their control has been an area of concern and research for many years. As alternative to conventional physicochemical air pollution control technologies, biological treatment processes which can transform sulfur compounds to harmless products by microbial activity, have gained in popularity due to their efficiency, cost-effectiveness and environmental acceptability. This paper provides an overview of the current biological techniques used for the treatment of air streams contaminated with sulfur compounds as well as the advances made in the past year. The discussion focuses on bioreactor configuration and design, mechanism of operation, insights into the overall biological treatment process, and the characterization of the microbial species present in bioreactors, their populations and their interactions with the environment. Some bioreactor case studies are also introduced. Finally, the perspectives on future research and development needs in this research area were also highlighted.

  16. Modeling the corrosion of high-level waste containers: CAM-CRM interface

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, J.C.; Bedrossian, P.J.; McCright, R.D.

    1998-06-01

    A key component of the Engineered Barrier System (EBS) being designed for containment of spent-fuel and high-level waste at the proposed geological respository at Yucca Mountain, Nevada is a two-layer canister. In this particular design, the inner barrier is made of a corrosion resistant material (CRM) such as Alloy 825, 625 or C-22, while the outer barrier is made of a corrosion-allowance material (CAM) such as A516 or Monel 400. At the present time, Alloy C-22 and A516 are favored. This publication addresses the development of models to account for corrosion of Alloy C-22 surfaces exposed directly to the Near Field Environmental (NFE), as well as to the exacerbated conditions in the CAM-CRM crevice.

  17. Solidification of aqueous tritium-containing wastes with calcium oxide and asphalt

    International Nuclear Information System (INIS)

    A simple method is proposed for solidifying aqueous tritium-containing wastes with calcium oxide and asphalt. We incorporated tritiated calcium hydroxide into molten asphalt at 100-210/degree/C and studied the evolution of tritium (T) oxides there from as well as the extent to which calcium and tritium are leached out of the solidified product. Depending on temperature and heating time, the evolution of HTO from a Ca(OH)OT-asphalt mixture was low (between 5.6 x 10/sup /minus/4/ and 5.9 x 10/sup /minus/4/ wt.% of the original amount). Tritium evolution rates and leaching coefficients of tritium and calcium showed the solidified product to have high stability in water. Conclusions were drawn as to the usefulness of the proposed method

  18. A review on soil cover in Waste and contaminant containment: design, monitoring, and modeling

    Institute of Scientific and Technical Information of China (English)

    Sheng PENG; Huilian JIANG

    2009-01-01

    Soil cover is a widely-used but relatively new method for solid waste containment. Standard while site-specific procedures for cover design, monitoring, and evluation are needed to insure reliable cover performance. This paper presents a review of soil cover types, design principles and procedures, cover monitoring, and long-term performance modeling. Cover types and cover design are introduced with the general concepts and discussed on their specific applicabilities in different circumstances. Detailed discussion is given on unsaturated flow system properties and their field measurements, including meth-ods, apparatuses/equipments and their advantages and disadvantages. Several unsaturated flow simulators are discussed and compared with regards to their simulation capacities for critical parameters closely related to soil cover performance such as runoff, infiltration and evaporation. Finally, research subjects are suggested for future work for better soil cover monitoring and modeling.

  19. Immobilization of simulated radioactive soil waste containing cerium by self-propagating high-temperature synthesis

    Energy Technology Data Exchange (ETDEWEB)

    Mao, Xianhe, E-mail: maoxianhe@hotmail.com; Qin, Zhigui; Yuan, Xiaoning; Wang, Chunming; Cai, Xinan; Zhao, Weixia; Zhao, Kang; Yang, Ping; Fan, Xiaoling

    2013-11-15

    A simulated radioactive soil waste containing cerium as an imitator element has been immobilized by a thermite self-propagating high-temperature synthesis (SHS) process. The compositions, structures, and element leaching rates of products with different cerium contents have been characterized. To investigate the influence of iron on the chemical stability of the immobilized products, leaching tests of samples with different iron contents with different leaching solutions were carried out. The results showed that the imitator element cerium mainly forms the crystalline phases CeAl{sub 11}O{sub 18} and Ce{sub 2}SiO{sub 5}. The leaching rate of cerium over a period of 28 days was 10{sup −5}–10{sup −6} g/(m{sup 2} day). Iron in the reactants, the reaction products, and the environment has no significant effect on the chemical stability of the immobilized SHS products.

  20. Immobilization of simulated radioactive soil waste containing cerium by self-propagating high-temperature synthesis

    Science.gov (United States)

    Mao, Xianhe; Qin, Zhigui; Yuan, Xiaoning; Wang, Chunming; Cai, Xinan; Zhao, Weixia; Zhao, Kang; Yang, Ping; Fan, Xiaoling

    2013-11-01

    A simulated radioactive soil waste containing cerium as an imitator element has been immobilized by a thermite self-propagating high-temperature synthesis (SHS) process. The compositions, structures, and element leaching rates of products with different cerium contents have been characterized. To investigate the influence of iron on the chemical stability of the immobilized products, leaching tests of samples with different iron contents with different leaching solutions were carried out. The results showed that the imitator element cerium mainly forms the crystalline phases CeAl11O18 and Ce2SiO5. The leaching rate of cerium over a period of 28 days was 10-5-10-6 g/(m2 day). Iron in the reactants, the reaction products, and the environment has no significant effect on the chemical stability of the immobilized SHS products.

  1. Hydrothermal transformations in an aluminophosphate glass matrix containing simulators of high-level radioactive wastes

    Science.gov (United States)

    Yudintsev, S. V.; Mal'kovsky, V. I.; Mokhov, A. V.

    2016-05-01

    The interaction of aluminophosphate glass with water at 95°C for 35 days results in glass heterogenization and in the appearance of a gel layer and various phases. The leaching rate of elements is low owing to the formation of a protective layer on the glass surface. It is shown that over 80% of uranium leached from the glass matrix occurs as colloids below 450 nm in size characterized by high migration ability in the geological environment. To determine the composition of these colloids is a primary task for further studies. Water vapor is a crystallization factor for glasses. The conditions as such may appear even at early stages of glass storage because of the failure of seals on containers of high-level radioactive wastes. The examination of water resistance of crystallized matrices and determination of the fraction of radionuclide in colloids are also subjects for further studies.

  2. Full-scale tests of sulfur polymer cement and non-radioactive waste in heated and unheated prototypical containers

    Energy Technology Data Exchange (ETDEWEB)

    Darnell, G.R.; Aldrich, W.C.; Logan, J.A.

    1992-02-01

    Sulfur polymer cement has been demonstrated to be superior to portland cement in the stabilization of numerous troublesome low- level radioactive wastes, notably mixed waste fly ash, which contains heavy metals. EG&G Idaho, Inc. conducted full-scale, waste-stabilization tests with a mixture of sulfur polymer cement and nonradioactive incinerator ash poured over simulated steel and ash wastes. The container used to contain the simulated waste for the pour was a thin-walled, rectangular, steel container with no appendages. The variable in the tests was that one container and its contents were at 65{degree}F (18{degree}C) at the beginning of the pour, while the other was preheated to 275{degree}F (135{degree}C) and was insulated before the pour. The primary goal was to determine the procedures and equipment deemed operationally acceptable and capable of providing the best probability of passing the only remaining governmental test for sulfur polymer cement, the Nuclear Regulatory Commission`s full-scale test. The secondary goal was to analyze the ability of the molten cement and ash mixture to fill different size pipes and thus eliminate voids in the resultant 24 ft{sup 3} monolith.

  3. Full-scale tests of sulfur polymer cement and non-radioactive waste in heated and unheated prototypical containers

    Energy Technology Data Exchange (ETDEWEB)

    Darnell, G.R.; Aldrich, W.C.; Logan, J.A.

    1992-02-01

    Sulfur polymer cement has been demonstrated to be superior to portland cement in the stabilization of numerous troublesome low- level radioactive wastes, notably mixed waste fly ash, which contains heavy metals. EG G Idaho, Inc. conducted full-scale, waste-stabilization tests with a mixture of sulfur polymer cement and nonradioactive incinerator ash poured over simulated steel and ash wastes. The container used to contain the simulated waste for the pour was a thin-walled, rectangular, steel container with no appendages. The variable in the tests was that one container and its contents were at 65{degree}F (18{degree}C) at the beginning of the pour, while the other was preheated to 275{degree}F (135{degree}C) and was insulated before the pour. The primary goal was to determine the procedures and equipment deemed operationally acceptable and capable of providing the best probability of passing the only remaining governmental test for sulfur polymer cement, the Nuclear Regulatory Commission's full-scale test. The secondary goal was to analyze the ability of the molten cement and ash mixture to fill different size pipes and thus eliminate voids in the resultant 24 ft{sup 3} monolith.

  4. Scale up issues involved with the ceramic waste form : ceramic-container interactions and ceramic cracking quantification.

    Energy Technology Data Exchange (ETDEWEB)

    Bateman, K. J.; DiSanto, T.; Goff, K. M.; Johnson, S. G.; O' Holleran, T.; Riley, W. P., Jr.

    1999-05-03

    Argonne National Laboratory is developing a process for the conditioning of spent nuclear fuel to prepare the material for final disposal. Two waste streams will result from the treatment process, a stainless steel based form and a ceramic based form. The ceramic waste form will be enclosed in a stainless steel container. In order to assess the performance of the ceramic waste form in a repository two factors must be examined, the surface area increases caused by waste form cracking and any ceramic/canister interactions that may release toxic material. The results indicate that the surface area increases are less than the High Level Waste glass and any toxic releases are below regulatory limits.

  5. CONTAINMENT OF LOW-LEVEL RADIOACTIVE WASTE AT THE DOE SALTSTONE DISPOSAL FACILITY

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, J.; Flach, G.

    2012-03-29

    As facilities look for permanent storage of toxic materials, they are forced to address the long-term impacts to the environment as well as any individuals living in affected area. As these materials are stored underground, modeling of the contaminant transport through the ground is an essential part of the evaluation. The contaminant transport model must address the long-term degradation of the containment system as well as any movement of the contaminant through the soil and into the groundwater. In order for disposal facilities to meet their performance objectives, engineered and natural barriers are relied upon. Engineered barriers include things like the design of the disposal unit, while natural barriers include things like the depth of soil between the disposal unit and the water table. The Saltstone Disposal Facility (SDF) at the Savannah River Site (SRS) in South Carolina is an example of a waste disposal unit that must be evaluated over a timeframe of thousands of years. The engineered and natural barriers for the SDF allow it to meet its performance objective over the long time frame. Some waste disposal facilities are required to meet certain standards to ensure public safety. These type of facilities require an engineered containment system to ensure that these requirements are met. The Saltstone Disposal Facility (SDF) at the Savannah River Site (SRS) is an example of this type of facility. The facility is evaluated based on a groundwater pathway analysis which considers long-term changes to material properties due to physical and chemical degradation processes. The facility is able to meet these performance objectives due to the multiple engineered and natural barriers to contaminant migration.

  6. Synthesis and Characterization of the Hybrid Clay- Based Material Montmorillonite-Melanoidin: A Potential Soil Model

    Energy Technology Data Exchange (ETDEWEB)

    V Vilas; B Matthiasch; J Huth; J Kratz; S Rubert de la Rosa; P Michel; T Schäfer

    2011-12-31

    The study of the interactions among metals, minerals, and humic substances is essential in understanding the migration of inorganic pollutants in the geosphere. A considerable amount of organic matter in the environment is associated with clay minerals. To understand the role of organic matter in the environment and its association with clay minerals, a hybrid clay-based material (HCM), montmorillonite (STx-1)-melanoidin, was prepared from L-tyrosine and L-glutamic acid by the Maillard reaction. The HCM was characterized by elemental analysis, nuclear magnetic resonance, x-ray photoelectron spectroscopy (XPS), scanning transmission x-ray microscopy (STXM), and thermal analysis. The presence of organic materials on the surface was confirmed by XPS and STXM. The STXM results showed the presence of organic spots on the surface of the STx-1 and the characterization of the functional groups present in those spots. Thermal analysis confirmed the existence of organic materials in the montmorillonite interlayer, indicating the formation of a composite of melanoidin and montmorillonite. The melanoidin appeared to be located partially between the layers of montmorillonite and partially at the surface, forming a structure that resembles the way a cork sits on the top of a champagne bottle.

  7. Electrochemical energy storage in montmorillonite K10 clay based composite as supercapacitor using ionic liquid electrolyte.

    Science.gov (United States)

    Maiti, Sandipan; Pramanik, Atin; Chattopadhyay, Shreyasi; De, Goutam; Mahanty, Sourindra

    2016-02-15

    Exploring new electrode materials is the key to realize high performance energy storage devices for effective utilization of renewable energy. Natural clays with layered structure and high surface area are prospective materials for electrical double layer capacitors (EDLC). In this work, a novel hybrid composite based on acid-leached montmorillonite (K10), multi-walled carbon nanotube (MWCNT) and manganese dioxide (MnO2) was prepared and its electrochemical properties were investigated by fabricating two-electrode asymmetric supercapacitor cells against activated carbon (AC) using 1.0M tetraethylammonium tetrafluroborate (Et4NBF4) in acetonitrile (AN) as electrolyte. The asymmetric supercapacitors, capable of operating in a wide potential window of 0.0-2.7V, showed a high energy density of 171Whkg(-1) at a power density of ∼1.98kWkg(-1). Such high EDLC performance could possibly be linked to the acid-base interaction of K10 through its surface hydroxyl groups with the tetraethylammonium cation [(C2H5)4N(+) or TEA(+)] of the ionic liquid electrolyte. Even at a very high power density of 96.4kWkg(-1), the cells could still deliver an energy density of 91.1Whkg(-1) exhibiting an outstanding rate capability. The present study demonstrates for the first time, the excellent potential of clay-based composites for high power energy storage device applications.

  8. Neutron measurements around storage casks containing spent fuel and vitrified high-level radioactive waste at ZWILAG.

    Science.gov (United States)

    Buchillier, T; Aroua, A; Bochud, F O

    2007-01-01

    Spectrometric and dosimetric measurements were made around a cask containing spent fuel and a cask containing high-level radioactive waste at the Swiss intermediate waste and spent fuel storage facility. A Bonner sphere spectrometer, an LB 6411 neutron monitor and an Automess Szintomat 6134A were used to characterise the n-gamma fields at several locations around the two casks. The results of these measurements show that the neutron fluence spectra around the cask containing radioactive waste are harder and higher in intensity than those measured in the vicinity of the spent fuel cask. The ambient dose equivalents measured with the LB 6411 neutron monitor are in good agreement with those obtained using the Bonner spheres, except for locations with soft neutron spectra where the monitor overestimates the neutron ambient dose equivalent by almost 50%. PMID:17494980

  9. Containment

    International Nuclear Information System (INIS)

    The primary mission of the Containment Group is to ensure that underground nuclear tests are satisfactorily contained. The main goal is the development of sound technical bases for containment-related methodology. Major areas of activity include siting, geologic description, emplacement hole stemming, and phenomenological predictions. Performance results of sanded gypsum concrete plugs on the Jefferson, Panamint, Cornucopia, Labquark, and Bodie events are given. Activities are also described in the following areas: computational capabilities site description, predictive modeling, and cavity-pressure measurement. Containment publications are listed. 8 references

  10. Nitrogen oxides from combustion of nitrogen-containing polymers in waste-derived fuels

    International Nuclear Information System (INIS)

    Usually, waste-derived fuels present nitrogen-containing fractions, which produce nitrogen oxides (NO) during combustion. This study was mainly concerned with poly amides (PA) (nylon), poly urethanes (PU), urea formaldehyde (UF) glue, sewage sludge and refuse-derived fuels (RDF). For control purposes, the authors chose a Polish sub-bituminous coal and a Finnish pine wood sample. An almost inverse trend between fuel nitrogen content and NO emissions was revealed through analysis of NO emissions at 850 Celsius, 1 bar, 7 per cent O2 in N2. It was not possible to derive a clear correlation to the amount of ash generated by the samples. PU foam decomposed through a two-step process, as suggested by thermochromatography, and PA6-containing samples yielded epsilon-caprolactam as a major decomposition product. Important decomposition products from PU, PA6, PA6/PE, sewage sludge and UF glue samples were greenhouse gases as demonstrated by pyrolysis-gas chromatography/mass spectroscopy. The work was carried out at Abo Akademi University and University of Helsinki, Finland. 5 refs., 2 tabs., 3 figs

  11. Feasibility assessment of copper-base waste package container materials in a tuff repository

    International Nuclear Information System (INIS)

    This report discussed progress made during the second year of a two-year study on the feasibility of using copper or a copper-base alloy as a container material for a waste package in a potential repository in tuff rock at the Yucca Mountain site in Nevada. Corrosion testing in potentially corrosive irradiated environments received emphasis during the feasibility study. Results of experiments to evaluate the effect of a radiation field on the uniform corrosion rate of the copper-base materials in repository-relevant aqueous environments are given as well as results of an electrochemical study of the copper-base materials in normal and concentrated J-13 water. Results of tests on the irradiation of J-13 water and on the subsequent formation of hydrogen peroxide are given. A theoretical study was initiated to predict the long-term corrosion behavior of copper in the repository. Tests were conducted to determine whether copper would adversely affect release rates of radionuclides to the environment because of degradation of the Zircaloy cladding. A manufacturing survey to determine the feasibility of producing copper containers utilizing existing equipment and processes was completed. The cost and availability of copper was also evaluated and predicted to the year 2000. Results of this feasibility assessment are summarized

  12. Fabrication development for high-level nuclear waste containers for the tuff repository

    International Nuclear Information System (INIS)

    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

  13. Localized corrosion of a candidate container material for high-level nuclear waste disposal

    International Nuclear Information System (INIS)

    Localized corrosion is one of the important considerations in the design of metallic containers used for the geologic disposal of high-level nuclear waste. This paper addresses the effect of environmental factors on the localized corrosion behavior of alloy 825, one of the candidate alloys for containers in the Yucca Mountain repository site. A two-level, full factorial experimental design was used to examine the main effects and interactions of chloride, sulfate, nitrate, fluoride, and temperature. This was augmented by additional experiments involving chloride and temperature at several levels. Cyclic, potentiodynamic polarization tests were used to determine the relative susceptibility of the alloy to localized corrosion. Crevice corrosion was detected at chloride levels as low as 20 ppm, and both pitting and crevice corrosion were observed at higher chloride levels. Among the environmental factors, chloride and sulfate were found to be promoters of localized corrosion, while nitrate and fluoride were inhibitors of localized corrosion. The experiments indicated that the electrochemical parameters (e.g., pitting potential, repassivation potential, or the difference between them) were not sufficient indicators of localized corrosion. Instead, the visual observation and electrochemical parameters were combined into an index, termed localized corrosion index (LCI), to quantify the extent of localized corrosion

  14. Corrosion of steel drums containing cemented ion-exchange resins as intermediate level nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    Duffó, G.S. [Departamento de Materiales, Comisión Nacional de Energía Atómica, Av. Gral. Paz 1499, 1650 Buenos Aires (Argentina); Universidad Nacional de San Martín, Av. Gral. Paz 1499, 1650 Buenos Aires (Argentina); Consejo Nacional de Investigaciones Científicas y Tecnológicas – CONICET, Av. Gral. Paz 1499, 1650 Buenos Aires (Argentina); Farina, S.B., E-mail: farina@cnea.gov.ar [Departamento de Materiales, Comisión Nacional de Energía Atómica, Av. Gral. Paz 1499, 1650 Buenos Aires (Argentina); Universidad Nacional de San Martín, Av. Gral. Paz 1499, 1650 Buenos Aires (Argentina); Consejo Nacional de Investigaciones Científicas y Tecnológicas – CONICET, Av. Gral. Paz 1499, 1650 Buenos Aires (Argentina); Schulz, F.M. [Consejo Nacional de Investigaciones Científicas y Tecnológicas – CONICET, Av. Gral. Paz 1499, 1650 Buenos Aires (Argentina)

    2013-07-15

    Highlights: • There are no works related to the corrosion of drums containing radioactive waste. • Chloride induces high corrosion rate and after 1 year it drops abruptly. • Decrease in the corrosion rate is due to the lack of water to sustain the process. • Cementated ion-exchange resins do not pose risks of corrosion of the steel drums. -- Abstract: Exhausted ion-exchange resins used in nuclear reactors are immobilized by cementation before being stored. They are contained in steel drums that may undergo internal corrosion depending on the presence of certain contaminants. The objective of this work is to evaluate the corrosion susceptibility of steel drums in contact with cemented ion-exchange resins with different aggressive species. The corrosion potential and the corrosion rate of the steel, and the electrical resistivity of the matrix were monitored for 900 days. Results show that the cementation of ion-exchange resins seems not to pose special risks regarding the corrosion of the steel drums.

  15. Vitrification of simulated radioactive Rocky Flats plutonium containing waste ash with a stir-melter system

    International Nuclear Information System (INIS)

    A demonstration trial has been completed in which a simulated Rocky Flats ash consisting of an industrial fly-ash material doped with cerium oxide was vitrified in an alloy tank Stir-Melter trademark System. The cerium oxide served as a substitute for plutonium oxide present in the actual Rocky Flats waste stream. The glass developed falls within the SiO2 +Al2O3 / ΣAlkali / B2O3 System. The glass batch contained approximately 40 wt % of ash, the ash was modified to contain ∼5 wt % CeO2 to simulate plutonium chemistry in the glass. The ash simulant was mixed with water and fed to the Stir-Melter as a slurry with a 60 wt % water to 40 wt % solids ratio. Glass melting temperature was maintained at approximately 1050 degrees C during the melting trials. Melting rates as functions of impeller speed and slurry feed rate were determined. An optimal melting rate was established through a series of evolutionary variations of the control variables' settings. The optimal melting rate condition was used for a continuous six hour steady state run of the vitrification system. Glass mass flow rates out of the melter were measured and correlated with the slurry feed mass flow. Melter off-gas was sampled for particulate and volatile species over a period of four hours during the steady state run. Glass composition and durability studies were run on samples collected during the steady state run

  16. Targeted Health Assessment for Wastes Contained at the Niagara Falls Storage Site to Guide Planning for Remedial Action Alternatives - 13428

    Energy Technology Data Exchange (ETDEWEB)

    Busse, John; Keil, Karen; Staten, Jane; Miller, Neil; Barker, Michelle [U.S. Army Corps of Engineers, Buffalo District, 1776 Niagara Street, Buffalo, NY (United States); MacDonell, Margaret; Peterson, John; Chang, Young-Soo; Durham, Lisa [Argonne National Laboratory, Environmental Science Division, 9700 S. Cass Ave., Argonne, IL 60439 (United States)

    2013-07-01

    The U.S. Army Corps of Engineers (USACE) is evaluating potential remedial alternatives at the 191-acre Niagara Falls Storage Site (NFSS) in Lewiston, New York, under the Formerly Utilized Sites Remedial Action Program (FUSRAP). The Manhattan Engineer District (MED) and Atomic Energy Commission (AEC) brought radioactive wastes to the site during the 1940's and 1950's, and the U.S. Department of Energy (US DOE) consolidated these wastes into a 10-acre interim waste containment structure (IWCS) in the southwest portion of the site during the 1980's. The USACE is evaluating remedial alternatives for radioactive waste contained within the IWCS at the NFSS under the Feasibility Study phase of the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) process. A preliminary evaluation of the IWCS has been conducted to assess potential airborne releases associated with uncovered wastes, particularly during waste excavation, as well as direct exposures to uncovered wastes. Key technical issues for this assessment include: (1) limitations in waste characterization data; (2) representative receptors and exposure routes; (3) estimates of contaminant emissions at an early stage of the evaluation process; (4) consideration of candidate meteorological data and air dispersion modeling approaches; and (5) estimates of health effects from potential exposures to both radionuclides and chemicals that account for recent updates of exposure and toxicity factors. Results of this preliminary health risk assessment indicate if the wastes were uncovered and someone stayed at the IWCS for a number of days to weeks, substantial doses and serious health effects could be incurred. Current controls prevent such exposures, and the controls that would be applied to protect onsite workers during remedial action at the IWCS would also effectively protect the public nearby. This evaluation provides framing context for the upcoming development and detailed

  17. Draft environmental assessment: Deaf Smith County site, Texas. Nuclear Waste Policy Act (Section 112). [Contains Glossary

    Energy Technology Data Exchange (ETDEWEB)

    1984-12-01

    In February 1983, the US Department of Energy identified a location in Deaf Smith County, Texas, as one of nine potentially acceptable sites for a mined geologic repository for spent nuclear fuel and high-level radioactive waste. The potentially acceptable site was subsequently narrowed to an area of 9 square miles. To determine their suitability, the Deaf Smith site and the eight other potentially acceptable sites have been evaluated in accordance with the DOE's General Guidelines for the Recommendation of Sites for Nuclear Waste Repositories. These evaluations are reported in this draft environmental assessment, which is being issued for public review and comment. The DOE findings and determinations that are based on these evaluations are preliminary and subject to public review and comment. A final EA will be prepared after considering the comments received. On the basis of the evaluations reported in this draft EA, the DOE has found that the Deaf Smith site is not disqualified under the guidelines. The site is in the Permian Basin, which is one of five distinct geohydrologic settings considered for the first repository. This setting contains one other potentially acceptable site - the Swisher site. Although the Swisher site appears to be suitable for site characterization, DOE has concluded that the Deaf Smith site is the preferred site. The DOE finds that the site is suitable for site characterization because the evidence does not support a conclusion that the site will not be able to meet each of the qualifying conditions specified in the guidelines. On the basis of these findings, the DOE is proposing to nominate the Deaf Smith site as one of five sites suitable for characterization. Having compared the Deaf Smith site with the other four sites proposed for nomination, the DOE has determined that the Deaf Smith site is one of the three preferred sites for recommendation to the President as candidates for characterization.

  18. Noble metal catalyzed hydrogen generation from formic acid in nitrite-containing simulated nuclear waste media

    International Nuclear Information System (INIS)

    The Hanford Waste Vitrification Plant (HWVP) is being designed by the U.S. Department of Energy to immobilize high-level nuclear waste. Simulants for the HWVP feed containing the major nonradioactive components Al, Cd, Fe, Mn, Nd, Ni, Si, Zr, Na, CO32-, NO3- and NO2- were used as media to evaluate the stability of formic acid towards hydrogen evolution by the reaction HCO2H→H2+/CO2 catalyzed by the noble metals Ru, Rh, and/or Pd found in significant quantities in uranium fission products. Small-scale experiments using 40-50 mL of feed simulant in closed glass reactors (250-550 mL total volume) at 80-100 degree C were used to study the effect of nitrite and nitrate ion on the catalytic activities of the noble metals for formic acid decomposition. Reactions were monitored using gas chromatography to analyze the CO2, H2, NO, and N2O in the gas phase as a function of time. Rhodium, which was introduced as soluble RhCl3.3H2O, was found to be the most active catalyst for hydrogen generation from formic acid above nearly 80 degree C in the presence of nitrite ion in accord with earlier observations. The apparent homogeneous nature of the nitrite-promoted Rh-catalyzed formic acid decomposition is consistent with the approximate pseudo-first-order dependence of the hydrogen production rate on Rh concentration. 24 refs., 7 figs., 2 tabs

  19. Preliminary technique assessment for nondestructive evaluation certification of the NNWSI [Nevada Nuclear Waste Storage Investigations] disposal container closure

    Energy Technology Data Exchange (ETDEWEB)

    Day, R.A.

    1988-12-31

    Under the direction of the Department of Energy`s (DOE) Office of Civilian Radioactive Waste Management (OCRWM) program, the Nevada Nuclear Waste Storage Investigations (NNWSI) project is evaluating a candidate repository site at Yucca Mountain, Nevada, for permanent disposal of high-level nuclear waste. The Lawrence Livermore National Laboratory (LLNL), a participant in the NNWSI project, is developing waste package designs to meet the NRC requirements. One aspect of this waste package is the nondestructive testing of the final closure of the waste container. The container closure weld can best be nondestructively examined (NDE) by a combination of ultrasonics and liquid penetrants. This combination can be applied remotely and can meet stringent quality control requirements common to nuclear applications. Further development in remote systems and inspection will be required to meet anticipated requirements for flaw detection reliability and sensitivity. New research is not required but might reduce cost or inspection time. Ultrasonic and liquid penetrant methods can examine all closure methods currently being considered, which include fusion welding and inertial welding, among others. These NDE methods also have a history of application in high radiation environments and a well developed technology base for remote operation that can be used to reduce development and design costs. 43 refs., 23 figs., 3 tabs.

  20. Preliminary technique assessment for nondestructive evaluation certification of the NNWSI [Nevada Nuclear Waste Storage Investigations] disposal container closure

    International Nuclear Information System (INIS)

    Under the direction of the Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM) program, the Nevada Nuclear Waste Storage Investigations (NNWSI) project is evaluating a candidate repository site at Yucca Mountain, Nevada, for permanent disposal of high-level nuclear waste. The Lawrence Livermore National Laboratory (LLNL), a participant in the NNWSI project, is developing waste package designs to meet the NRC requirements. One aspect of this waste package is the nondestructive testing of the final closure of the waste container. The container closure weld can best be nondestructively examined (NDE) by a combination of ultrasonics and liquid penetrants. This combination can be applied remotely and can meet stringent quality control requirements common to nuclear applications. Further development in remote systems and inspection will be required to meet anticipated requirements for flaw detection reliability and sensitivity. New research is not required but might reduce cost or inspection time. Ultrasonic and liquid penetrant methods can examine all closure methods currently being considered, which include fusion welding and inertial welding, among others. These NDE methods also have a history of application in high radiation environments and a well developed technology base for remote operation that can be used to reduce development and design costs. 43 refs., 23 figs., 3 tabs

  1. DISPOSAL OF TRU WASTE FROM THE PLUTONIUM FINISHING PLANT IN PIPE OVERPACK CONTAINERS TO WIPP INCLUDING NEW SECURITY REQUIREMENTS

    Energy Technology Data Exchange (ETDEWEB)

    Hopkins, A.M.; Sutter, C.; Hulse, G.; Teal, J.

    2003-02-27

    The Department of Energy is responsible for the safe management and cleanup of the DOE complex. As part of the cleanup and closure of the Plutonium Finishing Plant (PFP) located on the Hanford site, the nuclear material inventory was reviewed to determine the appropriate disposition path. Based on the nuclear material characteristics, the material was designated for stabilization and packaging for long term storage and transfer to the Savannah River Site or, a decision for discard was made. The discarded material was designated as waste material and slated for disposal to the Waste Isolation Pilot Plant (WIPP). Prior to preparing any residue wastes for disposal at the WIPP, several major activities need to be completed. As detailed a processing history as possible of the material including origin of the waste must be researched and documented. A technical basis for termination of safeguards on the material must be prepared and approved. Utilizing process knowledge and processing history, the material must be characterized, sampling requirements determined, acceptable knowledge package and waste designation completed prior to disposal. All of these activities involve several organizations including the contractor, DOE, state representatives and other regulators such as EPA. At PFP, a process has been developed for meeting the many, varied requirements and successfully used to prepare several residue waste streams including Rocky Flats incinerator ash, Hanford incinerator ash and Sand, Slag and Crucible (SS&C) material for disposal. These waste residues are packed into Pipe Overpack Containers for shipment to the WIPP.

  2. Effect Of Oxidation On Chromium Leaching And Redox Capacity Of Slag-Containing Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Almond, P. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Stefanko, D. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Langton, C. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2013-03-01

    The rate of oxidation is important to the long-term performance of reducing salt waste forms because the solubility of some contaminants, e.g., technetium, is a function of oxidation state. TcO4- in the salt solution is reduced to Tc(IV) and has been shown to react with ingredients in the waste form to precipitate low solubility sulfide and/or oxide phases [Shuh, et al., 1994, Shuh, et al., 2000, Shuh, et al., 2003]. Upon exposure to oxygen, the compounds containing Tc(IV) oxidize to the pertechnetate ion, Tc(VII)O4-, which is very soluble. Consequently the rate of technetium oxidation front advancement into a monolith and the technetium leaching profile as a function of depth from an exposed surface are important to waste form performance and ground water concentration predictions. An approach for measuring contaminant oxidation rate (effective contaminant specific oxidation rate) based on leaching of select contaminants of concern is described in this report. In addition, the relationship between reduction capacity and contaminant oxidation is addressed. Chromate was used as a non-radioactive surrogate for pertechnetate in simulated waste form samples. Depth discrete subsamples were cut from material exposed to Savannah River Site (SRS) field cured conditions. The subsamples were prepared and analyzed for both reduction capacity and chromium leachability. Results from field-cured samples indicate that the depth at which leachable chromium was detected advanced further into the sample exposed for 302 days compared to the sample exposed to air for 118 days (at least 50 mm compared to at least 20 mm). Data for only two exposure time intervals is currently available. Data for additional exposure times are required to develop an equation for the oxidation front progression. Reduction capacity measurements (per the Angus-Glasser method, which is a measurement of the ability of a material to chemically reduce Ce(IV) to Ce

  3. System of large transport containers for waste from dismantling light water and gas-cooled nuclear reactors. Volume 2

    International Nuclear Information System (INIS)

    This report deals with the operational, radiological and economic aspects of transport as well as conceptual designs of large containers for the transport of radioactive decommissioning wastes from nuclear power plants within the member states of the European Economic Community. The means of transport, the costs and radiological detriment are considered, and conceptual designs of containers are described. Recommendations are made for further studies. (U.K.)

  4. Hydrogen Concentration in the Inner-Most Container within a Pencil Tank Overpack Packaged in a Standard Waste Box Package

    Energy Technology Data Exchange (ETDEWEB)

    Marusich, Robert M.

    2013-08-15

    The purpose of this report is to evaluate hydrogen generation within Pencil Tank Overpacks (PTO) in a Standard Waste Box (SWB), to establish plutonium (Pu) limits for PTOs based on hydrogen concentration in the inner-most container and to establish required configurations or validate existing or proposed configurations for PTOs. The methodology and requirements are provided in this report.

  5. System of large transport containers for waste from dismantling light water and gas-cooled nuclear reactors. Volume 1

    International Nuclear Information System (INIS)

    The purpose of this volume is to introduce the main types of nuclear reactor in the European Community (EC), select reference plants for further study, estimate the waste streams from the reference reactors, survey the transport regulations and assess existing containers

  6. Stabilization of ZnCl2-Containing Waste Using Calcium Sulfoaluminate Cement

    International Nuclear Information System (INIS)

    The potential of calcium sulfoaluminate (CSA) cement was investigated to solidify and stabilize radwastes containing large amounts of soluble zinc chloride (a strong inhibitor of Portland cement hydration). Hydration of pastes and mortars prepared with a 0.5 mol/L ZnCl2 mixing solution was characterized over one year as a function of the gypsum content of the binder and the thermal history of the material. Blending the CSA clinker with 20% gypsum enabled rapid hydration, with only very small delay compared with a reference prepared with pure water. It also improved the compressive strength of the hardened material and significantly reduced its expansion under wet curing. Moreover, the hydrate assemblage was less affected by a thermal treatment at early age simulating the temperature rise and fall occurring in a large-volume drum of cemented waste. Fully hydrated materials contained ettringite, amorphous aluminum hydroxide, straetlingite, together with AFm phases (Kuzel's salt associated with monosulfoaluminate or Friedel's salt depending on the gypsum content of the binder), and possibly C-(A)-S-H. Zinc was readily insolubilized and could not be detected in the pore solution extracted from cement pastes, or in their leachates after 3 months of leaching by pure water at pH 7. The good retention of zinc by the cement matrix was mainly attributed to the precipitation of a hydrated and well crystallized phase with platelet morphology (which may belong to the layered double hydroxides family) at early age ≤ 1 day), and to chemisorption onto aluminum hydroxide at later age. (author)

  7. Conditioning of cladding waste for long-term storage by press compaction and encapsulation in lead containment

    International Nuclear Information System (INIS)

    The conditioning of compacted cladding waste has been based on the concept of a corrosion-resistant containment. The technique of press compaction in remote operation conditions has been demonstrated. Detailed specifications of the design of the container are discussed in the report. Testing procedures for the different containment parts have been developed and applied. The remote welding techniques of the stainless steel and lead-based parts of the containment have been investigated and reliable procedures are reported. A technique for remote leak-testing of welded containers is described. The report contains a series of pictures documenting the entire conditioning concept, starting from the dissolver basket at the reprocessing stage up to the final disposal container

  8. Containment of uranium in the proposed Egyptian geologic repository for radioactive waste using hydroxyapatite.

    Energy Technology Data Exchange (ETDEWEB)

    Moore, Robert Charles; Hasan, Ahmed Ali Mohamed; Headley, Thomas Jeffrey; Sanchez, Charles Anthony (University of Arizona, Yuma, AZ); Zhao, Hongting; Salas, Fred Manuel; Hasan, Mahmoud A. (Egyptian Atomic Energy Authority, Cairo, Egypt); Holt, Kathleen Caroline

    2004-04-01

    Currently, the Egyptian Atomic Energy Authority is designing a shallow-land disposal facility for low-level radioactive waste. To insure containment and prevent migration of radionuclides from the site, the use of a reactive backfill material is being considered. One material under consideration is hydroxyapatite, Ca{sub 10}(PO{sub 4}){sub 6}(OH){sub 2}, which has a high affinity for the sorption of many radionuclides. Hydroxyapatite has many properties that make it an ideal material for use as a backfill including low water solubility (K{sub sp}>10{sup -40}), high stability under reducing and oxidizing conditions over a wide temperature range, availability, and low cost. However, there is often considerable variation in the properties of apatites depending on source and method of preparation. In this work, we characterized and compared a synthetic hydroxyapatite with hydroxyapatites prepared from cattle bone calcined at 500 C, 700 C, 900 C and 1100 C. The analysis indicated the synthetic hydroxyapatite was similar in morphology to 500 C prepared cattle hydroxyapatite. With increasing calcination temperature the crystallinity and crystal size of the hydroxyapatites increased and the BET surface area and carbonate concentration decreased. Batch sorption experiments were performed to determine the effectiveness of each material to sorb uranium. Sorption of U was strong regardless of apatite type indicating all apatite materials evaluated. Sixty day desorption experiments indicated desorption of uranium for each hydroxyapatite was negligible.

  9. Ca(2+) and OH(-) release of ceramsites containing anorthite and gehlenite prepared from waste lime mud.

    Science.gov (United States)

    Qin, Juan; Yang, Chuanmeng; Cui, Chong; Huang, Jiantao; Hussain, Ahmad; Ma, Hailong

    2016-09-01

    Lime mud is a kind of solid waste in the papermaking industry, which has been a source of serious environmental pollution. Ceramsites containing anorthite and gehlenite were prepared from lime mud and fly ash through the solid state reaction method at 1050°C. The objective of this study was to explore the efficiency of Ca(2+) and OH(-) release and assess the phosphorus and copper ion removal performance of the ceramsites via batch experiments, X-ray diffraction (XRD) and scanning electron microscopy (SEM). The results show that Ca(2+) and OH(-) were released from the ceramsites due to the dissolution of anorthite, gehlenite and available lime. It is also concluded that gehlenite had stronger capacity for Ca(2+) and OH(-) release compared with anorthite. The Ca(2+) release could be fit well by the Avrami kinetic model. Increases of porosity, dosage and temperature were associated with increases in the concentrations of Ca(2+) and OH(-) released. Under different conditions, the ceramsites could maintain aqueous solutions in alkaline conditions (pH=9.3-10.9) and the release of Ca(2+) was not affected. The removal rates of phosphorus and copper ions were as high as 96.88% and 96.81%, respectively. The final pH values of both phosphorus and copper ions solutions changed slightly. The reuse of lime mud in the form of ceramsites is an effective strategy.

  10. Leaching studies of heavy concrete material for nuclear fuel waste immobilization containers

    International Nuclear Information System (INIS)

    The leaching behaviour of a high-density concrete was studied as part of a program to evaluate its potential use as a container material for nuclear fuel waste under conditions of deep geologic disposal. Samples of concrete material were leached in deionized distilled water, Standard Canadian Shield Saline Solution (SCSSS), SCSSS plus 20% Na-bentonite, and SCSSS plus granite and 20% Na-bentonite under static conditions at 100 degrees celsius for periods up to 365 days. The results of these leaching experiments suggest that the stability of concrete depends on the possible internal structural changes due to hydration reactions of unhydrated components, leading to the formation of C-S-H gel plus portlandite (Ca(OH)2). The factors controlling the concrete leaching process were the composition of the leachant and the concentration of elements in solution capable of forming precipitates on the concrete surface, e.g., silicon, Mg2+ and Ca2+. The main effect observed during leaching was an increase in groundwater pH (from 7 to 9). However, the addition of Na-bentonite suppressed the normal tendency of the pH of the groundwater in contact with concrete to rise rapidly. It was shown that the solution concentration of elements released from the concrete, particularly potassium, increased in the presence of Na-bentonite

  11. In-situ containment of buried waste at Brookhaven National Laboratory

    International Nuclear Information System (INIS)

    The primary objective of this project was to further develop close-coupled barrier technology 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 chemically resistant polymer layer. The technology has matured from a regulatory investigation of issues concerning barriers and barrier materials to a pilot-scale, multiple individual column injections at Sandia National Labs (SNL) to full scale demonstration. The feasibility of this barrier concept was successfully proven in a full scale 'cold test' demonstration at Hanford, WA. Consequently, a full scale deployment of the technology was conducted at an actual environmental restoration site at Brookhaven National Lab (BNL), Long Island, NY. This paper discusses the installation and performance of a technology deployment implemented at OU-1 an Environmental Restoration Site located at BNL

  12. NOx emission from incineration of organic hazardous liquid waste containing hexamethylendiamine in fluidized bed

    Institute of Scientific and Technical Information of China (English)

    别如山; 李季; 杨励丹

    2002-01-01

    Experiments have been conducted to investigate NOx concentration profiles along bed height and influences of temperature and excess air on NOx emission in the range from 700 ℃ to 900 ℃, when waste water containing 5% Hexamethylenediamine incinerated in a bench scale hot fluidized bed. The testing results indicate that the concentration of NO2 is larger than that of NO along bed height except in the freeboard at 900 ℃, where NO, NO2 concentrations are zero. Temperature and excess air play significant role on NOx emission. With increasing in temperature the NOx emission decreases very rapidly in the range from 700 ℃ to 900 ℃. With increasing in excess air, NOx emission increases considerably at 700 ℃, but it is almost independent of excess air at 800 ℃,and at 900 ℃ NOx emission is zero indicating that NH2 from NH2(CH2)6NH2 has strong effect on de-NOx with increasing in temperature and excess air. NOx concentration profiles decrease progressively with bed height because of reduction of NOx by NH2. The mechanism of NOx formation and destruction is presented in the paper.

  13. Ca(2+) and OH(-) release of ceramsites containing anorthite and gehlenite prepared from waste lime mud.

    Science.gov (United States)

    Qin, Juan; Yang, Chuanmeng; Cui, Chong; Huang, Jiantao; Hussain, Ahmad; Ma, Hailong

    2016-09-01

    Lime mud is a kind of solid waste in the papermaking industry, which has been a source of serious environmental pollution. Ceramsites containing anorthite and gehlenite were prepared from lime mud and fly ash through the solid state reaction method at 1050°C. The objective of this study was to explore the efficiency of Ca(2+) and OH(-) release and assess the phosphorus and copper ion removal performance of the ceramsites via batch experiments, X-ray diffraction (XRD) and scanning electron microscopy (SEM). The results show that Ca(2+) and OH(-) were released from the ceramsites due to the dissolution of anorthite, gehlenite and available lime. It is also concluded that gehlenite had stronger capacity for Ca(2+) and OH(-) release compared with anorthite. The Ca(2+) release could be fit well by the Avrami kinetic model. Increases of porosity, dosage and temperature were associated with increases in the concentrations of Ca(2+) and OH(-) released. Under different conditions, the ceramsites could maintain aqueous solutions in alkaline conditions (pH=9.3-10.9) and the release of Ca(2+) was not affected. The removal rates of phosphorus and copper ions were as high as 96.88% and 96.81%, respectively. The final pH values of both phosphorus and copper ions solutions changed slightly. The reuse of lime mud in the form of ceramsites is an effective strategy. PMID:27593276

  14. 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.

  15. Evaluation of polymer inclusion membranes containing crown ethers for selective cesium separation from nuclear waste solution.

    Science.gov (United States)

    Mohapatra, P K; Lakshmi, D S; Bhattacharyya, A; Manchanda, V K

    2009-09-30

    Transport behaviour of (137)Cs from nitric acid feed was investigated using cellulose triacetate plasticized polymer inclusion membrane (PIM) containing several crown ether carriers viz. di-benzo-18-crown-6 (DB18C6), di-benzo-21-crown-7 (DB21C7) and di-tert-butylbenzo-18-crown-6 (DTBB18C6). The PIM was prepared from cellulose triacetate (CTA) with various crown ethers and plasticizers. DTBB18C6 and tri-n-butyl phosphate (TBP) were found to give higher transport rate for (137)Cs as compared to other carriers and plasticizers. Effect of crown ether concentration, nitric acid concentration, plasticizer and CTA concentration on the transport rate of Cs was also studied. The Cs selectivity with respect to various fission products obtained from an irradiated natural uranium target was found to be heavily dependent on the nature of the plasticizer. The present work shows that by choosing a proper plasticizer, one can get either good transport efficiency or selectivity. Though TBP plasticized membranes showed good transport efficiency, it displayed poor selectivities. On the other hand, an entirely opposite separation behaviour was observed with 2-nitrophenyloctylether (NPOE) plasticized membranes suggesting the possible application of the later membranes for the removal of bulk (137)Cs from the nuclear waste. The stability of the membrane was tested by carrying out transport runs for nearly 25 days.

  16. Ceramic Coatings for Corrosion Resistant Nuclear Waste Container Evaluated in Simulated Ground Water at 90?C

    Energy Technology Data Exchange (ETDEWEB)

    Haslam, J J; Farmer, J C

    2004-03-31

    Ceramic materials have been considered as corrosion resistant coatings for nuclear waste containers. Their suitability can be derived from the fully oxidized state for selected metal oxides. Several types of ceramic coatings applied to plain carbon steel substrates by thermal spray techniques have been exposed to 90 C simulated ground water for nearly 6 years. In some cases no apparent macroscopic damage such as coating spallation was observed in coatings. Thermal spray processes examined in this work included plasma spray, High Velocity Oxy Fuel (HVOF), and Detonation Gun. Some thermal spray coatings have demonstrated superior corrosion protection for the plain carbon steel substrate. In particular the HVOF and Detonation Gun thermal spray processes produced coatings with low connected porosity, which limited the growth rate of corrosion products. It was also demonstrated that these coatings resisted spallation of the coating even when an intentional flaw (which allowed for corrosion of the carbon steel substrate underneath the ceramic coating) was placed in the coating. A model for prediction of the corrosion protection provided by ceramic coatings is presented. The model includes the effect of the morphology and amount of the porosity within the thermal spray coating and provides a prediction of the exposure time needed to produce a crack in the ceramic coating.

  17. Corrosion studies on containment materials for vitrified heat generating nuclear waste

    International Nuclear Information System (INIS)

    Progress is reported of the work undertaken to assess containment materials for the disposal of heat generating nuclear waste under geological disposal conditions. Mean corrosion rates of carbon steels, monitored by Rsub(p) measurements on specimens in on-going long term immersion tests which have been underway from 340 to 500 days, range between 20 to 37 μm/yr at 90 deg C, 9 to 32 μm/yr at 50 deg C and 2 to 10 μm/yr at 25 deg C. AC impedance measurements suggest that true corrosion rates may be a factor of ten lower than those at 90 deg C for the bentonite backfill. Recent tests with γ-radiation have shown that after 5000h in deaerated groundwater at 90 deg C, and at a dose rate of 105 Rads/h, the general corrosion rate of forged 0.2% carbon steel samples < 3 μm/yr. Improved electrochemical kinetic data, required for the mathematical model, is being determined experimentally for temperatures from 20 to 90 deg C and pH values 5 to 10. Preliminary results from studies to assess the effect of γ-radiation on the corrosion of Hastelloy C4 and Ti/0.2 Pd reference materials immersed in groundwater have provided visual evidence for pitting and crevice attack on the Hastelloy sample. (author)

  18. In-situ containment of buried waste at Brookhaven National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Dwyer, B.P. [Sandia National Labs., Albuquerque, NM (United States); Heiser, J. [Brookhaven National Lab., Upton, NY (United States); Stewart, W.; Phillips, S. [Applied Geotechnical Engineering and Construction, Inc., Richland, WA (United States)

    1997-12-31

    The primary objective of this project was to further develop close-coupled barrier technology 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 chemically resistant polymer layer. The technology has matured from a regulatory investigation of issues concerning barriers and barrier materials to a pilot-scale, multiple individual column injections at Sandia National Labs (SNL) to full scale demonstration. The feasibility of this barrier concept was successfully proven in a full scale {open_quotes}cold test{close_quotes} demonstration at Hanford, WA. Consequently, a full scale deployment of the technology was conducted at an actual environmental restoration site at Brookhaven National Lab (BNL), Long Island, NY. This paper discusses the installation and performance of a technology deployment implemented at OU-1 an Environmental Restoration Site located at BNL.

  19. Biodiesel Production from Waste Edible Oils and Grease Containing Free Fatty Acids

    Institute of Scientific and Technical Information of China (English)

    Huang Fenghong; Guo Pingmei; Huang Qingde

    2005-01-01

    Till now, most part of the biodiesel is produced from the refined vegetable oils using methanol as feedstock in the presence of an alkali catalyst. However, large amount of waste edible oils and grease are available. The difficulty with alkali-catalyzed esterification of these oils is that they often contain large amount of free fatty acids (FFA), polymers and decomposition products. These free fatty acids can quickly react with the alkali catalyst to produce soaps that inhibit the separation of the ester and glycerine. An esterification and transesterification process is developed to convert the high FFA oil to its monoesters. The first step, the acidcatalyzed esterification with glycerine and these FFA reduces the FFA content of the oil and grease to less than3%, and then an azeotropic distillation solvent is used to remove the water. The major factors affecting the conversion efficiency of the process such as glycerol to free fatty acid molar ratio, catalyst amount, reaction temperature and reaction duration are analyzed. The second step, alkali-catalyzed transesterification process converts the products of the first step to its monoesters and glycerol, and then the glycerol is recycled for utilization in the first step. Technical indicators of the biodiesel product can meet the ASTM 6751 standard.

  20. Containment barrier metals for high-level waste packages in a Tuff repository

    International Nuclear Information System (INIS)

    The Nevada Nuclear Waste Storage Investigations (NNWSI) Waste Package project is part of the US Department of Energy's Civilian Radioactive Waste Management (CRWM) Program. The NNWSI project is working towards the development of multibarriered packages for the disposal of spent fuel and high-level waste in tuff in the unsaturated zone at Yucca Mountain at the Nevada Test Site (NTS). The final engineered barrier system design may be composed of a waste form, canister, overpack, borehole liner, packing, and the near field host rock, or some combination thereof. Lawrence Livermore National Laboratory's (LLNL) role is to design, model, and test the waste package subsystem for the tuff repository. At the present stage of development of the nuclear waste management program at LLNL, the detailed requirements for the waste package design are not yet firmly established. In spite of these uncertainties as to the detailed package requirements, we have begun the conceptual design stage. By conceptual design, we mean design based on our best assessment of present and future regulatory requirements. We anticipate that changes will occur as the detailed requirements for waste package design are finalized. 17 references, 4 figures, 10 tables

  1. Road pavers' occupational exposure to asphalt containing waste plastic and tall oil pitch.

    Science.gov (United States)

    Väänänen, Virpi; Elovaara, Eivor; Nykyri, Erkki; Santonen, Tiina; Heikkilä, Pirjo

    2006-01-01

    Waste plastic (WP) and tall oil pitch (T), which are organic recycled industrial by-products, have been used as a binder with bitumen in stone mastic asphalt (SMA) and asphalt concrete (AC). We compared the exposure over one workday in 16 road pavers participating in a survey at four paving sites, using mixes of conventional asphalt (SMA, AC) or mixes containing waste material (SMA-WPT, AC-WPT). The concentrations of 11 aldehydes in air were 515 and 902 microg m(-3) at the SMA-WPT and AC-WPT worksites, being 3 and 13 times greater than at the corresponding worksites laying conventional asphalt. Resin acids (2-42 microg m(-3)), which are known sensitizers, were detected only during laying of AC-WPT. The emission levels (microg m(-3)) of total particulates (300-500), bitumen fumes (60-160), bitumen vapour (80-1120), naphthalene (0.59-1.2), phenanthrene (0.21-0.32), pyrene (<0.015-0.20), benzo(a)pyrene (<0.01) and the sum of 16 PAHs (polycyclic aromatic hydrocarbons, 1.28-2.00) were similar for conventional and WPT asphalts. The dermal deposition of 16 PAHs on exposure pads (on workers' wrist) was low in all pavers (0.7-3.5 ng cm(-2)). Eight OH-PAH biomarkers of naphthalene, phenanthrene and pyrene exposures were quantified in pre- and post-shift urine specimens. The post-shift concentrations (mean +/- SD, micromol mol(-1) creatinine) of 1- plus 2-naphthol; 1-,2-,3-,4- plus 9-phenanthrol; and 1-hydroxypyrene were, respectively, for asphalt workers: 18.1+/- 8.0, 2.41 +/- 0.71 and 0.66+/- 0.58 (smokers); 6.0+/- 2.3, 1.70+/- 0.72 and 0.27+/- 0.15 (non-smokers); WPT asphalt workers: 22.0+/- 9.2, 2.82+/- 1.11 and 0.76+/- 0.18 (smokers); 6.8+/- 2.6, 2.35+/- 0.69 and 0.46+/- 0.13 (non-smokers). The work-related uptake of PAHs was low in all pavers, although it was significantly greater in smokers than in non-smokers. The WPT asphalt workers complained of eye irritation and sore throat more than the pavers who had a much lower exposure to aldehydes and resin acids.

  2. 222-S radioactive liquid waste line replacement and 219-S secondary containment upgrade, Hanford Site, Richland, Washington

    International Nuclear Information System (INIS)

    The U.S. Department of Energy (DOE) is proposing to: (1) replace the 222-S Laboratory (222-S) radioactive liquid waste drain lines to the 219-S Waste Handling Facility (219-S); (2) upgrade 219-S by replacing or upgrading the waste storage tanks and providing secondary containment and seismic restraints to the concrete cells which house the tanks; and (3) replace the transfer lines from 219-S to the 241-SY Tank Farm. This environmental assessment (EA) has been prepared in compliance with the National Environmental Policy Act (NEPA) of 1969, as amended, the Council on Environmental Quality Regulations for Implementing the Procedural Provisions of NEPA (40 Code of Federal Regulations [CFR] 1500-1508), and the DOE Implementing Procedures for NEPA (10 CFR 1021). 222-S is used to perform analytical services on radioactive samples in support of the Tank Waste Remediation System and Hanford Site environmental restoration programs. Activities conducted at 222-S include decontamination of analytical processing and support equipment and disposal of nonarchived radioactive samples. These activities generate low-level liquid mixed waste. The liquid mixed waste is drained through pipelines in the 222-S service tunnels and underground concrete encasements, to two of three tanks in 219-S, where it is accumulated. 219-S is a treatment, storage, and/or disposal (TSD) unit, and is therefore required to meet Washington Administrative Code (WAC) 173-303, Dangerous Waste Regulations, and the associated requirements for secondary containment and leak detection. The service tunnels are periodically inspected by workers and decontaminated as necessary to maintain as low as reasonably achievable (ALARA) radiation levels. Although no contamination is reaching the environment from the service tunnels, the risk of worker exposure is present and could increase. 222-S is expected to remain in use for at least the next 30 years to serve the Hanford Site environmental cleanup mission

  3. Volume reduction and encapsulation process for water containing low level radioactive waste

    International Nuclear Information System (INIS)

    In encapsulating solutions or slurries of radio-active waste within polymeric material for disposal, the water is removed therefrom by adding a water insoluble liquid forming a low boiling azeotrope and evaporating the azeotrope, and then a polymerisable composition is dispersed throughout the dewatered waste and allowed to set. (author)

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

    International Nuclear Information System (INIS)

    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

  5. IDMS studies on sodalite - a candidate material for nuclear waste containment

    International Nuclear Information System (INIS)

    Nuclear waste management is one of the important aspects of nuclear fuel cycle from environmental and safety considerations. Different forms of waste storage matrices are known to be applicable for different kinds of nuclear wastes. Glass bonded sodalite (GBS) (Na8(AISiO4)6Cl2), a glass-ceramic, is a promising candidate for the immobilization of the chloride waste resulting from pyrometallurgical reprocessing of nuclear fuels. Characterization of individual components is essential for the development of this waste storage material which is expected to encounter different physicochemical conditions. For this purpose, we have undertaken studies to determine the concentrations of individual components in GBS employing Isotope Dilution Mass Spectrometry (IDMS) owing to its capability to ensure precise and accurate data for multi element analysis in a matrix

  6. Corrosion studies of container materials for radioactive waste disposal in granite formation

    International Nuclear Information System (INIS)

    This report describes the research carried out for the assessment of corrosion behaviour of materials selected for the manufacture of containers for disposal of radioactive waste in granite formation. Metals and alloys included in laboratory test program were: titanium and titanium + 0.2% palladium; zircaloy 4; Hastelloy C276; 625 and 825 alloys; 600 and 800 alloys; 316 L and 304 L stainless steels. Test solutions were designed on the basis of a synthetic ground-water, the concentration of some ions (e.g. Cl- and H3O+) being increased. Temperature was 800C. General corrosion rate was evaluated on non-pitted metals and was very low after a 4 month immersion test. Maximal rates are: - zircaloy 4: 0.03 μm/y; - hastelloy C276, titanium and titanium-palladium: 0.15 μm/y; - 625 alloy: 0.40 μm/y. Pitting corrosion is observed on stainless steels and on 600, 800 and 825 alloys. Zircaloy could be susceptible to this type of attack by coupling with more noble metals. Crevice corrosion is a consequence of the increase in Cl- and H3O+ concentration. Increasing resistance order gives the following list: 304L, 316L, 825, 625 and C276. Stress corrosion cracking susceptibility was assessed by low constant extension rate testing (CERT). Hastelloy C276 and 625 alloys develop S.C.C. in solutions used for crevice corrosion studies. Nevertheless, 825 alloy (eliminated after pitting studies) does not present S.C.C. when tested in the same way. The thermal treatment which can affect the welded zone increased the susceptibility to crevice corrosion of hastelloy C4 and hastelloy C276, but only if concerned areas are cold-worked. The behaviour of C4 is not better than C276 behaviour as far as this type of corrosion is concerned

  7. Strength of Blended Cement Sandcrete & Soilcrete Blocks Containing Cassava Waste Ash and Plantain Leaf Ash

    Directory of Open Access Journals (Sweden)

    L. O. Ettu

    2013-01-01

    Full Text Available This work investigated the compressive strength of binary and ternary blended cement sandcrete and soilcrete blocks containing cassava waste ash (CWA and plantain leaf ash (PLA. 135 solid sandcrete blocks and 135 solid soilcrete blocks of 450mm x 225mm x 125mm were produced with OPC-CWA binary blended cement, 135 with OPC-PLA binary blended cement, and 135 with OPC-CWA-PLA ternary blended cement, each at percentage OPC replacement with pozzolan of 5%, 10%, 15%, 20%, and 25%.Three sandcrete blocks and three soilcrete blocks for each OPC-pozzolan mix and the control were crushed to obtain their compressive strengths at 3, 7, 14, 21, 28, 50, 90, 120, and 150 days of curing. Sandcrete and soilcrete block strengths from binary and ternary blended cements were found to be higher than the control values beyond 90 days of hydration. The 150-day strength values for OPC-CWA-PLA ternary blended cement sandcrete and soilcrete blocks were respectively 5.90N/mm2and 5.10N/mm2for 5% replacement, 5.80N/mm2and 4.95N/mm2for 10% replacement, 5.65N/mm2and 4.85N/mm2for 15% replacement, 5.60N/mm2and 4.75N/mm2for 20% replacement, and 5.25N/mm2and 4.65N/mm2for 25% replacement; while the control values were 5.20N/mm2and 4.65N/mm2. Thus, OPC-CWA and OPC-PLA binary blended cements as well as OPC-CWA-PLA ternary blended cement could be used in producing sandcrete and soilcrete blocks with sufficient strength for use in building and minor civil engineering works where the need for high early strength is not a critical factor.

  8. Measurement and modelling of reactive transport in geological barriers for nuclear waste containment.

    Science.gov (United States)

    Xiong, Qingrong; Joseph, Claudia; Schmeide, Katja; Jivkov, Andrey P

    2015-11-11

    Compacted clays are considered as excellent candidates for barriers to radionuclide transport in future repositories for nuclear waste due to their very low hydraulic permeability. Diffusion is the dominant transport mechanism, controlled by a nano-scale pore system. Assessment of the clays' long-term containment function requires adequate modelling of such pore systems and their evolution. Existing characterisation techniques do not provide complete pore space information for effective modelling, such as pore and throat size distributions and connectivity. Special network models for reactive transport are proposed here using the complimentary character of the pore space and the solid phase. This balances the insufficient characterisation information and provides the means for future mechanical-physical-chemical coupling. The anisotropy and heterogeneity of clays is represented using different length parameters and percentage of pores in different directions. Resulting networks are described as mathematical graphs with efficient discrete calculus formulation of transport. Opalinus Clay (OPA) is chosen as an example. Experimental data for the tritiated water (HTO) and U(vi) diffusion through OPA are presented. Calculated diffusion coefficients of HTO and uranium species are within the ranges of the experimentally determined data in different clay directions. This verifies the proposed pore network model and validates that uranium complexes are diffusing as neutral species in OPA. In the case of U(vi) diffusion the method is extended to account for sorption and convection. Rather than changing pore radii by coarse grained mathematical formula, physical sorption is simulated in each pore, which is more accurate and realistic. PMID:26524292

  9. Thermal performance of a buried nuclear waste storage container storing a hybrid mix of PWR and BWR spent fuel rods

    International Nuclear Information System (INIS)

    Lawrence Livermore National Laboratory will design, model, and test nuclear waste packages for use at the Nevada Nuclear Waste Storage Repository at Yucca Mountain, Nevada. On such package would store tightly packed spent fuel rods from both pressurized and boiling water reactors. The storage container provides the primary containment of the nuclear waste and the spent fuel rod cladding provides secondary containment. A series of transient conduction and radiation heat transfer analyses was run to determine for the first 1000 yr of storage if the temperature of the tuff at the borehole wall ever falls below 97 degrees C and whether the cladding of the stored spent fuel ever exceeds 350 degrees C. Limiting the borehole to temperatures of 97 degrees C or greater helps minimize corrosion by assuring that no condensed water collects on the container. The 350 degrees C cladding limit minimizes the possibility of creep- related failure in the spent fuel rod cladding. For a series of packages stored in a 8 x 30 m borehole grid where each package contains 10-yr-old spent fuel rods generating 4.74 kW or more, the borehole wall stays above 97 degrees C for the full 10000-yr analysis period. For the 4.74-kW load, the peak cladding temperature rises to just below the 350 degrees C limit about 4 years after emplacement. If the packages are stored using the spacing specified in the Site Characterization Plan (15 ft x 126 ft), a maximum of 4.1 kW per container may be stored. If the 0.05-m-thick void between the container and the borehole wall is filled with loosely packed bentonite, the peak cladding temperature rises more than 40 degrees C above the allowed cladding limit. In all cases the dominant heat transfer mode between container components is thermal radiation

  10. INITIAL WASTE PACKAGE PROBABILISTIC CRITICALITY ANALYSIS: MULTI-PURPOSE CANISTER WITH DISPOSAL CONTAINER (TBV)

    Energy Technology Data Exchange (ETDEWEB)

    J.R. Massari

    1995-10-06

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide an assessment of the present waste package design from a criticality risk standpoint. The specific objectives of this initial analysis are to: (1) Establish a process for determining the probability of waste package criticality as a function of time (in terms of a cumulative distribution function, probability distribution function, or expected number of criticalities in a specified time interval) for various waste package concepts; (2) Demonstrate the established process by estimating the probability of criticality as a function of time since emplacement for an intact multi-purpose canister waste package (MPC-WP) configuration; (3) Identify the dominant sequences leading to waste package criticality for subsequent detailed analysis. The purpose of this analysis is to document and demonstrate the developed process as it has been applied to the MPC-WP. This revision is performed to correct deficiencies in the previous revision and provide further detail on the calculations performed. This analysis is similar to that performed for the uncanistered fuel waste package (UCF-WP, B00000000-01717-2200-00079).

  11. Conceptual design of retrieval systems for emplaced transuranic waste containers in a salt bed depository. Final report

    International Nuclear Information System (INIS)

    The US Department of Energy and the Nuclear Regulatory Commission have jurisdiction over the nuclear waste management program. Design studies were previously made of proposed repository site configurations for the receiving, processing, and storage of nuclear wastes. However, these studies did not provide operational designs that were suitable for highly reliable TRU retrieval in the deep geologic salt environment for the required 60-year period. The purpose of this report is to develop a conceptual design of a baseline retrieval system for emplaced transuranic waste containers in a salt bed depository. The conceptual design is to serve as a working model for the analysis of the performance available from the current state-of-the-art equipment and systems. Suggested regulations would be based upon the results of the performance analyses

  12. Corrosion behavior of carbon steel containers with organic coating during interim storage and disposal of low-level radioactive wastes

    International Nuclear Information System (INIS)

    In the Federal Republic of Germany low and intermediate level wastes (e.g., solids, concentrates) are conditioned in carbon steel canisters with organic coating. For this purpose waste drums and steel sheet containers are used. They serve as unshielded packagings during interim storage, transport and disposal in the Konrad mine or in the Gorleben salt dome. Considering the licensing situation for the planned repositories, interim storage periods of up to 20 years are possible. During this period, the transport to the repository and in the operation phase of the repository, the integrity of the waste packaging must be guaranteed. Therefore, special attention must be paid to the corrosion behavior of the steel sheet packagings described in this report. For these reasons, corrosion studies were made on epoxy resin coated or polyurethane coated and uncoated stell sheet specimens. In the investigations design details of the containers (e.g., roundings, screwed connections, gaps, welded seams) as well as damage due to handling (cracks in the organic coating) were taken into account. The specimens were stored for six and twelve months, respectively, both in waste form simulates (inner corrosion of container) and under simulated conditions of an interim storage (storage hall) and of a repository (storage galeries in Konrad and Asse, salt brines) in order to be able to describe external container corrosion. Under simplifying boundary conditions an extrapolation is made of the test results. It has been possible to show that the carbon steel containers described here, provided with a 150 μm epoxy resin coating on the inner and external sides, fulfil the requirements imposed on them as regards their corrosion behavior. (orig.)

  13. Method for contamination control and barrier apparatus with filter for containing waste materials that include dangerous particulate matter

    International Nuclear Information System (INIS)

    A container for hazardous waste materials that includes air or other gas carrying dangerous particulate matter has incorporated barrier material, preferably in the form of a flexible sheet, and one or more filters for the dangerous particulate matter sealably attached to such barrier material. The filter is preferably a HEPA type filter and is preferably chemically bonded to the barrier materials. The filter or filters are preferably flexibly bonded to the barrier material marginally and peripherally of the filter or marginally and peripherally of air or other gas outlet openings in the barrier material, which may be a plastic bag. The filter may be provided with a backing panel of barrier material having an opening or openings for the passage of air or other gas into the filter or filters. Such backing panel is bonded marginally and peripherally thereof to the barrier material or to both it and the filter or filters. A coupling or couplings for deflating and inflating the container may be incorporated. Confining a hazardous waste material in such a container, rapidly deflating the container and disposing of the container, constitutes one aspect of the method of the invention. The chemical bonding procedure for producing the container constitutes another aspect of the method of the invention. 3 figs

  14. Evaluation of dry-solids-blend material source for grouts containing 106-AN waste: September 1990 progress report

    Energy Technology Data Exchange (ETDEWEB)

    Gilliam, T.M.; Osborne, S.C.; Francis, C.L.; Scott, T.C.

    1993-09-01

    Stabilization/solidification (S/S) is the most widely used technology for the treatment and ultimate disposal of both radioactive and chemically hazardous wastes. Such technology is being utilized in a Grout Treatment Facility (GTF) by the Westinghouse Hanford Company (WHC) for the disposal of various wastes, including 106-AN wastes, located on the Hanford Reservation. The WHC personnel have developed a grout formula for 106-AN disposal that is designed to meet stringent performance requirements. This formula consists of a dry-solids blend containing 40 wt % limestone, 28 wt % granulated blast furnace slag (BFS), 28 wt % ASTM Class F fly ash, and 4 wt % Type I-II-LA Portland cement. The blend is mixed with 106-AN waste at a ratio of 9 lb of dry-solids blend per gallon of waste. This report documents progress made to date on efforts at Oak Ridge National Laboratory (ORNL) in support of WHC`s Grout Technology Program to assess the effects of the source of the dry-solids-blend materials on the resulting grout formula.

  15. Drug waste minimisation and cost-containment in Medical Oncology: Two-year results of a feasibility study

    Directory of Open Access Journals (Sweden)

    Mansutti Mauro

    2008-04-01

    Full Text Available Abstract Background Cost-containment strategies are required to face the challenge of rising drug expenditures in Oncology. Drug wastage leads to economic loss, but little is known about the size of the problem in this field. Methods Starting January 2005 we introduced a day-to-day monitoring of drug wastage and an accurate assessment of its costs. An internal protocol for waste minimisation was developed, consisting of four corrective measures: 1. A rational, per pathology distribution of chemotherapy sessions over the week. 2. The use of multi-dose vials. 3. A reasonable rounding of drug dosages. 4. The selection of the most convenient vial size, depending on drug unit pricing. Results Baseline analysis focused on 29 drugs over one year. Considering their unit price and waste amount, a major impact on expense was found to be attributable to six drugs: cetuximab, docetaxel, gemcitabine, oxaliplatin, pemetrexed and trastuzumab. The economic loss due to their waste equaled 4.8% of the annual drug expenditure. After the study protocol was started, the expense due to unused drugs showed a meaningful 45% reduction throughout 2006. Conclusion Our experience confirms the economic relevance of waste minimisation and may represent a feasible model in addressing this issue. A centralised unit of drug processing, the availability of a computerised physician order entry system and an active involvement of the staff play a key role in allowing waste reduction and a consequent, substantial cost-saving.

  16. The containment of toxic wastes: I. Long term metal movement in soils over a covered metalliferous waste heap at Parc lead-zinc mine, North Wales.

    Science.gov (United States)

    Shu, J; Bradshaw, A D

    1995-01-01

    In order to stabilise and contain a toxic metalliferous waste heap at Parc Mine, North Wales, it was covered with 30-40 cm layer of quarry waste in 1977-1978, and sown with a grass/clover seed mixture. This study has examined subsequent metal movement in the cover material and its effect on vegetation. The results, especially when compared with previous observations, give no evidence of upward migration of metals by capillarity in the cover material. Sideways movement of leachate, however, appears to be carrying the metals into the cover material on the sloping sides, giving rise to increasing concentrations of heavy metals in the vegetation and dieback in some places. Root growth on the flat top of the heap is greater than on the slope, but the roots have not penetrated the waste and the contents of Pb, Zn and Cd in surface vegetation remain low. Surface covering of toxic waste with coarse materials restricting capillary rise is therefore a valid reclamation technique so long as lateral movement of toxic leachate can be controlled.

  17. Development of a Lightweight Low-Carbon Footprint Concrete Containing Recycled Waste Materials

    OpenAIRE

    Talukdar, S.; Islam, S. T.; Banthia, N.

    2011-01-01

    Use of any recycled material helps to maintain a greener environment by keeping waste materials out of the landfills. Recycling practices also can decrease the environmental and economical impact of manufacturing the materials from virgin resources, which reduces the overall carbon footprint of industrial materials and processes. This study examined the use of waste materials such as crushed glass, ground tire rubber, and recycled aggregate in concrete. Compressive strength and elastic mod...

  18. Utilization of different waste proteins to create a novel PGPR-containing bio-organic fertilizer

    OpenAIRE

    Huang, Yan; Sun, Li; Zhao, Jianshu; Huang, Rong; Li, Rong; Shen, Qirong

    2015-01-01

    High-quality bio-organic fertilizers (BIOs) cannot be produced without the addition of some proteins, while many waste proteins are haphazardly disposed, causing serious environmental pollution. In this study, several waste proteins were used as additives to assist with the reproduction of the functional microbe (Bacillus amyloliquefaciens SQR9) inoculated into matured composts to produce BIOs. An optimized composition of solid-state fermentation (SSF) raw materials was predicted by response ...

  19. Development of a testing method for asbestos fibers in treated materials of asbestos containing wastes by transmission electron microscopy

    International Nuclear Information System (INIS)

    Highlights: • A high sensitive and selective testing method for asbestos in treated materials of asbestos containing wastes was developed. • Asbestos can be determined at a limits are a few million fibers per gram and a few μg g−1. • High temperature melting treatment samples were determined by this method. Asbestos fiber concentration were below the quantitation limit in all samples, and total fiber concentrations were determined as 47–170 × 106 g−1. - Abstract: Appropriate treatment of asbestos-containing wastes is a significant problem. In Japan, the inertization of asbestos-containing wastes based on new treatment processes approved by the Minister of the Environment is promoted. A highly sensitive method for testing asbestos fibers in inertized materials is required so that these processes can be approved. We developed a method in which fibers from milled treated materials are extracted in water by shaking, and are counted and identified by transmission electron microscopy. Evaluation of this method by using asbestos standards and simulated slag samples confirmed that the quantitation limits are a few million fibers per gram and a few μg/g in a sample of 50 mg per filter. We used this method to assay asbestos fibers in slag samples produced by high-temperature melting of asbestos-containing wastes. Fiber concentrations were below the quantitation limit in all samples, and total fiber concentrations were determined as 47–170 × 10−6 f/g. Because the evaluation of treated materials by TEM is difficult owing to the limited amount of sample observable, this testing method should be used in conjunction with bulk analytical methods for sure evaluation of treated materials

  20. Development of a testing method for asbestos fibers in treated materials of asbestos containing wastes by transmission electron microscopy

    Energy Technology Data Exchange (ETDEWEB)

    Yamamoto, Takashi, E-mail: tyama@nies.go.jp [Center for Material Cycles and Waste Management Research, National Institute for Environmental Studies, 16-2 Onogawa, Tsukuba, Ibaraki 305-8506 (Japan); Kida, Akiko [Faculty of Agriculture, Ehime University, 3-5-7 Tarumi, Matsuyama, Ehime 790-8566 (Japan); Noma, Yukio [Department of Environmental Science, Fukuoka Womens University, 1-1-1 Kasumigaoka, Higashiku, Fukuoka 813-8529 (Japan); Terazono, Atsushi [Center for Material Cycles and Waste Management Research, National Institute for Environmental Studies, 16-2 Onogawa, Tsukuba, Ibaraki 305-8506 (Japan); Sakai, Shin-ichi [Environmental Preservation Research Center, Kyoto University, Yoshidahonmachi, Sakyoku, Kyoto 606-8501 (Japan)

    2014-02-15

    Highlights: • A high sensitive and selective testing method for asbestos in treated materials of asbestos containing wastes was developed. • Asbestos can be determined at a limits are a few million fibers per gram and a few μg g{sup −1}. • High temperature melting treatment samples were determined by this method. Asbestos fiber concentration were below the quantitation limit in all samples, and total fiber concentrations were determined as 47–170 × 10{sup 6} g{sup −1}. - Abstract: Appropriate treatment of asbestos-containing wastes is a significant problem. In Japan, the inertization of asbestos-containing wastes based on new treatment processes approved by the Minister of the Environment is promoted. A highly sensitive method for testing asbestos fibers in inertized materials is required so that these processes can be approved. We developed a method in which fibers from milled treated materials are extracted in water by shaking, and are counted and identified by transmission electron microscopy. Evaluation of this method by using asbestos standards and simulated slag samples confirmed that the quantitation limits are a few million fibers per gram and a few μg/g in a sample of 50 mg per filter. We used this method to assay asbestos fibers in slag samples produced by high-temperature melting of asbestos-containing wastes. Fiber concentrations were below the quantitation limit in all samples, and total fiber concentrations were determined as 47–170 × 10{sup −6} f/g. Because the evaluation of treated materials by TEM is difficult owing to the limited amount of sample observable, this testing method should be used in conjunction with bulk analytical methods for sure evaluation of treated materials.

  1. Reduction of 68Ge activity containing liquid waste from 68Ga PET chemistry in nuclear medicine and radiopharmacy by solidification

    NARCIS (Netherlands)

    E. de Blois (Erik); H.S. Chan; K. Roy (Kamalika); E.P. Krenning (Eric); W.A.P. Breeman (Wouter)

    2011-01-01

    textabstractPET with68Ga from the TiO2- or SnO2- based68Ge/68Ga generators is of increasing interest for PET imaging in nuclear medicine. In general, radionuclidic purity (68Ge vs.68Ga activity) of the eluate of these generators varies between 0.01 and 0.001%. Liquid waste containing low amounts of6

  2. Measurements of Flammable Gas Generation from Saltstone Containing Actual Tank 48H Waste (Interim Report)

    Energy Technology Data Exchange (ETDEWEB)

    Cozzi, A. D.; Crowley, D. A.; Duffey, J. M.; Eibling, R. E.; Jones, T. M.; Marinik, A. R.; Marra, J. C.; Zamecnik, J. R

    2005-06-01

    The Savannah River National Laboratory was tasked with determining the benzene release rates in saltstone prepared with tetraphenylborate (TPB) concentrations ranging from 30 mg/L to 3000 mg/L in the salt fraction and with test temperatures ranging from ambient to 95 C. Defense Waste Processing Facility Engineering (DWPF-E) provided a rate of benzene evolution from saltstone of 2.5 {micro}g/L/h saltstone (0.9 {micro}g/kg saltstone/h [1.5 {micro}g/kg saltstone/h x 60%]) to use as a Target Rate of Concern (TRC). The evolution of benzene, toluene, and xylenes from saltstone containing actual Tank 48H salt solution has been measured as a function of time at several temperatures and concentrations of TPB. The Tank 48H salt solution was aggregated with a DWPF recycle simulant to obtain the desired TPB concentrations in the saltstone slurry. The purpose of this interim report is to provide DWPF-E with an indication of the trends of benzene evolution. The data presented are preliminary; more data are being collected and may alter the preliminary results. A more complete description of the methods and materials will be included in the final report. The benzene evolution rates approximately follow an increasing trend with both increasing temperature and TPB concentration. The benzene release rates from 1000 mg/L TPB at 95 C and 3000 mg/L TPB at 75 C and 95 C exceeded the recovery-adjusted 0.9 mg/kg saltstone/h TRC (2.5 {micro}g/L saltstone/h), while all other conditions resulted in benzene release rates below this TRC. The toluene evolution rates for several samples exceeded the TRC initially, but all dropped below the TRC within 2-5 days. The toluene emissions appear to be mainly dependent on the fly ash and are independent of the TPB level, indicating that toluene is not generated from TPB.

  3. SGSreco. Radiological characterization of waste containers by segmented gamma-Scan measurements

    International Nuclear Information System (INIS)

    Starting from 2021, low and intermediate level radioactive waste produced in the Federal Republic of Germany will be finally disposed at a depth from 800 m to 1300 m in the Konrad Repository, close to the city Salzgitter. A prerequisite for the final disposal of radioactive waste packages is their conformance with national acceptance criteria. These acceptance criteria include among others radiological requirements for waste packages. To ensure a conformance of waste packages with these radiological requirements, experimental techniques are applied to characterize their radionuclide inventories. For this purpose, segmented γ-scanning is used worldwide as the standard non-destructive assay for the radiological characterization of waste drums. Segmented γ-scanning investigates predefined parts of a waste drum independently of each other using γ-spectrometry with a collimated detection system. Radionuclides are identified by their characteristic γ-lines in each recorded γ-spectrum, and two-dimensional count rate distributions are determined depending on the positions of the investigated predefined parts. The reconstruction of radionuclide specific activities by conventional methods requires a homogeneous matrix and radionuclide distribution within the whole drum. Thus, radionuclide specific activities are estimated using an analytical model based on the average count rate of a characteristic γ-line over all investigated parts of the waste drum. However, only 25% of all waste drums meet these requirements. It is therefore expected that the radionuclide specific activities for the majority of waste drums are miscalculated by several orders of magnitude. In this work, an analysis framework known as SGSreco is presented. SGSreco aims to ensure an accurate and a reliable reconstruction of radionuclide specific activities for homogeneous and spatially concentrated (point sources) radionuclide inventories. SGSreco uses an inverse approach. Within a first

  4. Mobile neutron/gamma waste assay system for characterization of waste containing transuranics, uranium, and fission/activation products

    International Nuclear Information System (INIS)

    A new integrated neutron/gamma assay system has been built for measuring 55-gallon drums at Pacific Northwest Laboratory. The system is unique because it allows simultaneous measurement of neutrons and gamma-rays. This technique also allows measurement of transuranics (TRU), uranium, and fission/activation products, screening for shielded Special Nuclear Material prior to disposal, and critically determinations prior to transportation. The new system is positioned on a platform with rollers and installed inside a trailer or large van to allow transportation of the system to the waste site instead of movement of the drums to the scanner. The ability to move the system to the waste drums is particularly useful for drum retrieval programs common to all DOE sites and minimizes transportation problems on the site. For longer campaigns, the system can be moved into a facility. The mobile system consists of two separate subsystems: a passive Segmented Gamma Scanner (SGS) and a open-quotes clam-shellclose quotes passive neutron counter. The SGS with high purity germanium detector and 75Se transmission source simultaneously scan the height of the drum allowing identification of unshieled open-quotes hot spotsclose quotes in the drum or segments where the matrix is too dense for the transmission source to penetrate. Dense segments can flag shielding material that could be used to hide plutonium or uranium during the gamma analysis. The passive nuetron counter with JSR-12N Neutron Coincidence Analyzer measures the coincident neutrons from the spontaneous fission of even isotopes of plutonium. Because high-density shielding produces minimal absorption of neutrons, compared to gamma rays, the passive neutron portion of the system can detect shielded SNM. Measurements to evaluate the performance of the system are still underway at Pacific Northwest Laboratory

  5. Replacing fish meal by food waste to produce lower trophic level fish containing acceptable levels of polycyclic aromatic hydrocarbons: Health risk assessments.

    Science.gov (United States)

    Cheng, Zhang; Mo, Wing-Yin; Lam, Cheung-Lung; Choi, Wai-Ming; Wong, Ming-Hung

    2015-08-01

    This study aimed at using different types of food wastes (mainly containing cereal [food waste A] and meat meal [food waste B]) as major sources of protein to replace the fish meal used in fish feeds to produce quality fish. The traditional fish farming model used to culture low trophic level fish included: bighead, (Hypophthalmichthys nobilis), grass carp, (Ctenopharyngodon idellus), and mud carp, (Cirrhinus molitorella) of omnivorous chain. The results indicated that grass carp and bighead carp fed with food waste feeds were relatively free of PAHs. The results of health risk assessment showed that the fish fed with food waste feeds were safe for consumption from the PAHs perspective. PMID:25880597

  6. Closure of hazardous and mixed radioactive waste management units at DOE facilities. [Contains glossary

    Energy Technology Data Exchange (ETDEWEB)

    1990-06-01

    This is document addresses the Federal regulations governing the closure of hazardous and mixed waste units subject to Resource Conservation and Recovery Act (RCRA) requirements. It provides a brief overview of the RCRA permitting program and the extensive RCRA facility design and operating standards. It provides detailed guidance on the procedural requirements for closure and post-closure care of hazardous and mixed waste management units, including guidance on the preparation of closure and post-closure plans that must be submitted with facility permit applications. This document also provides guidance on technical activities that must be conducted both during and after closure of each of the following hazardous waste management units regulated under RCRA.

  7. Polymers in waste electric and electronic equipment (WEEE) contain PBDD/F in the ppb-range

    Energy Technology Data Exchange (ETDEWEB)

    Schlummer, M.; Brandl, F.; Maeurer, A.; Gruber, L.; Wolz, G. [Fraunhofer-Institut fuer Verfahrenstechnik und Verpackung (IVV), Freising (Germany)

    2004-09-15

    Waste electric and electronic equipment (WEEE) consists of metals (60%), polymers (20%) and residual materials as wood or glass (20%). Whereas state-of the art-technologies are able to recover most of the metals present, recovery rates for polymers and residuals are negligible low. Primarily, this is due to low disposal costs, which refers to landfill or incineration depending on geographic circumstances. The European WEEE directive, which assesses material recovery rates above 70%, and changes in the German disposal regulation, which will prohibit the landfill of organic materials starting 2005, currently alter the legislative conditions. This leads to an increased interest in polymer recovery strategies. Approaches discussed include polymer recycling and pyrolysis-based material recovery, both characterised by temperatures below 240 C or 600 C, respectively. Polybrominated biphenyls (PBB) and/or diphenyl ethers (PBDE) in these waste streams complicate waste treatment techniques, since they are known to form brominated dioxins and furans (PBDD/F) under thermal stress, either in polymer recyclates or in pyrolysis products. Additionally, polymer recycling is affected by European directive 2003/11/EC, restricting the distribution of products containing more than 0.1% of octa- or pentabrominated diphenyl ethers, respectively. Aim of this study was to determine concentration levels of polybrominated compounds including PBDD/F and brominated flame retardants in polymers from WEEE. Both, mixed polymer waste and pre-sorted polymer fractions consisting mainly of monitors, TV-sets or telecommunication housings, were examined. Furthermore, the dependency of PBDD/F concentrations on waste source, pre-treatment and flame retardant system was investigated, implication on waste treatment alternatives are discussed.

  8. Phase composition and elemental partitioning in glass-ceramics containing high-Na/Al high level waste

    Science.gov (United States)

    Stefanovsky, S. V.; Sorokaletova, A. N.; Nikonov, B. S.

    2012-05-01

    Mixtures of surrogates of high level waste with high sodium and aluminum contents and sodium-lithium borosilicate frit were melted in alumina crucibles in a resistive furnace followed by quenching of one portion of the melt and annealing of the residual material in a turned-off furnace. The annealed materials with waste loading of up to 45 wt.% contained minor spinel type phase and trace of nepheline (Na,K)AlSiO4. In the annealed materials contained waste oxides in amount of 50 wt.% and more nepheline and spinel were found to be major and minor phases, respectively. At high waste loadings two extra phases: Cs-aluminosilicate (CsAlSiO4) and mixed Na/Cs-aluminosilicate were found in amount of 3-5 vol.% each. The latter phase contains of up to ˜5.7 wt.% SO3 or 0.13 formula units S (Na0.75K0.05Cs0.29Ca0.02Sr0.02Al0.99Fe0.03Si0.76S0.13O4). Sulfur incorporation as S6+ or SO42- ions into crystal lattice may be facilitated in the presence of large-size Cs+ cations. Simplified suggested formula of this phase may be represented as Na0.8Cs0.3AlSi0.8S0.1O3.95. It was also synthesized by sintering of mixture of chemicals at 1300 °C and found to be instable at temperatures higher than 1300 °C.

  9. Advanced containment research for the Canadian Nuclear Fuel Waste Management Program

    International Nuclear Information System (INIS)

    This document outlines the program on the development of advanced containment systems for the disposal of used fuel in a vault deep in plutonic rock. Possible advanced containment concepts, the strategy adopted in selecting potential container materials, and experimental programs currently underway or planned are presented. Most effort is currently directed toward developing long-term containment systems based on non-metallic materials and massive metal containers. The use of additional independent barriers to extend the lifetime of simple containment systems is also being evaluated. 58 refs

  10. Survey of the degradation modes of candidate materials for high-level radioactive waste disposal containers. Final report

    International Nuclear Information System (INIS)

    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

  11. Steel corrosion resistance in model solutions and reinforced mortar containing wastes

    NARCIS (Netherlands)

    Koleva, D.A.; Van Breugel, K.

    2012-01-01

    This work reports on the corrosion resistance of steel in alkaline model solutions and in cement-based materials (mortar). The model solutions and the mortar specimens were Ordinary Portland Cement (OPC) based. Further, hereby discussed is the implementation of an eco-friendly approach of waste util

  12. Physical and mechanical properties of mortars containing PET and PC waste aggregates.

    Science.gov (United States)

    Hannawi, Kinda; Kamali-Bernard, Siham; Prince, William

    2010-11-01

    Non-biodegradable plastic aggregates made of polycarbonate (PC) and polyethylene terephthalate (PET) waste are used as partial replacement of natural aggregates in mortar. Various volume fractions of sand 3%, 10%, 20% and 50% are replaced by the same volume of plastic. This paper investigates the physical and mechanical properties of the obtained composites. The main results of this study show the feasibility of the reuse of PC and PET waste aggregates materials as partial volume substitutes for natural aggregates in cementitious materials. Despite of some drawbacks like a decrease in compressive strength, the use of PC and PET waste aggregates presents various advantages. A reduction of the specific weight of the cementitious materials and a significant improvement of their post-peak flexural behaviour are observed. The calculated flexural toughness factors increase significantly with increasing volume fraction of PET and PC-aggregates. Thus, addition of PC and PET plastic aggregates in cementitious materials seems to give good energy absorbing materials which is very interesting for several civil engineering applications like structures subjected to dynamic or impact efforts. The present study has shown quite encouraging results and opened new way for the recycling of PC waste aggregate in cement and concrete composites.

  13. Leaching due to hygroscopic water uptake in cemented waste containing soluble salts

    DEFF Research Database (Denmark)

    Brodersen, K.

    1992-01-01

    Considerable amounts of easily soluble salts such as sodium nitrate, sulphate, or carbonate are introduced into certain types of cemented waste. When such materials are stored in atmospheres with high relative humidity or disposed or by shallow land burial under unsaturated, but still humid...

  14. Stabilization of NaCl-containing cuttings wastes in cement concrete by in situ formed mineral phases.

    Science.gov (United States)

    Filippov, Lev; Thomas, Fabien; Filippova, Inna; Yvon, Jacques; Morillon-Jeanmaire, Anne

    2009-11-15

    Disposal of NaCl-containing cuttings is a major environmental concern due to the high solubility of chlorides. The present work aims at reducing the solubility of chloride by encapsulation in low permeability matrix as well as lowering its solubility by trapping into low-solubility phases. Both the studied materials were cuttings from an oil-based mud in oil drillings containing about 50% of halite, and cuttings in water-based mud from gas drilling containing 90% of halite. A reduction in the amount of dissolved salt from 41 to 19% according to normalized leaching tests was obtained by addition of potassium ortho-phosphate in the mortar formula of oil-based cuttings, while the aluminium dihydrogeno-phosphate is even more efficient for the stabilization of water-based cuttings with a NaCl content of 90%. Addition of ortho-phosphate leads to form a continuous and weakly soluble network in the cement matrix, which reduces the release of salt. The formed mineralogical phases were apatite and hydrocalumite. These phases encapsulate the salt grains within a network, thus lowering its interaction with water or/and trap chloride into low-solubility phases. The tested approaches allow to develop a confinement process of NaCl-containing waste of various compositions that can be applied to wastes, whatever the salt content and the nature of the drilling fluids (water or oil). PMID:19631465

  15. Environmental restoration and waste management site-specific plan for Richland Operations Office. [Contains glossary

    Energy Technology Data Exchange (ETDEWEB)

    1991-09-01

    This document was prepared to implement and support the US Department of Energy-Headquarters (DOE-HQ) national plan. The national plan, entitled Environmental Restoration and Waste Management Five-Year Plan (DOE 1990b) (hereinafter referred to as the DOE-HQ Five-Year Plan) is the cornerstone of the US Department of Energy's (DOE) long-term strategy in environmental restoration and waste management. The DOE-HQ Five-Year Plan addresses overall philosophy and environmental and waste-related activities under the responsibilities of the DOE Office of Environmental Restoration and Waste Management. The plan also reaffirms DOE-HQ goals to bring its nuclear sites into environmental compliance in cooperation with its regulators and the public, and to clean up and restore the environment by 2019 (the commitment for the Hanford Site is for one year sooner, or 2018). This document is part of the site-specific plan for the US Department of Energy-Richland Operations Office (DOE-RL). It is the first revision of the original plan, which was dated December 1989 (DOE-RL 1989a). This document is a companion document to the Overview of the Hanford Cleanup Five-Year Plan (DOE-RL 1989d) and The Hanford Site Environmental Restoration and Waste Management Five-Year Plan Activity Data Sheets (DOE-RL 1991). Although there are three documents that make up the complete DOE-RL plan, this detailed information volume was prepared so it could be used as a standalone document. 71 refs., 40 figs., 28 tabs.

  16. Sustainable Supply Chain Management: The Influence of Disposal Scenarios on the Environmental Impact of a 2400 L Waste Container

    Directory of Open Access Journals (Sweden)

    José Eduardo Galve

    2016-06-01

    Full Text Available This paper analyzes the influence of the supply chain management on the environmental impact of a 2400 L waste disposal container used in most cities of Spain. The studied functional unit, a waste disposal container, made up mostly of plastic materials and a metallic structure, and manufactured in Madrid (Spain, is distributed to several cities at an average distance of 392 km. A life cycle assessment of four different scenarios (SC has been calculated with the software EcoTool v4.0 (version 4.0; i+: Zaragoza, Spain, 2015 and using Ecoinvent v3.0 database (version 3.0; Swiss Centre for Life Cycle Inventories: St. Gallen, Switzerland, 2013. The environmental impact has been characterized with two different methodologies, recipe and carbon footprint. In order to reduce the environmental impact, several end of life scenarios have been performed, analyzing the influence of the supply chain on a closed-looped system that increases recycling. Closed loop management of the waste and reuse of parts allows companies to stop selling products and start selling the service that their products give to the consumers.

  17. Bioremediation of waste cooking oil using a novel lipase produced by Penicillium chrysogenum SNP5 grown in solid medium containing waste grease.

    Science.gov (United States)

    Kumar, Sunil; Mathur, Anisha; Singh, Varsha; Nandy, Suchismita; Khare, Sunil Kumar; Negi, Sangeeta

    2012-09-01

    The aim of present work was to bioremediate the waste cooking oil using a novel lipase produced in solid medium containing waste grease and wheat bran by Penicillium chrysogenum. Enzyme extracted with phosphate buffer was purified 10.6 and 26.28-fold after 90% ammonium sulfate precipitation and ion-exchange chromatography, respectively. The partial characterization of enzyme revealed its K(m) and V(max) value for p-nitrophenolpamitate as 0.4mM and 47.61 U/ml, respectively. The relative molecular mass of lipase was 40 kDa by SDS-PAGE and confirmed by zymogram. Purified lipase was most stable at 40°C and at 8.0 pH. Lipase activity was enhanced by metal ions such as Mg(2+), Fe(2+), Ca(2+) and non-ionic surfactant TritonX-100, while suppressed in the presence of SDS. Crude lipase was applied on cooking oil waste and the acid value was 26.92 mg/g. This showed that the enzyme could be employed for the bioremediation of used cooking oil. PMID:22770974

  18. Corrosion considerations of high-nickel alloys and titanium alloys for high-level radioactive waste disposal containers

    International Nuclear Information System (INIS)

    Corrosion resistant materials are being considered for the metallic barrier of the Yucca Mountain Project's high-level radioactive waste disposal containers. High nickel alloys and titanium alloys have good corrosion resistance properties and are considered good candidates for the metallic barrier. The localized corrosion phenomena, pitting and crevice corrosion, are considered as potentially limiting for the barrier lifetime. An understanding of the mechanisms of localized corrosion of how various parameters affect it will be necessary for adequate performance assessments of candidate container materials. Examples of some of the concerns involving candidate container materials. Examples of some of the concerns of involving localized corrosion are discussed. The effects of various parameters, such as temperature and concentration of halide species, on localized corrosion are given. In addition concerns about aging of the protective oxide layer in the expected service temperature range (50 to 250 degrees C) are presented. Also some mechanistic considerations of localized corrosion are given. 31 refs., 1 tab

  19. Leaching and comprehensive regulatory performance testing of an extruded bitumen containing a surrogate, sodium nitrate-based, low-level waste

    International Nuclear Information System (INIS)

    Performance test results obtained from laboratory testing of an extruded bitumen containing a surrogate, sodium nitrate-based waste are presented. A relatively viscous form of oxidized bitumen (ASTM D 312, Type III) has been tested and has been shown to meet all of the current regulatory performance criteria. Molded specimens were obtained using a 53-mm extruder. A surrogate, low-level, mixed, liquid waste was used. The surrogate waste contained ∼30 weight percent sodium nitrate, in addition to eight heavy metals, cold cesium, and strontium. Waste form specimens contained three levels of waste loading: 40, 50, and 60 weight percent salt. Results include thermal testing, extraction procedure toxicity tests, and 90-day American Nuclear Society 16.1 leach tests, as well as compressive strength tests

  20. Development of an immobilization process for heavy metal containing galvanic solid wastes by use of sodium silicate and sodium tetraborate

    International Nuclear Information System (INIS)

    Highlights: • A new physico-chemical process below 1000 °C for immobilization of galvanic sludges. • Sodium tetraborate and sodium silicate have been used as additives. • A strategy for adjustment of solid waste/additive mixture composition is presented. • Strategy is valid for wastes of hydrometallurgical and electro-plating processes. • Lower energy consumption and treated waste volume, shorter process time are provided. - Abstract: Heavy metal containing sludges from wastewater treatment plants of electroplating industries are designated as hazardous waste since their improper disposal pose high risks to environment. In this research, heavy metal containing sludges of electroplating industries in an organized industrial zone of Istanbul/Turkey were used as real-sample model for development of an immobilization process with sodium tetraborate and sodium silicate as additives. The washed sludges have been precalcined in a rotary furnace at 900 °C and fritted at three different temperatures of 850 °C, 900 °C and 950 °C. The amounts of additives were adjusted to provide different acidic and basic oxide ratios in the precalcined sludge-additive mixtures. Leaching tests were conducted according to the toxicity characteristic leaching procedure Method 1311 of US-EPA. X-ray diffraction (XRD), X-ray fluorescence (XRF), scanning electron microscope-energy dispersive spectrometer (SEM-EDS) and flame atomic absorption spectroscopy (FAAS) have been used to determine the physical and chemical changes in the products. Calculated oxide molar ratios in the precalcined sludge-additive mixtures and their leaching results have been used to optimize the stabilization process and to determine the intervals of the required oxide ratios which provide end-products resistant to leaching procedure of US-EPA. The developed immobilization-process provides lower energy consumption than sintering-vitrification processes of glass–ceramics

  1. Nondestructive assays of 55-gallon drums containing uranium and transuranic waste using passive-active shufflers

    International Nuclear Information System (INIS)

    A passive-active neutron shuffler for 55-gal. drums of waste has been characterized using more than 1500 active and 500 passive assays on drums with 28 different matrices. Flux-monitor corrections have been improved, the assay accuracy with localized fissile materials in a drum has been characterized, and improvements have been suggested. Minimum detectable masses for 235U with active assays and 240Pueff with passive assays are presented for the various amounts of moderators and absorbers studied

  2. Treatment of radioactive liquid waste containing Mn2+ by synthetic hydroxyapatite

    International Nuclear Information System (INIS)

    The removal of manganese from a simulated radioactive liquid waste was evaluated using a radiotracer, 54Mn in a batch method by synthetic Hydroxyapatite (HAP). The influences of different parameters such as solution pH, and sodium concentration, on manganese removal were studied. In the presence of sodium ions, the sorption of manganese on HAP was not affected. The sorption of manganese on HAP was pH independent in the range from 4 to 6, because of its buffering properties. (author)

  3. Growth of Pinus radiada in soil containing solid waste from the kraft pulp industry

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, M.; Vicuna, R.; Gonzalez, B.; Bronfman, M. [Pontificia Universidad Catolica de Chile, Facultad de Ciencias Biologicas, Santiago (Chile); Osses, M. [Celulosa Arauco y Constitucion, Arauco (Chile); Toro, J.; Balocchi, C.; Rodriguez, E. [Bioforest, S.A, Concepcion (Chile)

    2000-06-01

    The germination and growth of Pinus radials Don. plantlets in solid residues deriving from a Kraft pulp industry was evaluated. Plant conditions were monitored by histological studies of roots and shoot-tips, as well as by plant analyses of several essential and non essential elements. The solids employed consisted of ashes, fly-ashes, dregs, grits, primary sludge, brown stock screening rejects and various mixtures of them. Their addition, in a range of combinations to sandy/metamorphic or marine terrace/clay soils, resulted in effective and sustained growth under greenhouse conditions. Low proportions of wastes favored growth in most cases, indicating that they may act as fertilisers. In some experiments, especially in those where waste was added in proportions ranging from 50% to 60%, germination and/or development were slightly affected. Two-year old field experiments have confirmed that in spite of the high pH values, Na ion content or elevated water retention capacity exhibited by some of the solids tested, their use is beneficial for the growth of radiate pine. To date, we have not observed negative effects other than growth inhibition when some solids are present at concentrations above 60%. Our preliminary results suggest that an adequate use as fertiliser of solid waste from the Kraft pulp industry may constitute a profitable alternative in its management. (orig.)

  4. Blast furnace slag-cement grout blends for the immobilization of technetium-containing wastes

    International Nuclear Information System (INIS)

    Mixed low-level radioactive and chemically toxic process treatment wastes from the Portsmouth Gaseous Diffusion Plant are stabilized by solidification in cement-based grouts. Conventional portland cement and fly ash grouts are shown to be very effective for retention of hydrolyzable heavy metals (including lead, cadmium, uranium, and nickel), but are marginally acceptable for retention of radioactive 99Tc (which is present in the waste as the highly mobile pertechnate anion). Addition of ground blast furnace slag to the grout is shown to reduce the effective diffusivity of technetium by several orders of magnitude; retention of technetium is improved by decreasing the waste loading in the grout or by increasing the proportion of blast furnace slag in the grout dry mix. The selective effect of slag is believed to be due to its ability to reduce Tc(VIII) to the less soluble Tc(IV) species. The addition of other reductive grout admixtures (e.g., sodium sulfide, ferrous ion, and powdered iron metal) also appear to improve the retention of technetium in grout. 31 refs., 2 figs., 25 tabs

  5. Utilization of different waste proteins to create a novel PGPR-containing bio-organic fertilizer

    Science.gov (United States)

    Huang, Yan; Sun, Li; Zhao, Jianshu; Huang, Rong; Li, Rong; Shen, Qirong

    2015-01-01

    High-quality bio-organic fertilizers (BIOs) cannot be produced without the addition of some proteins, while many waste proteins are haphazardly disposed, causing serious environmental pollution. In this study, several waste proteins were used as additives to assist with the reproduction of the functional microbe (Bacillus amyloliquefaciens SQR9) inoculated into matured composts to produce BIOs. An optimized composition of solid-state fermentation (SSF) raw materials was predicted by response surface methodology and experimental validation. The results showed that 7.61% (w/w, DW, the same below) rapeseed meal, 8.85% expanded feather meal, 6.47% dewatered blue algal sludge and 77.07% chicken compost resulted in maximum biomass of strain SQR-9 and the maximum amount of lipopeptides 7 days after SSF. Spectroscopy experiments showed that the inner material structural changes in the novel SSF differed from the control and the novel BIO had higher dissolved organic matter. This study offers a high value-added utilization of waste proteins for producing economical but high-quality BIO.

  6. Closure development for high-level nuclear waste containers for the tuff repository

    International Nuclear Information System (INIS)

    This report summarizes Phase 1 activities for closure development of the high-level nuclear waste package task for the tuff repository. Work was conducted under U.S. Department of Energy (DOE) Contract 9172105, administered through the Lawrence Livermore National Laboratory (LLNL), as part of the Yucca Mountain Project (YMP), funded through the DOE Office of Civilian Radioactive Waste Management (OCRWM). The goal of this phase was to select five closure processes for further evaluation in later phases of the program. A decision tree methodology was utilized to perform an objective evaluation of 15 potential closure processes. Information was gathered via a literature survey, industrial contacts, and discussions with project team members, other experts in the field, and the LLNL waste package task staff. The five processes selected were friction welding, electron beam welding, laser beam welding, gas tungsten arc welding, and plasma arc welding. These are felt to represent the best combination of weldment material properties and process performance in a remote, radioactive environment. Conceptual designs have been generated for these processes to illustrate how they would be implemented in practice. Homopolar resistance welding was included in the Phase 1 analysis, and developments in this process will be monitored via literature in Phases 2 and 3. Work was conducted in accordance with the YMP Quality Assurance Program. 223 refs., 20 figs., 9 tabs

  7. Development of integraded mechanistically-based degradation-mode models for performance assessment of high-level waste containers

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, J. C., LLNL

    1998-06-01

    A key component of the Engineered Barrier System (EBS) being designed for containment of spent-fuel and high-level waste at the proposed geological repository at Yucca Mountain, Nevada is a two-tayer canister. In this particular design, the inner barrier is made of a corrosion resistant material (CRM) such as Alloy 825, 625 or C-22, while the outer barrier is made of a corrosion-allowance material (CAM) such as A516 Gr 55 or Monel 400. At the present time, Alloy C- 22 and A516 Gr 55 are favored.

  8. Containment at the Source during Waste Volume Reduction of Large Radioactive Components Using Oxylance High-Temperature Cutting Equipment - 13595

    Energy Technology Data Exchange (ETDEWEB)

    Keeney, G. Neil [Health Physicist, HazMat CATS, LLC (United States)

    2013-07-01

    As a waste-volume reduction and management technique, highly contaminated Control Element Drive Mechanism (CEDM) housings were severed from the Reactor Pressure Vessel Head (RPVH) inside the San Onofre Unit 2 primary containment utilizing Oxylance high-temperature cutting equipment and techniques. Presented are relevant data concerning: - Radiological profiles of the RPVH and individual CEDMs; - Design overviews of the engineering controls and the specialized confinement housings; - Utilization of specialized shielding; - Observations of apparent metallurgical-contamination coalescence phenomena at high temperatures resulting in positive control over loose-surface contamination conditions; - General results of radiological and industrial hygiene air sampling and monitoring; - Collective dose and personnel contamination event statistics; - Lessons learned. (author)

  9. Mechanical behavior of the asbestos-cement container for geological disposal of α level technological wastes from COGEMA reprocessing plants

    International Nuclear Information System (INIS)

    For the safety assessment of the SGN asbestos cement container concept selected by COGEMA for the conditioning of cemented technological wastes from the UP3-UP2 800 reprocessing plants, a general survey has been carried out to confirm both its confinement capacity and its mechanical strength. This safety assessment relates to the latter aspect. It implies two stages: first, the material characterization of asbestos cement and epoxide resin used in sealing and assembling; second, the finite element calculation of induced stresses and strains under storage conditions with regards to the experimented mechanical characteristics. The authors infer some damage in packaging materials in case of misoperation in conditioning process

  10. Mechanical Behaviour of the asbestos-cement container for geological disposal of α level technological wastes from Cogema reprocessing plants

    International Nuclear Information System (INIS)

    For the safety assessment of the SGN asbestos cement container concept selected by COGEMA for the conditionning of cemented α technological wastes from the UP3-UP2 800 reprocessing plants, a general survey has been carried out to confirm both its confinement capacity and its mechanical strength. This safety assessment relates to the latter aspect. It implies two stages: first, the material characterization of asbestos cement and epoxide resin used in sealing and assembling; second, the finite element calculation of induced stresses and strains under storage conditions with regards to the experimented mechanical characteristics. We infer some damage in packaging materials in case of misoperation in conditionning process

  11. 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.

  12. 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.

  13. Fire-resistance, physical, and mechanical characterization of particleboard containing Oceanic Posidonia waste

    Directory of Open Access Journals (Sweden)

    Saval, J. M.

    2014-06-01

    Full Text Available In this work, particleboards manufactured with Oceanic Posidonia waste and bonded with cement are investigated. The particleboards are made with 3/1.5/0.5 parts of cement per part of Posidonia waste. The physical properties of bulk density, swelling, surface absorption, and dimensional changes due to relative humidity as well as the mechanical properties of modulus of elasticity, bending strength, surface soundness, perpendicular tensile strength and impact resistance are studied. In terms of the above properties, the best results were obtained for particleboards with high cement content and when the waste “leaves” are treated (crushed before board fabrication, due to internal changes to the board structure under these conditions. Based on the results of fire tests, the particleboard is non-flammable without any fire-resistant treatment.En esta investigación se han diseñado y fabricado tableros con residuo de Posidonia Oceánica y cemento. Los tableros se han fabricado con 3/1.5/0.5 partes de cemento por cada parte de Posidonia estudiándose sus propiedades físicas (densidad, hinchazón, absorción superficial, variaciones dimensionales por humedad y mecánicas (módulo de elasticidad, resistencia a flexión, al arranque de superficie, al arranque de tornillo, a la tracción perpendicular y al choque. Se observa una mejora de los resultados de resistencia mecánica con el incremento de la cantidad de cemento y si la hoja del residuo es previamente tratada ya que proporciona una mejor estructura interna en el tablero. Además, tras los ensayos de reacción al fuego, se observa que el material es no inflamable sin ningún tipo de tratamiento ignifugante.

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

    International Nuclear Information System (INIS)

    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. Improvement of Murrah Buffalo Milk Production Fed Palm Oil Solid Waste Containing Ration

    OpenAIRE

    P. Mahyuddin

    2010-01-01

    A field trial was conducted to study the effect of dietary inclusion of palm oil solid waste on milk production of murrah buffalo raised under palm oil plantation. Two farms from different districts were involved in this study. Forty cows with 7–9 month pregnancy were selected from each farm and they were divided into control and treatment groups. Cows in control group were offered a mixed supplement of 1 kg copra meal + 2 kg fresh grated cassava root + mineral mix and treatments group wer...

  16. Effect and Removal Mechanisms of 6 Different Washing Agents for Building Wastes Containing Chromium

    OpenAIRE

    Wang Xing-run; Zhang Yan-xia; Wang Qi; Shu Jian-min

    2012-01-01

    With the building wastes contaminated by chromium in Haibei Chemical Plan in China as objects, we studied the contents of total Cr and Cr (VI) of different sizes, analyzed the effect of 6 different washing agents, discussed the removal mechanisms of 6 different washing agents for Cr in various forms, and finally selected applicable washing agent. As per the results, particle size had little impact on the contents of total Cr and Cr (VI); after one washing with water, the removal rate of total...

  17. Aluminothermic reduction of Cr2O3 contained in the ash of thermally treated leather waste

    OpenAIRE

    B. M. Wenzel; T. H. Zimmer; C. S. Fernandez; N. R. Marcilio; Godinho, M.

    2013-01-01

    In this study the viability of utilising ashes with high chromium oxide content, obtained by thermal treatment of footwear leather waste, in the production of low-carbon ferrochromium alloy (Fe-Cr-LC) by aluminothermic reduction was investigated. The following key-factors were selected for process modelling: the quantity of aluminium (Al) employed in the reaction, the iron amount added, the iron compound (Fe and/or Fe2O3) used, and the chromic acid addition. The process was investigated using...

  18. A reactive distillation process for the treatment of LiCl-KCl eutectic waste salt containing rare earth chlorides

    Science.gov (United States)

    Eun, H. C.; Choi, J. H.; Kim, N. Y.; Lee, T. K.; Han, S. Y.; Lee, K. R.; Park, H. S.; Ahn, D. H.

    2016-11-01

    The pyrochemical process, which recovers useful resources (U/TRU metals) from used nuclear fuel using an electrochemical method, generates LiCl-KCl eutectic waste salt containing radioactive rare earth chlorides (RECl3). It is necessary to develop a simple process for the treatment of LiCl-KCl eutectic waste salt in a hot-cell facility. For this reason, a reactive distillation process using a chemical agent was achieved as a method to separate rare earths from the LiCl-KCl waste salt. Before conducting the reactive distillation, thermodynamic equilibrium behaviors of the reactions between rare earth (Nd, La, Ce, Pr) chlorides and the chemical agent (K2CO3) were predicted using software. The addition of the chemical agent was determined to separate the rare earth chlorides into an oxide form using these equilibrium results. In the reactive distillation test, the rare earth chlorides in LiCl-KCl eutectic salt were decontaminated at a decontamination factor (DF) of more than 5000, and were mainly converted into oxide (Nd2O3, CeO2, La2O3, Pr2O3) or oxychloride (LaOCl, PrOCl) forms. The LiCl-KCl was purified into a form with a very low concentration (<1 ppm) for the rare earth chlorides.

  19. Analysis of factors influencing the reliability of retrievable storage canisters for containment of solid high-level radioactive waste

    International Nuclear Information System (INIS)

    The reliability of stainless steel type 304L canisters for the containment of solidified high-level radioactive wastes in the glass and calcine forms was studied. A reference system, drawn largely from information furnished by Battelle Northwest Laboratories and Atlantic Richfield Hanford Company is described. Operations include filling the canister with the appropriate waste form, interim storage at a reprocessing plant, shipment in water to a Retrievable Surface Storage Facility (RSSF), interim storage at the RSSF, and shipment to a final disposal facility. The properties of stainless steel type 304L, fission product oxides, calcine, and glass were reviewed, and mechanisms of corrosion were identified and studied. The modes of corrosion important for reliability were stress-corrosion cracking, internal pressurization of the canister by residual impurities present, intergranular attack at the waste-canister interface, and potential local effects due to migration of fission products. The key role of temperature control throughout canister lifetime is considered together with interactive effects. Methods of ameliorating adverse effects and ensuring high reliability are identified and described. Conclusions and recommendations are presented

  20. Pretreatment of Tc-Containing Waste and Its Effect on Tc-99 Leaching From Grouts

    International Nuclear Information System (INIS)

    A salt solution (doped with Tc-99), that simulates the salt waste stream to be processed at the Saltstone Production Facility, was immobilized in grout waste forms with and without (1) ground granulated blast furnace slag and (2) pretreatment with iron salts. The degree of immobilization of Tc-99 was measured through monolithic and crushed grout leaching tests. Although Fe (+2) was shown to be effective in reducing Tc-99 to the +4 state, the strong reducing nature of the blast furnace slag present in the grout formulation dominated the reduction of Tc-99 in the cured grouts. An effective diffusion coefficient of 4.75 x 10-12 (Leach Index of 11.4) was measured using the ANSI/ANS-16.1 protocol. The leaching results show that, even in the presence of a concentrated salt solution, blast furnace slag can effectively reduce pertechnetate to the immobile +4 oxidation state. The measured diffusivity was introduced into a flow and transport model (PORFLOW) to calculate the release of Tc-99 from a Saltstone Vault as a function of hydraulic conductivity of the matrix. (authors)

  1. Corrosion susceptibility of steel drums to be used as containers for intermediate level nuclear waste

    Directory of Open Access Journals (Sweden)

    Duffó G.

    2013-07-01

    Full Text Available The present work is a study of the corrosion susceptibility of steel drums in contact with cemented ion-exchange resins contaminated with different types and concentrations of aggressive species. A special type of specimen was manufactured to simulate the cemented ion-exchange resins in the drum. The evolution of the corrosion potential and the corrosion rate of the steel, as well as the electrical resistivity of the matrix were monitored over a time period of 900 days. The aggressive species studied were chloride ions (the main ionic species of concern and sulphate ions (produced during radiolysis of the cationic exchange-resins after cementation. The work was complemented with an analysis of the corrosion products formed on the steel in each condition, as well as the morphology of the corrosion products. When applying the results obtained in the present work to estimate the corrosion depth of the steel drumscontaining the cemented radioactive waste after a period of 300 years (foreseen durability of the Intermediate Level Radioactive Waste facility in Argentina , it is found that in the most unfavourable case (high chloride contamination, the corrosion penetration will be considerably lower than the thickness of the wall of the steel drums.

  2. Environmental performance and mechanical analysis of concrete containing recycled asphalt pavement (RAP) and waste precast concrete as aggregate.

    Science.gov (United States)

    Erdem, Savaş; Blankson, Marva Angela

    2014-01-15

    The overall objective of this research project was to investigate the feasibility of incorporating 100% recycled aggregates, either waste precast concrete or waste asphalt planning, as replacements for virgin aggregates in structural concrete and to determine the mechanical and environmental performance of concrete containing these aggregates. Four different types of concrete mixtures were designed with the same total water cement ratio (w/c=0.74) either by using natural aggregate as reference or by totally replacing the natural aggregate with recycled material. Ground granulated blast furnace slag (GGBS) was used as a mineral addition (35%) in all mixtures. The test results showed that it is possible to obtain satisfactory performance for strength characteristics of concrete containing recycled aggregates, if these aggregates are sourced from old precast concrete. However, from the perspective of the mechanical properties, the test results indicated that concrete with RAP aggregate cannot be used for structural applications. In terms of leaching, the results also showed that the environmental behaviour of the recycled aggregate concrete is similar to that of the natural aggregate concrete. PMID:24316812

  3. Stabilization of ZnCl2-containing wastes using calcium sulfoaluminate cement: cement hydration, strength development and volume stability.

    Science.gov (United States)

    Berger, Stéphane; Cau Dit Coumes, Céline; Le Bescop, Patrick; Damidot, Denis

    2011-10-30

    The potential of calcium sulfoaluminate (CSA) cement was investigated to solidify and stabilize wastes containing large amounts of soluble zinc chloride (a strong inhibitor of Portland cement hydration). Hydration of pastes and mortars prepared with a 0.5 mol/L ZnCl(2) mixing solution was characterized over one year as a function of the gypsum content of the binder and the thermal history of the material. Blending the CSA clinker with 20% gypsum enabled its rapid hydration, with only very small delay compared with a reference prepared with pure water. It also improved the compressive strength of the hardened material and significantly reduced its expansion under wet curing. Moreover, the hydrates assemblage was less affected by a thermal treatment at early age simulating the temperature rise and fall occurring in a large-volume drum of cemented waste. Fully hydrated materials contained ettringite, amorphous aluminum hydroxide, strätlingite, together with AFm phases (Kuzel's salt associated with monosulfoaluminate or Friedel's salt depending on the gypsum content of the binder), and possibly C-(A)-S-H. Zinc was readily insolubilized and could not be detected in the pore solution extracted from cement pastes. PMID:21889260

  4. Physical, Chemical and Structural Evolution of Zeolite-Containing Waste Forms Produced from Metakaolinite and Calcined Sodium Bearing Waste (HLW and/or LLW)

    Energy Technology Data Exchange (ETDEWEB)

    Grutzeck, Michael W.

    2005-06-27

    Zeolites are extremely versatile. They can adsorb liquids and gases and serve as cation exchange media. They occur in nature as well cemented deposits. The ancient Romans used blocks of zeolitized tuff as a building material. Using zeolites for the management of radioactive waste is not a new idea, but a process by which the zeolites can be made to act as a cementing agent is. Zeolitic materials are relatively easy to synthesize from a wide range of both natural and man-made substances. The process under study is derived from a well known method in which metakaolin (an impure thermally dehydroxylated kaolinite heated to {approx}700 C containing traces of quartz and mica) is mixed with sodium hydroxide (NaOH) and reacted in slurry form (for a day or two) at mildly elevated temperatures. The zeolites form as finely divided powders containing micrometer ({micro}m) sized crystals. However, if the process is changed slightly and only just enough concentrated sodium hydroxide solution is added to the metakaolinite to make a thick crumbly paste and then the paste is compacted and cured under mild hydrothermal conditions (60-200 C), the mixture will form a hard ceramic-like material containing distinct crystalline tectosilicate minerals (zeolites and feldspathoids) imbedded in an X-ray amorphous hydrated sodium aluminosilicate matrix. Due to its lack of porosity and vitreous appearance we have chosen to call this composite a ''hydroceramic''.

  5. Microbial studies in the Canadian nuclear fuel waste management program

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) has developed a concept for permanent geological disposal of nuclear fuel waste in Canada. An accelerated program was initiated in 1991 to address and quantify the potential effects of microbial action on the integrity of the disposal concept's multiple barrier system. This microbial program focuses on answering specific questions in areas such as the survival of bacteria in compacted clay-based buffer materials under relevant radiation and desiccation conditions; mobility of microbes in compacted buffer materials; the potential for microbially-influenced corrosion of containers; microbial gas production in backfill material; introduction of nutrients as a result of vault excavation and operation; the presence and activity of microbes in deep granitic groundwaters; and the effects of biofilms on radionuclide migration in the geosphere. This paper summarizes the current research activities at AECL in these areas. (author)

  6. [Removal of Waste Gas Containing Mixed Chlorinated Hydrocarbons by the Biotrickling Filter].

    Science.gov (United States)

    Chen, Dong-zhi; Miao, Xiao-ping; Ouyang, Du-juan; Ye, Jie-xu; Chen, Jian-meng

    2015-09-01

    An experimental investigation on purification of waste gas contaminated with a mixture of dichloromethane (DCM) and dichloroethane(1,2-DCA) was conducted in a biotrickling filter (BTF) inoculated with activated sludge of pharmaceuticals industry. Stable removal efficiency(RE) above 80% for DCM and above 75% for 1,2-DCA were achieved after 35 days, indicating that biofilm was developed. The best elimination capacity (EC) of DCM and 1,2-DCA were 13 g.(m3.h)-1 and 10 g.(m3.h)-1 respectively. And there was a linear relationship between the production of CO2 and mixed gas EC, the maximum mineralization rate of mixed gas stabled at 61. 2%. The interaction test indicated that DCM and 1,2-DCA would inhibit with each other. The changing of biomass of BTF during the operation process was also been studied. PMID:26717675

  7. Development of a Lightweight Low-Carbon Footprint Concrete Containing Recycled Waste Materials

    Directory of Open Access Journals (Sweden)

    S. Talukdar

    2011-01-01

    This study examined the use of waste materials such as crushed glass, ground tire rubber, and recycled aggregate in concrete. Compressive strength and elastic modulus were the primary parameters of interest. Results demonstrated that ground tire rubber introduced significant amounts of air into the mix and adversely affected the strength. The introduction of a defoamer was able to successfully remove part of the excess air from the mix, but the proportional strength improvements were not noted implying that air left in the defoamed mixture had undesirable characteristics. Freeze-thaw tests were next performed to understand the nature of air in the defoamed mixtures, and results demonstrated that this air is not helpful in resisting freeze-thaw resistance either. Overall, while lightweight, low-carbon footprint concrete materials seem possible from recycled materials, significant further optimization remains possible.

  8. Analysis of the optimization possibilities to recover the powdery wastes containing iron and carbon

    Science.gov (United States)

    Popescu, Darius-Alexandru; Vilceanu, Lucia; Socalici, Ana

    2016-06-01

    Most industrial activities result in one or more secondary products and wastes besides the primary product, with a variety of uses. The iron & steel industry is highly energy intensive, but it is also a major source of environmental pollution with gases and dusts, especially the extractive branch. The researches aimed the recovery of the dust from the sintering plants and blast furnaces through the briquetting technology. Its recovery is required either for preventing the pollution or for reducing the consumption of raw materials. The mechanical properties are important for the quality of briquettes. We presented in this paper a series of mathematical correlations among the mechanical properties and the components of the briquetting batch, obtained using Excel spreadsheet and MATLAB programs. After analysing the results, we choose the optimal variation limits for the briquetting batch components.

  9. Loading, moving, and shipping radioactive waste in reusable radioactive material containers

    International Nuclear Information System (INIS)

    While the dismantlement of systems and components at the Shoreham Nuclear Power Plant was a monumental task, the loading, movement, temporary storage, and shipping of over 2 1/2 million pounds of contaminated and/or activated material was nearly as difficult. Close coordination and teamwork between such diverse groups as craft labor, health physics, radiation controls, trucking companies and waste volume reducers were crucial elements in performing this work safely, cost effectively, and with particular attention to the station's very aggressive ALARA (As Low As Reasonably Achievable) goals. This paper discusses the actual work that was involved from the time the contaminated component was removed from its location in the plant through actual shipment offsite

  10. Cold gap grout formulation for waste containment at DOE site, Hanford, Washington

    International Nuclear Information System (INIS)

    This paper reports that WES developed a grout to be used as a cold (non-radioactive) cap or void-fill material between the solidified low-level waste and the cover blocks of near-surface disposal vaults at the U.S. Department of Energy (DOE) Hanford Facility. The project consisted of formulation and evaluation of candidate grout, followed by a physical scale-model test to verify grout performance under project-specific conditions and provide data to verify numerical models of stresses and isotherms inside the Hanford demonstration vault. Evaluation of unhardened grout included segregation, bleed, flow, and working time. For hardened grout, strength, volume stability, thermal heat rise, and geochemical compatibility with surrogate wasteform grout were examined

  11. Physical, Chemical and Structural Evolution of Zeolite - Containing Waste Forms Produced from Metakaolinite and Calcined HLW

    Energy Technology Data Exchange (ETDEWEB)

    Grutzeck, Michael

    2005-06-01

    During the seventh year of the current grant (DE-FG02-05ER63966) we completed an exhaustive study of cold calcination and began work on the development of tank fill materials to fill empty tanks and control residuals. Cold calcination of low and high NOx low activity waste (LAW) SRS Tank 44 and Hanford AN-107 simulants, respectively with metallic Al + Si powders was evaluated. It was found that a combination of Al and Si powders could be used as reducing agents to reduce the nitrate and nitrite content of both low and high NOx LAW to low enough levels to allow the LAW to be solidified directly by mixing it with metakaolin and allowing it to cure at 90 C. During room temperature reactions, NOx was reduced and nitrogen was emitted as N2 or NH3. This was an important finding because now one can pretreat LAW at ambient temperatures which provides a low-temperature alternative to thermal calcination. The significant advantage of using Al and Si metals for denitration/denitrition of the LAW is the fact that the supernate could potentially be treated in situ in the waste tanks themselves. Tank fill materials based upon a hydroceramic binder have been formulated from mixtures of metakaolinite, Class F fly ash and Class C flue gas desulphurization (FGD) ash mixed with various concentrations of NaOH solution. These harden over a period of hours or days depending on composition. A systematic study of properties of the tank fill materials (leachability) and ability to adsorb and hold residuals is under way.

  12. Physical, Chemical and Structural Evolution of Zeolite - Containing Waste Forms Produced from Metakaolinite and Calcined HLW

    International Nuclear Information System (INIS)

    During the seventh year of the current grant (DE-FG02-05ER63966) we completed an exhaustive study of cold calcination and began work on the development of tank fill materials to fill empty tanks and control residuals. Cold calcination of low and high NOx low activity waste (LAW) SRS Tank 44 and Hanford AN-107 simulants, respectively with metallic Al + Si powders was evaluated. It was found that a combination of Al and Si powders could be used as reducing agents to reduce the nitrate and nitrite content of both low and high NOx LAW to low enough levels to allow the LAW to be solidified directly by mixing it with metakaolin and allowing it to cure at 90 C. During room temperature reactions, NOx was reduced and nitrogen was emitted as N2 or NH3. This was an important finding because now one can pretreat LAW at ambient temperatures which provides a low-temperature alternative to thermal calcination. The significant advantage of using Al and Si metals for denitration/denitrition of the LAW is the fact that the supernate could potentially be treated in situ in the waste tanks themselves. Tank fill materials based upon a hydroceramic binder have been formulated from mixtures of metakaolinite, Class F fly ash and Class C flue gas desulphurization (FGD) ash mixed with various concentrations of NaOH solution. These harden over a period of hours or days depending on composition. A systematic study of properties of the tank fill materials (leachability) and ability to adsorb and hold residuals is under way

  13. New box container system for waste drums: dynamic tests and qualification

    International Nuclear Information System (INIS)

    A first technical report is presented on a new 20' box container, designed by the firm CORROBESCH/STM as a Type A package for the transport of radioactive and other dangerous materials, and having a carrying capacity of 22 tons. The container itself weighs only 4 tons, and it incorporates a proprietary corrosion finish that is highly resistant to mechanical wear, deformation, and radioactive contamination. The Type A Package, which also can be used as industrial package Type 2 and Type 3, is designed to withstand accelerations of up to 6g. This design criterion was established based on the European Railroad Associations requirement of considering dynamic loads of 4g arising during routine transport times a safety factor of 1.5. In contrast, the ISO Norm 1496, part 1, does not require explicit consideration of dynamic loads but only requires a static load test. Therefore, the main motivation behind the dynamic load design criteria was the complete lack of freight containers capable of withstanding dynamic loads of up to 6g. The container was subjected in Germany to a series of collision and drop tests, as specified by the International Atomic Energy Agency (IAEA), and it passed these tests to complete satisfaction. As a result, the container received certification by agencies such as the Germanischer Lloyd (GL), the Deutsche Bahn AG (DB), and the Bundesanstalt fuer Materialpruefung (BAM). At the same time, the authors have also developed a mathematical model of the container to predict its dynamic behaviour during service loads, and have been able to make motion predictions that are in good agreement with signatures recorded during tests. (author)

  14. POTENTIAL FOR STRESS CORROSION CRACKING OF A537 CARBON STEEL NUCLEAR WASTE TANKS CONTAINING HIGHLY CAUSTIC SOLUTIONS

    Energy Technology Data Exchange (ETDEWEB)

    Lam, P.; Stripling, C.; Fisher, D.; Elder, J.

    2010-04-26

    The evaporator recycle streams of nuclear waste tanks may contain waste in a chemistry and temperature regime that exceeds the current corrosion control program, which imposes temperature limits to mitigate caustic stress corrosion cracking (CSCC). A review of the recent service history found that two of these A537 carbon steel tanks were operated in highly concentrated hydroxide solution at high temperature. Visual inspections, experimental testing, and a review of the tank service history have shown that CSCC has occurred in uncooled/un-stress relieved tanks of similar construction. Therefore, it appears that the efficacy of stress relief of welding residual stress is the primary corrosion-limiting mechanism. The objective of this experimental program is to test A537 carbon steel small scale welded U-bend specimens and large welded plates (30.48 x 30.38 x 2.54 cm) in a caustic solution with upper bound chemistry (12 M hydroxide and 1 M each of nitrate, nitrite, and aluminate) and temperature (125 C). These conditions simulate worst-case situations in these nuclear waste tanks. Both as-welded and stress-relieved specimens have been tested. No evidence of stress corrosion cracking was found in the U-bend specimens after 21 days of testing. The large plate test was completed after 12 weeks of immersion in a similar solution at 125 C except that the aluminate concentration was reduced to 0.3 M. Visual inspection of the plate revealed that stress corrosion cracking had not initiated from the machined crack tips in the weld or in the heat affected zone. NDE ultrasonic testing also confirmed subsurface cracking did not occur. Based on these results, it can be concluded that the environmental condition of these tests was unable to develop stress corrosion cracking within the test periods for the small welded U-bends and for the large plates, which were welded with an identical procedure as used in the construction of the actual nuclear waste tanks in the 1960s. The

  15. Isolating 241Am from waste solutions containing Al, Ca, Fe, and Cr

    International Nuclear Information System (INIS)

    About 2.4 kg of 241Am contaminated with calcium and aluminum had been recovered from low-activity waste during recycle of 11% 240Pu. A process was developed and demonstrated to purify the americium before shipment as 241AmO2. The americium and some of the calcium were batch extracted into 50% TBP-n-paraffin from 2.2M Al(NO3)3 - 0.3M HNO3 solution in a canyon tank. Pregnant solvent was scrubbed first with 2.1M Al3+-0.3M Li+-6.7M NO3- and then with 7M LiNO3 to reduce the calcium content and to displace the aluminum. Americium was then stripped from the solvent with water and concentrated by evaporation. Before precipitating the americium with oxalic acid, the nitric acid was adjusted with NH4OH to yield a 1M NH4NO3 solution. Recovery across the batch extraction step was 97.8%, while 93% of the calcium and >99% of the aluminum was rejected. Recovery across precipitation averaged >96% while producing a product which was >99.3% pure 241AmO2. The major impurities were water, carbon, calcium, iron, and zinc

  16. Pressurized grout applications in fractured tuff for containment of radioactive wastes

    International Nuclear Information System (INIS)

    Currently under study by the Department of Energy are the geologic and hydrologic characteristics of the ash-flow deposits under Yucca mountain at the Nevada test site. Of interest at this site is the potential for disposal of high-level radioactive wastes in the unsaturated zone of the densely welded portions of the tuffs. These studies include the performance-assessment of barriers and seals for boreholes, ramps, drifts and shafts at the Yucca mountain site. In-situ tests on standard Type II Portland cement and microfine cement as grout materials have been performed on a similar rock type to Yucca Mountain's near Superior Arizona. The tests were performed in a vertical borehole drilled in highly fractured and densely welded tuff (brown unit of Apache Leap) through a series of pressurized grout applications. Packer flow tests prior to and after each grout application measure the effectiveness of the grout application in reducing the permeability of the rock surrounding the borehole. Overall the grout applications have reduced the permeability of the test hole by three orders of magnitude. (author)

  17. Aluminothermic reduction of Cr2O3 contained in the ash of thermally treated leather waste

    Directory of Open Access Journals (Sweden)

    B. M. Wenzel

    2013-03-01

    Full Text Available In this study the viability of utilising ashes with high chromium oxide content, obtained by thermal treatment of footwear leather waste, in the production of low-carbon ferrochromium alloy (Fe-Cr-LC by aluminothermic reduction was investigated. The following key-factors were selected for process modelling: the quantity of aluminium (Al employed in the reaction, the iron amount added, the iron compound (Fe and/or Fe2O3 used, and the chromic acid addition. The process was investigated using a 2(4 full factorial design where the percentage of Cr2O3 reduced was used as the response. Variance analysis was employed to determine the significant effects and to validate the obtained model. The model was useful for finding the optimal operating conditions, including the maximisation of chromium conversion and the gross margin. Both resulted in similar process conditions, with 76.8±12.3% of chromium being reduced to the metallic phase, and 1.65±0.52 USD (kg ash-1 as the gross margin. The qualities of some alloys obtained were investigated by scanning electron microscopy coupled with energy dispersive X-ray spectroscopy analysis (SEM/EDS. The results showed that the main problem for these alloys in a standard specification was the P and S content, suggesting that a pre-treatment is required.

  18. Production of ultrafine zinc powder from wastes containing zinc by electrowinning in alkaline solution

    Directory of Open Access Journals (Sweden)

    Zhao Youcai

    2013-12-01

    Full Text Available Production of ultrafine zinc powder from industrial wastes by electrowinning in alkaline solution was studied. Stainless steel and magnesium electrodes were used as anode and cathode, respectively. Morphology, size distribution and composition of the Zn particles were characterized by Scanning Electron Microscopy, Laser Particle Size Analyzer, and Inductive Coupled Plasma Emission Spectrometer. The required composition of the electrolyte for ultrafine particles was found to be 25-35 g/L Zn, 200-220 g/L NaOH and 20-40 mg/L Pb. The optimal conditions were a current density of 1000-1200 A/m² and an electrolyte temperature of 30-40 °C. The results indicated that the lead additive exerted a beneficial effect on the refining of the particles, by increasing the cathodic polarization. Through this study, ultrafine zinc powder with a size distribution of around 10 μm could be produced, and considerably high current efficiencies (97-99 % were obtained.

  19. Study of the impact behaviour of packages containing intermediate level radioactive waste coming from nuclear installations

    International Nuclear Information System (INIS)

    The following describes primarily an experimental study into the benefits, for impact resistance, to be gained by incorporating a welded lid into the design of the cement filled drum type of intermediate level waste package. Tests on packages which were not provided with a lid showed that matrix material began to be expelled from drop heights of about 16m. This damage threshold was similar for packages composed of both high and low strength matrix. Above the damage threshold, however, the rate of increase of expelled mass with drop height was greater for the packages filled with a low strength matrix. Similar tests were conducted with specimens to which a lid had been attached by welding. Even from the greatest drop height available at the test facility (28m) only one package showed a significant amount of drum tearing but even then little matrix was lost. The benefits of incorporating a welded lid into package design were thus clearly established. Simple calculations were performed to predict the local deformations and deceleration/time histories of the packages. By optimisation of the impact resistive stress used in the computer model, final knockback areas were predicted to an accuracy of 30%. The average deceleration predicted for four of the six tests for which deceleration histories were available were also within 30% of measured values

  20. Study of the reuse of treated wastewater on waste container washing vehicles.

    Science.gov (United States)

    Vaccari, Mentore; Gialdini, Francesca; Collivignarelli, Carlo

    2013-02-01

    The wheelie bins for the collection of municipal solid waste (MSW) shall be periodically washed. This operation is usually carried out by specific vehicles which consume about 5000 L of water per day. Wastewater derived from bins washing is usually stored on the same vehicle and then discharged and treated in a municipal WWTP. This paper presents a study performed to evaluate the reuse of the wastewater collected from bins washing after it has been treated in a small plant mounted on the vehicle; the advantage of such a system would be the reduction of both vehicle dimension and water consumption. The main results obtained by coagulation-flocculation tests performed on two wastewater samples are presented. The addition of 2 mL/L of an aqueous solution of aluminum polychloride (18% w/w), about 35 mL/L of an aqueous solution of CaO (4% w/w) and 25 mL/L of an aqueous solution of an anionic polyelectrolyte (1 ‰ w/w) can significantly reduce turbidity and COD in treated water (to about 99% and 42%, respectively); the concomitant increase of UV transmittance at 254 nm (up to 15%) enables UV disinfection application by a series of two ordinary UV lamps. Much higher UV transmittance values (even higher than 80%) can be obtained by dosing powdered activated carbon, which also results in a greater removal of COD. PMID:23142511

  1. Technology obtaining of nitrogen fertilizer from the calcium is containing waste of production of calcium saltpetre

    OpenAIRE

    Власян, Світлана Варужанівна; Шестозуб, Анатолій Борисович; Волошин, Микола Дмитрович

    2013-01-01

    The new technology of obtaining nitrogen fertilizer from calcium-containing sludge of calcium saltpeter production is considered in the paper. The main objective of the research is the development of processing technology of sludge of calcium saltpeter production into alkaline nitrogen fertilizer, analysis of the composition of initial material and finished product, testing of fertilizer by means of vegeta­tive studies and determination of expenditure of drying agent that is exhaust gases of ...

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

    Energy Technology Data Exchange (ETDEWEB)

    Dabbs, Daniel M.; Aksay, I.A.

    2005-12-01

    In the first phase of our study, we focused on the use of simple organics to raise the solubility of aluminum oxyhydroxides in high alkaline aqueous solvents. In a limited survey of common organic acids, we determined that citric acid had the highest potential to achieve our goal. However, our subsequent investigation revealed that the citric acid appeared to play two roles in the solutions: first, raising the concentration of aluminum in highly alkaline solutions by breaking up or inhibiting ''seed'' polycations and thereby delaying the nucleation and growth of particles; and second, stabilizing nanometer-sized particles in suspension when nucleation did occur. The results of this work were recently published in Langmuir: D.M. Dabbs, U. Ramachandran, S. Lu, J. Liu, L.-Q. Wang, I.A. Aksay, ''Inhibition of Aluminum Oxyhydroxide Precipitation with Citric Acid'' Langmuir, 21, 11690-11695 (2005). The second phase of our work involved the solvation of silicon, again in solutions of high alkalinity. Citric acid, due to its unfavorable pKa values, was not expected to be useful with silicon-containing solutions. Here, the use of polyols was determined to be effective in maintaining silicon-containing particles under high pH conditions but at smaller size with respect to standard suspensions of silicon-containing particles. There were a number of difficulties working with highly alkaline silicon-containing solutions, particularly in solutions at or near the saturation limit. Small deviations in pH resulted in particle formation or dissolution in the absence of the organic agents. One of the more significant observations was that the polyols appeared to stabilize small particles of silicon oxyhydroxides across a wider range of pH, albeit this was difficult to quantify due to the instability of the solutions.

  3. Rotating biological contractor treatment of 2-nitrophenol and 2-chlorophenol containing hazardous wastes

    International Nuclear Information System (INIS)

    Rotating Biological Contactors (RBCs) have a number of advantages over other biological treatment systems. For example, they can provide high treatment efficiencies of activated sludge systems with much lower energy inputs. Organic shock loads are handled well because large biomass is present. No bulking, foaming, or floating of sludge occurs and sludge has good settleability and dewaterability. Another advantage of RBC systems is the minimal labor requirement for operation and maintenance. Even though RBC systems have these advantages, their acceptance was slow mainly due to operational problems with the earlier units (such as shaft failures) and the lack of considerable design and operation data. A review of literature shows that there is only limited information available on the wastewater treatment with RBCs. Recently, there has been considerable contributions to the knowledge on RBC technology. However, information on the treatment of organic hazardous wastes using RBCs is still very limited. This paper reports that a considerable number of studies on the biological treatment of organic hazardous compounds was sponsored by U.S. Environmental Protection Agency (EPA). For example, an EPA sponsored study examined the effect of such compounds on the performance of activated sludge process. Bench-scale continuous-flow and batch units were used. Influent was settled municipal wastewater to which toxic compounds were added. In batch operations, 2-chlorophenol and pentachlorophenol caused an increase in the effluent Chemical Oxygen Demand (COD) at an influent concentration of 5 mg/L. No adverse effect of 2-nitrophenol on the batch system was reports. 2-Chlorophenol was one of the compounds that upset the performance of continuous-flow activated sludge units, yielding higher than normal levels of effluent suspended solids

  4. An approach to study the corrosion behaviour of stainless steel containers for packaging of intermediate level radioactive waste during atmospheric storage

    Energy Technology Data Exchange (ETDEWEB)

    Padovani, C.G.; Wood, P. [Nuclear Decommissioning Authority (United Kingdom); Smart, N.R.; Winsley, R.J. [Serco Technical and Assurance Services (United Kingdom); Charles, A.; Albores-Silva, O. [Newcastle upon Tyne Univ. (United Kingdom); Krouse, D. [Industrial Research Limited (New Zealand)

    2009-07-01

    Full text of publication follows: In the UK, intermediate level radioactive waste (ILW) arising from the decommissioning of power stations and other nuclear installations is generally encapsulated in cement waste forms and packaged within stainless steel containers. The function of the waste package is to immobilise and physically contain the waste in a stable form and to allow its safe storage, transport, handling and eventual disposal in a geological disposal facility. Given such a function, it is important to ensure that the corrosion resistance of the waste container is sufficient to ensure its integrity for long times. This paper discusses the expected corrosion behaviour of ILW containers manufactured in stainless steel 304L and 316L within the current disposal concept, with specific focus on the behaviour of the material during atmospheric storage. In an indoor atmosphere, localised corrosion and stress corrosion cracking may develop on waste containers only if aggressive hygroscopic salts (e.g. MgCl{sub 2}) accumulate on the container surfaces in certain quantities and in certain humidity ranges. Experimental observation is being carried out in order to better identify conditions in which corrosion damage develops. This type of analysis, together with laboratory and field observation, is being used to identify suitable storage conditions for the packages. On the other hand, extrapolation of short-term data on pit depth in aggressive environments (e.g. marine atmospheres) suggests that penetration of the container walls by pitting over long-time scales is unlikely. Experimental observation and modelling are progressing in order to better understand the mechanistic aspects of propagation and to evaluate whether container penetration by pitting may occur over long timescales. Outstanding uncertainties (e.g. related to the effect of ionising radiation on the atmospheric corrosion behaviour of the packages) will also be outlined.

  5. The possibilities of the microwave utilization of wastes on the example of materials containing the asbestos

    Directory of Open Access Journals (Sweden)

    M. Pigiel

    2010-04-01

    Full Text Available The presented paper introduce some of the results of the investigations in the utilization of the materials containing asbestos in the existingin Wroclaw University of Technology Institute’s of Technology of Machines and the Automation Foundry and Automation Group themicrowave reactor. In the reactor’s heating chamber there is possible to recycle from 3 up to 5 kg of the batch at once. The temperaturewith which is possible to receive in it is approx. 1400 oC. The time of it’s achievement (in dependence from utilized material can take outfrom 25 up to 40 minutes.

  6. Prediction of corrosion depth of selected materials for the container of high-level wastes under a repository condition

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Soo; Chun, Kwan Sik; Kang, Chul Hyung; Choi, Jong Won; Han, Kyung Won [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-05-01

    The corrosion depth of selected materials for container of high-level wastes in an underground disposal condition was predicted by analyzing the corrosion behaviors and corrosion rates of copper/copper alloys, carbon steel, titanium/titanium alloys, stainless steel and nickel alloys. Their corrosion rates depend on the amount of oxygen and microbes in bentonite at the bore hole, and local corrosion in addition to general corrosion. However, the effect of radiation and the oxygen dissolved in groundwater would be insignificant. To calculate the corrosion depth, it is assumed that the total amount of oxygen contained in the pore and surface of a bentonite block, and in the gaps among container, rock and bentonite block at a borehole is 300 moles. Assuming that all organic compounds in a bentonite block are presumed as lactate, they would produce 2,100 moles of HS-. The corrosion depths were calculated based on the above assumptions and the wall thickness of copper, carbon steel, titanium, stainless steel and nickel alloys of at least 2.6, 25, 1.3, 5 and 0.3 mm would be required for their corrosion allowances that guarantee their desired service life of 1,000 years. 94 refs., 13 figs., 6 tabs. (Author)

  7. Immobilisation of nuclear waste materials containing different alkali elements into single-phase NZP-based ceramics

    Science.gov (United States)

    Pet'kov, V. I.; Orlova, A. I.; Trubach, I. G.; Asabina, Y. A.; Demarin, V. T.; Kurazhkovskaya, V. S.

    2003-01-01

    A single-phase host matrix based upon the sodium zirconium phosphate (NZP) structure and designed to immobilise commercial nuclear waste was investigated. In comparison with other waste forms the important advantage of the NZP ceramics is its ability to incorporate, at crystallographic levels, alkali elements without significant deterioration of the physical and chemical matrix stability. Studies on the incorporation of different alkali elements into the NZP host structure were performed. Single-phase phosphates corresponding to crystalline solutions (continuous and limited) with a structure similar to NZP were found in the series of compounds with the general formula A1-x+4yA'xE2-y(PO4)3 (y=0, 0.5 and 1, and 0≤x≤1+4y), where A-A' are different alkali elements (Li, Na, K, Rb, and Cs) and E are Ti or Zr. Leaching studies with alkali containing samples revealed reasonable resistance towards the release of the constituents.

  8. Effect of microbial action on the corrosion potential of austenitic alloy containers for high-level nuclear waste

    International Nuclear Information System (INIS)

    The safe disposal of high-level nuclear waste (HLW) entails the ability to ensure the integrity of waste containers for prolonged time periods. It is generally accepted that under certain conditions, microbial action may change local benign environments to those in which localized corrosion can be actively promoted. The use of repassivation potential (Erp) in relation to the value of the corrosion potential (Ecorr) has been proposed as a means of assessing the propensity of a metallic material to localized corrosion. Microbial activity is known to influence Ecorr however, the precise mechanism is unresolved. Shewanella putrefaciens, a bacteria with many of the characteristics of sulfate-reducing bacteria (SRB), are being grown under controlled conditions on 316L stainless steel (SS) surfaces to understand the relationship between Ecorr and metabolic activity. It has been observed that the growth of the bacteria under aerobic conditions, without the production of metabolic sulfide, leads to only minor variation in Ecorr. These changes possibly correlate to the periods of active bacterial growth

  9. Effect of microbial action on the corrosion potential of austenitic alloy containers for high-level nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    Angell, P.; Dunn, D.S.; Cragnolino, G.A. [Southwest Research Inst., San Antonio, TX (United States). Center for Nuclear Waste Regulatory Analyses

    1996-08-01

    The safe disposal of high-level nuclear waste (HLW) entails the ability to ensure the integrity of waste containers for prolonged time periods. It is generally accepted that under certain conditions, microbial action may change local benign environments to those in which localized corrosion can be actively promoted. The use of repassivation potential (E{sub rp}) in relation to the value of the corrosion potential (E{sub corr}) has been proposed as a means of assessing the propensity of a metallic material to localized corrosion. Microbial activity is known to influence E{sub corr} however, the precise mechanism is unresolved. Shewanella putrefaciens, a bacteria with many of the characteristics of sulfate-reducing bacteria (SRB), are being grown under controlled conditions on 316L stainless steel (SS) surfaces to understand the relationship between E{sub corr} and metabolic activity. It has been observed that the growth of the bacteria under aerobic conditions, without the production of metabolic sulfide, leads to only minor variation in E{sub corr}. These changes possibly correlate to the periods of active bacterial growth.

  10. Composting of waste paint sludge containing melamine resin and the compost's effect on vegetable growth and soil water quality.

    Science.gov (United States)

    Tian, Yongqiang; Chen, Liming; Gao, Lihong; Michel, Frederick C; Keener, Harold M; Klingman, Michael; Dick, Warren A

    2012-12-01

    Melamine resin (MR) is introduced to the environment from many industrial effluents, including waste paint sludge (WPS) from the automobile industry. Melamine resin contains a high nitrogen (N) content and is a potential N source during composting. In this study, two carbon sources, waste paper (WP) and plant residue (PR), were used to study their effects on composting of WPS. Additional work tested the WPS-composts effects on plant growth and soil water quality. After 84 days of composting, 85% and 54% of the initial MR was degraded in WP- and PR-composts, respectively. The limiting factor was that the MR created clumps during composting so that decomposition was slowed. Compared to the untreated control, both WP- and PR-composts increased growth of cucumber (Cucumis sativus), radish (Raphanus sativus) and lettuce (Lactuca sativa). Concentrations of trace elements in plants and soil water did not rise to a level that would preclude WPS-composts from being used as a soil amendment.

  11. Using Modified Sorbents for Reducing Negative Impact of Oil-Containing Industrial Wastes on Natural and Artificial Waterways

    Directory of Open Access Journals (Sweden)

    Ljudmila Anatolievna Marchenko

    2015-09-01

    Full Text Available The study is aimed at developing and using co-precipitated hydroxides (CPH of aluminum and magnesium as sorbents, in order to reduce the negative impact of oil-containing industrial wastes on natural and artificial waterways. A new method of synthesizing a modified sorbent has been developed, featuring high sorption capacity to a wide range of pollutants in low-acid, neutral, and low-alkaline environments, able to extract complex compounds. A possibility to use this sorbent has been shown, and its sorption capacity has been studied. Sorption parameters have been defined. It has been shown that the value of the maximum achievable concentration efficiency in extracting the said ions on a co-precipitated sorbent is about ten times higher than the corresponding values that characterize sorption on analogous sorbents. In performing analytical studies, standard methods were used, as well as modern methods of physical and chemical analysis: x-ray phase, x-ray fluorescent, atomic absorption, spectral, chemical, thin layer chromatography, and chromato-mass-spectrometry. The specific surface of the samples was defined by the temperature of nitrogen absorption, using the chromatographic method, followed by processing the obtained results using the Brunauer-Emmett-Teller method. Porosity was defined using mercury porometry. The performed research has made it possible to obtain new highly efficient sorbents and assess their economic efficiency. The performed research will make it possible to resolve ecological and social problems by preventing the damage caused by environmental pollution with anthropogenous wastes.

  12. Immobilisation of nuclear waste materials containing different alkali elements into single-phase NZP-based ceramics

    International Nuclear Information System (INIS)

    A single-phase host matrix based upon the sodium zirconium phosphate (NZP) structure and designed to immobilise commercial nuclear waste was investigated. In comparison with other waste forms the important advantage of the NZP ceramics is its ability to incorporate, at crystallographic levels, alkali elements without significant deterioration of the physical and chemical matrix stability. Studies on the incorporation of different alkali elements into the NZP host structure were performed. Single-phase phosphates corresponding to crystalline solutions (continuous and limited) with a structure similar to NZP were found in the series of compounds with the general formula A1-x+4yA'xE2-y(PO4)3 (y = 0, 0.5 and 1, and 0 ≤ x ≤ 1+4y), where A-A' are different alkali elements (Li, Na, K, Rb, and Cs) and E are Ti or Zr. Leaching studies with alkali containing samples revealed reasonable resistance towards the release of the constituents. (author)

  13. Corrosion behaviour of container materials for geological disposal of high-level waste. Joint annual progress report 1983

    International Nuclear Information System (INIS)

    Within the framework of the Community R and D programme on management and storage of radioactive waste (shared-cost action), a research activity is aiming at the assessment of corrosion behaviour of potential container materials for geological disposal of vitrified high-level wastes. In this report, the results obtained during the year 1983 are described. Research performed at the Studiecentrum voor Kernenergie/Centre d'Etudes de l'Energie Nucleaire (SCK/CEN) at Mol (B), concerns the corrosion behaviour in clay environments. The behaviour in salt is tested by the Kernforschungszentrum (KfK) at Karlsruhe (D). Corrosion behaviour in granitic environments is being examined by the Commissariat a l'Energie Atomique (CEA) at Fontenay-aux-Roses (F) and the Atomic Energy Research Establishment (AERE) at Harwell (UK); the first is concentrating on corrosion-resistant materials and the latter on corrosion-allowance materials. Finally, the Centre National de la Recherche Scientifique (CNRS) at Vitry (F) is examining the formation and behaviour of passive layers on the metal alloys in the various environments

  14. The future supply of and demand for candidate materials for the fabrication of nuclear fuel waste disposal containers

    International Nuclear Information System (INIS)

    This report summarizes the findings of a literature survey carried out to assess the future world supply of and demand for titanium, copper and lead. These metals are candidate materials for the fabrication of containers for the immobilization and disposal of Canada's nuclear used-fuel waste for a reference Used-fuel Disposal Centre. Such a facility may begin operation by approximately 2020, and continue for about 40 years. The survey shows that the world has abundant supplies of titanium minerals (mostly in the form of ilmenite), which are expected to last up to at least 2110. However, for copper and lead the balance between supply and demand may warrant increased monitoring beyond the year 2000. A number of factors that can influence future supply and demand are discussed in the report

  15. Technical studying on design and manufacturing of the container for low level radioactive solid waste from the KRR 1 and 2 decommissioning

    Energy Technology Data Exchange (ETDEWEB)

    Park, Seung Kook; Chung, Un Soo; Yang, Sung Hong; Lee, Dong Gyu; Jung Ki Jung

    2000-12-01

    The design requirement and manufacturing criteria have been proposed on the container for the package, storage and transportation of low level radioactive solid waste from decommissioning of KRR 1 and 2. The structure analysis was carried out based on the design criteria, and the safety of the container was assessed. The container with its capacity of 4m{sup 3} was selected for the radioactive solid waste storage. The proposed container was satisfied the criteria of ISO 1496/1 and the packaging standard of Atomic Energy Act. Manufacturing and testing standards of IAEA were also applied to the container. Stress distribution and deformation were analyzed under given condition using ANSYS code, and the maximum stress was verified to be within the yield stress without any structural deformation. From the results of lifting tests which were lifting from the four top corner fittings and fork-lift pockets, it was verified that this container was safe.

  16. Technical studying on design and manufacturing of the container for low level radioactive solid waste from the KRR 1 and 2 decommissioning

    International Nuclear Information System (INIS)

    The design requirement and manufacturing criteria have been proposed on the container for the package, storage and transportation of low level radioactive solid waste from decommissioning of KRR 1 and 2. The structure analysis was carried out based on the design criteria, and the safety of the container was assessed. The container with its capacity of 4m3 was selected for the radioactive solid waste storage. The proposed container was satisfied the criteria of ISO 1496/1 and the packaging standard of Atomic Energy Act. Manufacturing and testing standards of IAEA were also applied to the container. Stress distribution and deformation were analyzed under given condition using ANSYS code, and the maximum stress was verified to be within the yield stress without any structural deformation. From the results of lifting tests which were lifting from the four top corner fittings and fork-lift pockets, it was verified that this container was safe

  17. Thermochemical destruction of asbestos-containing roofing slate and the feasibility of using recycled waste sulfuric acid

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Seong-Nam, E-mail: namsn76@gmail.com [Engineering Research Institute, Seoul National University, Daehak-dong, Gwanak-gu 151-744 (Korea, Republic of); Jeong, Seongkyeong [Environmental Resource Recirculation Division, National Institute of Environmental Research, Environmental Research Complex, Kyeongseo-dong, Seo-gu, Incheon 404-708 (Korea, Republic of); Lim, Hojoo [Indoor Environment and Noise Division, National Institute of Environmental Research, Environmental Research Complex, Kyeongseo-dong, Seo-gu, Incheon 404-708 (Korea, Republic of)

    2014-01-30

    Highlights: • Asbestos-containing roofing slates (ACS) were thermochemically treated. • 5 N H{sub 2}SO{sub 4} with 100 °C heating for 10–24 h showed complete disappearance. • Asbestiform of ACS was changed to non-asbestiform after treatment. • Favorable destruction was occurred at the Mg(OH){sub 2} layer rather than SiO{sub 2} sheet. • Equivalent treatability of waste acid brightened the feasibility of this approach. -- Abstract: In this study, we have investigated the feasibility of using a thermochemical technique on ∼17% chrysotile-containing roofing sheet or slate (ACS), in which 5 N sulfuric acid-digestive destruction was incorporated with 10–24-h heating at 100 °C. The X-ray diffraction (XRD) and the polarized light microscopy (PLM) results have clearly shown that raw chrysotile asbestos was converted to non-asbestiform material with no crystallinity by the low temperature thermochemical treatment. As an alternative to the use of pricey sulfuric acid, waste sulfuric acid discharged from a semiconductor manufacturing process was reused for the asbestos-fracturing purpose, and it was found that similar removals could be obtained under the same experimental conditions, promising the practical applicability of thermochemical treatment of ACWs. A thermodynamic understanding based on the extraction rates of magnesium and silica from a chrysotile structure has revealed that the destruction of chrysotile by acid-digestion is greatly influenced by the reaction temperatures, showing a 80.3-fold increase in the reaction rate by raising the temperature by 30–100 °C. The overall destruction is dependent upon the breaking-up of the silicon-oxide layer – a rate-limiting step. This study is meaningful in showing that the low temperature thermochemical treatment is feasible as an ACW-treatment method.

  18. Ceramicrete stabilization of radioactive-salt-containing liquid waste and sludge water. Final CRADA report.

    Energy Technology Data Exchange (ETDEWEB)

    Ehst, D.; Nuclear Engineering Division

    2010-08-04

    It was found that the Ceramicrete Specimens incorporated the Streams 1 and 2 sludges with the adjusted loading about 41.6 and 31.6%, respectively, have a high solidity. The visible cracks in the matrix materials and around the anionite AV-17 granules included could not obtain. The granules mentioned above fixed by Ceramicrete matrix very strongly. Consequently, we can conclude that irradiation of Ceramecrete matrix, goes from the high radioactive elements, not result the structural degradation. Based on the chemical analysis of specimens No.462 and No.461 used it was shown that these matrix included the formation elements (P, K, Mg, O), but in the different samples their correlations are different. These ratios of the content of elements included are about {+-} 10%. This information shows a great homogeneity of matrix prepared. In the list of the elements founded, expect the matrix formation elements, we detected also Ca and Si (from the wollastonite - the necessary for Ceramicrete compound); Na, Al, S, O, Cl, Fe, Ni also have been detected in the Specimen No.642 from the waste forms: NaCl, Al(OH){sub 3}, Na{sub 2}SO{sub 4}. Fe(OH){sub 3}, nickel ferrocyanide and Ni(NO{sub 3})2. The unintelligible results also were found from analysis of an AV-17 granules, in which we obtain the great amount of K. The X-ray radiographs of the Ceramicrete specimens with loading 41.4 % of Stream 1 and 31.6% of Stream 2, respectively showed that the realization of the advance technology, created at GEOHKI, leads to formation of excellent ceramic matrix with high amount of radioactive streams up to 40% and more. Really, during the interaction with start compounds MgO and KH{sub 2}PO{sub 4} with the present of H{sub 3}BO{sub 3} and Wollastonite this process run with high speed under the controlled regimes. That fact that the Ceramicrete matrix with 30-40% of Streams 1 and 2 have a crystalline form, not amorphous matter, allows to permit that these matrix should be very stable, reliable

  19. Application of archaeological analogues for a repository safety case: arguments supporting the waste container lifetime

    International Nuclear Information System (INIS)

    In the Japanese HLW safety case, a carbon-steel container (overpack) was designed to have a 1 000 year lifetime, based on a corrosion allowance of 40 mm derived from laboratory data obtained under anaerobic conditions. Analogue studies have been conducted to increase confidence in the robustness of this design basis. Using X-ray computer tomography (X-CT) to measure corrosion rate, around 40 samples of archaeological iron artifacts were analysed; these were found at ancient Japanese monuments and had been buried underground for periods of several hundred to one thousand years. The corrosion rates are more than one order of magnitude less than the design allowance of 40 mm/ka, which supports the argument that the designed corrosion allowance is conservative. (authors)

  20. Creep properties of welded joints in OFHC copper for nuclear waste containment

    International Nuclear Information System (INIS)

    In Sweden it has been suggested that copper canisters are used for containment of spent nuclear fuel. These canisters will be subjected to temperatures up to 100 degrees C and external pressures up to 15 MPa. Since the material is pure (OFHC) copper, creep properties must be considered when the canisters are dimensioned. The canisters are sealed by electron beam welding which will affect the creep properties. Literature data for copper - especially welded joints - at the temperatures of interest is very scare. Therefore uniaxial creep tests of parent metal, weld metal, and simulated HAZ structures have been performed at 110 degrees C. These tests revealed considerable differences in creep deformation and rupture strength. The weld metal showed creep rates and rupture times ten times higher and ten times shorter, respectively, than those of the parent metal. The simulated HAZ was equally strongen than the parent metal. These differences were to some extent verified by results from creep tests of cross-welded specimens which, however, showed even shorter rupture times. Constitutive equations were derived from the uniaxial test results. To check the applicability of these equations to multiaxial conditions, a few internal pressure creep tests of butt-welded tubes were performed. Attemps were made to simulate their creep behaviour by constitutive equations were used. These calculations failed due to too great differences in creep deformation behaviour across the welded joint. (authors)

  1. Nuclear waste shipping container response to severe accident conditions, A brief critique of the modal study

    Energy Technology Data Exchange (ETDEWEB)

    Audin, L.

    1990-12-01

    The Modal Study (NUREG/CR-4829) attempts to upgrade the analysis of spent nuclear fuel transportation accidents, and to verify the validity of the present regulatory scheme of cask performance standards as a means to minimize risk. While an improvement over many prior efforts in this area (such as NUREG-0170), it unfortunately fails to create a realistic simulation either of a shipping cask, the severe conditions to which it could be subjected, or the potential damage to the spent fuel cargo during an accident. There are too many deficiencies in its analysis to allow acceptance of its results for the presumed cask design, and many pending changes in new containers, cargoes and shipping patterns will limit applicability of the Modal Study to future shipments. In essence, the Modal Study is a good start, but is too simplistic, incomplete, outdated and open to serious question to be used as the basis for any present-day environmental or risk assessment of spent fuel transportation. It needs to be redone, with peer review during its production and experimental verification of its assumptions, before it has any relevance to the shipments planned to Yucca Mountain. Finally, it must be expanded into a full risk assessment by inputing its radiological release fractions and probabilities into a valid dispersal simulation to properly determine the impact of its results. 51 refs.

  2. Nuclear waste shipping container response to severe accident conditions, A brief critique of the modal study

    International Nuclear Information System (INIS)

    The Modal Study (NUREG/CR-4829) attempts to upgrade the analysis of spent nuclear fuel transportation accidents, and to verify the validity of the present regulatory scheme of cask performance standards as a means to minimize risk. While an improvement over many prior efforts in this area (such as NUREG-0170), it unfortunately fails to create a realistic simulation either of a shipping cask, the severe conditions to which it could be subjected, or the potential damage to the spent fuel cargo during an accident. There are too many deficiencies in its analysis to allow acceptance of its results for the presumed cask design, and many pending changes in new containers, cargoes and shipping patterns will limit applicability of the Modal Study to future shipments. In essence, the Modal Study is a good start, but is too simplistic, incomplete, outdated and open to serious question to be used as the basis for any present-day environmental or risk assessment of spent fuel transportation. It needs to be redone, with peer review during its production and experimental verification of its assumptions, before it has any relevance to the shipments planned to Yucca Mountain. Finally, it must be expanded into a full risk assessment by inputing its radiological release fractions and probabilities into a valid dispersal simulation to properly determine the impact of its results. 51 refs

  3. Investigation of pharmaceuticals and medical devices containing 90Y extracted from high radioactive liquid waste in spent-fuel reprocessing

    International Nuclear Information System (INIS)

    Pharmaceuticals and medical devices containing radioactive 90Y are realized, approved and placed on the international market where three products are available in Europe and the United States, and one product in Japan. These products are used not for diagnosis but for treatment by internal irradiation. It was estimated from the deliberative report of the approval in Japan that 90Y was extracted in Europe from high radioactive liquid waste (HALW) yielded in spent-fuel reprocessing. In this report, products placed on the market and physical properties were reviewed, reasons of the realization and conditions to realize succeeding products were estimated, extraction method was compared with other methods, technical subjects, and relevant regulations were investigated. Although a medical device containing radioactive 90Y has been studied in Japan and one pharmaceutical product was approved, a breakthrough would be necessary to put 90Y utilization beyond alternative treatments. The breakthrough would become be promising; for example, if conventional treatments could be supported by technical development to deliver 90Y more sharply to the target with shorter serum half-life. Extraction of 90Y nuclide from HALW has advantages over thermal neutron irradiation of natural nuclide, a system is envisioned where 90Sr as a parent nuclide is separated in the reprocessing then transported to and stored in a factory of radiopharmaceuticals followed by 90Y extraction on demand. (author)

  4. In situ corrosion studies on candidate container materials for the underground disposal of high level radioactive waste in Boom Clay

    International Nuclear Information System (INIS)

    SCK·CEN has developed in the early 1980's, with the support of NIRAS/ONDRAF and EC, an extensive in situ corrosion program to evaluate the long-term corrosion behavior of various candidate container materials for the disposal of conditioned high-level radioactive waste and spent fuel. The in situ corrosion experiments were performed in the underground research facility, HADES, situated in the Boom Clay formation at a depth of 225 meters below ground level. These experiments place the samples either in direct contact with clay (type I), in a humid clay atmosphere (type 2), or in a concrete saturated clay atmosphere (type 3). During the period 1985--1994, twelve in situ corrosion experiments were installed in the underground laboratory. The exploitation of these experiments ended in 1996. All samples were recuperated and analyzed. The purpose of this paper is to summarize and discuss the results from the type 1 corrosion experiments (samples in direct contact with Boom Clay). Surface analyses tend to indicate that the so-called corrosion-resistant materials, e.g. stainless steels, Ni- and Ti-alloys, remain intact after exposure to Boom Clay between 16 and 170 C, whereas carbon steel presents significant pitting corrosion. Carbon steel seems to be unsuitable for the Belgian repository concept (pits up to 240microm deep are detected after direct exposure to the argillaceous environment for 2 years at 90 C). The stainless steels look very promising candidate container materials

  5. A prototype scintillating-fibre tracker for the cosmic-ray muon tomography of legacy nuclear waste containers

    Energy Technology Data Exchange (ETDEWEB)

    Mahon, D.F., E-mail: David.Mahon@Glasgow.ac.uk [Nuclear Physics Group, University of Glasgow, Kelvin Building, University Avenue, Glasgow, G12 8QQ Scotland (United Kingdom); Clarkson, A.; Hamilton, D.J.; Hoek, M.; Ireland, D.G. [Nuclear Physics Group, University of Glasgow, Kelvin Building, University Avenue, Glasgow, G12 8QQ Scotland (United Kingdom); Johnstone, J.R. [National Nuclear Laboratory, Central Laboratory, Sellafield, Seascale, Cumbria, CA20 1PG England (United Kingdom); Kaiser, R.; Keri, T.; Lumsden, S.; McKinnon, B.; Murray, M.; Nutbeam-Tuffs, S. [Nuclear Physics Group, University of Glasgow, Kelvin Building, University Avenue, Glasgow, G12 8QQ Scotland (United Kingdom); Shearer, C.; Staines, C. [National Nuclear Laboratory, Central Laboratory, Sellafield, Seascale, Cumbria, CA20 1PG England (United Kingdom); Yang, G. [Nuclear Physics Group, University of Glasgow, Kelvin Building, University Avenue, Glasgow, G12 8QQ Scotland (United Kingdom); Zimmerman, C. [National Nuclear Laboratory, Central Laboratory, Sellafield, Seascale, Cumbria, CA20 1PG England (United Kingdom)

    2013-12-21

    Cosmic-ray muons are highly penetrative charged particles observed at sea level with a flux of approximately 1 cm{sup −2} min{sup −1}. They interact with matter primarily through Coulomb scattering which can be exploited in muon tomography to image objects within industrial nuclear waste containers. A prototype scintillating-fibre detector has been developed for this application, consisting of two tracking modules above and below the volume to be assayed. Each module comprises two orthogonal planes of 2 mm fibres. The modular configuration allows the reconstruction of the initial and scattered muon trajectories which enable the container content, with respect to atomic number Z, to be determined. Fibre signals are read out by Hamamatsu H8500 MAPMTs with two fibres coupled to each pixel via dedicated pairing schemes developed to avoid space point ambiguities and retain the high spatial resolution of the fibres. A likelihood-based image reconstruction algorithm was developed and tested using a GEANT4 simulation of the prototype system. Images reconstructed from this simulation are presented in comparison with experimental results taken with test objects. These results verify the simulation and show discrimination between the low, medium and high-Z materials imaged.

  6. Hanford Site annual dangerous waste report: Volume 1, Part 1, Generator dangerous waste report, dangerous waste

    International Nuclear Information System (INIS)

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

  7. Hanford Site annual dangerous waste report: Volume 1, Part 1, Generator dangerous waste report, dangerous waste

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-12-31

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

  8. Research on changes of nitrate by interactions with metals under the wastes disposal environment containing TRU nuclide. 2

    International Nuclear Information System (INIS)

    In TRU wastes, wastes containing nitrate ion as salt exist. In the disposal site environment, this nitrate ion changes into nitrite ion, ammonia, etc., and possibly affects disposal site environmental changes or nuclide migration parameters. In the present research, evaluation was carried out on the chemical interaction between nitrate ion and carbon steel, which is primary reducing agent, under the low-oxygen conditions simulating a disposal site. (1) In the electrochemical test, test data were generated in order to supplement influence parameters required for improvement of the accuracy of the nitrate reaction model (NEON). As the results, it was found that the influence of potential and pH is remarkable, also that of initial nitrate concentration is significant, while the temperature is not remarkable to the nitrate and nitrite reaction themselves. Besides, it was found that the difference in the surface condition of the electrodes is not remarkable. (2) Several long-term reaction tests were carried out to assume the effects of important parameters on the nitrate behavior with carbon steel under low-oxygen high-alkaline type simulated groundwater conditions using glass sealed apparatus (ampoule tests). As the results, it was found that initial nitrate ion concentration and temperature causes the increase of hydrogen generation as well as ammonia generation, while it was found that the difference of carbon steel composition doesn't affect significantly. (3) The parameter fitting NEON was reexamined to improve accuracy, gathering data of electrochemical tests and ampoule tests conducted in 2003 and 2000 through 2002. In addition by comparing the calculation results with experimental results, applicability of NEON was investigated. (4) Implementation of NEON to the mass transfer calculation code was carried out in order to enable the calculation of the nitrate ion behavior including incomings and outgoings of substance to and from the system, resulting in the

  9. Proposed design requirements for high-integrity containers used to store, transport, and dispose of high-specific-activity, low-level radioactive wastes from Three Mile Island Unit II

    International Nuclear Information System (INIS)

    This report develops proposed design requirements for high integrity containers used to store, transport and/or dispose of high-activity, low-level radioactive wastes from Three Mile Island Unit II. The wastes considered are the dewatered resins produced by the EPICOR II waste treatment system used to clean-up the auxiliary building water. The radioactivity level of some of these EPICOR II liners is 1300 curies per container. These wastes may be disposed of in an intermediate depth burial (10 to 20 meter depth) facility. The proposed container design requirements are directed to ensure isolation of the waste and protection of the public health and safety

  10. Corrosion kinetics of alloy Ni-22Cr-13Mo-3W as structural material in high level nuclear waste containers

    International Nuclear Information System (INIS)

    Alloy Ni-22Cr-13Mo-3W (also known as C-22) is one of the candidates to fabricate high level nuclear waste containers. These containers are designed to maintain isolation of the waste for a minimum of 10,000 years. In this period, the material must be resistant to corrosion. If the containers were in contact with water, it is assumed that alloy C-22 may undergo three different corrosion mechanisms: general corrosion, localized corrosion and stress corrosion cracking. This thesis discusses only the first two types of degradation. Electrochemical techniques such as amperometry, potentiometry, potentiodynamic polarization and electrochemical impedance spectroscopy (EIS) and non-electrochemical techniques such as microscopic observation, X-ray fluorescence (XRF) and X-ray photoelectron spectroscopy (XPS) were applied to study the corrosion behavior of alloy C-22 in 1 M NaCl, 25 C degrees saturated NaF (approximately 1 M) and 0,5 M NaCl + 0,5 M NaF solutions. Effects of temperature, pH and alloy thermal aging were analyzed. The corrosion rates obtained at 90 C degrees were low ranging from 0.04 μm/year to 0.48 μm /year. They increased with temperature and decreased with solution pH. Most of the impedance measurements showed a simply capacitive behavior. A second high-frequency time constant was detected in some cases. It was attributed to the formation of a nickel oxide and/or hydroxide at potentials near the reversible potential for this reaction. The active/passive transition detected in some potentiodynamic polarization curves was attributed to the same process. The corrosion potential showed an important increase after 24 hours of immersion. This increase in the corrosion potential was associated with an improvement of the passive film. The corrosion potential was always lower than the re-passivation potential for the corresponding media. The trans passive behavior of alloy C-22 was mainly influenced by temperature and solution chemistry. A clear trans passive peak

  11. Hanford Site annual dangerous waste report: Volume 3, Part 1, Waste Management Facility report, dangerous waste

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  13. Hanford Site annual dangerous waste report: Volume 3, Part 1, Waste Management Facility report, dangerous waste

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-12-31

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

  14. Investigations on the populations of introduced and resident micro-organisms in deep repositories and their effects on containment of radioactive wastes

    International Nuclear Information System (INIS)

    The present study has sought to establish basic facts concerning the importance of microbial presence in a radioactive waste repository. These are: (1) establishing the presence of microbes in relevant geological formations; (2) defining the importance of isolated groups to radioactive waste containment; (3) establishing the ability of sampled microbes to tolerate repository environmental conditions; (4) conducting preliminary work on microbial ability to influence radionuclide migration characteristics; (5) attempting basic modelling of the effects of micro-organisms on containment and radionuclide migration. Results have shown that micro-organisms of significance (e.g. sulphur cycle bacteria) are present in relevant formations. These groups can tolerate repository conditions and influence migration studies. Their growth, however, would seem to be limited by the nutrient availability. Simple modelling has concentrated on conservative calculations on the constraints of maximum effects of microbial contamination of a high-level waste repository

  15. Identifying optimal regional solid waste management strategies through an inexact integer programming model containing infinite objectives and constraints.

    Science.gov (United States)

    He, Li; Huang, Guo-He; Zeng, Guang-Ming; Lu, Hong-Wei

    2009-01-01

    The previous inexact mixed-integer linear programming (IMILP) method can only tackle problems with coefficients of the objective function and constraints being crisp intervals, while the existing inexact mixed-integer semi-infinite programming (IMISIP) method can only deal with single-objective programming problems as it merely allows the number of constraints to be infinite. This study proposes, an inexact mixed-integer bi-infinite programming (IMIBIP) method by incorporating the concept of functional intervals into the programming framework. Different from the existing methods, the IMIBIP can tackle the inexact programming problems that contain both infinite objectives and constraints. The developed method is applied to capacity planning of waste management systems under a variety of uncertainties. Four scenarios are considered for comparing the solutions of IMIBIP with those of IMILP. The results indicate that reasonable solutions can be generated by the IMIBIP method. Compared with IMILP, the system cost from IMIBIP would be relatively high since the fluctuating market factors are considered; however, the IMILP solutions are associated with a raised system reliability level and a reduced constraint violation risk level. PMID:18406594

  16. Resource Utilization of Chromium Containing Leather Waste%含铬革屑的资源化利用

    Institute of Scientific and Technical Information of China (English)

    罗凤香

    2014-01-01

    Heavy metal chromium containing leather waste shavings, if not effectively treated, will cause environmental pollution and human health damage. In the increasingly serious environmental problems and increasingly scarce resource today, the resource utilization of above mentioned chrome tanned leather shavings has enormous environmental and economic value. This paper describes the methods to utilize chromium leather shavings in recent years.%含有重金属铬的含铬革屑制革废弃物,如果不能有效处理,将会造成环境污染及人类等健康损害。在环境问题日益严重和资源日趋贫乏的今天,对含铬革屑的资源化利用具有巨大的环境和经济价值。本文简述了近年来对含铬革屑的利用方法。

  17. Development and characterization of new high-level waste form containing LiCl KCl eutectic salts for achieving waste minimization from pyroprocessing

    International Nuclear Information System (INIS)

    The purpose of this project is to develop new high level waste (HLW) forms and fabrication processes to dispose of active metal fission products that are removed from electrorefiner salts in the pyroprocessing based fuel cycle. The current technology for disposing of active metal fission products in pyroprocessing involves non selectively discarding of fission product loaded salt in a glass-bonded sodalite ceramic waste form. Selective removal of fission products from the molten salt would greatly minimize the amount of HLW generated and methods were developed to achieve selective separation of fission products during a previous I NERI research project (I NERI 2006 002 K). This I NERI project proceeds from the previous project with the development of suitable waste forms to immobilize the separated fission products. The Korea Atomic Energy Research Institute (KAERI) has focused primarily on developing these waste forms using surrogate waste materials, while the Idaho National Laboratory (INL) has demonstrated fabrication of these waste forms using radioactive electrorefiner salts in hot cell facilities available at INL. Testing and characterization of these radioactive materials was also performed to determine the physical, chemical, and durability properties of the waste forms

  18. Effect of the activation of a clay-base paper industry by-product on cement matrix behaviour

    Directory of Open Access Journals (Sweden)

    García, R.

    2008-12-01

    Full Text Available The present study addresses variations in the calcination temperature (600-750 ºC and kiln time (two to five hours applied to activate coated paper waste and their effect on the rheological, physical and mechanical behaviour of cement matrices containing these active additions.The results obtained showed that the conditions under which kaolinite was activated had a direct effect on the subsequent behaviour of the calcined products. At activating temperatures of over 700 ºC, pozzolanic activity and mechanical strength were observed to be lower, setting time shorter and the mortar less workable.El presente trabajo de investigación aborda la influencia de las condiciones de activación (600-750 ºC y 2-5 horas de permanencia en el horno de los lodos de papel procedente de la fabricación de papel estucado en el comportamiento reológico, físico y mecánico de las matrices de cementos elaboradas con este tipo de adiciones activas.Los resultados obtenidos muestran una influencia directa entre las condiciones de activación de la caolinita y el comportamiento posterior de los productos calcinados. Así, en condiciones de activación superiores a 700 ºC se observa una menor actividad puzolánica, tiempo de fraguado más corto, disminución de la trabajabilidad de los morteros mezcla y resistencia mecánica más baja.

  19. Energy recovery of combustible fraction from shredding of wastes containing metals; Energiaatervinning av braennbar fraktion fraan fragmentering av metallhaltigt avfall

    Energy Technology Data Exchange (ETDEWEB)

    Gyllenhammar, Marianne [Stena Metall, Goeteborg (Sweden); Victoren, Anders; Niemi, Jere [Metso Power, Tammerfors (Finland); Johansson, Andreas [SP Technical Research Inst. of Sweden, Boraas (Sweden)

    2009-01-15

    Combustible products from fragmentation are not allowed to be deposited on landfills any more in Sweden. These products have to be material recovered or energy recovered. The combustible fraction from recovered metal scrap, SLF (shredder light fraction), contains metals and the chlorine content is relatively high. Due to this there could be a risk with deposits and corrosion on convection surfaces in combustion plants. Co-combustion with sewage sludge could be a solution for solving problems with the difficult contents in SLF. The aim of the project was to do a theoretical judgment of how sewage sludge could affect deposit formation and corrosion when co-combusted with SLF. Due to the high amount of water in the sewage sludge the percentage of sewage sludge in the fuel mixture was limited. The maximum percentage of energy used was 3.5 % (ca 13% on weight basis). The thermodynamic calculations showed that at combustion with 100% SLF the lead and zinc chlorides in gaseous form increased 5-6 times in comparative with combustion with ordinary waste combustion in Boraas. But as the thermodynamic equilibrium calculations will not consider the kinetics and just calculate independent of time the results should be considered as indicative and not directly comparative to actual boiler conditions. All lead and zinc were assumed reactive which will probably not be the case in a boiler. In the calculations the aluminum was removed from the calculations (not taken into account) and the alkali-phosphor reactions are incomplete due to lack of reliable thermodynamic data. These defiance's should be considered when evaluating the results from the thermodynamic chemical equilibrium calculations as well as the fact that the calculations cannot yet take into account the possible erosive effect the high ash amount could have on the deposits. The calculations showed that co-combusting with SLF (ca 20%) gave high amounts of gaseous lead chlorides. Also high amount of zinc chlorides

  20. Gel Explosive Containing Waste HTPB Propellants%含废弃丁羟推进剂的凝胶炸药

    Institute of Scientific and Technical Information of China (English)

    王鹏; 魏晓安; 肖学海; 何卫东

    2011-01-01

    为安全处理和再利用废弃固体推进剂,通过添加单基药将丁羟推进剂再利用制备了灌注式凝胶炸药.采用验证板试验及电离探针法研究了不同装药配比、推进剂颗粒尺寸及装药直径对炸药爆轰性能的影响.结果表明,丁羟推进剂难以发生爆轰,若添加适量单基药,能显著提高炸药的爆轰感度,并降低其临界直径;该凝胶炸药密度为1.6 g/cm3,直径为70 mm时爆速可达6 500 m/s,可做大直径的露天工业炸药使用.%In order to dispose and reuse waste solid propellants safely,a type of gel explosive containing waste HTPB propellant and single-base propellant was proposed. The effects of charge ratio between the single-base propellants and HTPB propellants, particle size of HTPB propellant and charge diameter on the detonation properties of the gel explosive were studied by the witness plate test and ionization probes. The results show that the HTPB propellant itself is hard to detonate, but if some single-base propellants are added and mixed with HTPB propellant in the charge, the impact sensitivity of explosive can be significantly improved, the critical diameter would also be reduced. The density of the gel explosive is about 1. 6g/cm3. And the detonation velocity is more than 6 500m/s when the charge diameter is 70 mm. Thus it can be used as an open-air commercial gel explosive with big charge diameter.

  1. MODELO ACO PARA LA RECOLECCIÓN DE RESIDUOS POR CONTENEDORES ACO MODEL APPLIED TO THE WASTE COLLECTION BY CONTAINERS

    Directory of Open Access Journals (Sweden)

    Eduardo Salazar Hornig

    2009-08-01

    Full Text Available ACO es una metaheurística inspirada en el comportamiento de las colonias de hormigas para solucionar problemas de optimización combinatoria, por medio de la utilización de agentes computacionales simples que trabajan de manera cooperativa y se comunican mediante rastros de feromona artificiales. En este trabajo se presenta un modelo para resolver el Problema de Recolección de Residuos Domiciliarios por Contenedores, el que aplica un concepto de secuencias parciales de recolección que deben ser unidas para minimizar la distancia total de recolección. El problema de unir las secuencias parciales se representa como un TSP, el que es resuelto mediante un algoritmo ACO. En base a recomendaciones de la literatura, se calibran experimentalmente los parámetros del algoritmo y se recomiendan rangos de valores que representan buenos rendimientos promedio. El modelo se aplica a un sector de recolección de la comuna de San Pedro de la Paz, Chile, obteniéndose rutas de recolección que reducen la distancia total recorrida respecto de la actual ruta utilizada y de la solución obtenida con otro modelo desarrollado previamente.ACO is a metaheuristic inspired in the behavior of natural ant colonies to solve combinatorial optimization problems, based on simple agents that work cooperatively communicating by artificial pheromone trails. In this paper a model to solve the municipal waste collection problem by containers is presented, which applies a concept of partial collection sequences that must be joined to minimize the total collection distance. The problem to join the partial collection sequences is represented as a TSP, which is solved by an ACO algorithm. Based on the literature, algorithm parameters are experimentally calibrated and range of variations that represents good average solutions are recommended. The model is applied to a waste collection sector of the San Pedro de la Paz commune in Chile, obtaining recollection routes with less total

  2. Idaho National Engineering Laboratory response to the December 13, 1991, Congressional inquiry on offsite release of hazardous and solid waste containing radioactive materials from Department of Energy facilities

    International Nuclear Information System (INIS)

    This report is a response to the December 13, 1991, Congressional inquiry that requested information on all hazardous and solid waste containing radioactive materials sent from Department of Energy facilities to offsite facilities for treatment or disposal since January 1, 1981. This response is for the Idaho National Engineering Laboratory. Other Department of Energy laboratories are preparing responses for their respective operations. The request includes ten questions, which the report divides into three parts, each responding to a related group of questions. Part 1 answers Questions 5, 6, and 7, which call for a description of Department of Energy and contractor documentation governing the release of waste containing radioactive materials to offsite facilities. ''Offsite'' is defined as non-Department of Energy and non-Department of Defense facilities, such as commercial facilities. Also requested is a description of the review process for relevant release criteria and a list of afl Department of Energy and contractor documents concerning release criteria as of January 1, 1981. Part 2 answers Questions 4, 8, and 9, which call for information about actual releases of waste containing radioactive materials to offsite facilities from 1981 to the present, including radiation levels and pertinent documentation. Part 3 answers Question 10, which requests a description of the process for selecting offsite facilities for treatment or disposal of waste from Department of Energy facilities. In accordance with instructions from the Department of Energy, the report does not address Questions 1, 2, and 3

  3. Efficiency Assessment of Using Flammable Compounds from Water Treatment and Methanol Production Waste for Plasma Synthesis of Iron-Containing Pigments

    Science.gov (United States)

    Shekhovtsova, Anastasia P.; Karengin, Alexander G.

    2016-08-01

    This article describes the possibility of applying the low-temperature plasma for obtaining iron-containing pigments from water purification and flammable methanol production waste. In this paper were calculated combustion parameters of water-saltorganic compositions (WSOC) with different consists. Authors determined the modes of energy- efficient processing of the previously mentioned waste in an air plasma. Having considered the obtained results there were carried out experiments with flammable dispersed water-saltorganic compositions on laboratory plasma stand. All the experimental results are confirmed by calculations.

  4. Experimental Simulation of the Radionuclide Behaviour in the Process of Creating Additional Safety Barriers in Solid Radioactive Waste Repositories Containing Irradiated Graphite

    Science.gov (United States)

    Pavliuk, A. O.; Kotlyarevskiy, S. G.; Bespala, E. V.; Zakarova, E. V.; Rodygina, N. I.; Ermolaev, V. M.; Proshin, I. M.; Volkova, A.

    2016-08-01

    Results of the experimental modeling of radionuclide behavior when creating additional safety barriers in solid radioactive waste repositories are presented. The experiments were run on the repository mockup containing solid radioactive waste fragments including irradiated graphite. The repository mockup layout is given; the processes with radionuclides that occur during the barrier creation with a clayey solution and during the following barrier operation are investigated. The results obtained confirm high anti-migration and anti-filtration properties of clay used for the barrier creation even under the long-term excessive water saturation of rocks confining the repository.

  5. Het composteren van groente- en tuinafval in containers voor particulier gebruik = Composting of vegetable and garden waste in containers for household use

    OpenAIRE

    Riem Vis, F.

    1985-01-01

    In three experiments composting in different types of containers was studied. No differences between the containers were observed. Seedling tests showwed that with up to 50 % compost added to a sandy soil growth was stimulated. In the case of a 100 % compost crop damage occurred. Verschillende typen containers voor het composteren van tuin- en groenteafval voor particulier gebruik werden in dit onderzoek vergeleken. Uit het verloop van de temperatuur en van het zuurstofgehalte van de lucht bl...

  6. Research on the treatment of liquid waste containing cesium by an adsorption-microfiltration process with potassium zinc hexacyanoferrate

    Energy Technology Data Exchange (ETDEWEB)

    Zhang Changping, E-mail: melindazhang@yahoo.com.cn [School of Environmental Science and Engineering, Tianjin University, Tianjin, 300072 (China); Gu Ping, E-mail: guping@tju.edu.cn [School of Environmental Science and Engineering, Tianjin University, Tianjin, 300072 (China); Zhao Jun; Zhang Dong; Deng Yue [Institute of Nuclear Physics and Chemistry, Chinese Academy of Engineering Physics, Mianyang 621900 (China)

    2009-08-15

    The removal of cesium from an aqueous solution by an adsorption-microfiltration (AMF) process was investigated in jar tests and lab-scale tests. The adsorbent was K{sub 2}Zn{sub 3}[Fe(CN){sub 6}]{sub 2}. The obtained cesium data in the jar test fit a Freundlich-type isotherm well. In the lab-scale test, the mean cesium concentration of the raw water and the effluent were 106.87 {mu}g/L and 0.59 {mu}g/L, respectively, the mean removal of cesium was 99.44%, and the mean decontamination factors (DF) and concentration factors (CF) were 208 and 539, respectively. The removal of cesium in the lab-scale test was better than that in the jar test because the old adsorbents remaining in the reactor still had adsorption capacity with the premise of no significant desorption being observed, and the continuous renewal of the adsorbent surface improved the adsorption capacity of the adsorbent. Some of the suspended solids were deposited on the bottom of the reactor, which would affect the mixing of adsorbents with the raw water and the renewing of the adsorbent surface. Membrane fouling was the main physical fouling mechanism, and the cake layer was the main filtration resistance. Specific flux (SF) decreased step by step during the whole period of operation due to membrane fouling and concentration polarization. The quality of the effluent was good and the turbidity remained lower than 0.1 NTU, and the toxic anion, CN{sup -}, could not be detected because of its low concentration, this indicated that the effluent was safe. The AMF process was feasible for practical application in the treatment of liquid waste containing cesium.

  7. Application of ferrate(VI) in the treatment of industrial wastes containing metal complexed cyanides : A green treatment

    Institute of Scientific and Technical Information of China (English)

    SEUNG-MOK Lee; DIWAKAR Tiwari

    2009-01-01

    The higher oxidation state of iron (Fe(VI)) was employed for the oxidation of cyanide (CN) and the simultaneous removal of copper or nickel in the mixed/complexed systems of CN-Cu, CN-Ni, or CN-Cu-Ni. The degradation of CN (1.00 mmol/L) and removal of Cu (0.095 mmol/L) were investigated as a function of Fe(VI) doses from 0.3-2.00 mmol/L at pH 10.0. It was found that Fe(VI) could readily oxidize CN and the reduction of Fe(VI) into Fe(III) might serve efficiently for the removal of free copper ions. The increase in Fe(VI) dose apparently favoured the CN oxidation as well Cu removal. Moreover, the pH dependence study (pH 10.0-13.0) revealed that the oxidation of CN was almost unaffected in the studied pH range (10.0-13.0), however, the maximum removal efficiency of Cu was obtained at pH 13.0. Similarly, treatment was carried out for CN-Ni system having the initial Ni concentration of 0.170 mmol/L and CN concentration of 1.00 mmol with Fe(VI) dose 2.00 mmol at various pH values (10.0-12.0). Results showed a partial oxidation of CN and partial removal of Ni occurred. It can be observed that Fe(VI) can partially degrade the CN-Ni complex in this pH range. Further, Fe(VI) was applied for the treatment of simulated industrial waste/effluent waters treatment containing CN, Cu, and Ni.

  8. The disposal of Canada's nuclear fuel waste: engineered barriers alternatives

    International Nuclear Information System (INIS)

    The concept for disposal of Canada's nuclear fuel waste involves emplacing the waste in a vault excavated at a depth of 500 to 1000 m in plutonic rock of the Canadian Shield. The solid waste would be isolated from the biosphere by a multibarrier system consisting of engineered barriers, including long-lived containers and clay and cement-based sealing materials, and the natural barrier provided by the massive geological formation. The technical feasibility of this concept and its impact on the environment and human health are being documented in an Environmental Impact Statement (EIS), which will be submitted for review under the federal Environmental Assessment and Review Process. This report, one of nine EIS primary references, describes the various alternative designs and materials for engineered barriers that have been considered during the development of the Canadian disposal concept and summarizes engineered barrier concepts being evaluated in other countries. The basis for the selection of a reference engineered barrier system for the EIS is presented. This reference system involves placing used CANDU (Canada Deuterium Uranium) fuel bundles in titanium containers, which would then be emplaced in boreholes drilled in the floor of disposal rooms. Clay-based sealing materials would be used to fill both the space between the containers and the rock and the remaining excavations. In the section on waste forms, the properties of both used-fuel bundles and solidified high-level wastes, which would be produced by treating wastes resulting from the reprocessing of used fuel, are discussed. Methods of solidifying the wastes and the chemical durability of the solidified waste under disposal conditions are reviewed. Various alternative container designs are reviewed, ranging from preliminary conceptual designs to designs that have received extensive prototype testing. Results of structural performance, welding and inspection studies are also summarized. The corrosion of

  9. Thermal performance of a buried nuclear waste storage container storing a hybrid mix of PWR and BWR spent fuel rods; Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, G.L.

    1991-11-01

    Lawrence Livermore National Laboratory will design, model, and test nuclear waste packages for use at the Nevada Nuclear Waste Storage Repository at Yucca Mountain, Nevada. One such package would store tightly packed spent fuel rods from both pressurized and boiling water reactors. The storage container provides the primary containment of the nuclear waste and the spent fuel rod cladding provides secondary containment. A series of transient conduction and radiation heat transfer analyses was run to determine for the first 1000 yr of storage if the temperature of the tuff at the borehole wall ever falls below 97{degrees}C and whether the cladding of the stored spent fuel ever exceeds 350{degrees}C. Limiting the borehole to temperatures of 97{degrees}C or greater helps minimize corrosion by assuring that no condensed water collects on the container. The 350{degrees}C cladding limit minimizes the possibility of creep- related failure in the spent fuel rod cladding. For a series of packages stored in a 8 {times} 30 m borehole grid where each package contains 10-yr-old spent fuel rods generating 4.74 kW or more, the borehole wall stays above 97{degrees}C for the full 10000-yr analysis period. For the 4.74-kW load, the peak cladding temperature rises to just below the 350{degrees}C limit about 4 years after emplacement. If the packages are stored using the spacing specified in the Site Characterization Plan (15 ft {times} 126 ft), a maximum of 4.1 kW per container may be stored. If the 0.05-m-thick void between the container and the borehole wall is filled with loosely packed bentonite, the peak cladding temperature rises more than 40{degrees}C above the allowed cladding limit. In all cases the dominant heat transfer mode between container components is thermal radiation.

  10. Multi-point injection: A general purpose delivery system for treatment and containment of hazardous and radiological waste

    Energy Technology Data Exchange (ETDEWEB)

    Kauschinger, J.L. [Ground Environmental Services, Alpharetta, GA (United States); Kubarewicz, J. [Jacobs Engineering, Oak Ridge, TN (United States); Van Hoesen, S.D. [Lockheed Martin Energy Systems, Oak Ridge, TN (United States)

    1997-12-31

    The multi-point injection (MPI) technology is a proprietary jetting process for the in situ delivery of various agents to treat radiological and/or chemical wastes. A wide variety of waste forms can be treated, varying from heterogeneous solid waste dumped into shallow burial trenches, bottom sludge (heel material) inside of underground tanks, and contaminated soils with widely varying soil composition (gravel, silts/clays, soft rock). The robustness of the MPI system is linked to the use of high speed mono-directional jets to deliver various types of agents for a variety of applications, such as: pretreatment of waste prior to insitu vitrification, solidification of waste for creating low conductivity monoliths, oxidants for insitu destruction of organic waste, and grouts for creating barriers (vertical, inclined, and bottom seals). The only strict limitation placed upon the MPI process is that the material can be pumped under high pressure. This paper describes the procedures to inject ordinary grout to form solidified monoliths of solid wastes.

  11. Multi-point injection: A general purpose delivery system for treatment and containment of hazardous and radiological waste

    International Nuclear Information System (INIS)

    The multi-point injection (MPI) technology is a proprietary jetting process for the in situ delivery of various agents to treat radiological and/or chemical wastes. A wide variety of waste forms can be treated, varying from heterogeneous solid waste dumped into shallow burial trenches, bottom sludge (heel material) inside of underground tanks, and contaminated soils with widely varying soil composition (gravel, silts/clays, soft rock). The robustness of the MPI system is linked to the use of high speed mono-directional jets to deliver various types of agents for a variety of applications, such as: pretreatment of waste prior to insitu vitrification, solidification of waste for creating low conductivity monoliths, oxidants for insitu destruction of organic waste, and grouts for creating barriers (vertical, inclined, and bottom seals). The only strict limitation placed upon the MPI process is that the material can be pumped under high pressure. This paper describes the procedures to inject ordinary grout to form solidified monoliths of solid wastes

  12. Selection of candidate container materials for the conceptual waste package design for a potential high level nuclear waste repository at Yucca Mountain

    Energy Technology Data Exchange (ETDEWEB)

    Van Konynenburg, R.A.; Halsey, W.G.; McCright, R.D.; Clarke, W.L. Jr. [Lawrence Livermore National Lab., CA (United States); Gdowski, G.E. [KMI, Inc., Albuquerque, NM (United States)

    1993-02-01

    Preliminary selection criteria have been developed, peer-reviewed, and applied to a field of 41 candidate materials to choose three alloys for further consideration during the advanced conceptual design phase of waste package development for a potential high level nuclear waste repository at Yucca Mountain, Nevada. These three alloys are titanium grade 12, Alloy C-4, and Alloy 825. These selections are specific to the particular conceptual design outlined in the Site Characterization Plan. Other design concepts that may be considered in the advanced conceptual design phase may favor other materials choices.

  13. Selection of candidate container materials for the conceptual waste package design for a potential high level nuclear waste repository at Yucca Mountain

    International Nuclear Information System (INIS)

    Preliminary selection criteria have been developed, peer-reviewed, and applied to a field of 41 candidate materials to choose three alloys for further consideration during the advanced conceptual design phase of waste package development for a potential high level nuclear waste repository at Yucca Mountain, Nevada. These three alloys are titanium grade 12, Alloy C-4, and Alloy 825. These selections are specific to the particular conceptual design outlined in the Site Characterization Plan. Other design concepts that may be considered in the advanced conceptual design phase may favor other materials choices

  14. Solid waste reclamation and recycling: Packaging and containers. (Latest citations from the NTIS bibliographic database). Published Search

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-08-01

    The bibliography contains citations concerning techniques and management of packaging and container recycling. References discuss recycling of tin and aluminum cans, reverse vending machines, reusable packaging and containers, and the future of containers. Environmental aspects, government programs, and development of recycling markets are covered. (Contains 50-250 citations and includes a subject term index and title list.) (Copyright NERAC, Inc. 1995)

  15. Comparison of long-term stability of containment systems for residues and wastes contaminated with naturally occurring radionuclides at an arid site and two humid sites

    International Nuclear Information System (INIS)

    The long-term stability of near-surface containment systems designed for the management of radioactive wastes and residues contaminated with naturally occurring radionuclides are compared at the three different sites. The containment designs are: (1) a diked 8.9-m high mound, including a 3.2-m layered cap at a site (humid) near Lewiston, New York, (2) a 6.8-m-high mound, including a similar 3.2-m cap at a site (humid) near Oak Ridge, Tennessee, and (3) 4.8-m deep trenches with 3.0-m backfilled caps at a site (arid) near Hanford, Washington. Geological, hydrological, and biological factors affecting the long-term (1000-year) integrity of the containment systems at each site are examined, including: erosion, flooding, drought, wildfire, slope and cover failure, plant root penetration, burrowing animals, other soil-forming processes, and land-use changes. For the containment designs evaluated, releases of radon-222 at the arid site are predicted to be several orders of magnitude higher than at the two humid sites - upon initial burial and at 1000 years (after severe erosion). Transfer of wastes containing naturally occurring radionuclides from a humid to an arid environment offers little or no advantage relative to long-term stability of the containment system and has a definite disadvantage in terms of gaseous radioactive releases. 26 references, 3 figures, 4 tables

  16. Beneficiation Research on Waste Rocks Containing Iron Ores of a Iron Mine%某铁矿排岩含铁废石选矿试验研究

    Institute of Scientific and Technical Information of China (English)

    郭晗曙

    2011-01-01

    目前国内部分铁矿采场都有一定量的排岩,有的采场排岩当做废石处理,而该废石亦含有部分可选性较好的磁性铁矿物,对国内某铁矿采场的排岩进行了阶段干选、干选精矿磨选试验研究,结果表明在干选抛弃大量的干选尾矿后,获得的干选精矿经过磨矿弱磁选试验获得了质量较好的铁精矿。%At present,some of iron ore stopes at domestic have a certain amount of waste rocks.Some rocks are treated as waste rock,but the waste rock also contains magnetic iron minerals with good dressing ability.Experiments on waste rocks from an Iron Mine at domestic are made through the process of stage dry magnetic separation and dry concentrate separation after grinding.The results show that after a large number of waste rocks tailings are cast by dry magnetic separation,and dry concentrates are treated by low intensity magnetic separation after grinding,high-quality iron concentrates are achieved.

  17. FOURTH ANNUAL REPORT. PHYSICAL, CHEMICAL AND STRUCTURAL EVOLUTION OF ZEOLITE-CONTAINING WASTE FORMS PRODUCED FROM METAKAOLINITE AND CALCINED SODIUM BEARING WASTE (HLW AND/OR LLW)

    Science.gov (United States)

    Using zeolites for the management of radioactive waste is not new, but a process by which the zeolites can be made to act as a cementing agent is. Zeolitic materials are relatively easy to synthesize from a wide range of both natural and man-made precursors. The process under st...

  18. Purification of LiCl-KCl eutectic waste salt containing rare earth chlorides delivered from the pyrochemical process of used nuclear fuel using a reactive distillation process

    International Nuclear Information System (INIS)

    In the pyrochemical process of used nuclear fuel, the purification of waste salts containing radioactive nuclides can greatly contribute to a radioactive waste reduction. For this reason, the purification of LiCl-KCl eutectic salt containing rare earth chlorides was performed using a series of the phosphorylation process and the distillation process. LiCl-KCl eutectic salt recovered from the purification had a very low concentration (<1 ppm) for the rare earth chlorides. The recycling feasibility of the recovered salt was verified through a uranium electro-deposition test using LiCl-KCl eutectic salt as the electrolyte. Based on these results, one body type of reactive distillation equipment with two top covers was designed. (author)

  19. Thermoelastic/plastic analysis of waste-container sleeve. II. Influence of large displacements on sleeve loading. Technical memorandum report RSI-0017

    International Nuclear Information System (INIS)

    Modification of the thermoelastic/plastic finite element program to account for large displacements possibly associated with the development of an extensive plastic zone about a radioactive waste container emplaced in a typical repository room (SALT-4/T model) has been completed. Comparisons of radial stresses acting on the waste container and borehole wall displacements computed by the modified and conventional analyses techniques reveal little difference between the two sets of results over a 10 year heating period for salt strengths 25 percent of original values. Because no significant differences in results arise even under these exaggerated conditions, the more costly large displacement option need be used only sparingly as an occasional control check on the conventional procedure. As a consequence, economy of computer run time can be maintained without sacrifice of accuracy

  20. Method of extracting copper sulfate from waste residue and waste liquor containing copper%从含铜废渣废液中提取硫酸铜的方法研究

    Institute of Scientific and Technical Information of China (English)

    曾琦斐

    2012-01-01

    研究从含铜废渣、废液中提取硫酸铜的方法,以减少污染、回收资源.利用含铜废渣、废液生产海绵铜,再通过置换、氧化、酸化、结晶以及重结晶等步骤制备五水硫酸铜晶体.通过上述方法由含铜废渣、废液制备出硫酸铜晶体.该方法所用设备简单,操作简便,铜的回收率高,硫酸铜产品质量达到化学纯(CP)等级.%In order to reduce pollution and recycle the resources,a method of extracting copper sulfate from waste residue and waste liquor containing copper was explored. Spongy copper was produced from waste residue and waste liquor containing copper,and then copper sulphate pentahydrate crystal was produced after the steps of displacing, oxidizing,acidifying, crystallizing, and recrystallizing. Copper sulfate crystal was produced by the above-mentioned method. This method had advantages, such as simple equipment, easy operation, and high recovery rate of copper. Product quality of copper sulfate is up to chemical reagents grade.

  1. The Waste Package Project. Final report, July 1, 1995--February 27, 1996: Volume 1, The structural performance of the shell and fuel rods of a high level nuclear waste container

    Energy Technology Data Exchange (ETDEWEB)

    Ladkany, S.G.; Rajagopalan, R.

    1996-06-01

    This dissertation proposal covers research work that started in the spring of 1992. The aim of the research has been to study the structural performance and stability of proposed nuclear waste containers and the enclosed fuel rods to be used in the long term storage of High Level Nuclear Waste (HLNW). This research is in two phases, computational and experimental. The computational phase deals with the linear and nonlinear Finite Element Analysis of the different containers due to various loading conditions during normal handling conditions and due to the effect of long term corrosion while the canister is stored in the drift of a backfilled geological repository. The elastoplastic stability of the nuclear fuel rods were studied under body forces resulting from acceleration vectors at varying angles, resulting from a sudden drop of the canister at an angle onto a hard surface.

  2. Immobilization of antimony in waste-to-energy bottom ash by addition of calcium and iron containing additives.

    Science.gov (United States)

    Van Caneghem, Jo; Verbinnen, Bram; Cornelis, Geert; de Wijs, Joost; Mulder, Rob; Billen, Pieter; Vandecasteele, Carlo

    2016-08-01

    The leaching of Sb from waste-to-energy (WtE) bottom ash (BA) often exceeds the Dutch limit value of 0.32mgkg(-1) for recycling of BA in open construction applications. From the immobilization mechanisms described in the literature, it could be concluded that both Ca and Fe play an important role in the immobilization of Sb in WtE BA. Therefore, Ca and Fe containing compounds were added to the samples of the sand fraction of WtE BA, which in contrast to the granulate fraction is not recyclable to date, and the effect on the Sb leaching was studied by means of batch leaching tests. Results showed that addition of 0.5 and 2.5% CaO, 5% CaCl2, 2.5% Fe2(SO4)3 and 1% FeCl3 decreased the Sb leaching from 0.62±0.02mgkgDM(-1) to 0.20±0.02, 0.083±0.044, 0.25±0.01, 0.27±0.002 and 0.29±0.02mgkgDM(-1), respectively. Due to the increase in pH from 11.41 to 12.53 when 2.5% CaO was added, Pb and Zn leaching increased and exceeded the respective leaching limits. Addition of 5% CaCO3 had almost no effect on the Sb leaching, as evidenced by the resulting 0.53mgkgDM(-1) leaching concentration. This paper shows a complementary enhancement of the effect of Ca and Fe, by comparing the aforementioned Sb leaching results with those of WtE BA with combined addition of 2.5% CaO or 5% CaCl2 with 2.5% Fe2(SO4)3 or 1% FeCl3. These lab scale results suggest that formation of romeites with a high Ca content and formation of iron antimonate (tripuhyite) with a very low solubility are the main immobilization mechanisms of Sb in WtE BA. Besides the pure compounds and their mixtures, also addition of 10% of two Ca and Fe containing residues of the steel industry, hereafter referred to as R1 and R2, was effective in decreasing the Sb leaching from WtE BA below the Dutch limit value for reuse in open construction applications. To evaluate the long term effect of the additives, pilot plots of WtE BA with 10% of R1 and 5% and 10% of R2 were built and samples were submitted to leaching tests at

  3. The Design and Performance of a Scintillating-Fibre Tracker for the Cosmic-ray Muon Tomography of Legacy Nuclear Waste Containers

    CERN Document Server

    Clarkson, Anthony; Hoek, Matthias; Ireland, David G; Johnstone, Russell; Kaiser, Ralf; Keri, Tibor; Lumsden, Scott; Mahon, David F; McKinnon, Bryan; Murray, Morgan; Nutbeam-Tuffs, Sian; Shearer, Craig; Staines, Cassie; Yang, Guangliang; Zimmerman, Colin

    2013-01-01

    Tomographic imaging techniques using the Coulomb scattering of cosmic-ray muons are increasingly being exploited for the non-destructive assay of shielded containers in a wide range of applications. One such application is the characterisation of legacy nuclear waste materials stored within industrial containers. The design, assembly and performance of a prototype muon tomography system developed for this purpose are detailed in this work. This muon tracker comprises four detection modules, each containing orthogonal layers of Saint-Gobain BCF-10 2mm-pitch plastic scintillating fibres. Identification of the two struck fibres per module allows the reconstruction of the incoming and Coulomb-scattered muon trajectories. These allow the container content, with respect to the atomic number Z of the scattering material, to be determined through reconstruction of the scattering location and magnitude. On each detection layer, the light emitted by the fibre is detected by a single Hamamatsu H8500 MAPMT with two fibre...

  4. A Brief Discussion on the Application of the Ideas of Cleaner Production to the Dispose of Waste Materials Containing Zine%浅谈"清洁生产"思维在含锌废料处理中的应用

    Institute of Scientific and Technical Information of China (English)

    张德华; 魏昶; 戴永年

    2000-01-01

    The cleaner production is applied to quantitatively investigate the situation of pollution and waste materials produced by every procedure in whole production of enterprise in order to find out the reasons of high consumption and pollution discharge and the treatment program,The destination goal of cleaner production is to optimize the usage of resource minimize the discharge of waste materials and upgrade the capability of treatment of pollution of the enterprise.The technical principle are produced to dispose the waste material containing zinc using cleaner production,After the materials containing zinc can be recovered,therefore the amount of waste materials discharge can be decreased greatly.

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

    International Nuclear Information System (INIS)

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

  6. Interim report on status of containment integrity studies for continued in-tank storage of Hanford high-level defense waste

    International Nuclear Information System (INIS)

    This interim report supplements technical information reported in RHO-LD-52, Status of Containment Integrity Studies for Continued In-Tank Storage of Hanford Defense High-Level Waste, September 1978. Only new data from the continuing laboratory programs and those studies initiated in the past year are included. Analyses of waste tank concrete integrity continued through the year. Laboratory tests to determine the effect of long-term elevated temperatures on the strength and elastic properties of concrete showed that the modulus of elasticity, compressive strength, and splitting tensile strength continued to decrease as a function of temperature; Poisson's ratio was relatively unchanged. The durability tests of reinforced concrete specimens exposed to simulated waste chemicals showed no evidence of deterioration after 6 months exposure. Temperature cycling effects after 17 cycles showed little change in the compressive strength but a large reduction in the modulus of elasticity. A structural failure mode analysis was initiated to estimate the effect of constant dead load, elevated temperature, and aggressive chemicals on tank structural integrity after 100 years of waste storage. No results are presently available. A report is scheduled for completion in early 1980. An electron microscopy analysis was initiated to determine if microstructural changes in concrete can be detected which would provide a key for correlating relatively short-term laboratory data to predicting long-term structural behavior. Documented results are scheduled for late 1980

  7. System of large transport containers for waste from dismantling light water and gas-cooled nuclear reactors. Volume 1

    International Nuclear Information System (INIS)

    General descriptions of the main types of reactors in the European Economic Community are given, a series of reference plants selected for further study. Estimates are made of the radioactive decommissioning wastes for each, including neutron-activated and contaminated materials. Regulations governing the transport of radioactive materials, both international and national, are reviewed. (U.K.)

  8. An evaluation of supercompaction of drums containing solid low-level waste from Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Supercompaction and grouting technologies were demonstrated with solid LLW from Oak Ridge National Laboratory at the solid Waste Storage Area 5 (SWSA 5). The subcontractor used its mobile supercompaction system operating at 220 tons of compressive force to volume reduce 300 55-gal drums of solid LLW. The supercompaction of these drums resulted in a disposal capacity savings of about 85% of the original disposal capacity needs. The packaging of the compacted drums into 47 overpacks decreased the disposal capacity savings by about 19%. The net disposal capacity savings from the demonstration project is about 66% of the original, uncompacted waste volume. Based on the approximately $95K in direct costs, the supercompaction of the 2304 ft3 of waste processed cost about $41ft3 of uncompacted waste. Once the supercompaction unit was set up and operating, the incremental cost for the supercompaction services was only about $4ft3. The economic assessment for this project revealed that the cost-effectiveness of on-site demonstrations is very sensitive to the on-site support (non-vendor-related) costs. The minimum disposal costs for cost-effectiveness in this demonstration project was calculated to be about $18ft3 for no on-site support costs and about $180ft3 when the on-site support costs represented about 90% of the total demonstration project cost. (2 refs., 14 figs., 9 tabs.)

  9. Waste management with recourse: an inexact dynamic programming model containing fuzzy boundary intervals in objectives and constraints.

    Science.gov (United States)

    Tan, Q; Huang, G H; Cai, Y P

    2010-09-01

    The existing inexact optimization methods based on interval-parameter linear programming can hardly address problems where coefficients in objective functions are subject to dual uncertainties. In this study, a superiority-inferiority-based inexact fuzzy two-stage mixed-integer linear programming (SI-IFTMILP) model was developed for supporting municipal solid waste management under uncertainty. The developed SI-IFTMILP approach is capable of tackling dual uncertainties presented as fuzzy boundary intervals (FuBIs) in not only constraints, but also objective functions. Uncertainties expressed as a combination of intervals and random variables could also be explicitly reflected. An algorithm with high computational efficiency was provided to solve SI-IFTMILP. SI-IFTMILP was then applied to a long-term waste management case to demonstrate its applicability. Useful interval solutions were obtained. SI-IFTMILP could help generate dynamic facility-expansion and waste-allocation plans, as well as provide corrective actions when anticipated waste management plans are violated. It could also greatly reduce system-violation risk and enhance system robustness through examining two sets of penalties resulting from variations in fuzziness and randomness. Moreover, four possible alternative models were formulated to solve the same problem; solutions from them were then compared with those from SI-IFTMILP. The results indicate that SI-IFTMILP could provide more reliable solutions than the alternatives. PMID:20580864

  10. ENVIRONMENTAL RESEARCH BRIEF: WASTE REDUCTION ACTIVITIES AND OPTIONS FOR A MANUFACTURER OF PLASTIC CONTAINERS BY INJECTION MOLDING.

    Science.gov (United States)

    The U.S. Environmental Protection Agency (EPA) funded a project with the New Jersey Department of Environmental Protection and Energy (NJDEPE) to assist in conducting waste minimization assessments at thirty small- to medium-sized businesses in the state of New Jersey. ne of the ...

  11. Composite sulfidation process for treatment of waste water containing mercury%含汞废水复合硫化法处理工艺

    Institute of Scientific and Technical Information of China (English)

    杨兴娟; 史志伟; 李开明

    2013-01-01

    介绍了工业含汞废水常用的除汞方法---活性炭吸附法、化学沉淀法、离子交换法、还原法、微生物法,阐述了北京中科国益环保工程有限公司研发的复合硫化法(引入复合助剂IAMD-08)处理含汞废水的原理及工艺流程,列举了该工艺在化工企业的实际运行数据,含汞废水经处理后汞的质量浓度小于0.005 mg/L。%Commonly used processes for the treatment of industrial waste water containing mercury, such as active carbon adsorption method, chemical precipitation method, ion exchange method, reduc-tion method and microorganism method were introduced.The principle and flow of composite sulfidation process ( introducing complex additive IAMD-08 ) developed by Beijing China Sciences Environment Protection Co., Ltd.for the treatment of waste water containing mercury were discussed.The actual operation data of this process in chemical enterprises were listed, showing that the mercury content in the treated waste water was less than 0.005 mg/L.

  12. Investigations of the populations of introduced and resident micro-organisms in deep repositories and their effects on containment of radioactive wastes

    International Nuclear Information System (INIS)

    The potential role of microbes in the containment of radioactive waste has been described in several papers but can be briefly summarised as: (1) near field effects ((a) biodeterioration of repository structural materials; (b) mobilisation of waste); (2) far field effects ((a) change in ground water chemistry eg pH, Eh; (b) changes in radionuclide sorption characteristics (speciation, retardation)). The present study has sought to establish the presence of microbes in relevant formations, determine their potential role in disposal and make preliminary investigations on their ability to tolerate extreme environmental conditions and change migration characteristics. As the ultimate goal of this work is to try and model the effect of microorganisms on containment and radionuclide migration, an attempt to define the geochemical constraints on their growth has been attempted within the present study. Several sites of relevance to radioactive waste disposal have been sampled for microbiological content according to a protocol defined in pilot studies. These sites have been in Britain and in mainland Europe. Results have shown that the microbiota of each site is unique but that certain groups are of more importance eg sulphate reducing bacteria, sulphur oxidisers, and that their presence in future sampling should be ascertained as a first priority. (author)

  13. Hanford Site annual dangerous waste report: Volume 1, Part 2, Generator dangerous waste report, dangerous waste

    International Nuclear Information System (INIS)

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

  14. Hanford Site annual dangerous waste report: Volume 1, Part 2, Generator dangerous waste report, dangerous waste

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-12-31

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

  15. Effects of Clear and Amber Cullet on Physical and Mechanical Properties of Glass-Ceramics Containing Zinc Hydrometallurgy Waste

    Science.gov (United States)

    Hanpongpun, Wilasinee; Jiemsirilers, Sirithan; Thavorniti, Parjaree

    The effect of glass cullet on physical and mechanical properties of glass-ceramics developed from zinc hydrometallurgy waste and glass cullet was investigated. The glass-ceramics were prepared by mixing zinc hydrometallurgy waste with glass cullet through vitrification process. Two difference types of glass cullet (clear and amber cullet) were used. The parent glasses were ground and pressed into bars and sintered at low temperature (850°C) for 2 hours. The obtained glass-ceramics had low porosity. The glass-ceramics with clear cullet exhibited higher density and strength, comparing with the glass-ceramics with amber cullet. The type and the amount of the glass cullet present in the glass-ceramics have strong effect on their properties.

  16. Application of Recycled Concrete Aggregates Containing Waste Glass Powder/Suspension and Bottom Ash as a Cement Component in Concrete

    OpenAIRE

    Kara, P

    2013-01-01

    The growing environmental concerns and the increasing scarcity of landfills encourage the recycling of industrial wastes and adopting environmentally friendly practices by rational usage of natural resources. The production of concrete with recycled aggregate and reduced cement volume is the most desirable form of achieving a closed life cycle as an ecological constructional material. This paper describes results of a study undertaken to examine the influence of recycled aggregates obta...

  17. Surfactants containing radioactive run-offs: Ozone treatment, influence on nuclear power plants water waste special treatment

    International Nuclear Information System (INIS)

    The authors discuss the problems encountered in the efficiency of radioactive waste treatment in nuclear power plants in Kursk. The ozonization of aqueous solutions of surfactants was carried out in the laboratory's ozonization system. The surfactants which are discharged to the ion exchangers deteriorate resins, clog up the ion exchangers, and decrease filtration velocity. Therefore, this investigation focused on finding a method to increase the efficiency of this treatment process

  18. Defense Waste Processing Facility (DWPF) Viscosity Model: Revisions for Processing 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)

    2016-08-30

    Radioactive high level waste (HLW) at the Savannah River Site (SRS) has successfully been vitrified into borosilicate glass in the Defense Waste Processing Facility (DWPF) since 1996. Vitrification requires stringent product/process (P/P) constraints since the glass cannot be reworked once it is 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 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 guarantee, 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.

  19. The potential for using slags activated with near neutral salts as immobilisation matrices for nuclear wastes containing reactive metals

    Energy Technology Data Exchange (ETDEWEB)

    Bai, Y. [Immobilisation Science Laboratory, Department of Engineering Materials, Sir Robert Hadfield Building, University of Sheffield, Mappin Street, Sheffield S1 3JD (United Kingdom); Collier, N.C., E-mail: nick.collier@nnl.co.uk [Immobilisation Science Laboratory, Department of Engineering Materials, Sir Robert Hadfield Building, University of Sheffield, Mappin Street, Sheffield S1 3JD (United Kingdom); Milestone, N.B. [Immobilisation Science Laboratory, Department of Engineering Materials, Sir Robert Hadfield Building, University of Sheffield, Mappin Street, Sheffield S1 3JD (United Kingdom); Yang, C.H. [Department of Building Materials and Engineering, College of Materials and Engineering, Chongqing University, Chongqing 400045 (China)

    2011-06-30

    The UK currently uses composite blends of Portland cement and other inorganic cementitious material such as blastfurnace slag and pulverised fuel ash to encapsulate or immobilise intermediate and low level radioactive wastes. Typically levels up 9:1 blast furnace slag:Portland cement or 4:1 pulverised fuel ash:Portland cement are used. Whilst these systems offer many advantages, their high pH causes corrosion of various metallic intermediate level radioactive wastes. To address this issue, lower pH/weakly alkaline cementitious systems have to be explored. While the blast furnace slag:Portland cement system is referred to as a composite cement system, the underlying reaction is actually an indirect activation of the slag hydration by the calcium hydroxide generated by the cement hydration, and by the alkali ions and gypsum present in the cement. However, the slag also can be activated directly with activators, creating a system known as alkali-activated slag. Whilst these activators used are usually strongly alkaline, weakly alkaline and near neutral salts can also be used. In this paper, the potential for using weakly alkaline and near neutral salts to activate slag in this manner is reviewed and discussed, with particular emphasis placed on the immobilisation of reactive metallic nuclear wastes.

  20. Closure development for high-level nuclear waste containers for the tuff repository; Phase 1, Final report

    Energy Technology Data Exchange (ETDEWEB)

    Robitz, E.S. Jr.; McAninch, M.D. Jr.; Edmonds, D.P. [Babcock and Wilcox Co., Lynchburg, VA (USA). Nuclear Power Div.]|[Babcock and Wilcox Co., Alliance, OH (USA). Research and Development Div.

    1990-09-01

    This report summarizes Phase 1 activities for closure development of the high-level nuclear waste package task for the tuff repository. Work was conducted under U.S. Department of Energy (DOE) Contract 9172105, administered through the Lawrence Livermore National Laboratory (LLNL), as part of the Yucca Mountain Project (YMP), funded through the DOE Office of Civilian Radioactive Waste Management (OCRWM). The goal of this phase was to select five closure processes for further evaluation in later phases of the program. A decision tree methodology was utilized to perform an objective evaluation of 15 potential closure processes. Information was gathered via a literature survey, industrial contacts, and discussions with project team members, other experts in the field, and the LLNL waste package task staff. The five processes selected were friction welding, electron beam welding, laser beam welding, gas tungsten arc welding, and plasma arc welding. These are felt to represent the best combination of weldment material properties and process performance in a remote, radioactive environment. Conceptual designs have been generated for these processes to illustrate how they would be implemented in practice. Homopolar resistance welding was included in the Phase 1 analysis, and developments in this process will be monitored via literature in Phases 2 and 3. Work was conducted in accordance with the YMP Quality Assurance Program. 223 refs., 20 figs., 9 tabs.

  1. Research on changes of nitrate by interactions with metals under the wastes disposal environment containing TRU nuclide

    International Nuclear Information System (INIS)

    There exists the waste including a nitrate ion as a salt in the TRU waste materials. This nitrate ion can transferred to the nitrite ion and/or ammonia by reducing materials such as metals in the waste disposal environment, and has the possibility to affect on the disposal environment and nuclide transfer parameters. Therefore, electrochemical tests were conducted to evaluate the reaction rate parameters of the nitrate ion and metals under the low oxygen environment. The long-term reaction test using the glass-seal vessel was also conducted to grasp precisely the nitrate ion transition reaction rate and the gas generation rate caused by the reaction of metal and the nitrate ion coexist solution. (1) Reaction rate constants under various environments were obtained performing the potentiostatic holding tests with the parameters of the solution pH, temperature, and the nitrate and nitrite ion concentrations. The formula of the nitrate ion transition reaction rate was also examined based on these obtained data. (2) Conducting the immersion tests under the environment of the low oxygen and high-pH rainfall underground water site, the long-term reaction rate data were obtained on the reaction products (ammonia, hydrogen gas etc.) of metals (carbon steel, stainless steel and zircaloy etc.) with nitrate ion. The tests under the same conditions as in the past were also conducted to evaluate the test accuracy and error range of the long-term reaction test with the glass-seal vessels. (author)

  2. Process stability and microbial community structure in anaerobic hydrogen-producing microflora from food waste containing kimchi.

    Science.gov (United States)

    Jo, Ji Hye; Jeon, Che Ok; Lee, Dae Sung; Park, Jong Moon

    2007-09-15

    Hydrogen production by the dark fermentation of food wastes is an economic and environmentally friendly technology to produce the clean energy source as well as to treat the problematic wastes. However, the long-term operations of the continuous anaerobic reactor for fermentative hydrogen production were frequently unstable. In this study, the structure of microbial community within the anaerobic reactor during unstable hydrogen production was examined by denaturing gradient gel electrophoresis (DGGE) and terminal restriction fragment length polymorphism (T-RFLP) techniques. The changes in microbial community from H(2)-producing Clostridium spp. to lactic acid-producing Lactobacillus spp. were well coincident with the unexpected process failures and the changes of metabolites concentrations in the effluent of the anaerobic reactor. As the rate of hydrogen production decreased, effluent lactic acid concentration increased. Low rate of hydrogen production and changes in microbial community were related to the 'kimchi' content and storage temperature of food waste feed solution. After low temperature control of the storage tank of the feed solution, any significant change in microbial community within the anaerobic reactor did not occur and the hydrogen production was very stably maintained for a long time.

  3. The Canadian approach to microbial studies in nuclear waste management and disposal

    International Nuclear Information System (INIS)

    Many countries considering radioactive waste disposal have, or are considering programs to study and quantify microbial effects in terms of their particular disposal concept. Although there is an abundance of qualitative information, there is a need for quantitative data. Quantitative research should cover topics such as the kinetics of microbial activity in geological media, microbial effects on radionuclide migration in host rock (including effects of biofilms), tolerance to extreme conditions of radiation, heat and desiccation, microbially-influenced corrosion of waste containers and microbial gas production. The research should be performed in relevant disposal environments with the ultimate objective to quantify those effects that need to be included in models for predictive and safety assessment purposes. The Canadian approach to dealing with microbial effects involves a combination of pertinent, quantitative measurements from carefully designed laboratory studies and from large scale engineering experiments in AECL's Underground Research Laboratory (URL). The validity of these quantitative data is measured against observations from natural environments and analogues. An example is the viability of microbes in clay-based scaling materials. Laboratory studies have shown that the clay content of these barriers strongly affects microbial activity and movement. This is supported by natural environment and analogue observations that show clay deposits to contain very old tree segments and dense clay lenses in sediments to contain much smaller, less diverse and less active microbial populations than more porous sediments. This approach has allowed for focused, quantitative research on microbial effects in Canada. (author)

  4. Analysis on Prevention and Treatment Technology of Mercury-containing Waste Regeneration Pollution%含汞废物再生污染防治技术分析

    Institute of Scientific and Technical Information of China (English)

    甘露

    2014-01-01

    The paper briefly introduces utilization of pollution prevention and treatment technology during the application of the existing mercury-containing waste regeneration in China, and puts forward some suggestions for shortcomings of environmental protection measures in construction and operation of regeneration mercury enterprises.%简述目前我国含汞废物再生利用过程的污染防治技术应用情况,并对再生汞企业建设运营过程中环保措施方面的不足提出了建议。

  5. Structure formation of aerated concrete containing waste coal combustion products generated in the thermal vortex power units

    Science.gov (United States)

    Ivanov, A. I.; Stolboushkin, A. Yu; Temlyanstev, M. V.; Syromyasov, V. A.; Fomina, O. A.

    2016-10-01

    The results of fly ash research, generated in the process of waste coal combustion in the thermal vortex power units and used as an aggregate in aerated concrete, are provided. It is established that fly ash can be used in the production of cement or concrete with low loss on ignition (LOI). The permitted value of LOI in fly ash, affecting the structure formation and operational properties of aerated concrete, are defined. During non-autoclaved hardening of aerated concrete with fly ash aggregate and LOI not higher than 2%, the formation of acicular crystals of ettringite, reinforcing interporous partitions, takes place.

  6. Use of waste water containing hydrocarbons to produce energy in a modified gas turbine process; Energetische Verwendung kohlenwasserstoffhaltiger Abwaesser in einem modifizierten Gasturbinenprozess

    Energy Technology Data Exchange (ETDEWEB)

    Haep, S. [Inst. fuer Umwelttechnologie und Umweltanalytik e.V., Duisburg (Germany); Klaassen, S. [Schmeink und Cofreth Energie-Management GmbH, Bocholt (Germany); Leclaire, T. [Inst. fuer Umwelttechnologie und Umweltanalytik e.V., Duisburg (Germany); Roth, H. [Schmeink und Cofreth Energie-Management GmbH, Bocholt (Germany)

    1996-10-01

    The modified steam injection gas turbine process presented here, the so-called Cheng Cycle, makes it possible in principle to use waste water containing hydrocarbons to produce energy. In addition to being a means of waste disposal, this process reduces consumption of fossil fuel resources. The example dealt with in this article concerning the use of waste contaminated with alcohol from the production of non-alcoholic beer illustrates that the process under consideration can also have commercial advantages over conventional disposal. As the alcohol used is produced in a biological fermentation process and its exploitation for energy production is accordingly practically CO{sub 2} neutral, the use of the process in this application also represents a contribution towards reducing CO{sub 2} emissions that are harmful to the climate of the planet. (orig.) [Deutsch] Die energetische Verwertung kohlenwasserstoffhaltigen Abwassers ist grundsaetzlich durch Einsatz dieses Abwassers in dem hier vorgestellten modifizierten Gasturbinenprozess mit Dampfeinspritzung, dem sogenannten `Cheng-Cycle`, moeglich. Neben dem Entsorgungsaspekt lassen sich mit diesem Verfahren fossile Ressourcen schonen. Das in diesem Beitrag behandelte Beispiel der energetischen Nutzung von alkoholhaltigem Abwasser, das bei der Entalkoholisierung von Bier anfaellt, zeigt auf, dass das vorgestellte Verfahren auch unter wirtschaftlichen Aspekten gegenueber der konventionellen Entsorgung vorteilhaft sein kann. Da der verwertete Alkohol ueber einen biologischen Gaerprozess erzeugt wurde, seine energetische Nutzung damit praktisch CO{sub 2}-neutral ist, traegt die Anwendung des vorgestellten Verfahrens auf diesen Einsatzfall zusaetzlich zur Entlastung der Umwelt von klimaschaedigenden CO{sub 2}-Emissionen bei. (orig.)

  7. A glass-encapsulated calcium phosphate wasteform for the immobilization of actinide-, fluoride-, and chloride-containing radioactive wastes from the pyrochemical reprocessing of plutonium metal

    Science.gov (United States)

    Donald, I. W.; Metcalfe, B. L.; Fong, S. K.; Gerrard, L. A.; Strachan, D. M.; Scheele, R. D.

    2007-03-01

    Chloride-containing radioactive wastes are generated during the pyrochemical reprocessing of Pu metal. Immobilization of these wastes in borosilicate glass or Synroc-type ceramics is not feasible due to the very low solubility of chlorides in these hosts. Alternative candidates have therefore been sought including phosphate-based glasses, crystalline ceramics and hybrid glass/ceramic systems. These studies have shown that high losses of chloride or evolution of chlorine gas from the melt make vitrification an unacceptable solution unless suitable off-gas treatment facilities capable of dealing with these corrosive by-products are available. On the other hand, both sodium aluminosilicate and calcium phosphate ceramics are capable of retaining chloride in stable mineral phases, which include sodalite, Na 8(AlSiO 4) 6Cl 2, chlorapatite, Ca 5(PO 4) 3Cl, and spodiosite, Ca 2(PO 4)Cl. The immobilization process developed in this study involves a solid state process in which waste and precursor powders are mixed and reacted in air at temperatures in the range 700-800 °C. The ceramic products are non-hygroscopic free-flowing powders that only require encapsulation in a relatively low melting temperature phosphate-based glass to produce a monolithic wasteform suitable for storage and ultimate disposal.

  8. Sets of Reports and Articles Regarding Cement Wastes Forms Containing Alpha Emitters that are Potentially Useful for Development of Russian Federation Waste Treatment Processes for Solidification of Weapons Plutonium MOX Fuel Fabrication Wastes for

    Energy Technology Data Exchange (ETDEWEB)

    Jardine, L J

    2003-06-12

    This is a set of nine reports and articles that were kindly provided by Dr. Christine A. Langton from the Savannah River Site (SRS) to L. J. Jardine LLNL in June 2003. The reports discuss cement waste forms and primarily focus on gas generation in cement waste forms from alpha particle decays. However other items such as various cement compositions, cement product performance test results and some cement process parameters are also included. This set of documents was put into this Lawrence Livermore National Laboratory (LLNL) releasable report for the sole purpose to provide a set of documents to Russian technical experts now beginning to study cement waste treatment processes for wastes from an excess weapons plutonium MOX fuel fabrication facility. The intent is to provide these reports for use at a US RF Experts Technical Meeting on: the Management of Wastes from MOX Fuel Fabrication Facilities, in Moscow July 9-11, 2003. The Russian experts should find these reports to be very useful for their technical and economic feasibility studies and the supporting R&D activities required to develop acceptable waste treatment processes for use in Russia as part of the ongoing Joint US RF Plutonium Disposition Activities.

  9. GEANT4 Simulation of a Scintillating-Fibre Tracker for the Cosmic-ray Muon Tomography of Legacy Nuclear Waste Containers

    CERN Document Server

    Clarkson, Anthony; Hoek, Matthias; Ireland, David G; Johnstone, Russell; Kaiser, Ralf; Keri, Tibor; Lumsden, Scott; Mahon, David F; McKinnon, Bryan; Murray, Morgan; Nutbeam-Tuffs, Sian; Shearer, Craig; Staines, Cassie; Yang, Guangliang; Zimmerman, Colin

    2013-01-01

    Cosmic-ray muons are highly penetrative charged particles that are observed at sea level with a flux of approximately one per square centimetre per minute. They interact with matter primarily through Coulomb scattering, which is exploited in the field of muon tomography to image shielded objects in a wide range of applications. In this paper, simulation studies are presented that assess the feasibility of a scintillating-fibre tracker system for use in the identification and characterisation of nuclear materials stored within industrial legacy waste containers. A system consisting of a pair of tracking modules above and a pair below the volume to be assayed is simulated within the GEANT4 framework using a range of potential fibre pitches and module separations. Each module comprises two orthogonal planes of fibres that allow the reconstruction of the initial and Coulomb-scattered muon trajectories. A likelihood-based image reconstruction algorithm has been developed that allows the container content to be det...

  10. Rice Husk Ash to Stabilize Heavy Metals Contained in Municipal Solid Waste Incineration Fly Ash: First Results by Applying New Pre-treatment Technology

    Directory of Open Access Journals (Sweden)

    Laura Benassi

    2015-10-01

    Full Text Available A new technology was recently developed for municipal solid waste incineration (MSWI fly ash stabilization, based on the employment of all waste and byproduct materials. In particular, the proposed method is based on the use of amorphous silica contained in rice husk ash (RHA, an agricultural byproduct material (COSMOS-RICE project. The obtained final inert can be applied in several applications to produce “green composites”. In this work, for the first time, a process for pre-treatment of rice husk, before its use in the stabilization of heavy metals, based on the employment of Instant Pressure Drop technology (DIC was tested. The aim of this work is to verify the influence of the pre-treatment on the efficiency on heavy metals stabilization in the COSMOS-RICE technology. DIC technique is based on a thermomechanical effect induced by an abrupt transition from high steam pressure to a vacuum, to produce changes in the material. Two different DIC pre-treatments were selected and thermal annealing at different temperatures were performed on rice husk. The resulting RHAs were employed to obtain COSMOS-RICE samples, and the stabilization procedure was tested on the MSWI fly ash. In the frame of this work, some thermal treatments were also realized in O2-limiting conditions, to test the effect of charcoal obtained from RHA on the stabilization procedure. The results of this work show that the application of DIC technology into existing treatment cycles of some waste materials should be investigated in more details to offer the possibility to stabilize and reuse waste.

  11. Understanding long-term corrosion of Alloy 22 container in the potential Yucca Mountain repository for high-level nuclear waste disposal

    International Nuclear Information System (INIS)

    Alloy 22 (Ni-22Cr-13Mo-3W-4Fe) is the candidate material for the waste package outer container in a potential geologic repository for high-level nuclear waste disposal at Yucca Mountain, Nevada. This alloy exhibits very low corrosion rates in the absence of environmental conditions promoting crevice corrosion. However, there are uncertainties regarding Alloy 22's corrosion performance when general corrosion rates and susceptibility to crevice corrosion are extrapolated to a geological time period (e.g. 105 years). This paper presents an analysis of available literature information relevant to the long-term extrapolation of general corrosion processes and the crevice corrosion behavior of Alloy 22, under potential repository environments. For assessment of general corrosion rates, potential degradation processes causing the loss of the long-term persistence of passive film formed are considered. For crevice corrosion, induction time, and the extent of susceptibility and opening area, are considered. Disclaimer: The US Nuclear Regulatory Commission (NRC) staff views expressed herein are preliminary and do not constitute a final judgment or determination of the matters addressed nor of the acceptability of a license application for a geologic repository at Yucca Mountain. The paper describes work performed by the Center for Nuclear Waste Regulatory Analyses (CNWRA) for NRC under Contract Number NRC-02-02-012. The activities reported here were performed by CNWRA on behalf of the NRC office of Nuclear Material Safety and Safeguards, Division of High Level Waste Repository Safety. This paper is an independent product of the CNWRA and does not necessarily reflect the view or regulatory position of the NRC

  12. A New Comprehensive Recovery Technology for Gold- containing Waste Residue%含金废渣综合回收新技术研究

    Institute of Scientific and Technical Information of China (English)

    吴在玖

    2012-01-01

    Focusing on gold - containing waste residue produced in the terminal process of gold be-neficiation and metallurgy such as high carbon mud, carbon powder, waste slag and wastewater treatment pond mud etc, the technology of roasting - acid leaching - cyanidation was studied, the optimized condition was obtained and the process match with industrial technology of fluidized roasting - acid leaching - cyanidation was researched. The results showed that these two processes could match perfectly. By controlling the ratio of gold - containing waste residue to gold concentrate and matching corresponding dicyclic atomization feeding technology and calcination atmosphere, the recovery rate of Au and Cu was 94. 37% and 85. 07% respectively, which realized high efficiency, energy saving and environmental production.%针对选冶末端产生的高泥炭、碎炭末、废水沉渣、废水处理塘泥等含金废渣,研究开发了焙烧—酸浸—氰化工艺,获得了优化的工艺条件;并研究了该工艺与现有金精矿沸腾焙烧—酸浸—氰化工业生产工艺的匹配.结果表明,两种工艺能完美匹配,控制含金废渣与金精矿的合适配比,同时匹配相应的双环式雾化进料技术及焙烧气氛,Au、Cu的回收率分别达94.37%和85.07%,实现了高效、节能、环保工业化生产.

  13. Leaching characteristics of encapsulated controlled low-strength materials containing arsenic-bearing waste precipitates from refractory gold bioleaching.

    Science.gov (United States)

    Bouzalakos, S; Dudeney, A W L; Chan, B K C

    2016-07-01

    We report on the leaching of heavy elements from cemented waste flowable fill, known as controlled low-strength materials (CLSM), for potential mine backfill application. Semi-dynamic tank leaching tests were carried out on laboratory-scale monoliths cured for 28 days and tested over 64 days of leaching with pure de-ionised water as leachant. Mineral processing waste include flotation tailings from a Spanish nickel-copper sulphide concentrate, and two bioleach neutralisation precipitates (from processing at 35°C and 70°C) from a South African arsenopyrite concentrate. Encapsulated CLSM formulations were evaluated to assess the reduction in leaching by encapsulating a 'hazardous' CLSM core within a layer of relatively 'inert' CLSM. The effect of each bioleach waste in CLSM core and tailings in CLSM encapsulating medium, are assessed in combination and in addition to CLSM with ordinary silica sand. Results show that replacing silica sand with tailings, both as core and encapsulating matrix, significantly reduced leachability of heavy elements, particularly As (from 0.008-0.190 mg/l to 0.008-0.060 mg/l), Ba (from 0.435-1.540 mg/l to 0.050-0.565 mg/l), and Cr (from 0.006-0.458 mg/l to 0.004-0.229 mg/l), to below the 'Dutch List' of groundwater contamination intervention values. Arsenic leaching was inherently high from both bioleach precipitates but was significantly reduced to below guideline values with encapsulation and replacing silica sand with tailings. Tailings proved to be a valuable encapsulating matrix largely owing to small particle size and lower hydraulic conductivity reducing diffusion transport of heavy elements. Field-scale trials would be necessary to prove this concept of encapsulation in terms of scale and construction practicalities, and further geochemical investigation to optimise leaching performance. Nevertheless, this work substantiates the need for alternative backfill techniques for sustainable management of hazardous finely-sized bulk

  14. Resistance of Coatings for Boiler Components of Waste-to-Energy Plants to Salt Melts Containing Copper Compounds

    Science.gov (United States)

    Galetz, Mathias Christian; Bauer, Johannes Thomas; Schütze, Michael; Noguchi, Manabu; Cho, Hiromitsu

    2013-06-01

    The accelerating effect of heavy metal compounds on the corrosive attack of boiler components like superheaters poses a severe problem in modern waste-to-energy plants (WTPs). Coatings are a possible solution to protect cheap, low alloyed steel substrates from heavy metal chloride and sulfate salts, which have a relatively low melting point. These salts dissolve many alloys, and therefore often are the limiting factor as far as the lifetime of superheater tubes is concerned. In this work the corrosion performance under artificial salt deposits of different coatings, manufactured by overlay welding, thermal spraying of self-fluxing as well as conventional systems was investigated. The results of our studies clearly demonstrate the importance of alloying elements such as molybdenum or silicon. Additionally, the coatings have to be dense and of a certain thickness in order to resist the corrosive attack under these severe conditions.

  15. Utilization of sludge waste from natural rubber manufacturing process as a raw material for clay-ceramic production.

    Science.gov (United States)

    Vichaphund, S; Intiya, W; Kongkaew, A; Loykulnant, S; Thavorniti, P

    2012-12-01

    The possibility of utilization of the sludge waste obtained from the natural rubber manufacturing process as a raw material for producing clay ceramics was investigated. To prepared clay-based ceramic, the mixtures of traditional clay and sludge waste (10-30 wt%) were milled, uniaxilly pressed and sintered at a temperature between 1000 and 1200 degrees C. The effect of sludge waste on the properties of clay-based ceramic products was examined. The results showed that the amount of sludge waste addition had an effect on both sinterability and properties of the clay ceramics. Up to 30 wt% of sludge waste can be added into the clay ceramics, and the sintered samples showed good properties. PMID:23437647

  16. STARCH/PULP-FIBER BASED PACKAGING FOAMS AND CAST FILMS CONTAINING ALASKAN FISH BY-PRODUCTS (WASTE

    Directory of Open Access Journals (Sweden)

    Syed H. Imam

    2008-08-01

    Full Text Available Baked starch/pulp foams were prepared from formulations containing zero to 25 weight percent of processed Alaskan fish by-products that consisted mostly of salmon heads, pollock heads, and pollock frames (bones and associated remains produced in the filleting operation. Fish by-products thermoformed well along with starch and pulp fiber, and the foam product (panels exhibited useful mechanical properties. Foams with all three fish by-products, ranging between 10 and 15 wt%, showed the highest flexural modulus (500-770 Mpa. Above 20% fiber content, the modulus dropped considerably in all foam samples. Foam panels with pollock frames had the highest flexural modulus, at about 15% fiber content (770 Mpa. Foams with salmon heads registered the lowest modulus, at 25% concentration. Attempts were also made to cast starch-glycerol-poly (vinyl alcohol films containing 25% fish by-product (salmon heads. These films showed a tensile strength of 15 Mpa and elongation at break of 78.2%. All foams containing fish by-product degraded well in compost at ambient temperature (24oC, loosing roughly between 75-80% of their weight within 7 weeks. The films degraded at a much higher rate initially. When left in water, foams prepared without fish by-product absorbed water much more quickly and deteriorated faster, whereas, water absorption in foams with fish by-product was initially delayed and/or slowed for about 24 h. After this period, water absorption was rapid.

  17. Hazard of Fluorine Contained Waste Gas and Its Recovery/ Utilization%含氟废气的危害及其回收利用

    Institute of Scientific and Technical Information of China (English)

    周帼红; 柳惠平; 徐旺生

    2011-01-01

    针对磷化工生产排放的大量含氟废气对环境的污染问题,提出了以含氟废气为原料制取纳米白炭黑的实验研究方法;论述了氨化反应的基本原理、工艺流程特点以及白炭黑质量的影响因素;实验结果表明:用氟化铵溶液和氨吸收含氟废气得到氟硅酸铵溶液,进一步氨解得到高浓度的氟化铵溶液,由高浓度氟化铵可以制取系列高附加值的无机氟化物;该法克服了传统方法的不足,且2种吸收剂可以从后续加工产品中回收得到。%In allusion to pollution problems on environment from fluorine contained waste gas largely vented from chemical production of phosphor,author has presented the experimental and researching method to make the nanometer white carbon black taking the fluorine contained waste gas as raw material;has discussed the basic principle for ammonia reaction,process flow feature and effecting factors of white carbon black quality;experimental result indicates that the ammonium fluo silicate solution can be produced by using of absorbing fluorine contained waste gas by ammonium fluoride solution and ammonia solution,ammonium fluoride solution with high concentration can be gotten by ammonia decomposition,non-organic fluoride with highly added value can be made from ammonium fluoride solution with high concentration;this method can overcome the shortage of the traditional method,moreover the two kinds of absorbents can be obtained by recovery from consequent processing

  18. [alpha]-Decay damage effects in curium-doped titanate ceramic containing sodium-free high-level nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    Mitamura, Hisayoshi; Matsumoto, Seiichiro; Tsuboi, Takashi; Hashimoto, Masaaki; Togashi, Yoshihiro; Kanazawa, Hiroyuki (Japan Atomic Energy Research Inst., Ibaraki (Japan)); Stewart, M.W.A.; Vance, E.R.; Hart, K.P.; Ball, C.J. (Australian Nuclear Science and Technology Organization, Lucas Heights, New South Wales (Australia). Lucas Heights Research Labs.); White, T.J.

    1994-09-01

    A polyphase titanate ceramic incorporating sodium-free simulated high-level nuclear waste was doped with 0.91 wt% of [sup 224]Cm to accelerate the effects of long-term self-irradiation arising from [alpha] decays. The ceramic included three main constituent minerals: hollandite, perovskite, and zirconolite, with some minor phases. Although hollandite showed the broadening of its X-ray diffraction lines and small lattice parameter changes during damage in growth, the unit cell was substantially unaltered. Perovskite and zirconolite, which are the primary hosts of curium, showed 2.7% and 2.6% expansions, respectively, of their unit cell volumes after a dose of 12 [times] 10[sup 17] [alpha] decays[center dot]g[sup [minus]1]. Volume swelling due to damage in growth caused an exponential (almost linear) decrease in density, which reached 1.7% after a dose of 12.4 [times] 10[sup 17] [alpha] decays[center dot]g[sup [minus]1]. Leach tests on samples that had incurred doses of 2.0 [times] 10[sup 17] and 4.5 [times] 10[sup 17] [alpha] decays[center dot]g[sup [minus]1] showed that the rates of dissolution of cesium and barium were similar to analogous leach rates from the equivalent cold ceramic, while strontium and calcium leach rates were 2--15 times higher. Although the cerium, molybdenum, strontium, and calcium leach rates in the present material were similar to those in the curium-doped sodium-bearing titanate ceramic reported previously, the cesium leach rate was 3--8 times lower.

  19. Notice of construction work in tank farm waste transfer pit 244-TX double contained receiver-tank

    International Nuclear Information System (INIS)

    The following description and any attachments and references are provided to the Washington State Department of Health (WDOH), Division of Radiation Protection, Air Emissions and Defense Waste Section as a notice of construction (NOC) in accordance with Washington Administrative Code (WAC) 246-247, Radiation Protection - Air Emissions. WAC 246-247-060, ''Applications, registration, and licensing'', states ''This section describes the information requirements for approval to construct, modify, and operate an emission unit. Any NOC requires the submittal of information listed in Appendix A,'' Appendix A (WAC 246-247-1 10) lists the requirements that must be addressed. Additionally, the following description, attachments, and references are provided to the U.S. Environmental Protection Agency (EPA) as an NOC, in accordance with Title 40 Code of Federal Regulations (CFR), Part 61, ''National Emission Standards for Hazardous Air Pollutants.'' The information required for submittal to the EPA is specified in 40 CFR 61.07. The potential emissions from this activity are estimated to provide less than 0.1 millired year total effective dose equivalent to the hypothetical offsite maximally exposed individual, and commencement is needed within a short time. Therefore, this application also is intended to provide notification of the anticipated date of initial startup in accordance with the requirement listed in 40 CFR 61.09(a)(1), and it is requested that approval of this application also will constitute EPA acceptance of this initial startup notification. Written notification of the actual date of initial startup, in accordance with the requirement listed in 40 CFR 61.09(a)(2), will be provided later. The activities described in this NOC are estimated to provide a potential offsite (unabated) total effective dose equivalent (TEDE) to the hypothetical maximally exposed individual (MEI) of 2.36 E-02 millirem per year

  20. Use Of Ferrihydrite-coated Pozzolana And Biogenic Green Rust To Purify Waste Water Containing Phosphate And Nitrate

    Energy Technology Data Exchange (ETDEWEB)

    Ruby, Christian; Naille, Sebastien; Ona-Nguema, Georges; Morin, Guillaume; Mallet, M.; Guerbois, Delphine; Barthelemy, Kevin; Etique, Marjorie; Zegeye, Asfaw; Zhang, Yuhai; Boumaiza, Hella; Al-Jaberi, Muayad; Renard, Aurelien; Noel, Vincent; Binda, Paul; Hanna, Khalil; Despas, Christelle; Abdelmoula, Mustapha; Kukkadapu, Ravi K.; Sarrias, Joseph; Albignac, Magali; Rocklin, Pascal; Nauleau, Fabrice; Hyvrard, Nathalie; Genin, Jean-Marie

    2016-04-29

    The activated sludge treatments combined to the addition of ferric chloride is commonly used to eliminate nitrate and phosphate from waste water in urban area. These processes that need costly infrastructures are not suitable for rural areas and passive treatments (lagoons, reed bed filters…) are more frequently performed. Reed bed filters are efficient for removing organic matter but are not suitable for treating phosphate and nitrate as well. Passive water treatments using various materials (hydroxyapatite, slag…) were already performed, but those allowing the elimination of both nitrate and phosphate are not actually available. The goal of this work is to identify the most suitable iron based materials for such treatments and to determine their optimal use conditions, in particular in hydrodynamic mode. The reactivity of the iron based minerals was measured either by using free particles in suspension or by depositing these particles on a solid substrate. Pouzzolana that is characterized by a porous sponge-like structure suits for settling a high amount of iron oxides. The experimental conditions enabling to avoid any ammonium formation when green rust encounters nitrate were determined within the framework of a full factorial design. The process is divided into two steps that will be performed inside two separated reactors. Indeed, the presence of phosphate inhibits the reduction of nitrate by green rust and the dephosphatation process must precede the denitrification process. In order to remove phosphate, ferrihydrite coated pouzzolana is the best materials. The kinetics of reaction of green rust with nitrate is relatively slow and often leads to the formation of ammonium. The recommendation of the identified process is to favor the accumulation of nitrite in a first step, these species reacting much more quickly with green rust and do not transform into ammonium.