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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. Long-term modeling of glass waste in portland cement- and clay-based matrices

    International Nuclear Information System (INIS)

    A set of ''templates'' was developed for modeling waste glass interactions with cement-based and clay-based matrices. The templates consist of a modified thermodynamic database, and input files for the EQ3/6 reaction path code, containing embedded rate models and compositions for waste glass, cement, and several pozzolanic materials. Significant modifications were made in the thermodynamic data for Th, Pb, Ra, Ba, cement phases, and aqueous silica species. It was found that the cement-containing matrices could increase glass corrosion rates by several orders of magnitude (over matrixless or clay matrix systems), but they also offered the lowest overall solubility for Pb, Ra, Th and U. Addition of pozzolans to cement decreased calculated glass corrosion rates by up to a factor of 30. It is shown that with current modeling capabilities, the ''affinity effect'' cannot be trusted to passivate glass if nuclei are available for precipitation of secondary phases that reduce silica activity

  4. Long-term modeling of glass waste in portland cement- and clay-based matrices

    Energy Technology Data Exchange (ETDEWEB)

    Stockman, H.W.; Nagy, K.L. [Sandia National Labs., Albuquerque, NM (United States); Morris, C.E. [Wollongong Univ., NSW (Australia). Dept. of Civil and Mining Engineering

    1995-12-01

    A set of ``templates`` was developed for modeling waste glass interactions with cement-based and clay-based matrices. The templates consist of a modified thermodynamic database, and input files for the EQ3/6 reaction path code, containing embedded rate models and compositions for waste glass, cement, and several pozzolanic materials. Significant modifications were made in the thermodynamic data for Th, Pb, Ra, Ba, cement phases, and aqueous silica species. It was found that the cement-containing matrices could increase glass corrosion rates by several orders of magnitude (over matrixless or clay matrix systems), but they also offered the lowest overall solubility for Pb, Ra, Th and U. Addition of pozzolans to cement decreased calculated glass corrosion rates by up to a factor of 30. It is shown that with current modeling capabilities, the ``affinity effect`` cannot be trusted to passivate glass if nuclei are available for precipitation of secondary phases that reduce silica activity.

  5. 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

    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 of...

  6. Practical designs for clay based engineered barrier systems for heat emitting radioactive wastes

    International Nuclear Information System (INIS)

    Many of the present designs of repositories for radioactive wastes derive from generic feasibility studies which emphasize post-closure safety. These include little (or no) treatment of the practicality of safe and quality-assured construction of engineered barriers under the conditions (humidity, dust, etc.) and requirements (emplacement rate, remote handling, etc.) of an operational underground facility. Indeed, as soon as attempts are made to demonstrate such concepts in-situ at full scale, considerable practical problems are encountered and, in many cases, additional engineering components are introduced (liners, borehole caps, grouts, rock-bolts, drainage systems, etc.) which could be detrimental to - or at leas t complicate - the long-term safety case. As the discrepancy between the idealized concepts illustrated in performance assessment and the actual systems which are shown to be feasible grows, there is a critical need for design rationalization. Such a process needs to include careful balancing of factors influencing safety during the operational phase - which should not be compromised - with those which contribute to potential hazards which occur only in the distant future. Apart from such almost philosophical considerations, the robustness of the EBS construction procedure to possible operational perturbations needs serious consideration. Even if closed and sealed repositories are very insensitive to disruptive events such as earthquakes, hurricanes, industrial actions and terrorist actions, the operational system may be more vulnerable to perturbation. Designs should be introduced which, to the greatest extent possible, not only fail safe, but are also easy to remedy (or reverse) in case the assurance of EBS quality is lost. This paper will expand on ideas for a second generation of clay-based EBS designs, which are both practical and safe. Associated requirements for R and D and performance assessment model development will also be outlined, with a

  7. 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

  8. 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...

  9. 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

  10. Underground nuclear waste containments

    International Nuclear Information System (INIS)

    In the United States, about a hundred million gallons of high-level nuclear waste are stored in underground containments. Basically, these containments are of two different designs: single-shell and double-shell structures. The single-shell structures consist of reinforced concrete cylindrical walls seated on circular mats and enclosed on top with torispherical domes or circular flat roofs. The walls and the basemats are lined with carbon steel. The double-shell structures provide another layer of protection and constitute a completely enclosed steel containment within the single-shell structure leaving an annular space between the two walls. Single-shell containments are of earlier vintage and were built in the period 1945-1965. Double-shell structures were built through the 1960s and 1970s. Experience gained in building and operating the single-shell containments was used in enhancing the design and construction of the double-shell structures. Currently, there are about 250 underground single-shell and double-shell structures containing the high-level waste with an inventory of about 800 million curies. During their service lives, especially in early stages, these structures were subjected to thermal excursions of varying extents; also, they have aged in the chemical environment. Furthermore, in their remaining service lives, the structures may be subjected to loads for which they were not designed, such as larger earthquakes or chemical explosions. As a result, the demonstration of safety of these underground nuclear containments poses a challenge to structural engineers, which increases with time. Regardless of current plans for gradual retrieval of the waste and subsequent solidification for disposal, many of these structures are expected to continue to contain the waste through the next 20-40 years. In order to verify their structural capabilities in fulfilling this mission, several studies were recently performed at Brookhaven National Laboratory

  11. 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 ...

  12. Occurrence and identification of microorganisms in compacted clay-based buffer material designed for use in a nuclear fuel waste disposal vault

    International Nuclear Information System (INIS)

    A full-scale nuclear fuel waste disposal container experiment was carried out 240 m below ground in an underground granitic rock research laboratory in Canada. An electric heater was surrounded by buffer material composed of sand and bentonite clay and provided heat equivalent to what is anticipated in a Canadian nuclear fuel waste repository. During the experiment, the heat caused a mass transport of water and moisture content gradients developed in the buffer ranging from 13% closest to the beater to 23% at the rock wall of the deposition hole. Upon decommissioning after 2.5 years, microorganisms could be cultured from all samples having a moisture content above 15% but not from samples with a moisture content below 15%. Heterotrophic aerobic and anaerobic bacteria were found in numbers ranging from 101 to 106 cells/g dry weight buffer. Approximately 102, or less, sulphate-reducing bacteria and methanogens per grain of dry weight buffer were also found. Identification of buffer population members was performed using Analytical Profile Index (API) strips for isolated bacteria and 16S rRNA gene sequencing for in situ samples. A total of 79 isolates from five buffer layers were identified with API strips as representing the beta, gamma and delta groups of Proteobacteria and Gram-positive bacteria. Sixty-seven 16S rRNA clones that were obtained from three buffer layers were classified into 21 clone groups representing alpha and gamma groups of Proteobacteria, Gram-positive bacteria, and a yeast. Approximately 20% of the population comprised Gram-positive bacteria. Members of the genera Amycolatopsis, Bacillus, and Nocardia predominated. Among Gram-negative bacteria, the genera Acinetobacter and Pseudomonas predominated. Analysis of lipid biomarker signatures and in situ leucine uptake demonstrated that the buffer population was viable. The results suggest that a nuclear fuel waste buffer will be populated by active microorganisms only if the moisture content is above

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

    International Nuclear Information System (INIS)

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

  14. Properties of radioactive wastes and waste containers

    International Nuclear Information System (INIS)

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

  15. 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)

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

    DEFF Research Database (Denmark)

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

    2015-01-01

    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. The...

  17. 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.

  18. Treatment of solid waste containing 226Ra

    International Nuclear Information System (INIS)

    This work is directed to the treatment of radioactive solid waste containing mainly radium (226Ra) produced from oil and gas production industries in Egypt. The treatment process has been carried out by suspending the clay fraction content in the solid waste in suitable leaching solutions. These compremise aqueous saline solution and aqueous saline solutions containing certain additives, namely, Washing Powder (W.P.), Shell and Span 20 surfactants. Treatment with saline solution containing either W.P. or Shell surfactants, showed an enhancement in the removal of 226Ra compared to that with saline solution alone or containing Span 20. Factors affect the treatment process have been investigated and discussed. The removal percentage of 226Ra was found to depend on the clay fines content in the solid waste. Further sequential treatment schemes have been tested and optimized

  19. Characterization of Hanford tank wastes containing ferrocyanides

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-02-01

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

  20. Characterization of Hanford tank wastes containing ferrocyanides

    International Nuclear Information System (INIS)

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

  1. 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.

  2. 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)

  3. 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

  4. 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

  5. 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

  6. Burning treatment of phosphorus-containing wastes

    International Nuclear Information System (INIS)

    Granular solids such as active carbon wastes containing tributylphosphate (TBP) generated from a nuclear power plant are drained if they are wetted and then basic compounds represented by calcium salts such as Ca(OH)2 or CaO and magnesium salts such as Mg(OH)2 or MgO are excessively deposited at the outer surface of the wastes. Deposition may be attained by charging the glanular solids and the basic compounds into a stirrer and mixing and stirring them homogenously. They are put to a movable stirring type incinerator using a screw feeder and burnt out at 600 to 800degC under stirring. Then, the active carbon is burnt and the phosphorous compounds such as TBP can be converted into stable and safe materials such as calcium phosphate and magnesium phosphate. A great devoluming ratio can be obtained, as well as the off-gas systems are not corroded. (T.M.)

  7. 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)

  8. 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

  9. An analysis of the factors affecting the hydraulic conductivity and swelling pressure of Kyungju ca-bentonite for use as a clay-based sealing material for a high level waste repository

    International Nuclear Information System (INIS)

    The buffer and backfill are important components of the engineered barrier system in a high-level waste repository, which should be constructed in a hard rock formation at a depth of several hundred meters below the ground surface. The primary function of the buffer and backfill is to seal the underground excavation as a preferred flow path for radionuclide migration from the deposited high-level waste. This study investigates the hydraulic conductivity and swelling pressure of Kyungju Ca-bentonite, which is the candidate material for the buffer and backfill in the Korean reference high-level waste disposal system. The factors that influence the hydraulic conductivity and swelling pressure of the buffer and backfill are analyzed. The factors considered are the dry density, the temperature, the sand content, the salinity and the organic carbon content. The possibility of deterioration in the sealing performance of the buffer and backfill is also assessed.

  10. Radioactive waste containing method and containing vessel for underground processing

    International Nuclear Information System (INIS)

    An overpack has spherical seats at the openings of a vessel main body, and a lid is fitted to the spherical seat as a bottom and mounted by a welded portion. The bottom and the outer circumferential surface of the overpack are previously covered by a body covering material, an end covering material, a lid covering material and an annular closing plate. A containing vessel (canister) after glass-solidification of radioactive liquid wastes is charged from an upper opening of the overpack. A lid is fitted to the upper spherical seat, and a welded portion is formed to seal the canister. An end covering material is disposed on the upper end face of the vessel main body, and a seal-welding portion is formed to the joining portion with the body covering material to integrate them. In addition, the annular closing plate is fitted between the end covering material and the lid covering material, and then a seal-welding portion is formed to cover the entire surface of the vessel main body and the lids. (I.N.)

  11. 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.

  12. Waste container fabrication from recycled DOE metal

    International Nuclear Information System (INIS)

    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 reuseclose 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

  13. 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.

  14. 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

  15. MAVL wastes containers functional demonstration and associated tests program

    International Nuclear Information System (INIS)

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

  16. 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

  17. Process for treating waste water containing radioactive substances

    International Nuclear Information System (INIS)

    A process for treating waste water containing radioactive substances comprising treating the waste water by reverse osmosis in the presence of at least one organic surfactant selected from the group consisting of anionic surfactants, cationic surfactants and nonionic surfactants

  18. 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)

  19. 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)

  20. 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.

  1. 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.

  2. 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.

  3. Defense High Level Waste Disposal Container System Description Document

    Energy Technology Data Exchange (ETDEWEB)

    N. E. Pettit

    2001-07-13

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

  4. Corrosion of metal containers containing cemented radioactive wastes

    International Nuclear Information System (INIS)

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

  5. 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

  6. 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

  7. Corrosion process studies in a nuclear waste container

    International Nuclear Information System (INIS)

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

  8. Container for transport and for storage of biologically damaging waste

    International Nuclear Information System (INIS)

    A container is described for the transport and storage of biologically damaging, particularly radioactive waste, in which it is possible to fill the inside of the container with gas without using complicated shut-off devices. (orig./PW)

  9. 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

  10. 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

  11. 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

  12. 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)

  13. 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)

  14. 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.)

  15. 1995 solid waste 30-year container volume summary

    International Nuclear Information System (INIS)

    This report describes a 30-year forecast of the solid waste volumes by container category. The volumes described are low-level mixed waste (LLMW) and transuranic/transuranic mixed (TRU-TRUM) waste. These volumes and their associated container categories will be generated or received at the US Department of Energy Hanford Site for storage, treatment, and disposal at Westinghouse Hanford Company's Solid Waste Operations Complex (SWOC) during a 30-year period from FY 1995 through FY 2024. The data presented in this report establish a baseline for solid waste management both in the present and future. With knowledge of the volumes by container type, decisions on the facility handling and storage requirements can be adequately made. It is recognized that the forecast estimates will vary as facility planning and missions continue to change and become better defined; however, the data presented in this report still provide useful insight into Hanford's future solid waste management requirements

  16. EVALUATION OF HDPE CONTAINERS FOR MACROENCAPSULATION OF MIXED WASTE DEBRIS

    Energy Technology Data Exchange (ETDEWEB)

    Eaton, David; Carlson, Tim; Gardner, Brad; Bushmaker, Robert; Battleson, Dan; Shaw, Mark; Bierce, Lawrence

    2003-02-27

    Macroencapsulation is currently available at facilities permitted by the U.S. Environmental Protection agency for the treatment of radioactively contaminated hazardous waste. The U.S. Department of Energy is evaluating the use of high-density polyethylene containers to provide a simpler means of meeting macroencapsulation requirements. Macroencapsulation is used for the purpose of isolating waste from the disposal environment in order to meet the Land Disposal Restriction treatment standards for debris-like waste. The containers being evaluated have the potential of providing a long-term reduction in the leachability and subsequent mobility of both the hazardous and radioactive contaminants in this waste while at the same allowing treatment by the generator as the waste is being generated. While the testing discussed in this paper shows that further developmental work is necessary, these tests also indicate that these containers have the potential to reduce the cost, schedule, and complexity of meeting the treatment standard for mixed waste debris.

  17. Technologies for the containment, immobilization and disposal of radioactive wastes

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited is developing methods for the management and safe disposal of radioactive wastes. These wastes range from the highly radioactive (high-level) UO2 fuel arising from the nuclear generation of electrical power to the low- and intermediate-level wastes arising from research in various Canadian institutions using radioactive isotopes. This report will review the research programs in place on materials and processes for the immobilization and containment of UO2 fuel wastes and the technical aspects of programs demonstrating the various technologies needed for implementing a disposal program for low-level wastes. 90 refs

  18. 3.6. Chlorination of alumina containing waste products

    International Nuclear Information System (INIS)

    Chlorination of alumina containing waste products is considered in this article. Based on conducted studies following optimal conditions of chlorination of alumina containing waste products with reducer - coal were found: temperature - 750-850 deg C, chlorination duration -1-1,5 hours, quantity of reducer - 30% and size of particles - 0,1 mm. Based on conducted studies following optimal conditions of chlorination of alumina containing waste products with reducer - natural gas were found: temperature - 650-750 deg C, chlorination duration - 2 hours, chlorine to methane ratio is 4:1 and size of particles - 0,2-0,3 mm.

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

    International Nuclear Information System (INIS)

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

  20. 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

  1. Organic material in clay-based buffer materials and its potential impact on radionuclide transport

    International Nuclear Information System (INIS)

    AECL has submitted an Environmental Impact Statement (EIS) to evaluate the concept of nuclear fuel disposal at depth in crystalline rock of the Canadian Shield. In this disposal concept used fuel would be emplaced in corrosion-resistant containers which would be surrounded by clay-based buffer and backfill materials. Once groundwater is able to penetrate the buffer and corrosion-resistant container, radionuclides could be transported from the waste form to the surrounding geosphere, and eventually to the biosphere. The release of radionuclides from the waste form and their subsequent transport would be determined by the geochemistry of the disposal vault and surrounding geosphere. Organic substances affect the geochemistry of radionuclides through complexation reactions that increase solubility and alter mobility, by affecting the redox of certain radionuclides and by providing food for microbes. The purpose of this study was to determine whether the buffer and backfill materials proposed for use in a disposal vault contain organics that could be leached by groundwater in large enough quantities to complex with radionuclides and affect their mobility within the disposal vault and surrounding geosphere. Buffer material, made from a mixture of 50 wt.% Avonlea sodium bentonite and 50 wt.% silica sand, was extracted with deionized water to determine the release of dissolved organic carbon, humic acid and fulvic acid. The effect of radiation and heat from the used fuel was simulated by treating samples of buffer before leaching to various amounts of heat (60 deg C and 90 deg C) for periods of 2, 4 and 6 weeks, and to ionizing radiation with doses of 25 kGy and 50 kGy. Humic substances were isolated from the leachates to determine the concentrations of humic and fulvic acids and to determine their functional group content by acid-base titrations. The results showed that groundwater would leach significant amounts of organics that would complex with radionuclides such as

  2. Device for filling a container with radioactive wastes

    International Nuclear Information System (INIS)

    Object: To fill a container with radioactive wastes while lowering a level in the container in response to the filled quantity of the radioactive wastes in the container to thereby prevent destruction of the pellet-like radioactive wastes and prevent the scattering of powdery material. Structure: Pellet-like radioactive wastes conveyed by a conveyor are thrown into a drum can located at the top end through a hopper. Load of the drum can is detected by a load cell to transmit a signal to an adjuster. The adjuster causes a vertically driving electric motor to be run in response to said detection signal. When the rotative shaft is rotated, a threaded rod is rotated through a bevel gear so that a threaded nut in engagement with the threaded rod is moved down. Accordingly, a bed for receiving the drum can moves down as the pellet-like radioactive wastes is filled to maintain the dropping height of pellets constant. (Yoshino, Y.)

  3. 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

  4. Correlation of container exposure rates with radioisotopic inventory of waste

    International Nuclear Information System (INIS)

    The Radioactive Waste Management Department (GRR) at the Nuclear and Energy Research Institute (IPEN) develops methods for determining the radioactive inventory of waste packages and waste streams. These methods involve radioanalysis of waste samples, gamma scanning of waste drums and calculations. The determination of the radioisotopic content of nuclear waste is a basic step in the waste characterization process and is essential in the treatment, in the transportation, and in the disposal of the waste. While radiochemical analysis of the wastes yields the most accurate results, an expedite method based on the measurement of exposure rates around waste packages can provide good approximations when previous information on gamma emitters present are available. Results of exposure rates can be used as a quick check of content in waste drum consignment and as an initial estimate of activity of fresh waste packages. The point-kernel source method is used to correlate exposure rate measurement results with gamma emission rate. Exposure rates calculated with this method are compared with results of measurements of drums containing spent ion-exchange resin replaced from the IEA-R1 research reactor water treatment system. (author)

  5. 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.

  6. 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)

  7. 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.

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

    International Nuclear Information System (INIS)

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

  9. Research into Smell Emitted by Containers for Public Waste

    Directory of Open Access Journals (Sweden)

    Tadas Lukauskas

    2012-12-01

    Full Text Available 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 and three underground Molok containers at an outside temperature of 0 °C and depending on the replenishment of containers. The performed analysis has shown that the strongest smell of household waste is detected from completely replenished containers. Open containers have a distinctive feature of releasing a strong smell - in one of those, the odour strength of 119 OUE/m3 has been determined.Article in Lithuanian

  10. 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

  11. 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...

  12. 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...

  13. 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.

  14. Optimized concrete containers for waste management

    International Nuclear Information System (INIS)

    The least eight years, Siempelkamp verified different types of containers, partly with concrete shielding according to the requirements of German Transportation Regulations as well as to the acceptance criteria for the planned German final storage site, Schacht Konrad. After the process of verifying, more than 1,000 containers of different cubic and cylindrical shape with concrete shielding were produced by Siempelkamp. In the course of using the containers, certain improvements possibilities became obvious in order to fulfil the handling and transportation requirements. These improvements are based on: Protection against improper operational handling on site. Improvement of surface protection. Optimization of the lid fixation. Furthermore, Siempelkamp developed a concept to adjust non-qualified containers to achieve the requirements for transportation and storage. In consideration of a long term interim storage of the containers, improving the containers allows not only a better handling but it also provides an adequate basis for a safe depostion in an interim storage as well as for later final storage

  15. 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 ...

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

    International Nuclear Information System (INIS)

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

  17. 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

  18. The design of high level radioactive waste disposal container

    International Nuclear Information System (INIS)

    The utilization of nuclear energy can produce radioactive wastes. The disposal of wastes is highly concerned, especially for the high level radioactive wastes (HLW) which are characterized by nuclides of very high initial radioactivity, large thermal emissivity and the long life-term. The deep geological disposal is regarded as the most reasonable and effective way to safely dispose HLW in the world. The conceptual model of HLW geological disposal in China is based on a multi-barrier system which combines an isolating geological environment with an engineering barrier system (EBS). The EBS include the HLW, HLW canister, disposal container, buffer materials and backfill materials. The disposal container is the most important barrier for isolating the HLW from the surroundings owing to the integrity and corrosion proof of the container. According to the character of the content, two types of container BV and BG are discussed in this paper, focusing on the material, structure and disposal method of these containers. (authors)

  19. Combined decontamination processes for wastes containing PCBs

    International Nuclear Information System (INIS)

    This project has focused on the development of a complex assembly of mutually corresponding technological units:-a low temperature thermal process for the desorption of PCBs and other organics from soils and other contaminated solid wastes;-the extraction of PCBs from soils by an ecological friendly aqueous solution of selected surfactants;-the chemical decontamination of PCBs in oils and in oil-in-water emulsions by metallic sodium and potassium in polyethylene glycols in the presence of aluminum powder;-the modified alkaline catalyzed chemical decontamination of PCBs in oils and in oil-in-water dispersions in a solid-state reactor (in a film of reacting emulsion on solid carriers); and-the breakdown of PCBs in aqueous emulsions with activated hydroxyl radicals enhanced by UV radiation The processes operate in a closed loop configuration with effluents circulating among the process unit. These technologies have been verified at laboratory and pilot-plant scales

  20. Method of processing nitrate-containing radioactive liquid wastes

    International Nuclear Information System (INIS)

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

  1. Method of refining uranium-containing waste oils

    International Nuclear Information System (INIS)

    Purpose: To economically refine radioactive waste oils containing uranium fluorides as reprocessed oils usable as lubricating oils for an oil-lubricated rotational vacuum pump. Method: Waste lubircating oils containing uranium fluorides are filtered through column-like filtration material in a shape of ultra-fine long fibers such as α-cellulose or the like to remove contamination substances of a size above several microns. Then, the filtered waste oils are incorporated with active white clay, stirred and filtered again. Since the white clay used in the above processing is inexpensive, reprocessed oils reusable as lubricating oils for the oil-lubricated rotational vacuum pump can be refined economically from radioactive waste oils containing uranium fluorides. (Moriyama, K.)

  2. 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.

  3. 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

  4. Microanalysis of clay-based pigments by XRD techniques

    Czech Academy of Sciences Publication Activity Database

    Hradil, David; Bezdička, Petr; Hradilová, J.

    Catania : Technart, 2015. "O-48". [Technart 2015 : non-destructive and microanalytical techniques in art and cultural heritage. 27.04.2015-30.04.2015, Catania] R&D Projects: GA ČR GA14-22984S Keywords : micro-XRD * clay-based pigments * paintings Subject RIV: CA - Inorganic Chemistry http://technart2015.lns.infn.it/images/BoA.pdf

  5. 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

  6. Leach and radiolysis data for FUETAP concretes containing SRP wastes

    International Nuclear Information System (INIS)

    This supplement to ORNL/TM-8579 contains experimental results for leach tests and alpha-radiolysis tests made on FUETAP concretes containing Savannah River Plant waste. The results, presented in two sections, consist of both the raw data and calculated values for individual experiments. This information is summarized and analyzed in Sections 5 and 7 of ORNL/TM-8579

  7. 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

  8. 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.)

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

    International Nuclear Information System (INIS)

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

  10. 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

  11. Acid digestion of chlorine-containing wastes, (2)

    International Nuclear Information System (INIS)

    In the Plutonium Fuel Fabrication Facility, about 40% of the alpha-contaminated solid wastes contains organic chlorides, mainly PVC sheets and chloroprene rubber gloves. Acid digestion has been developed to reduce the volume of alpha-contaminated wastes, while converting the wastes into stable nonreactive residue. Based on the results of the basic studies on the acid digestion for nonradioactive chlorine-containing wastes, a non-radioactive pilot plant equipped with a 200 l digester was designed and constructed to confirm the performance in scaled-up process and the engineering problems. The following matters are described: the pilot plant of nonradioactive acid digestion and the experiments performed with it to confirm the reproducibility of the results of basic studies and to verify the safety of the processes. Many useful results were able to be obtained for the nitric acid oxidation processes. (J.P.N.)

  12. 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.)

  13. 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

  14. LLNL/YMP Waste Container Fabrication and Closure Project

    International Nuclear Information System (INIS)

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

  15. Disposal of water treatment wastes containing arsenic - A review

    International Nuclear Information System (INIS)

    Solid waste management in developing countries is often unsustainable, relying on uncontrolled disposal in waste dumps. Particular problems arise from the disposal of treatment residues generated by removing arsenic (As) from drinking water because As can be highly mobile and has the potential to leach back to ground and surface waters. This paper reviews the disposal of water treatment wastes containing As, with a particular emphasis on stabilisation/solidification (S/S) technologies which are currently used to treat industrial wastes containing As. These have been assessed for their appropriateness for treating As containing water treatment wastes. Portland cement/lime mixes are expected (at least in part) to be appropriate for wastes from sorptive filters, but may not be appropriate for precipitative sludges, because ferric flocs often used to sorb As can retard cement hydration. Brine resulting from the regeneration of activated alumina filters is likely to accelerate cement hydration. Portland cement can immobilise soluble arsenites and has been successfully used to stabilise As-rich sludges and it may also be suitable for treating sludges generated from precipitative removal units. Oxidation of As(III) to As(V) and the formation of calcium-arsenic compounds are important immobilisation mechanisms for As in cements. Geopolymers are alternative binder systems that are effective for treating wastes rich in alumina and metal hydroxides and may have potential for As wastes generated using activated alumina. The long-term stability of cemented, arsenic-bearing wastes is however uncertain, as like many cements, they are susceptible to carbonation effects which may result in the subsequent re-release of As.

  16. 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.

  17. 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...

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  20. 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

  1. 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.

  2. Precipitation and Deposition of Aluminum Containing Species in Tank Wastes

    International Nuclear Information System (INIS)

    Aluminum-containing phases represent the most prevalent solids that can appear or disappear during the processing of radioactive tank wastes. Processes such as sludge washing and leaching are designed to dissolve Al-containing phases and, thereby, minimize the volume of high-level waste glass required to encapsulate radioactive sludges. Unfortunately, waste-processing steps that include evaporation can involve solutions that are supersaturated with respect to cementitious aluminosilicates that result in unwanted precipitation and scale formation. Of all the constituents of tank waste, limited solubility cementitious aluminosilicates have the greatest potential for clogging pipes and transfer lines, fouling highly radioactive components such as ion exchangers, and completely shutting down processing operations. For instance, deposit buildup and clogged drain lines experienced during the tank waste volume-reduction process at Savannah River Site (SRS) required an evaporator to be shut down in October 1999. The Waste Processing Technology Section (WPTS) of Westinghouse Savannah River Company (WSRC) at SRS is now collaborating with team members from Pacific Northwest National Laboratory (PNNL) to verify the thermodynamic stability of aluminosilicate compounds under waste tank conditions in an attempt to solve the deposition and clogging problems. The primary goals of this study are to understand the (1) the major factors controlling precipitation, heterogeneous nucleation and growth phenomena, of relatively insoluble aluminosilicates, (2) role of organics for inhibiting aluminosilicate formation, and (3) to develop a predictive tool to control precipitation, scale formation, and cementation under tank waste processing conditions. The results obtained from this will provide crucial information for (1) avoiding problematical sludge processing steps, and (2) identifying and developing effective technologies to process retrieved sludges and supernatants before ultimate

  3. Precipitation and Deposition of Aluminum Containing Species in Tank Wastes

    International Nuclear Information System (INIS)

    Aluminum-containing phases represent the most prevalent solids that can appear or disappear during the processing of radioactive tank wastes. Processes such as sludge washing and leaching are designed to dissolve Al-containing phases and, thereby, minimize the volume of high-level waste glass required to encapsulate radioactive sludges. Unfortunately, waste-processing steps that include evaporation can involve solutions that are supersaturated with respect to cementitious aluminosilicates that result in unwanted precipitation and scale formation. Of all the constituents of tank waste, limited solubility cementitious aluminosilicates have the greatest potential for clogging pipes and transfer lines, fouling highly radioactive components such as ion exchangers, and completely shutting down processing operations. For instance, deposit buildup and clogged drain lines experienced during the tank waste volume-reduction process at Savannah River Site (SRS) required an evaporator to b e shut down in October 1999. The Waste Processing Technology Section (WPTS) of Westinghouse Savannah River Company (WSRC) at SRS is now collaborating with team members from Pacific Northwest National Laboratory (PNNL) to verify the thermodynamic stability of aluminosilicate compounds under waste tank conditions in an attempt to solve the deposition and clogging problems. The primary goals of this study are to understand the (1) the major factors controlling precipitation, heterogeneous nucleation and growth phenomena, of relatively insoluble aluminosilicates, (2) role of organics for inhibiting aluminosilicate formation, and (3) to develop a predictive tool to control precipitation, scale formation, and cementation under tank waste processing conditions. The results obtained from this will provide crucial information for (1) avoiding problematical sludge processing steps, and (2) identifying and developing effective technologies to process retrieved sludges and supernatants before ultimate

  4. Russian Containers for Transportation of Solid Radioactive Waste

    International Nuclear Information System (INIS)

    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

  5. 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.

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

    International Nuclear Information System (INIS)

    During the 1950's ferrocyanide and nitrate bearing wastes were added to the Hanford radioactive waste tanks as a result of a need to increase the amount of free tank space. To provide additional tank space, the aqueous wastes were decontaminated by coprecipitating the radiocesium with other alkali nickel ferrocyanides and disposing of the decontaminated liquids separately. In recent years a concern has arisen because of the potential explosive reactions of ferrocyanides with nitrates and the nitrate radiolytic degradation product nitrite. Because of this concern the Department of Energy (DOE) and the Hanford site have established a program to insure continued safe operation of the Hanford radioactive waste storage tanks containing ferrocyanides. This paper discusses briefly the history of the ferrocyanide wastes and the four basis components of the program with special emphasis on the chemical reactivity and explosivity research and development studies. These studies with synthetic ferrocyanide bearing wastes indicate that reactions begin at temperatures > 200 degrees C and that reactions within these wastes are not initiated by mechanical impact, friction, and low energy electrical spark but are initiated by a very high energy spark

  7. 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.)

  8. Possible combustion hazards in 3013 plutonium waste container

    International Nuclear Information System (INIS)

    Are there combustion hazards in plutonium-contaminated waste containers caused by combustible gas generation? Current gas generation models in which the only reaction considered is radiolysis must inevitably predict eventual complete dissociation of any water present into hydrogen and oxygen. Waste prepared for the 3013 container should be less subject to this problem because organic material and most of the absorbed water should have been removed. Depending on the waste form, moisture content, organic content, temperature, and container material, the pressure rise due to gas generation will be bounded by backreactions, recombination of the hydrogen and oxygen, absorption of the oxygen by plutonium oxide, and possibly other chemical reactions. Examination of a variety of food pack waste containers at Los Alamos National Laboratory (LANL) has shown little pressure rise, indeed often subatmospheric pressures. In a few cases large hydrogen concentrations up to 47% mole fraction were observed, but with negligible oxygen content. The only fuel seen in significant quantities was H2 and, in one case, CO; the only oxidizer seen in significant quantities was O2. Considerable work on measuring gas generation is being done at Westinghouse Savannah River Company and LANL. In a mixture of H2, O2, and other diluent gases, if the hydrogen concentration is below the value at the lean flammability limit, or if the oxygen concentration is below that at the rich flammability limit, a flame will not propagate from an ignition source. Assuming H2 is the only fuel present in significant quantities, a mixture leaner than the lean limit will get only leaner if mixed with air and is therefore no combustion hazard. However, when a mixture containing large amounts of H2 is nonflammable because there is insufficient O2, there is a hazard. If the mixture should leak into a volume containing O2, or the container is opened into the surrounding air, the mixture will pass through the flammable

  9. Migration of bacteria in compacted clay-based material

    International Nuclear Information System (INIS)

    Buffer (a mixture of 50 wt.% Na-bentonite and 50 wt.% silica sand compacted to a dry density of about 1.68 g/cm3) would surround waste containers in a Canadian nuclear fuel waste disposal vault. The initial heat and radiation output from these containers would likely prevent significant microbial activity at or near container surfaces for some time after disposal, thereby limiting effects such as microbially-influenced corrosion. Microbial repopulation of buffer as conditions improve with time may not occur because of its small pores. Experiments with irradiated buffer plugs (2.2 cm in diameter and 5-cm long; at dry densities of 1.68 and 1.80 g/cm3) were performed to assess the ability of microbes to migrate in buffer. Viable bacteria (Pseudomonas stutzeri), in a suitable growth medium, were brought in contact with one end of the plugs. After 2, 4, 8, 16 and 20 weeks, the plugs were sectioned and tested for moisture content and viable bacteria. Results showed that the plugs were slowly wetting and that moisture levels were sufficient to sustain microbes. No evidence of P. stutzeri was found, however, in all but the first 0.5 cm of the plugs (smallest distance sampled) over a 20-week period. The results suggest that microbial migration in buffer is limited or even completely prevented because of its relatively small pores. (author)

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

    International Nuclear Information System (INIS)

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

  14. Plasma melting of miscellaneous solid waste containing rubber

    International Nuclear Information System (INIS)

    Miscellaneous solid waste contains burnable and/or non-burnable materials besides metallic waste. When non-metallic waste is treated by transferred arc plasma, plasma stability may be decreased by the low electrical conductivity of the materials. This paper describes the result of an experiment in which rubber as a typical burnable waste was fed into the crucible of an operating experimental plasma furnace. The power of the plasma was 20 to 50 kW. The base materials of the simulated waste were 10 kg of iron and 3 kg of fly-ash. Up to 10 g of rubber was fed after the base materials had melted. The voltage of the arc plasma increased just after the feed, and the increase was almost proportional to the weight of the fed rubber. The intrusion of thermally decomposed products of the rubber into the plasma dominated the voltage increase. The voltage increase was investigated for several operating conditions of the furnace. The increase of air injection into the furnace and that of arc current controlled the voltage increase during feeding, but the increase of plasma working gas aggravated it. (author)

  15. 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

  16. Experience of reprocessing of cadmium sulfide-containing waste materials

    International Nuclear Information System (INIS)

    Technique of cadmium extraction from sulfide-containing wastes using the method of oxidizing leaching was developed and subjected to industrial testing. Reagents containing manganese dioxide - manganese ore or manganese slime of electrolytic shop usually used in zinc production - are advisable to be used as oxidizers. Factors of cadmium extraction into solution appeared to be close to ones, obtained during laboratory investigation. If the yield of leaching residual equals ∼38% and the content of cadmium, being in insoluble form, equals ∼0.40%, metal losses with this residual are equal to 0.37%

  17. Fate of metals contained in waste electrical and electronic equipment in a municipal waste treatment process.

    Science.gov (United States)

    Oguchi, Masahiro; Sakanakura, Hirofumi; Terazono, Atsushi; Takigami, Hidetaka

    2012-01-01

    In Japan, waste electrical and electronic equipment (WEEE) that is not covered by the recycling laws are treated as municipal solid waste. A part of common metals are recovered during the treatment; however, other metals are rarely recovered and their destinations are not clear. This study investigated the distribution ratios and substance flows of 55 metals contained in WEEE during municipal waste treatment using shredding and separation techniques at a Japanese municipal waste treatment plant. The results revealed that more than half of Cu and most of Al contained in WEEE end up in landfills or dissipate under the current municipal waste treatment system. Among the other metals contained in WEEE, at least 70% of the mass was distributed to the small-grain fraction through the shredding and separation and is to be landfilled. Most kinds of metals were concentrated several fold in the small-grain fraction through the process and therefore the small-grain fraction may be a next target for recovery of metals in terms of both metal content and amount. Separate collection and pre-sorting of small digital products can work as effective way for reducing precious metals and less common metals to be landfilled to some extent; however, much of the total masses of those metals would still end up in landfills and it is also important to consider how to recover and utilize metals contained in other WEEE such as audio/video equipment. PMID:21963338

  18. Development of electrochemical denitrification from waste water containing ammonium nitrate

    International Nuclear Information System (INIS)

    The authors developed processes to dentrify waste water containing ammonium nitrate discharged from the nuclear fuel manufacturing works and to recover nitric acid and ammonia. For denitrification they applied the operating method and the conditions of operation to make 0.4mM or less from NH4NO3 waste water of 1.5 M by 3 stages of electrodialysis cells. To recover nitric acid and ammonium water, they separated HNO3 solution of 6 M and NH4OH solution with one unit of electrolysis cell, then absorbed NH3 gas from NH4OH solution with water and applied the condition of operation to recover 8 M NH4OH solution. The authors demonstrated that treatment and recovery can be carried out stably with actual waste water with a system through the combination of previously mentioned electrodialysis cells, electrolysis cells and an ammonia gas absorber. At present they are planning a plant where NH4NO3 waste water of 4,500 mol can be treated per day

  19. In-situ containment and stabilization of buried waste

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  1. Large waste containers made of fibre reinforced concrete

    International Nuclear Information System (INIS)

    The production of large-sized metallic waste by dismantling operations, and the evolution of the specifications on the waste to be stored in the different European countries will create a need for large standard containers for the transport and final disposal of the corresponding waste. The research conducted during the 1984/1988 programme, supported by the Commission of European Communities, and based on a comparative study of high-grade concrete materials, reinforced with organic or metallic fibres, led to the development of a high performance container meeting international transport recommendations as well as French requirements for shallow-ground disposal. The material selected, consisting of high-performance mortar with metal fibre reinforcement, was the subject of an intensive programme of characterization tests conducted in close cooperation with LaFarge Company, demonstrating the achievement of mechanical and physical properties comfortably above the regulatory requirements. The construction of an industrial prototype and the subsequent economic analysis served to guarantee the industrial feasibility and cost of this system, in which attempts were made to optimize the 'finished package' product, including its closure system. (author)

  2. 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

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

    International Nuclear Information System (INIS)

    Incinerator ash from the combustion of general trash and ion exchange resins was immobilized in cement and bitumen. Tests were conducted on the resulting waste forms to provide a data base for the acceptability of actual low-level waste forms. The testing was done in accordance with the US Nuclear Regulatory Commission Technical Position on Waste Form. Bitumen had a measured compressive strength of 130 psi and a leachability index of 13 as measured with the ANS 16.1 leach test procedure. Cement demonstrated a compressive strength of 1400 psi and a leachability index of 7. Both waste forms easily exceed the minimum compressive strength of 50 psi and leachability index of 6 specified in the Technical Position. Irradiation to 108 Rad and exposure to 31 thermal cycles ranging from +600) to -300C did not significantly impact these properties. Neither waste form supported bacterial or fungal growth as measured with ASTM G21 and G22 procedures. However, there is some indication of biodegradation due to co-metabolic processes. Concentration of organic complexants in leachates of the ash, cement and bitumen were too low to significantly affect the release of radionuclides from the waste forms. Neither bitumen nor cement containing incinerator ash caused any corrosion or degradation of potential container materials including steel, polyethylene and fiberglass. However, moist ash did cause corrosion of the steel

  4. 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.

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

    International Nuclear Information System (INIS)

    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

  6. Influence of clay mineralogy on clay based ceramic products

    International Nuclear Information System (INIS)

    Clay-based ceramic products can either be produced directly from a suitable clay source without the need further addition or such products can be produced from a ceramic body formulated by additions of other raw materials such as feldspar and silica sand. In either case, the mineralogical make-up of the clay component plays a dominating role in the fabrication and properties of the ceramic product. This study was sparked off by a peculiar result observed in one of five local ball clay samples that were used to reformulate a ceramic body. Initial characterisation tests conducted on the clays indicated that these clays can be classified as kaolinitic. However, one of these clays produced a ceramic body that is distinctively different in terms of whiteness, smoothness and density as compared to the other four clays. Careful re-examination of other characterisation data, such as particle size distribution and chemical analysis, failed to offer any plausible explanation. Consequently, the mineralogical analysis by x-ray diffraction was repeated by paying meticulous attention to specimen preparation. Diffraction data for the clay with anomalous behaviour indicated the presence of a ∼ 10A peak that diminished when the same specimen was re-tested after heating in an oven at 12O degree C whilst the other four clays only exhibit the characteristic kaolinite (Al sub 2 O sub 3. 2SiO sub 2. 2H sub 2 0) and muscovite peaks at ∼ 7A and ∼ 10A before and after heat treatment. This suggests the presence of the mineral halloysite (A1 sub 2 0 sub 3. 2SiO sub 2.4H sub 2 0) in that particular clay. This difference in mineralogy can be attributed to account for the variations in physical properties of the final product. Consequently, this paper reviews in general the precautionary measures that must be adhered to during any mineralogical investigation of clay minerals or clay-based materials. The common pitfalls during specimen preparation, machine settings and interpretation of

  7. 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

  8. Processing method for metallic aluminum-containing-radioactive solid wastes

    International Nuclear Information System (INIS)

    Metallic Al-containing radioactive solid wastes are reacted with an alkali-solution to generate hydrogen gas, and then obtained reaction liquid and a solidifying material mainly comprising a latent water-hardenable material are mixed and solidified. In this case, since the metallic Al is transformed into Al hydroxide or alkali aluminate, the solidifying material even if it is solidified, it is not reacted with the metallic Al, and generation of voids and cracks in the solids is suppressed, and the solidification material has excellent mechanical strength and leaching proof property against radioactive nuclides. (T.M.)

  9. 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

  10. Clay based superior aggregate for making light construction

    International Nuclear Information System (INIS)

    Aggregate is defined as material consisting of solid minerals with grain size ranging from sand to gravel. This material is usually used for filling components of concrete. Clay based aggregate has special properties that can be useful for making light construction in wet environment. Concrete build using the aggregate is not impermeable, but it can adsorb water in significant amount, so that is very useful to be used as paving block for walker zone (trotoir) and zone surrounding swimming pool. Laboratory test results show that the aggregate has grain size in zone 1 type with specific gravity between 1360 - 1840 kg/cm3. Its abrasive factor about 35.32 % in average and it can be used as building material for light construction. Concrete with mixing ratio cement/aggregate of 1 to 6 with slump condition 3 to 6 cm and curing time of 28 days indicates compressive strength of 10 N/cm3. Qualitative cost analysis reflected that the paving block production cost is relatively low and it will be more profitable if they are product from failed pentile having fail between 10 to 30 %

  11. Mathematical Modelling of Leachate Production from Waste Contained Site

    Directory of Open Access Journals (Sweden)

    Ojolo S. Joshua

    2012-07-01

    Full Text Available In this work, mathematical models of leachate production from Waste Contained Site (WCS was developed and validated using the existing experimental data with aid of MATLAB, 2007a. When the leachate generation potentials (Lo were 100m3, 80m3 and 50m3, the maximum amount of leachate generated were about 2920m3, 2338m3 and 1461m3 for about 130 days respectively. It was noted that as the leachate percolates through a selected distance, the concentration keeps decreasing for one-dimensional flow in all the cases considered. Decreasing in concentration continues until a point was reached when the concentration was almost zero and later constant. The effects of diffusivity, amount of organic content present within the waste and gravity, as cases, were also considered in various occasions during the percolation. Comparison of their effects was also taken into account. In case of gravity at constant diffusivity, decrease in concentration was not rapid but gradually while much organic content in the waste caused the rate of leachate production to be rapid; hence, giving rise to a sharp sloped curve. It can be concluded that gravity influences the rate of change in the concentration of the leachate generation as the leachate percolate downward to the underground water. When the diffusivity and gravity are put into consideration, the concentration of the leachate decreases gradually and slowly.

  12. Method for treating sulfuric acid-containing waste

    International Nuclear Information System (INIS)

    When sulfuric acid-containing wastes are treated at a high temperature, since sulfuric acid ingredients (SO4) are converted into gaseous sulfur oxides (SOx), the volume can be remarkably reduced. However, if Na2SO4 generated from a nuclear power plant is treated by using this method, since the pyrolysis temperature of Na2SO4 is higher than 2000degC, it has been impossible to use an incinerator or a high temperature melting reactor (highest temperature: lower than 1500degC) ordinarily used in nuclear power plants. In the present invention, Ba(OH)2 is added under stirring to liquid wastes mainly composed of Na2SO4 to convert it into insoluble BaSO4 (pyrolysis temperature: 1200degC), thereafter, decomposition treatment is applied. With such procedures, sulfuric ingredients are gasified and the radioactive wastes are reduced as less as possible. Since the generated radioactive residues are chemically stable, the storage and disposal thereof are facilitated. (T.M.)

  13. 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)

  14. Container materials for isolation of radioactive waste in salt

    International Nuclear Information System (INIS)

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

  15. 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

  16. Assay of contained waste using active and passive computed tomography

    International Nuclear Information System (INIS)

    The US Department of Energy (DOE) has more than 600,000 transuranic waste drums temporarily stored at nearly 40 sites within the US. Contents of these drums must be characterized before they are transported for permanent disposal. Opening drums for examination is expensive mainly because of the safety precautions that must be taken. Current nonintrusive methods of characterizing waste in sealed drums are often inaccurate where assay errors are related to nonuniform measurement responses associated with unknown radioactive-source and waste-matrix-material distributions. These errors can be reduced by the application of imaging techniques that better measure the spatial locations of sources and matrix attenuation. Lawrence Livermore National Laboratory (LLNL) has developed an emerging gamma-ray nondestructive analysis (NDA) technology, called active and passive computed tomography (A and PCT), that identifies and accurately quantifies all detectable radioisotopes in closed containers of waste. The performance of the A and PCT technology has been determined by several open and blind tests. Several 15-replicate studies were completed for three of the four required activity ranges. The three ranges were measured by acquiring A and PCT data for three separate placements of radioactive standards within an empty-matrix drum. The standards had a total mass of 0.93, 9.3, and 33.48 g of 239Pu positioned within the drum and required 4, 0.75, and 0.5 h total assay time per replicate, respectively. The performance results are summarized in Table 1. Additional research is being performed to maintain requirements while decreasing assay time

  17. 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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  20. The stress corrosion cracking of copper nuclear waste containers

    International Nuclear Information System (INIS)

    The extent of stress corrosion cracking (SCC) of copper nuclear waste containers is being predicted on the basis of a limited propagation argument. In this argument, it is accepted that crack initiation may occur, but it is argued that the environmental conditions and material properties required for a through-wall crack to propagate will not be present. In this paper, the effect of one environmental parameter, the supply of oxidant (JOX), on the crack growth rate is examined. Experiments have been conducted on two grades of Cu in NaNO2 environments using two loading techniques. The supply of oxidant has been varied either electrochemically in bulk solution using different applied current densities or by embedding the loaded test specimens in compacted buffer material containing O2 as the oxidant. Measured and theoretical crack growth rates as a function of JOX are compared with the predicted oxidant flux to the containers in a disposal vault and an estimate of the maximum crack depth on a container obtained

  1. The stress corrosion cracking of copper nuclear waste containers

    International Nuclear Information System (INIS)

    The extent of stress corrosion cracking (SCC) of copper nuclear waste containers is being predicted on the basis of a 'limited propagation' argument. In this argument, it is accepted that crack initiation may occur, but it is argued that the environmental conditions and material properties required for a through-wall crack to propagate will not be present. In this paper, the effect of one environmental parameter, the supply of oxidant (Jox), on the crack growth rate is examined. Experiments have been conducted on two grades of Cu in NANO2 environments using two loading techniques. The supply of oxidant has been varied either electrochemically in bulk solution using different applied current densities or by embedding the loaded test specimens in compacted buffer material containing O2 as the oxidant. Measured and theoretical crack growth rates as a function of Jox are compared with the predicted oxidant flux to the containers in a disposal vault and an estimate of the maximum crack depth on a container obtained. (author)

  2. Production of New Biomass/Waste-Containing Solid Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Glenn A. Shirey; David J. Akers

    2005-09-23

    CQ Inc. and its industry partners--PBS Coals, Inc. (Friedens, Pennsylvania), American Fiber Resources (Fairmont, West Virginia), Allegheny Energy Supply (Williamsport, Maryland), and the Heritage Research Group (Indianapolis, Indiana)--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 is applicable to a variety of combinations of biomass, wastes, and coal; economically competitive with current fuels; and provides environmental benefits compared with coal. During Phase I of this project (January 1999 to July 2000), several biomass/waste materials were evaluated for potential use in a composite fuel. As a result of that work and the team's commercial experience in composite fuels for energy production, paper mill sludge and coal were selected for further evaluation and demonstration

  3. 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)

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

    International Nuclear Information System (INIS)

    Two main by-products or waste from the production of glass fibre are following: sewage sludge containing montmorillonite clay as sorbent material and ca 50 % of organic matter as well as waste glass from aluminium borosilicate glass fibre with relatively high softening temperature (> 600 degree centigrade). In order to elaborate different new ceramic products (porous or dense composites) the mentioned by-products and illicit clay from two different layers of Apriki deposit (Latvia) with illite content in clay fraction up to 80-90 % was used as a matrix. The raw materials were investigated by differential-thermal (DTA) and XRD analysis. Ternary compositions were prepared from mixtures of 15 - 35 wt % of sludge, 20 wt % of waste glass and 45 - 65 wt % of clay and the pressed green bodies were thermally treated in sintering temperature range from 1080 to 1120 degree centigrade in different treatment conditions. Materials produced in temperature range 1090 - 1100 degree centigrade with the most optimal properties - porosity 38 - 52 %, water absorption 39 - 47 % and bulk density 1.35 - 1.67 g/cm3 were selected for production of porous ceramics and materials showing porosity 0.35 - 1.1 %, water absorption 0.7 - 2.6 % and bulk density 2.1 - 2.3 g/cm3 - for dense ceramic composites. Obtained results indicated that incorporation up to 25 wt % of sewage sludge is beneficial for production of both ceramic products and glass-ceramic composites according to the technological properties. Structural analysis of elaborated composite materials was performed by scanning electron microscopy(SEM). By X-ray diffraction analysis (XRD) the quartz, diopside and anorthite crystalline phases were detected. (Author) 16 refs.

  5. Corrosion processes studies in a radioactive waste container

    International Nuclear Information System (INIS)

    Full text: In high activity nuclear waste containers design, a number of metals have been considered, among them carbon steels, stainless steels, high alloy steels, copper and titanium should be mentioned. Lead may also be useful for this purpose due to its good radiological and corrosion resistance in natural water properties. As disadvantage, its low creep resistance and toxicity should be mentioned. Corrosion tests results in commercial lead and lead-tin alloys, with tin contents ranging from 1 to 9 %, lead-antimony and lead bismuth with alloy content ranging from 1 to 3.5 % are reported. Those tests were performed in low salinity carbonated groundwater and in artificial seawater at 60 and 75 degree C. This test program is aimed to select an alloy with better corrosion resistance than commercial lead. It was found that lead-tin alloys fulfill this condition when tin concentration is equal to 3.5 % or higher. A lower corrosion resistance was found in Lead-Antimony and Lead-Bismuth alloys tests. Besides, a corrosion processes modeling in a container placed in a repository is intended. In a first stage, taking into account that for this particular container design, where a carbon steel liner is the external container wall, an iron corrosion process based on oxygen diffusion through a porous media (bentonite backfill) is proposed. No water radiolysis is taken into account because this phenomenon is negligible due to the internal thick lead container main barrier. The results of this model predict uniform corrosion and a low corrosion kinetics of the carbon steel external liner. It is due to the limited oxygen amount retained in the backfill pores and its low diffusion kinetics

  6. Processing method for contaminated water containing radioactive waste

    International Nuclear Information System (INIS)

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

  7. Assessment of materials for containment of nuclear fuel waste

    International Nuclear Information System (INIS)

    As part of the assessment program for container materials for the long-term disposal of nuclear fuel waste, a test involving slow tensile straining at 10-5 mm/s has been developed to study the effect of hydrogen on the fracture behaviour of titanium. Compact tension specimens of Grade-2 and Grade-12 material have been precracked in fatigue before charging with hydrogen at various concentrations up to 3000 wppm and then straining to failure. At low hydrogen concentrations, propagation of a slow crack may be followed at higher stress intensities by fast brittle failure, whereas at high concentrations the majority of materials suffer only fast brittle failure at stress intensity factors that decrease with increasing hydrogen content. The resistance of both Grade-2 and Grade-12 titanium to embrittlement by hydrogen is more dependent upon microstructure, and particularly the distribution of β-phase, than upon the concentration of basal poles lying perpendicular to the principal fracture plane

  8. 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

  9. 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

  10. Proposed Objective Odor Control Test Methodology for Waste Containment

    Science.gov (United States)

    Vos, Gordon

    2010-01-01

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

  11. Overview of European concepts for high-level waste and spent fuel disposal with special reference waste container corrosion

    International Nuclear Information System (INIS)

    This paper provides a brief overview of current repository and engineered barrier system (EBS) designs in selected high-level waste (HLW) and spent fuel (SF) disposal concepts from European countries, with special reference to key metallic waste containers and corrosion processes. The paper discusses assessments of copper, iron and steel container corrosion behaviour under the environmental conditions expected, given likely repository host rocks and groundwaters, and comments on the significance of corrosion processes, the choice of waste container materials, and areas of research. Most of the HLW and/or SF disposal programmes in European countries are pursuing disposal options in which the primary waste container is designed, in conjunction with the surrounding EBS materials, to provide complete containment of the waste for at least the period when temperatures in the disposal system are significantly raised by radioactive decay

  12. 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)

  13. 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

  14. 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.

  15. Biofilm treatment of soil for waste containment and remediation

    International Nuclear Information System (INIS)

    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 ksat using a flexible-wall permeameter. Saturated hydraulic conductivity (ksat) was reduced from 1 x 10-5 cm/sec to 2 x 10-8 cm/sec: by biofilm treatment. Permeation with saline, acidic, and basic solutions following formation of a biofilm was found to have negligible effect on the reduced ksat, 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

  16. Study on hazardous substances contained in radioactive waste

    International Nuclear Information System (INIS)

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

  17. 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.

  18. 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

  19. Novel On-Site Cupric Oxide Recovery Process from Waste Containing Copper

    Science.gov (United States)

    Kobayashi, Takuya; Kano, Kazunori; Suzuki, Toshihiro; Kobayashi, Atsushi

    2013-05-01

    Although the copper-containing waste from semiconductor or printed circuit board (PCB) manufacturing contains a high concentration of copper, it is usually transported and treated outside of the factories. We studied a novel treatment technology for on-site recycling in the factories. In this technology, cupric oxide with a low-chloride-content was obtained from waste with a high copper concentration, such as cupric chloride etchant waste and cupric sulfate plating waste. In the proposed method, copper-containing waste mixed with H2O2 solution is added to NaOH solution by stepwise addition. In laboratory experiments, we optimized the reaction conditions and obtained low-chloride-content CuO from actual cupric chloride etchant waste and cupric sulfate plating waste. Based on the laboratory experiments, we constructed the first practical plant at a PCB factory and obtained low-chloride-content CuO.

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

    International Nuclear Information System (INIS)

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

  1. Radioactive package container system for remote handling of low-level radioactive waste

    International Nuclear Information System (INIS)

    Remotely operated handling systems are employed for safe processing and transfer of low level radioactive wastes at nuclear generating plants. These systems minimize or preclude personnel radiation exposure while expediting waste handling operations. A remotely operated waste handling and transfer system containing several unique features has been designed, fabricated and tested for installation oat Arizona Public Service's, Palo Verde Nuclear Generating Station. The system incorporates modular subcomponents such as a waste processing shield, bottom and top loading shielded cask, and remote grappling equipment, making it adaptable to multi-task waste handling operations. The system has been designed to be operationally flexible, and contributes significantly to reducing waste processing personnel exposure

  2. 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...

  3. Gamma-ray computed tomography of waste containers and conditioned radioactive waste in nuclear industry

    International Nuclear Information System (INIS)

    The properties of a computer tomograph for NDE applications using energies up to 1 MeV are presented. As examples tomograms of containers for high level waste containing simulated waste are shown. Further the possibilities of detecting corrosion defects at tubes or containers and examples of imaging objects of highly complicated geometric structure are given. After the successful development of computer tomography in medicine a wide range of applications in NDE is expected. The main advantages of CT are the ability of imaging only one object slice without cross talk from other object slices and the elimination of scattered radiation, resulting in visualizing small local density variations as low as some percent of the average density. This is still possible, even if the material is surrounded by material with higher absorption. Finally CT can image larger cracks. At BAM (''Federal Institute for Materials Testing'') a computer tomograph for various applications has been established. It's main features are : several types of sources, i.e. a 400 kV x-ray tube, a 500 Ci Co-60 source and in the future an electron accelerator. The manipulator, which performs the scanning movement of the object, has a maximum load of 1000 kg and an overall accuracy of 0.1 mm. The maximum object diameter is 1 m, under some constraints 1.4 m. The collimators are built for a spatial resolution of 1 mm at radiation energies up to 2 MeV. The thickness of the slice scanned is variable from 8 to 1 mm. The detector system consists of 32 photomultiplier tubes with plastic scintillators. The image reconstruction and display is made with a minicomputer

  4. Neutron shielding analysis for remote handled transuranic waste containers in facility casks at the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Neutron shielding characteristics of the Waste Isolation Pilot Plant facility cask have been quantified for a variety of combinations of neutron sources and waste matrices which would potentially be handled in waste containers. The neutron attenuation and neutron environment of the waste container and the facility cask have been analyzed to ensure that the design requirement of neutron dose rate will be met under the combinations of the source and waste matrix conditions. The analyses considered the ranges of neutron source spectrum and waste matrices which combine to produce the minimum neutron shielding worth of the facility cask. One-dimensional analyses were performed with discrete ordinate transport theory methods using multigroup neutron cross section data. The results discussed in this report demonstrate the effect of source spectrum and waste container matrix on predicted neutron dose rates adjacent to the unshielded waste container and the surface of the facility cask. An evaluation of the uncertainties in predicted neutron dose rates is provided which results in an assessment of the maximum measured neutron dose rate external to the facility cask. A description of the analytical models developed, the analysis methodology, the neutron source spectra, and the detailed results are described in this report. 10 refs., 50 figs., 39 tabs

  5. 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

  6. 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

  7. 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.

  8. Design and testing of the TX-4 Type A container for packaging radioactive waste

    International Nuclear Information System (INIS)

    The Toxic Waste Control Group at the Lawrence Livermore National Laboratory has designed and tested the TX-4, a Type A steel container for shipping and storing radioactive waste. We designed the TX-4 to eliminate the safety, maneuverability, weight, and cost problems experienced by other hazards waste containers. Our design meets the test criteria set by the Department of Transportation (49 CFR 173.398). The TX-4 container passed all tests when loaded to 7000 lb gross weight and effectively solved the above problems. Its simplicity of design, low weight, and ease in handling have proved to be timesaving and cost-effective. This report summarizes our testing of the TX-4 and past and present radioactive waste containers used by defense-related operations. Based on our results, we believe the TX-4 is a superior container for the hazardous waste industry. 10 figures, 1 table

  9. Substance Flow Analysis of Wastes Containing Polybrominated Diphenyl Ethers

    DEFF Research Database (Denmark)

    Vyzinkarova, Dana; Brunner, Paul H.

    2013-01-01

    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...... 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...... recommend establishing a new, goal-oriented data set by additional analyses of waste constituents and plastic recycling samples, as well as establishing reliable mass balances of polybrominated diphenyl ethers’ flows and stocks by means of SFA....

  10. 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

  11. Stabilization Using Phosphate Bonded Ceramics. Salt Containing Mixed Waste Treatment. Mixed Waste Focus Area. OST Reference #117

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    1999-09-01

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

  12. 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.

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

    International Nuclear Information System (INIS)

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

  14. 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

  15. Assay of TRU wastes containing (α,n) sources

    International Nuclear Information System (INIS)

    We have studied methods of determining alpha activity in transuranic waste from the gamma rays produced by the reactions (α,nγ), (α,pγ), and (α,α'γ) or produced directly by the alpha-emitting isotopes. Gamma-ray spectra were acquired for 20 drums of waste at the Stored Waste Examination Pilot Plant at Idaho National Engineering Laboratory with a high-purity germanium detector. To assist in the analysis, gamma-ray spectra were also acquired from the following standard neutron sources: 238Pu/Li, 241Am/Li, 241Am/Be, 241Am/B, 241Am/13C, 238PuO2, and 241AmF. Neutron measurements on the 20 drums were made in a combined passive and active neutron assay system for transuranic waste in Idaho. Representative gamma-ray spectra, line intensities, and calculated alpha activities for the drums are presented. In most matrix types, the alpha activity can be estimated from the neutron assay system results alone or from the reaction gamma-ray outputs. This work demonstrates a method of certifying the alpha activity, as well as the fissile content, of transuranic waste for the Waste Isolation Pilot Plant

  16. 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.

  17. 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.

  18. Iron Phosphate Glass-Containing Hanford Waste Simulant

    International Nuclear Information System (INIS)

    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.

  19. Stress corrosion cracking of candidate waste container materials; Final report

    Energy Technology Data Exchange (ETDEWEB)

    Park, J.Y.; Maiya, P.S.; Soppet, W.K.; Diercks, D.R.; Shack, W.J.; Kassner, T.F. [Argonne National Lab., IL (United States)

    1992-06-01

    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, Incoloy 825, phosphorus-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) and fracture-mechanics crack-growth-rate (CGR) tests was performed at 93{degree}C and 1 atm of pressure in simulated J-13 well water. This water is representative, prior to the widespread availability of unsaturated-zone water, of the groundwater present at the Yucca Mountain site. Slow-strain-rate tests were conducted on 6.35-mm-diameter cylindrical specimens at strain rates of 10-{sup {minus}7} and 10{sup {minus}8} s{sup {minus}1} under crevice and noncrevice conditions. All tests were interrupted after nominal elongation strain of 1--4%. Scanning electron microscopy revealed some crack initiation in virtually all the materials, as well as weldments made from these materials. A stress- or strain-ratio cracking index ranks these materials, in order of increasing resistance to SCC, as follows: Type 304 SS < Type 316L SS < Incoloy 825 < Cu-30%Ni < Cu and Cu-7%Al. Fracture-mechanics CGR tests were conducted on 25.4-mm-thick compact tension specimens of Types 304L and 316L stainless steel (SS) and Incoloy 825. Crack-growth rates were measured under various load conditions: load ratios M of 0.5--1.0, frequencies of 10{sup {minus}3}-1 Hz, rise nines of 1--1000s, and peak stress intensities of 25--40 MPa{center_dot}m {sup l/2}.

  20. The conceptual design of waste repository for radioactive waste from medical, industrial and research facilities containing comparatively high radioactivity. 3

    International Nuclear Information System (INIS)

    Advisory Committee on Nuclear Fuel Cycle Backend Policy reported the basic approach to the RI and Institute etc. wastes on March 2002. According to it, radioactive waste form medical, industrial and research facilities should be classified by their radioactivity properties and physical and chemical properties, and should be disposed in the appropriate types of repository with that classification. For the radioactive waste containing comparatively high radioactivity generated from reactors, NSC has established the Concentration limit for disposal. NSC is now discussing about the limit for the radioactive waste from medical, industrial and research facilities containing comparatively high radioactivity. Japan Nuclear Cycle Development Institute (JNC) has studied about the feasibility and the cost of the disposal for radioactive waste from medical, industrial and research facilities. This study was started to renew to latest data of the radioactive waste. And at the point of shielding from radiation, the waste was categorized by activity of nuclide in waste container. Then the safety assessment and the prediction of cost of the disposal performed. The result of this study showed as follow; (1) According to groundwater scenario, the summed does for the repository are below of the regulatory guideline of 10μSv/year. (2) A rough estimate values of a disposal cost under the assumed situation were indicated with the arguments. (author)

  1. 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.

  2. Containment of Radioactive Waste for Sea Disposal and Fisheries Off the Canadian Pacific Coast

    International Nuclear Information System (INIS)

    Low-level radioactive waste consists largely of contaminated laboratory trash, larger items of equipment, animal carcasses from biological tracer experiments, active liquids, and some reactor-irradiated materials. A total of 16,288 55-gallon drums of low-level waste has been dumped off the coast of California from 1946 to 1957, inclusive. The major objectives of the criteria for US sea disposal operations have been (1) safety in handling between waste originator and disposal site, and (2) adequate sinking of wastes in the sea. There has been no requirement for integrity of container or contents at depth. Other techniques used or previously suggested for radioactive waste containment, such as canning, adsorption on clay, and integration into glass are described. In order to render a radioactive waste harmless to fish and other aquatic organisms, it must be either (1) isolated from the environment, or (2) dispersed to levels of permissible concentrations. A spherical vault of suitable engineering design, hermetically sealed and able to withstand high pressures, is proposed for low-level and intermediate-level solid waste disposal, as a means of isolating the waste from the environment. Fish may be affected by radioactive waste in their environment through: (1) direct radiation from the disposed radioactive material; (2) ingestion of food organisms containing concentrated radioisotopes; (3) irradiation by water containing radioactive ions or particles; and (4) contamination by bottom materials rich in precipitated radioisotopes. Research necessary before widespread marine disposal of radioactive waste should be permitted is suggested. (author)

  3. Clay-based materials for engineered barriers: a review

    International Nuclear Information System (INIS)

    The potential importance of backfilling and plugging in underground radioactive waste repositories has led different research institutions to carry out extensive studies of swelling clay materials for the development of engineered barriers in underground conditions. These materials should combine a variety of hydro-thermo-mechanical and geochemical properties: impermeability, swelling ability in order to fill all void space, heat transfer and retention capacity for the most noxious radionuclides. Smectite clays best exhibit these properties and most of the research effort has been devoted to this type of materials. In this paper, mineralogical composition, sodium or calcium content, thermo-hydro-mechanical properties, swelling pressure, hydraulic and thermal conductivity, and chemical properties of five smectite clays selected by five major nuclear countries are reviewed: Avonseal montmorillonite (Canada), MX 80 montmorillonite (Sweden), Montigel montmorillonite (Switzerland), S-2 montmorillonite (Spain), and Fo-Ca inter stratified kaolinite/beidellite (France). (J.S.). 29 refs., 5 figs., 3 tabs

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

    International Nuclear Information System (INIS)

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

  5. 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.

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

    International Nuclear Information System (INIS)

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

  7. 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)

  8. Radiological hazards of waste containing alpha-emitting radionuclides

    International Nuclear Information System (INIS)

    The radiological hazards of alpha-contaminated wastes are discussed in this overview in terms of two components of hazard: radiobiological hazard and radioecological hazard. Radiobiological hazard refers to human uptake of alpha-emitters by inhalation and ingestion, and the resultant dose to critical organs of the body. Radioecological hazard refers to the processes of release from buried wastes, transport in the environment, and translocation to man through the food chain. Besides detailing the sources and magnitude of hazards, this brief review identifies the uncertainties in their estimation, and implications for the regulatory process

  9. 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)

  10. 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

  11. 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)

  12. 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.

  13. Process for cleaning waste gases containing hydrogen fluoride and fluorine

    International Nuclear Information System (INIS)

    The waste gases are taken into an aqueous solution of uranium(IV) sulphate, where insoluble uranium tetrafluoride hydrate is formed. This can be returned to the manufacturing process for uranium hexafluoride. The uranyl sulphate which is also formed is best reduced to uranium(IV) where the reduction need not be complete, because of the excess of uranium(IV). (PW)

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

    International Nuclear Information System (INIS)

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

  15. 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

  16. 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

  17. Assessing efficiency of radioactive waste isolation in containers of different materials

    International Nuclear Information System (INIS)

    The mathematical model of dynamics of radionuclide release from the protective containers into host geological massif has been developed for the deep disposal of vitrified high- level radioactive waste. According to this model, the comparable efficiency estimation of high- activity waste isolation for two variants of container construction, in particularly, steel and dense corrosion- and radiation-resistant ceramics, has been realized. The considerable advantage of ceramic container in comparison with steel container was shown. The capacity of long-lived radionuclide containment has confirmed the perspectives of the ceramics as a barrier material

  18. 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...

  19. 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

  20. Microanalysis of clay-based pigments in painted artworks by the means of Raman spectroscopy

    Czech Academy of Sciences Publication Activity Database

    Košařová, V.; Hradil, David; Němec, I.; Bezdička, Petr; Kanický, V.

    2013-01-01

    Roč. 44, č. 11 (2013), s. 1570-1577. ISSN 0377-0486 Institutional support: RVO:61388980 Keywords : Raman spectroscopy * clay-based pigments * clay minerals * iron oxides * microanalysis of paintings Subject RIV: CB - Analytical Chemistry, Separation Impact factor: 2.519, year: 2013

  1. 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.

  2. 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

  3. Immobilization of nitrate waste streams containing small amounts of organic solvents

    International Nuclear Information System (INIS)

    The influence of organic solvents in nitrate waste streams is investigated concerning the physical, chemical and mechanical properties of the full size waste forms when ordinary Portland cement is used as a binder matrix. Simulated waste streams containing sodium nitrate varying from 0 to about 26 wt %, including tributyl phosphate/dodecane, 30/70, as the organic phase varying from 0 to 10 wt %, were assayed. (author)

  4. 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)

  5. Incineration of urban solid waste containing radioactive sources

    International Nuclear Information System (INIS)

    Incineration of urban solid waste accidentally contaminated by orphan sources or radioactive material is a potential risk for environment and public health. Moreover, production and emission of radioactive fumes can cause a heavy contamination of the plant, leading to important economic detriment. In order to prevent such a hazard, in February 2004 a radiometric portal for detection of radioactive material in incoming waste has been installed at AMSA (Azienda Milanese per i Servizi Ambientali) 'Silla 2' urban solid waste incineration plant of Milan. Radioactive detections performed from installation time up to December 2006 consist entirely of low-activity material contaminated from radiopharmaceuticals (mainly 131I). In this work an estimate of the dose that would have been committed to population, due to incineration of the radioactive material detected by the radiometric portal, has been evaluated. Furthermore, public health and environmental effects due to incineration of a high-activity source have been estimated. Incineration of the contaminated material detected appears to have negligible effects at all; the evaluated annual collective dose, almost entirely conferred by 131I, is indeed 0.1 man mSv. Otherwise, incineration of a 3.7 x 1010 Bq (1 Ci) source of 137Cs, assumed as reference accident, could result in a light environmental contamination involving a large area. Although the maximum total dose, owing to inhalation and submersion, committed to a single individual appears to be negligible (less than 10-8 Sv), the environmental contamination leads to a potential important exposure due to ingestion of contaminated foods. With respect to 'Silla 2' plant and to the worst meteorological conditions, the evaluated collective dose results in 0.34 man Sv. Performed analyses have confirmed that radiometric portals, which are today mainly used in foundries, represent a valid public health and environmental protection also in urban waste incineration plants.

  6. 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.

  7. 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.……

  8. 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.

  9. Method of deactivation by filtering of radioactive liquid wastes containing organic solvent

    International Nuclear Information System (INIS)

    Liquid radioactive wastes at each stage of processing are filtered with at least one filter layer containing at least one of the following sorbents: silica gel, glass or kieselguhr. The said sorbents have higher decontamination efficiency than ion exchangers and hydrophobic sorbents. The present decontamination method may also be used after the distillation of liquid wastes. (J.P.)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  13. 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

  14. Aluminum-Containing Phases in Tank Waste: Precipitation and Deposition of Aluminum-Containing Phases

    International Nuclear Information System (INIS)

    Aluminosilicate deposit buildup experienced during the tank waste volume-reduction process at the Savannah River Site (SRS) required an evaporator to be shut down in October 1999. Recent investigations illustrated the accumulation 7 wt% uranium, 3% was 235U and absent of neutron poisons, within these deposits and presented a criticality concern. The Waste Processing Technology Section of Westinghouse Savannah River Company at SRS is now collaborating with a team from Pacific Northwest National Laboratory in efforts to identify the phases controlling uranium solubility and understand the conditions under which they precipitate

  15. Inhibition of nuclear waste solutions containing multiple aggressive anions

    International Nuclear Information System (INIS)

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

  16. Lead localized corrosion evaluation in high activity nuclear waste containers

    International Nuclear Information System (INIS)

    In Argentina, the conceptual basis for the final disposal of high activity nuclear waste was set according to the multiple barrier concept. A lead layer was preselected due to its good radiological protection and corrosion resistance. The focus of present experiments is to assess on the high purity lead corrosion kinetics due to localized attack by aggressive anions, such as chlorides, nitrates and acetates. Hence, high purity lead corrosion potentials with time, potential kinetic and current time measurements at constant potential were performed in synthetic groundwater contaminated with such anions. (author)

  17. 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)

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

    International Nuclear Information System (INIS)

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

  19. Low-level waste shallow land burial source term container breach and waste form leaching model development

    International Nuclear Information System (INIS)

    A general computer model has been developed to predict the release and transport (i.e. source term) of radionuclides from shallow land burial facilities. This model predicts the processes of unsaturated water flow, metallic container degradation, leaching of radionuclides from the waste form, and their movement away from the waste form. This paper discusses model development work for the container degradation and leaching aspects of the source term model. Application of these models and the sensitivity of release rates to model parameters, e.g. diffusion coefficients, corrosion rates, etc., are also discussed. 14 refs., 2 figs

  20. 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.)

  1. Physical and chemical studying of cryolite-, alumina containing wastes of aluminium industry

    International Nuclear Information System (INIS)

    The purpose of present work is investigation of compositions and properties of cryolite-, alumina containing wastes of aluminium industry and determination of chemistry changes taking place during reprocessing process. After studying of above mentioned process authors became to conclusion that physical and chemical studying of cryolite-, alumina containing wastes of aluminium industry and products of their reprocessing by roentgen-phase and derivative-graphic methods showed that in the mud composition and dump screening of solid wastes Al2O3, Na3AlF6, C, Na2CO3, NaHCO3, NaF, SlO2 present

  2. Method to produce bodies containing high-radioactive waste and/or actinides in granulated glass

    International Nuclear Information System (INIS)

    According to the main patent no. 2524169, a corrosion-resistant container is filled with granulated glass which contains high-activity wastes or actinides, and a liquid metal or metal alloy is then filled in which confines the radioactive granulate when cooled down. The patent claim deals with wastes whose activity is not so high that heat must be removed at once by liquid metal. The inventor proposes to fill the metal matrix in the form of rods or shredded metal and melt it afterwards, making sure that the waste material is completely covered and enclosed, if necessary by filling in additional matrix material. (orig./MM)

  3. 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.)

  4. 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.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    Procedures and techniques were evaluated to mitigate thermal stress fracture in waste glass as the glass cools after casting. The two principal causes of fracture identified in small-scale testing are internal thermal stresses arising from excessive thermal gradients when cooled too fast, and shear fracturing in the surface of the glass because the stainless steel canister shrinks faster than the glass on cooling. Acoustic emission and ceramographic techniques were used to outline an annealing schedule that requires at least three weeks of controlled cooling below 5500C to avoid excessive thermal gradients and corresponding stresses. Fracture arising from canister interactions cannot be relieved by slow cooling, but can be eliminated for stainless steel canisters by using ceramic paper, ceramic or graphite paste linings, or by choosing a canister material with a thermal expansion coefficient comparable to, or less than, that of the glass

  8. Containment canister for capturing hazardous waste debris during piping modifications

    Science.gov (United States)

    Dozier, Stanley B.

    2001-07-24

    The present invention relates to a capture and containment canister which reduces the risk of radiation and other biohazard exposure to workers, the need for a costly containment hut and the need for the extra manpower associated with the hut. The present invention includes the design of a canister having a specially designed magnetic ring that attracts and holds the top of the canister in place during modifications to gloveboxes and other types of radiological and biochemical hoods. The present invention also provides an improved hole saw that eliminates the need for a pilot bit.

  9. 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

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

    International Nuclear Information System (INIS)

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

  11. Properties of radioactive wastes and waste containers. Quarterly progress report, April-June 1981

    International Nuclear Information System (INIS)

    An empirical relationship has been developed to estimate the cumulative fractional releases of 137Cs from simulated waste forms as a function of leaching time and the geometric surface-to-volume ratios. Data from an ongoing leaching study were used. The simulated waste forms consisted of organic cation exchange resins solidified in Portland I cement at a waste-to-cement ratio of 0.6 and water-to-cement ratio of 0.4. The nominal specimen dimensions were: 1-inch diameter x 1-inch high, 2-inch diameter x 2-inch high, 2-inch diameter x 4-inch high, 3-inch diameter x 3-inch high, 6-inch diameter x 6-inch high, 6-inch diameter x 12-inch high, and 12-inch diameter x 12-inch high. The waste forms were leached in deionized water using a modified IAEA leaching procedure. A study designed to evalate the leachability of 137Cs, 85Sr, and 60Co from simulated boric acid waste solidified in Portland III cement and to measure the compressive strength of the ensuing waste forms before and after leaching was concluded. Leaching data extending over 229 days are presented. The simulated waste forms were leached in deionized water using a modified IAEA leaching procedure. The compressive strength of the specimens was measured initially and after their exposure to a leaching environment for 352 days

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

    International Nuclear Information System (INIS)

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

  13. 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.

  14. Acid digestion of chlorine-containing wastes, (3)

    International Nuclear Information System (INIS)

    Along with the R and D efforts on the treatment of plutonium-contaminated waste materials for volume reduction and stabilization at PNC, studies have been conducted for the development of the technology related to the acid digestion treatment utilizing sulfuric acid and nitric acid. This report is described on the treatment capacity, the treatment of decomposed gas, and the properties of the reduced residues following the treatment. The process involves two steps; the decomposition employing sulfuric acid and the oxidation utilizing nitric acid. The treatment capacity during reaction is influenced by the foaming of the acid digestion solution for the H2SO4 decomposition process. A series of experiments were carried out on H2SO4 decomposition reaction by feeding the crushed samples of 10 mm size into hot sulfuric acid (250 degree C) continuously under a pressure of 100 mm aq. The foaming of the acid digestion solution is closely related to the concentration of solids such as carbon and inorganic materials. Addition of nitric acid to the H2SO4 decomposition process is effective for controlling the foaming of the acid digestion solution. The influence of initial nitric acid concentration upon sulfur oxide oxidation-absorption is very small when the nitric acid concentration is in the range from 14 to 32 percent. (Kato, T.)

  15. 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

  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. Device for conditioning toxic or radioactive wastes containing borate ions and its fabrication process

    International Nuclear Information System (INIS)

    Wastes, especially spent ion exchange resins, are mixed in a container with Portland cement, aluminous cement and water to form a cement matrix containing stable phases of borated ettringite type and/or calcium monoboroaluminate. Mechanical resistance is obtained by subsequent encapsulation of the whole in a mortar

  19. 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

  20. 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.

  1. 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.)

  2. 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

  3. 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

  4. 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.

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

    International Nuclear Information System (INIS)

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

  6. Test container design/fabrication/function for the Waste Isolation Pilot Plant gas generation experiment glovebox

    International Nuclear Information System (INIS)

    The gas generation experiments (GGE) are being conducted at Argonne National Laboratory-West (ANL0W) with contact handled transuranic (CH-TRU) waste in support of the Waste Isolation Pilot Plant (WIPP) in Carlsbad, New Mexico. The purpose of the GGE is to determine the different quantities and types of gases that would be produced and the gas-generation rates that would develop if brine were introduced to CH-TRU waste under post-closure WIPP disposal room conditions. The experiment requires that a prescribed matrix of CH-TRU waste be placed in a 7.5 liter test container. After loaded with the CH-TRU waste, brine and inoculum mixtures (consisting of salt and microbes indigenous to the Carlsbad, New Mexico region) are added to the waste. The test will run for an anticipated time period of three to five years. The test container itself is an ASME rated pressure vessel constructed from Hastelloy C276 to eliminate corrosion that might contaminate the experimental results. The test container is required to maintain a maximum 10% head space with a maximum working pressure of 17.25 MPa (2,500 psia). The test container is designed to provide a gas sample of the head space without the removal of brine. Assembly of the test container lid and process valves is performed inside an inert atmosphere glovebox. Glovebox mockup activities were utilized from the beginning of the design phase to ensure the test container and associated process valves were designed for remote handling. In addition, test container processes (including brine addition, sparging, leak detection, and test container pressurization) are conducted inside the glovebox

  7. Studies on long term leaching behaviour of vitrified waste product containing sulphate bearing high level radioactive waste

    International Nuclear Information System (INIS)

    Borosilicate glass system is adopted in India and world-wide as a matrix for immobilization of high level radioactive liquid waste (HLW). Sulphate bearing HLW is generated during reprocessing of spent fuel from research reactors at BARC, Trombay. The presently stored HLW at Trombay contains uranium, sodium and sulphate in addition to fission products, corrosion products and small amount of other actinides. Presence of sulphate in HLW is attributed to the usage of ferrous sulphamate as a reducing agent in earlier reprocessing flow sheets for valency adjustment of plutonium during partitioning stage. A barium borosilicate based glass matrix is developed for vitrification of sulphate bearing HLW. Assessment of long term chemical durability is one of the critical aspects for evaluation of conditioned products from containment and environmental protection point of view. Chemical durability of waste form is evaluated by studying the leaching behaviour of the conditioned product. Leaching, being the only pathway through which radionuclide can migrate to human environment, is one of the most important properties of vitrified waste product which depends on various factors like composition of waste, glass matrix, type of test method, flow rate, composition of leachant, effect of radiation etc. The present paper reports the details of leaching studies of the glass products made with chemically simulated waste. Efforts were also made to understand the mechanism of leaching and to study the alteration layer formed on the leached surface of the glass products. (author)

  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 (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. 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.

  10. 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...

  11. 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.

  12. Comparison of land and ocean disposal alternatives for bulk wastes containing naturally occurring radionuclides

    International Nuclear Information System (INIS)

    Land and ocean disposal alternatives for a large volume of wastes and residues containing naturally occurring radionuclides are assessed. These wastes and residues are currently stored at the US Department of Energy's Niagara Falls Storage Site near Lewiston, New York. Both land and ocean disposal are considered for the 180,000 m3 of slightly contaminated wastes (average 36 pCi/g radium-226), whereas only land disposal is considered for the 11,000 m3 of residues (average 67,000 pCi/g radium-226). The land and ocean disposal alternatives share similar engineering considerations, occupational and transportation risks, and radiological risks. Impacts from placement of the wastes in the ocean would be negligible. However, the land-based activities required to transport the wastes to the ocean would account for most of the potential impacts associated with the ocean disposal alternatives. Thus, the land and ocean disposal alternatives are comparable in terms of potential environmental impacts

  13. 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

  14. The disposal of Canada's nuclear fuel waste: a study of postclosure safety of in-room emplacement of used CANDU fuel in copper containers in permeable plutonic rock. Volume 2: vault model

    International Nuclear Information System (INIS)

    A study has been undertaken to evaluate the design and long-term performance of a nuclear fuel waste disposal vault based on a concept of in-room emplacement of copper containers at a depth of 500 m in plutonic rock in the Canadian Shield. The containers, each with 72 used CANDU fuel bundles, would be surrounded by clay-based buffer and backfill materials in an array of parallel rooms, with the excavation boundary assumed to have an excavation-disturbed zone (EDZ) with a higher permeability than the surrounding rock. In the anoxic conditions of deep rock of the Canadian Shield, the copper containers are expected to survive for >106 a. Thus container manufacturing defects, which are assumed to affect approximately 1 in 5000 containers, would be the only potential source of radionuclide release in the vault. The vault model is a computer code that simulates the release of radionuclides that would occur upon contact of the used fuel with groundwater, the diffusive transport of these radionuclides through the defect in the container shell and the surrounding buffer, and their dispersive and convective transport through the backfill and EDZ into the surrounding rock. The vault model uses a computationally efficient boundary integral model (BIM) that simulates radionuclide mass transport in the engineered barrier system as a point source (representing the defective container) that releases radionuclides into concentric cylinders, that represent the buffer, backfill and EDZ. A 3-dimensional finite-element model is used to verify the accuracy of the BIM. The results obtained in the present study indicates the effectiveness of a design using in-room emplacement of long-lived containers in providing a safe disposal system even under permeable geosphere conditions. (author). refs., tabs., figs

  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. 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.

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

    International Nuclear Information System (INIS)

    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

  18. Application analysis of high integrity container on domestic radioactive waste management

    International Nuclear Information System (INIS)

    This paper simply described three kinds of material high integrity containers, and accordingly emphasized the cross linked polyethylene HIC used in the domestic projects under construction, focusing on the waste treatment proposal coupling with HIC model and the advantages and disadvantages comparing with the cement solidification proposal. Many aspects are analyzed including waste filling and HIC lifting, transportation, and final disposal. The potential solutions are pointed out for the issues and the post actions as well. (authors)

  19. 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.)

  20. Production of aromatic compounds containing nitrogen during high level radioactive waste processing

    International Nuclear Information System (INIS)

    The Savannah River Site Defense Waste Processing Facility will use a slurry-fed melter to immobilize high-level radioactive waste in borosilicate glass for permanent storage. During the melter feed operation, some nitrogen-containing organic species are generated due to the nitrite in the sludge and the organic by-products in the aqueous product. The chemistry involved is summarized in this paper

  1. 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 ...

  2. 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)

  3. 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

  4. 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.

  5. Ocean disposal option for bulk wastes containing naturally occurring radionuclides: an assessment case history

    International Nuclear Information System (INIS)

    There are 180,000 m3 of slightly contaminated radioactive wastes (36 pCi/g radium-226) currently stored at the US Department of Energy's Niagara Falls Storage Site (NFSS), near Lewiston, New York. These wastes resulted from the cleanup of soils that were contaminated above the guidelines for unrestricted use of property. An alternative to long-term management of these wastes on land is dispersal in the ocean. A scenario for ocean disposal is presented for excavation, transport, and emplacement of these wastes in an ocean disposal site. The potential fate of the wastes and impacts on the ocean environment are analyzed, and uncertainties in the development of two worst-case scenarios for dispersion and pathway analyses are discussed. Based on analysis of a worst-case pathway back to man, the incremental dose from ingesting fish containing naturally occurring radionuclides from ocean disposal of the NFSS wastes is insignificant. Ocean disposal of this type of waste appears to be a technically promising alternative to the long-term maintenance costs and eventual loss of containment associated with management in a near-surface land burial facility

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

    Directory of Open Access Journals (Sweden)

    Cahit GÜRER

    2016-03-01

    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.

  7. 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)

  8. 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)

  9. 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

  10. 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.

  11. Corrosion susceptibility of steel drums containing cemented intermediate level nuclear wastes

    International Nuclear Information System (INIS)

    Cementation processes are used as immobilization techniques for low or intermediate level radioactive waste for economical and safety reasons and for being a simple operation. In particular, ion-exchange resins commonly used for purification of radioactive liquid waste from nuclear reactors are immobilized before being stored to improve the leach resistance of the waste matrix and to maintain mechanical stability. Combustible solid radioactive waste can be incinerated and the resulting ashes can also be immobilized before storage. The immobilized resins and ashes are then contained in steel drums that may undergo corrosion depending on the presence of certain contaminants. The work described in this paper was aimed at evaluating the corrosion susceptibility of steel drums in contact with cemented ion-exchange resins and incineration ashes containing different concentrations of aggressive species (mostly chloride and sulphate ions). A special type of specimen was designed to simulate the cemented waste in the drum. The evolution of the corrosion potential and the corrosion current density of the steel, as well as the electrical resistivity of the matrix were monitored over a time period of 1 year. The results show the deleterious effect of chloride on the expected lifespan of the waste containers.

  12. 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)

  13. Design and fabrication of demonstration containers for long term storage and geological disposal of radioactive waste

    International Nuclear Information System (INIS)

    In the framework of topics 2 and 3 of the 1991 radioactive waste management act, containers suitable both for long-term interim storage and for disposal in deep geological strata have been designed and fabricated since 2002. Their main characteristics and the results obtained are subject of this article. These containers are meant for B type waste, i.e. Long-Lived Medium Activity waste, resulting mainly from spent fuel reprocessing (e.g.: hulls and end-pieces) and from the operation of research centers. The wastes are packed in primary containers, resulting from the last stage of packaging by the waste producer. The common storage/disposal container is an overpack or external container, housing several primary waste packages. According to the management scenarios identified in the ANDRA Design Basis Inventory Model (MID), the total number of primary containers to be taken into account is about 200 000. In the MID these primary packages were classified in 42 categories of various characteristics and types, corresponding to 8 standard packages. These reference packages representing the overall inventory, serve as a basis for dimensioning the external containers. The results presented here are based on reference bituminized waste packages (called B2) and on reference packages for compacted hulls and end-pieces (called B5). The full-scale demonstration prototypes, or 'demonstrators' were based on these two reference packages: B2 for the reference concept and B5 for the enhanced containment concept. These two concepts will be further explained below. The paper has the following contents: Introduction; Container functions; Design requirements and joint container standard concept; Choice of materials; Standard container geometry; Justification and advantages of stainless steel metal reinforcement in the concrete containers; Variant of the reference container; Exploratory variant:Container with enhanced containment capability (application to B5); Concrete formulation; Full

  14. 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

  15. Continuous extraction of uranium from actual uranium-containing liquid wastes using an 'emulsion flow' extractor

    International Nuclear Information System (INIS)

    A newly developed liquid waste treatment system using an 'emulsion flow' extractor has been applied to actual uranium-containing liquid wastes that originated from the decontamination of used gas centrifuges at Ningyo-toge Environmental Engineering Center of Japan Atomic Energy Agency. The emulsion flow extractor performs efficient liquid-liquid extraction by supplying solutions without additional stirring or shaking. The solvent used in this system is an isooctane solution containing TnOA and 1-octanol, which is effective in the selective extraction of uranium, without the formation of the third phase, from dilute sulfuric acid solutions containing a large amount of Fe. With the use of this system, 90% or more of uranium is extracted from actual and simulated decontamination liquid wastes under such mild emulsion flow conditions that fine drops of organic phases do not leak outside the apparatus. (author)

  16. 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

  17. Fabrication and closure development of nuclear waste disposal containers for the Yucca Mountain Project: Status report

    International Nuclear Information System (INIS)

    In GFY 89, a project was underway to determine and demonstrate a suitable method for fabricating thin-walled monolithic waste containers for service within the potential repository at Yucca Mountain. A concurrent project was underway to determine and demonstrate a suitable closure process for these containers after they have been filled with high-level nuclear waste. Phase 1 for both the fabrication and closure projects was a screening phase in which candidate processes were selected for further laboratory testing in Phase 2. This report describes the final results of the Phase 1 efforts. It also describes the preliminary results of Phase 2 efforts

  18. 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

  19. Decontamination flowsheet development for a waste oil containing mixed radioactive contaminants

    International Nuclear Information System (INIS)

    The majority of waste oils contaminated with both radioactive and hazardous components are generated in nuclear power plant, research lab. and uranium-refinery operations. The waste oils are complex, requiring a detailed examination of the waste management strategies and technology options. It may appear that incineration offers a total solution, but this may not be true in all cases. An alternative approach is to decontaminate the waste oils to very low contaminant levels, so that the treated oils can be reused, burned as fuel in boilers, or disposed of by commercial incineration. This paper presents selected experimental data and evaluation results gathered during the development of a decontamination flowsheet for a specific waste oil stores at Chalk River Labs. (CRL). The waste oil contains varying amounts of lube oils, grease, paint, water, particulates, sludge, light chloro- and fluoro-solvents, polychlorinated biphenyls (PCB), complexing chemicals, uranium, chromium, iron, arsenic and manganese. To achieve safe management of this radioactive and hazardous waste, several treatment and disposal methods were screened. Key experiments were performed at the laboratory-scale to confirm and select the most appropriate waste-management scheme based on technical, environmental and economic criteria. The waste-oil-decontamination flowsheet uses a combination of unit operations, including prefiltration, acid scrubbing, and aqueous-leachage treatment by precipitation, microfiltration, filter pressing and carbon adsorption. The decontaminated oil containing open-quotes de minimisclose quotes levels of contaminants will undergo chemical destruction of PCBs and final disposal by incineration. The recovered uranium will be recycled to a uranium milling process

  20. Precipitation and Deposition of Aluminum-Containing Species in Tank Wastes

    International Nuclear Information System (INIS)

    Aluminum-containing phases represent the most prevalent solids that can appear or disappear during the processing of radioactive tank wastes. Processes such as sludge washing and leaching are designed to dissolve Al-containing phases and thereby minimize the volume of high-level waste glass required to encapsulate radioactive sludges. Unfortunately, waste-processing steps that include evaporation can involve solutions that are supersaturated with respect to cementitious aluminosilicates that result in unwanted precipitation and scale formation. Of all the constituents of tank waste, limited solubility cementitious aluminosilicates have the greatest potential for clogging pipes and transfer lines, fouling highly radioactive components such as ion exchangers, and completely shutting down processing operations. For instance, deposit buildup and clogged drain lines experienced during the tank waste volume-reduction process at the Savannah River Site (SRS) required an evaporator to be shut down in October 1999. The Waste Processing Technology Section of Westinghouse Savannah River Company at SRS now is collaborating with team members from Pacific Northwest National Laboratory (PNNL) to verify the thermodynamic stability of aluminosilicate compounds under waste tank conditions in an attempt to solve the deposition and clogging problems. The primary objectives of this study are (1) to understand the major factors controlling precipitation, heterogeneous nucleation, and growth phenomena of relatively insoluble aluminosilicates; (2) to determine the role of organics for inhibiting aluminosilicate formation, and (3) to develop a predictive tool to control precipitation, scale formation, and cementation under tank waste processing conditions. The results of this work will provide crucial information for (1) avoiding problematical sludge processing steps and (2) identifying and developing effective technologies to process retrieved sludges and supernatants before ultimate

  1. 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.

  2. 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.

  3. 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

  4. Design report for the interim waste containment facility at the Niagara Falls Storage Site

    International Nuclear Information System (INIS)

    Low-level radioactive residues from pitchblende processing and thorium- and radium-contaminated sand, soil, and building rubble are presently stored at the Niagara Falls Storage Site (NFSS) in Lewiston, New York. These residues and wastes derive from past NFSS operations and from similar operations at other sites in the United States conducted during the 1940s by the Manhattan Engineer District (MED) and subsequently by the Atomic Energy Commission (AEC). The US Department of Energy (DOE), successor to MED/AEC, is conducting remedial action at the NFSS under two programs: on-site work under the Surplus Facilities Managemnt Program and off-site cleanup of vicinity properties under the Formerly Utilized Sites Remedial Action Program. On-site remedial action consists of consolidating the residues and wastes within a designated waste containment area and constructing a waste containment facility to prevent contaminant migration. The service life of the system is 25 to 50 years. Near-term remedial action construction activities will not jeopardize or preclude implementation of any other remedial action alternative at a later date. Should DOE decide to extend the service life of the system, the waste containment area would be upgraded to provide a minimum service life of 200 years. This report describes the design for the containment system. Pertinent information on site geology and hydrology and on regional seismicity and meteorology is also provided. Engineering calculations and validated computer modeling studies based on site-specific and conservative parameters confirm the adequacy of the design for its intended purposes of waste containment and environmental protection

  5. 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.

  6. 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

  7. 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

  8. 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.)

  9. 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.

  10. Safety requirements for the design of large containers for decommissioning waste

    International Nuclear Information System (INIS)

    About 50% of the total volume of conditioned radioactive waste from nuclear power generation will finally result from the decommissioning of nuclear power plants (NPPs). Higher activated metallic waste from the core region of the reactors would, according to the International Transport Regulations (IAEA), require a type B container. This would, however, bring a significant increase in the costs of the management of such waste. A considerably cheaper solution of this problem can be achieved by separating the protection requirements for type B packaging (i.e. mechanical integrity and tightness). By using industrial packaging (IP) designed for a higher mechanical integrity, it is possible to cope with higher IAEA protection goals for the safe transport of dismantling waste without strictly following the extremely high and also expensive tightness requirements for type B packaging. This is because the radioactivity is bound in the metallic lattice of such activated decommissioning waste.The above mentioned strong IP for decommissioning waste can also be used for other highly activated waste from the core region (e.g. fuel element boxes, control rods or other activated equipment) during the operation of the NPPs. (orig.)

  11. 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

  12. 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.)

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

    International Nuclear Information System (INIS)

    Six alloys are being considered as possible materials for the fabrication of containers for the disposal of high-level radioactive waste. Three of these candidate materials are copper-based alloys: CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni). The other three are iron- to nickel-based austenitic materials: Types 304L and 316L stainless steels and Alloy 825. Radioactive waste will include spent-fuel assemblies from reactors as well as waste in borosilicate glass and will be sent to the prospective site at Yucca Mountain, Nevada, for disposal. The waste-package 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, the containers must be retrievable from the disposal site. Shortly after emplacement of the containers in the repository, they will be exposed to high temperatures and high gamma radiation fields from the decay of high-level waste. This radiation will promote the radiolytic decomposition of moist air to hydrogen. This volume surveys the available data on the effects of hydrogen on the six candidate alloys for fabrication of the containers. For copper, the mechanism of hydrogen embrittlement is discussed, and the effects of hydrogen on the mechanical properties of the copper-based alloys are reviewed. The solubilities and diffusivities of hydrogen are documented for these alloys. For the austenitic materials, the degradation of mechanical properties by hydrogen is documented. The diffusivity and solubility of hydrogen in these alloys are also presented. For the copper-based alloys, the ranking according to resistance to detrimental effects of hydrogen is: CDA 715 (best) > CDA 613 > CDA 102 (worst). For the austenitic alloys, the ranking is: Type 316L stainless steel ∼ Alloy 825 > Type 304L stainless steel (worst). 87 refs., 19 figs., 8 tabs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1988-08-01

    Six alloys are being considered as possible materials for the fabrication of containers for the disposal of high-level radioactive waste. Three of these candidate materials are copper-based alloys: CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni). The other three are iron- to nickel-based austenitic materials: Types 304L and 316L stainless steels and Alloy 825. Radioactive waste will include spent-fuel assemblies from reactors as well as waste in borosilicate glass and will be sent to the prospective site at Yucca Mountain, Nevada, for disposal. The waste-package 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, the containers must be retrievable from the disposal site. Shortly after emplacement of the containers in the repository, they will be exposed to high temperatures and high gamma radiation fields from the decay of high-level waste. This radiation will promote the radiolytic decomposition of moist air to hydrogen. This volume surveys the available data on the effects of hydrogen on the six candidate alloys for fabrication of the containers. For copper, the mechanism of hydrogen embrittlement is discussed, and the effects of hydrogen on the mechanical properties of the copper-based alloys are reviewed. The solubilities and diffusivities of hydrogen are documented for these alloys. For the austenitic materials, the degradation of mechanical properties by hydrogen is documented. The diffusivity and solubility of hydrogen in these alloys are also presented. For the copper-based alloys, the ranking according to resistance to detrimental effects of hydrogen is: CDA 715 (best) > CDA 613 > CDA 102 (worst). For the austenitic alloys, the ranking is: Type 316L stainless steel {approx} Alloy 825 > Type 304L stainless steel (worst). 87 refs., 19 figs., 8 tabs.

  15. 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

  16. Risks attached to container- and bunker-storage of nuclear waste

    International Nuclear Information System (INIS)

    The results are presented of a literature study into the risks attached to the two dry-storage options selected by the Dutch Central Organization For Radioactive Waste (COVRA): the container- and the bunker-storage for irradiated nuclear-fuel elements and nuclear waste. Since the COVRA does not make it clear how these concepts should have to be realized, the experiences abroad with dry interim-storage are considered. In particular the Castor-container-storage and the bunker storage proposed in the committee MINSK (Possibilities of Interim-storage in the Netherlands of Irradiated nuclear-fuel elements and Nuclear waste) are studied further in depth. The committee MINSK has performed a study into the technical realizability of various interim-storage facilities, among which a storage in bunkers. (author). 75 refs.; 14 figs.; 16 tabs

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

    International Nuclear Information System (INIS)

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

  18. Preparation of magnesium phosphate cement by recycling the product of thermal transformation of asbestos containing wastes

    OpenAIRE

    Viani, A; Gualtieri, A.F.

    2014-01-01

    Asbestos containing wastes have been employed for the first time in the formulation of magnesium phosphate cements. Two samples were mixed with magnesium carbonate and calcined at 1100 and 1300 C. Under these conditions, complete destruction of asbestos minerals is known to occur. The product, containing MgO, after reaction with water-soluble potassium di-hydrogen phosphate, led to the formation of hydrated phases at room temperature. Crystalline and amorphous reaction products were detected,...

  19. Technology of obtaining of cryolite and aluminium fluoride from alumina- and fluoride containing wastes of aluminium production

    International Nuclear Information System (INIS)

    This article is devoted to technology of obtaining of cryolite and aluminium fluoride from alumina- and fluoride containing wastes of aluminium production. Thus, the investigations on elaboration of technology of obtaining of cryolite and aluminium fluoride from alumina- and fluoride containing wastes of aluminium production by means of sulfuric acid decomposition method are carried out. The optimal parameters of technological processes are found. The physicochemical analysis of fluoride containing wastes is conducted. The flowsheet of obtaining of cryolite, aluminium fluoride and alumina from alumina- and fluoride containing wastes is presented.

  20. Kinetics of sulfuric acid decomposition of solid fluorine containing wastes of aluminium production

    International Nuclear Information System (INIS)

    The results of researches of kinetics of sulfuric acid decomposition of solid fluorine containing wastes of aluminium production were considered in this work. The apparent activation energy was determined. The rate constant and pre-exponential factor of Arrhenius equation were defined as well.

  1. 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.

  2. 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

  3. 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.

  4. 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.

  5. 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

  6. Biochemical process of low level radioactive liquid simulation waste containing detergent

    International Nuclear Information System (INIS)

    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. Development of electrochemical ion exchange electrodes for the treatment of wastes containing chromium or cesium ions

    International Nuclear Information System (INIS)

    Nowadays, environmental preservation using technologies that do not attack it, generating non-toxic residues and reduced volumes, has been discussed. Hazardous effluents, containing metals, as chromium, have been poured in the soils and rivers, degrading the water. Not different are the problems originated from some nuclear activities that generate wastes, as in chemical research laboratories. Although those wastes are not poured in the environment, sometimes they are inadequately stored, what can cause serious accidents. With the purpose of solving this problem, there are some techniques to waste treatment, between them there is the electrochemical ion exchange (EIX). EIX is an advanced process that has advantages over traditional ion exchange and the fact of using the electron as the only reagent, reduces the volume of the solution to be treated. This technique consists of development of an electrode, where an ion exchanger is physically incorporated in an electrode structure with a binder. In this study, cationic resin Amberlite CG-50 and zirconium phosphate have been chosen for the separation of chromium and cesium from waste, respectively. They were chosen because they present high chemical stability in oxidizing media and at ionizing radiation. The quantity of charcoal, graphite and binder used in formulation of electrode have been studied either. Before choosing the best electrode, it was verified sorption percentage of 99,3% for chromium and 99,8% for cesium. The greater advantage of this process is the total elution of chromium as much as cesium, without reagents addition, being possible to reuse the electrode without losing its capacity. Beside on the results, a continuous process for the wastes containing Cr and Cs, using a flux electrolytic cell (CELFLUX) of high retention capacity, was presented. The high efficiency of this cell for both retention and elution, leading to an important reduction of waste volume, and, every more, making possible the

  8. 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.

  9. 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

  10. 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)

  11. Possible recipes for conditioning low radioactive asbestos-containing waste material

    International Nuclear Information System (INIS)

    Cementing seems to be the easiest way to condition asbestos-containing low radioactive waste material, but it must be considered that in practice the term 'asbestos waste' can subsume a lot of materials with very different properties. In this case the asbestos waste could be assigned to 3 different categories, cement bound asbestos, gypsum-bound fibres and asbestos contaminated fibre mats. For a cement recipe, several preliminary tests were performed with plastic fibers or rock wool in order to mimic the behaviour of asbestos waste. The high water demand makes it impossible to find a flowable recipe for cementing the cement-bound blue asbestos with a reasonable waste and water content. Despite this, nine recipes to solidify the cement-bound asbestos were tested as to whether they fulfill the guideline B05 of the Swiss Federal Nuclear Safety Inspectorate. Three of the nine mixtures had 90-days compressive strengths of more than the required 10 MPa, but had a plastic or stiff consistency. The best solution to solidify the gypsum waste seemed to use Calcium Sulpho-aluminate cement. Four of seven gypsum recipes fulfilled the requirements of the B05. One had a supple, almost flowable consistency and could be foreseen for the solidification process. (authors)

  12. Development and assessment of a fiber reinforced HPC container for radioactive waste

    International Nuclear Information System (INIS)

    As part of its research into solutions for concrete disposal containers for long-lived radioactive waste, Andra defined requirements for high-performance concretes with enhanced porosity, diffusion, and permeability characteristics. It is the starting point for further research into severe conditions of containment and durability. To meet these objectives, Eiffage TP consequently developed a highly fibered High Performance Concrete (HPC) design mix using CEM V cement and silica fume. Then, mockups were produced to characterize the performance various concepts of containers with this new concrete mix. These mockups helped to identify possible manufacturing problems, and particularly the risk of cracking due to restrained shrinkage. (authors)

  13. Processing method for liquid waste containing ammonia nitrogen and organic material

    International Nuclear Information System (INIS)

    For removing scales (iron oxide, copper oxide) contaminated by radioactivity such as of a steam generator in a nuclear power plant, a decontaminating liquid containing a chelating agent which forms stable water soluble chelating compounds with the metals contained in the scales is used. In this case, NH4OH is added to control the pH of the liquid wastes to 9. Therefore, radioactive materials, complexes of the chelate and NH3 are contained in the decontaminated liquid wastes in addition to the scales. Such liquid wastes are electrolyzed to liberate the NH3 which formed complexes. The free NH3 is removed by scattering using air blown from a scattering tube disposed to the bottom of an electrolysis vessel to reduce the volume of the decontaminated materials. Namely, ammonium sulfate (which increases the amount of solidification products since it remains as solids) formed by neutralization by the addition of H2SO4 to the liquid wastes after the electrolysis treatment can be suppressed. (T.M.)

  14. 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

  15. Experimental drop testing of waste containers for the Konrad repository - 59269

    International Nuclear Information System (INIS)

    Document available in abstract form only. Full text of publication follows: The Konrad repository for not heat generating radioactive wastes was licensed in 2002 primarily. Due to legal actions the final confirmation of this license took place not until 2007. Subsequently, the Federal Office for Radiation Protection (BfS) began scheduling backfitting of the former iron ore mine into a repository. The licensed repository volume is 303, 000 m3 considering estimations of expected waste volumes to be disposed off. The mine itself would offer a much larger volume. Waste packages can be disposed off as recently as the repository is ready for operation what is expected not before the end of this decade. Nevertheless, there is high interest of qualified and certified waste conditioning and packaging for disposal today, what for from BAM and BfS tested, evaluated and certified containers are needed. In recent years numerous container prototypes made of steel, concrete and ductile cast iron have been tested by BAM, the Federal Institute for Materials Research and Testing in Germany. To cover the Konrad test requirements in a conservative manner container drop tests are performed mostly onto the unyielding IAEA target of BAMs large drop test facility instead of a representative foundation of the repository

  16. 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.

  17. In-Drift Accumulation of Fissile Material From Waste Packages Containing Plutonium Disposition Waste Forms

    International Nuclear Information System (INIS)

    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

  18. 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

  19. 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

    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 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...

  20. NEATER robot carries out remote contamination monitoring of high-level vitrified waste containers

    International Nuclear Information System (INIS)

    A NEATER Nuclear Engineered Robot, has been successfully installed during a planned shutdown in the High Level Waste Plant at British Nuclear Fuels plc (BNFL) Sellafield site. The robot swabs vitrified waste containers to ensure they are free of surface contamination before they are put into store. Engineering of the NEATER Robot system was carried out by the Remote Handling and Robotics Department of AEA Technology (RHRD), working closely with BNFL. The success of the NEATER robot system demonstrates how planned use of mock up trials can minimise difficulties encountered during active commissioning and give increased confidence of successful installations. (author)

  1. 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

  2. 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.

  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. 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 high-level radioactive-waste disposal containers. The waste will include spent fuel assemblies from reactors as well as high-level waste in borosilicate glass and will be sent to the prospective site at Yucca Mountain, Nevada, for disposal. The copper-based alloy materials are CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni). The austenitic materials are Types 304L and 316L stainless steels and Alloy 825. The waste-package containers must maintain substantially complete containment for at least 300 yr and perhaps as long as 1000 yr, and they must be retrievable from the disposal site during the first 50 yr after emplacement. The containers will be exposed to high temperatures and high gamma radiation fields from the decay of high-level waste. This volume surveys the available data on the phase stability of both groups of candidate alloys. The austenitic alloys are reviewed in terms of the physical metallurgy of the iron-chromium-nickel system, martensite transformations, carbide formation, and intermetallic-phase precipitation. The copper-based alloys are reviewed in terms of their phase equilibria and the possibility of precipitation of the minor alloying constituents. For the austenitic materials, the ranking based on phase stability is: Alloy 825 (best), Type 316L stainless steel, and then Type 304L stainless steel (worst). For the copper-based materials, the ranking is: CDA 102 (oxygen-free copper) (best), and then both CDA 715 and CDA 613. 75 refs., 24 figs., 6 tabs

  5. 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

  6. Preliminary analysis of the creep behaviour of nuclear fuel-waste container materials

    International Nuclear Information System (INIS)

    In the Canadian Nuclear Fuel Waste Management Program, it is proposed that nuclear fuel waste be placed in a durable container and disposed of in a deep underground vault. Consideration of various disposal-container designs has identified either titanium or copper as the material suitable for constructing the container shell. As part of the R and D program to examine the structural integrity of the container, creep tests are being conducted on commercially pure titanium and oxygen-free copper. This report presents the preliminary data obtained. It also describes the evaluation of various constitutive equations to represent the creep curves, thus providing the basis for extrapolation of the creep behaviour over the design lifetime of the container. In this regard, a specific focus is placed on equations derived from the 0-Projection Concept. Recognizing that the container lifetime will be determined by the onset of tertiary creep leading to creep rupture, we present the results of the metallographic examination of creep damage. This shows that the tertiary stage in titanium is associated with the formation of transgranular cavities within the region of localized necking of the creep specimens. In contrast, creep damage in copper is in the form of intergranular cavities uniformly distributed throughout the gauge length. These results are analyzed within the context of the extant literature, and their implications for future container design are discussed. (author)

  7. 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

  8. 5.3. Obtaining of cryolite-alumina concentrate from carbon-, and fluorine containing wastes by burning method

    International Nuclear Information System (INIS)

    The method of obtaining of cryolite-alumina concentrate from carbon-, and fluorine containing wastes by means of burning method was elaborated. The flowsheet of obtaining of cryolite-alumina concentrate from carbon-, and fluorine containing wastes by means of burning method was considered and presented in this article.

  9. Technology for Efficient Usage of Hydrocarbon-Containing Waste in Production of Multi-Component Solid Fuel

    OpenAIRE

    B. M. Khroustalev; A. N. Pekhota

    2016-01-01

    The paper considers modern approaches to usage of hydrocarbon-containing waste as energy resources and presents description of investigations, statistic materials, analysis results on formation of hydrocarbon-containing waste in the Republic of Belarus. Main problems pertaining to usage of waste as a fuel and technologies for their application have been given in the paper. The paper describes main results of the investigations and a method for efficient application of viscous hydrocarbon-cont...

  10. 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

  11. 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.)

  12. Thermal integrity of packages containing vitrified high-level radioactive wastes under sea surface fire

    International Nuclear Information System (INIS)

    Some spent fuels from light-water reactors have been reprocessed in the U.K and France, and some of the high-level radioactive wastes generated by such reprocessing have been returned to Japan. In order to ensure the safety sea transport of vitrified high-level radioactive wastes, thermal analyses of the packages were conducted under sea surface fire accidents. According to thermal analyses results of an exclusive ship using the thermal characteristic test results for materials which compose hatch cover members in a cargo hold, the thermal integrity of packages containing vitrified high-level radioactive wastes under sea surface fire accidents is consequently maintained both in the cases that the emergency water flooding system operates and does not operate. (author)

  13. 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)

  14. Researches on the production of self-reducing briquettes from waste containing iron and carbon

    International Nuclear Information System (INIS)

    The extension of the raw material basis for the steel making industry represents a priority nowadays, within the context of sustainable development. The issue is of utmost importance particularly now when the environment legislation sets strict dumping conditions on the one hand and, on the other hand, when huge quantities of waste, already deposited in dumps and ponds raise serious problems in terms of meeting such conditions. The paper introduces some researches on exploiting waste containing iron and carbon in steel making; this powdery and small-grain waste is to be processed into briquettes. The self-reducing briquettes can be used in steel elaboration in electric arc furnaces, replacing the scrap, which is scarce

  15. The nuclear waste containment and some aspects of the deep disposal concept

    International Nuclear Information System (INIS)

    The French agency for the management of nuclear waste, ANDRA, is in charge of investigating the feasibility of deep disposal of high level waste in at least two types of geologic formation, leading to the validation of disposal concepts with and without retrievability. Plans to build two underground laboratories are afoot. Meanwhile, parametric modelling studies have been performed, with interesting results, some of which are shown here in graphic form. It is proved that if the overpack surrounding waste containers can be made to last for a thousand years, the dose resulting from Sr-90 and Cs-137 is nil. Conversely, the dose from actinides such as americium and Th-229 is largely unaffected by the package, being determined by their own low solubilities and underground water flow. Temperature rise in a granite host formation was modelled as a function of the distance between disposal boreholes. Finite element two dimensional calculations of water flow through backfill were also performed. 1 ref., 8 figs

  16. 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.

  17. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1988-08-01

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

  19. 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

  20. Viscoplasticity of simulated high-level radioactive waste glass containing platinum group metal particles

    International Nuclear Information System (INIS)

    The shear rate dependency of the viscosity of three simulated high-level radioactive waste glasses containing 0, 1.2 and 4.5 wt% platinum group metals (PGMs) was examined at a temperature range of 1173–1473 K by a rotating viscometer. Shear stress when the shear rate equals zero, i.e. yield stress, was also measured by capillary method. The viscosity of the glass containing no PGM was shear rate-independent Newtonian fluid. On the other hand, the apparent viscosity of the glasses containing PGMs increased with decreasing shear rate, and nonzero amount of yield stresses were detected from both glasses. The viscosity and yield stress of the glass containing 4.5 wt% PGMs was roughly one to two orders of magnitude greater than the glass containing 1.2 wt% PGMs. These viscoplastic properties were numerically expressed by Casson equation

  1. 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

  2. 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).

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

    International Nuclear Information System (INIS)

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

  4. 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.

  5. 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

  6. 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

  7. Accelerated damage studies of titanate ceramics containing simulated PW-4b and JW-A waste

    International Nuclear Information System (INIS)

    Ceramic waste forms are affected by radiation damage, primarily arising from aloha-decay processes that can lead to volume expansion and amorphization of the component crystalline phases. The understanding of the extent and impact of these effects on the overall durability of the waste form is critical to the prediction of their long-term performance under repository conditions. Since 1985 ANSTO and JAERI have carried out joint studies on the use of 244Cm to simulate alpha-radiation damage in ceramic waste forms. These studies have focussed on synroc formulations doped with simulated PW-4b and JW-A wastes. The studies have established the relationship between density change and irradiation levels for Synroc containing JW-A and PW-4b wastes. The storage of samples at 200 C halves the rate of decrease in the density of the samples compared to that measured at room temperature. This effect is consistent with that found for natural samples where the amorphization of natural samples stored under crustal conditions is lower, by factors between 2 and 4, than that measured for samples from accelerated doping experiments stored at room temperature. (J.P.N.)

  8. Electrochemical incineration of C-14-containing liquid wastes. First results and outlook

    International Nuclear Information System (INIS)

    Liquid radioactive wastes containing carbon-14 are not acceptable for radioactive waste repository. At present incineration is the only approved way of treatment in Germany. But capacities are limited and incineration by itself is known to be a technically complex and rather expansive process. Therefore within an experimental proof of concept it should be examined whether an electrochemical incineration process is also applicable for this purpose. The R and D activities mainly comprised the gathering of informations about the chemical nature and constitution of such liquid wastes from enterprises, research laboratories and a number of the state owned collecting facilities, numerous electrochemical examinations with typical organic substances and finally electrolysis tests with original carbon-14 waste solution at lab-scale. We were able to demonstrate that electrochemical oxidation permits almost quantitative mineralization of the organic waste substances. The carbon dioxide released during electrolysis was completely fixed as solid calcium carbonate which is acceptable for final repository. An estimate shows that in comparison to conventional incineration a substantial decrease in the costs of treatment and repository can be expected. With regard to technical application further steps of development are necessary. (orig.)

  9. Method of processing boric acid-containing liquid wastes discharged from nuclear power plants

    International Nuclear Information System (INIS)

    Purpose: To process boric acid-containing liquid wastes discharged from nuclear power plants by ultrafiltration, further process the filtrates with ion exchange resins and recover the processed liquid as primary coolants. Method: Boric acid-containing liquid wastes are sent from a storage tank into a circulation tank, removed with suspended substances through filters and then supplied to an ultrafiltration film module. The liquid suspensions separated in the ultrafiltration film module are returned to the circulation tank. While on the other hand, the filtrates from the ultrafiltration film module are sent into an ion exchange resin column, where ionic impurities are removed. The liquid discharged from the ion exchange resin column is a pure boric acid solution, which is recovered as primary coolants and used again. The suspension substances concentrated in the ultrafiltration film module are then solidified by cements or the likes. (Moriyama, K.)

  10. 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

  11. 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.

  12. 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

  13. 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...

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

    OpenAIRE

    MAJID MATOUQ ASSAS

    2016-01-01

    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 a...

  16. Theoretical modeling of crevice and pitting corrosion processes in relation to corrosion of radioactive waste containers

    International Nuclear Information System (INIS)

    A mathematical and numerical model for evaluation of crevice and pitting corrosion in radioactive waste containers is presented. The model considers mass transport, mass transfer at the metal/solution interface, and chemical speciation in the corrosion cavity. The model is compared against experimental data obtained in artificial crevices. Excellent agreement is found between modeled and experimental values. The importance of full consideration of complex ion formation in the aqueous solution is emphasized and illustrated. 10 refs., 5 figs

  17. Determination of the amount of ion exchange resin in concrete containing radioactive wastes

    International Nuclear Information System (INIS)

    A method for determining the amount of ion exchange resin in waste concrete was tested on a (approximately 2 g) piece of concrete containing a known amount of ion exchange resin. The difference between the reference and the analysis values was less than ten per cent, and it seems likely that the reproducibility is considerably better. It was concluded that the method is suitable for homogeneity determinations, although some further experiments are needed before it can be used as a standard method. (Auth.)

  18. Long-term storage of waste ion-exchange resins in high-density polyethylene containers

    International Nuclear Information System (INIS)

    This research effort is being carried out for the Empire Static Electric Energy Research Corporation (ESEERCO) to evaluate the use of high-density polyethylene (HDPE) as a container material for the extended storage of low-level radioactive waste ion-exchange resins. This need arises because of the possibility that a new disposal site may not be commissioned soon enough for wastes to be shipped away. If the wastes are to be stored for periods greater than five years application to the NRC for extended storage must include an analysis of the integrity of the container and its retrievability. In order to ensure that New York State utilities have the necessary database to include in such an application, ESEERCO has contracted with Brookhaven National Laboratory (BNL) to quantify the integrity of HDPE containers for resin storage periods in excess of five years. Based on past BNL research it is considered that three procedural steps should be implemented if a container is to qualify for long-term use. These are: (a) Determine storage conditions and state which degradation modes are likely to be applicable; (b) Specify container performance criteria that must be met for safe storage; and (c) Carry out testing and analyses to demonstrate that these requirements will be met over the storage period. It is envisioned that the study will provide the information needed to determine if the current, inexpensive HDPE containers in use will serve effectively for extended storage. This will be done by quantifying anticipated degradation rates and specifying means to retard their progress

  19. 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

  20. Study on hydrogen embrittlement property of titanium for nuclear fuel waste container

    International Nuclear Information System (INIS)

    In geologic disposal system of high-level radioactive waste, confinement by waste container must be assured over a thousand years. Titanium is one of the candidate materials, so it is important to clarify hydrogen embrittlement property under geological environment for the container lifetime prediction. The purpose of this study is to investigate hydrogen embrittlement behavior of titanium under reducing condition. Hydrogen was absorbed into titanium test pieces by electrochemical method, and tensile bending and impact tests were performed for mechanical property research. Under 1000 ppm concentration of hydrogen, while distinct degradation of mechanical properties by hydrogen embrittlement occurred on dynamic stress, micro cracks induced by hydride were observed in fracture, but distinct degradation of mechanical properties by hydrogen embrittlement did not occur on static stress. Under low oxygen circumstances, corrosion rates of titanium were estimated 10-2 micrometer/year by hydrogen absorption method, on the contrary to 10-4 micrometer/year by gas evolution method. These results indicated hydrogen generated by corrosion of titanium under reducing condition, is almost absorbed into material. Carbon steel is regarded as reinforcement of the titanium nuclear fuel waste container. Magnetite, corrosion product of carbon steel, is considered to accelerate corrosion rate. Contribution of hydrogen evolution reaction to its acceleration is estimated to ca.60%. (author)

  1. 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.

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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

  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. 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

  9. 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

  10. Determination of the calibration curve for an ST5 box containing baled waste

    International Nuclear Information System (INIS)

    The Trash Monitor Station (TMS), a nondestructive examination facility at the Oak Ridge Y-12 Plant, utilizes gamma scintillation detectors for the isotopic characterization of possible radioactive contamination and measurement of the quantity of potential depleted uranium contamination present in solid waste. Experimental data has been collected and analyzed for construction of a calibration curve for an ST5 box containing baled waste. The raw data consists of gross counts per second derived from five minute counts from three TMS sodium iodide detectors when known amounts of depleted uranium have been placed in a predetermined ''worst case'' location. Extensive analysis of the data indicates that a function of the form y = a + bx + cx1/2, where x and y represent sources and net counts per minute respectively, provides the best characterization of a calibration curve. While this equation provides the best prediction, an established method for calculating the minimum detectable amount (MDA) for this type of equation has notbeen found. Therefore, a piece-wise fit of two lines may be preferable at this time. This study provides two linear and two nonlinear models, any one of which may be appropriate for use as the calibration equation for an ST5 box containing baled waste. Two equations for both the linear and nonlinear model types have been supplied to provide a choice between using cpm from a third sodium iodide detector as background or using cpm from a different region of interest as background

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

    International Nuclear Information System (INIS)

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

  12. Valorisation of different types of boron-containing wastes for the production of lightweight aggregates

    International Nuclear Information System (INIS)

    Four boron-containing wastes (BW), named as Sieve (SBW), Dewatering (DBW), Thickener (TBW) and Mixture (MBW) waste, from Kirka Boron plant in west Turkey were investigated for the formation of artificial lightweight aggregates (LWA). The characterisation involved chemical, mineralogical and thermal analyses as well as testing of their bloating behaviour by means of heating microscopy. It was found that SBW and DBW present bloating behaviour whereas TBW and MBW do not. Following the above results two mixtures M1 and M2 were prepared with (in wt.%): 20 clay mixture, 40 SBW, 40 DBW and 20 clay mixture, 35 SBW, 35 DBW, 10 quartz sand, respectively. Two different firing modes were applied: (a) from room temperature till 760 deg. C and (b) abrupt heating at 760 deg. C. The obtained bulk density for M1 and M2 pellets is 1.2 g/cm3 and 0.9 g/cm3, respectively. The analysis of microstructure with electron microscopy revealed a glassy phase matrix and an extended formation of both interconnected and isolated, closed pores. The results indicate that SBW and DBW boron-containing wastes combined with a clay mixture and quartz sand can be valorised for the manufacturing of lightweight aggregates.

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

    International Nuclear Information System (INIS)

    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 Cr2O3, have been investigated. All of the iron alumina phosphate glasses had an outstanding chemical durability as measured by their small dissolution rate (1 . 10-9 g/(cm2 . 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 2 . d) after 7 d at 200 C by the vapor hydration test (VHT). The solubility limit for Cr2O3 in the iron phosphate melts was estimated at 4.1 wt%, but all of the as-annealed melts contained a few percent of crystalline Cr2O3 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 Cr2O3 which is at least three times larger than that for borosilicate glasses. (orig.)

  14. Proceedings of a workshop on corrosion of Nuclear fuel waste containers

    International Nuclear Information System (INIS)

    The 23 papers presented at this conference review the technical merits, and particularly corrosion performance, of the three main materials used for nuclear fuel waste containers: titanium and its alloys, copper and its alloys, and iron and carbon steels. The specific questions posed to the Workshop were: 1) Can we predict the lifetime of container materials in a variety of vault environments? 2) Is there a limiting range of conditions beyond which a specific material cannot be used? 3) Do we have the necessary corrosion rate data and/or mechanistic models required to make predictions? 4) Can we justify the use of titanium on the basis of propagation rate measurements for crevice corrosion, or do we need to prove initiation cannot occur? 5) Will the pitting of copper be significant? 6) How thick a carbon steel container would be required, and can it be fabricated and stress-relieved? 7) Are radiation fields of any consequence at the dose rates expected?

  15. 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

  16. 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

  17. A biological process effective for the conversion of CO-containing industrial waste gas to acetate.

    Science.gov (United States)

    Kim, Tae Wan; Bae, Seung Seob; Lee, Jin Woo; Lee, Sung-Mok; Lee, Jung-Hyun; Lee, Hyun Sook; Kang, Sung Gyun

    2016-07-01

    Acetogens have often been observed to be inhibited by CO above an inhibition threshold concentration. In this study, a two-stage culture consisting of carboxydotrophic archaea and homoacetogenic bacteria is found to be effective in converting industrial waste gas derived from a steel mill process. In the first stage, Thermococcus onnurineus could grow on the Linz-Donawitz converter gas (LDG) containing ca. 56% CO as a sole energy source, converting the CO into H2 and CO2. Then, in the second stage, Thermoanaerobacter kivui could grow on the off-gas from the first stage culture, consuming the H2 and CO in the off-gas completely and producing acetate as a main product. T. kivui alone could not grow on the LDG gas. This work represents the first demonstration of acetate production using steel mill waste gas by a two-stage culture of carboxydotrophic hydrogenogenic microbes and homoacetogenic bacteria. PMID:27106591

  18. Corrosion of iron-base waste package container materials in SALT environments

    International Nuclear Information System (INIS)

    Low-carbon ferrous materials are being considered for waste package container materials in high-level nuclear waste salt repositories. The short-term corrosion rates of ASTM Type A216 Grade WCA steel have been determined under both brine-only and moist-salt conditions at 1500C for times ranging from 1 to 12 months. Tests run in moist salt with low Mg content brine yielded relatively low corrosion rates, below and adjusted value of 0.032 mm (1.3 mils) per year at 1500C. Post-test examinations have shown that the corrosion product is a complex Fe-Mg hydroxide of amakinite structure, as opposed to the Fe/sub 3/O/sub 4/ observed in the low-Mg brines. The Mg content of the environment is believed to be a major factor leading to the higher corrosion rates and studies to understand the operative corrosion mechanisms are in progress

  19. 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)

  20. 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...

  1. 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

  2. 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

  3. A model for predicting the lifetimes of Grade-2 titanium nuclear waste containers

    International Nuclear Information System (INIS)

    The development of a model to predict the lifetimes of Grade-2 titanium containers for nuclear fuel waste is described. This model assumes that the corrosion processes most likely to lead to container failure are crevice corrosion, hydrogen-induced cracking and general corrosion. Because of the expected evolution of waste vault conditions from initially warm (<∼ 100 deg C) and oxidizing to eventually cool (<30 deg C) and non-oxiding, the period for which crevice corrosion can propagate will be limited by repassivation, and long container lifetimes will be achieved since the rate of general corrosion is extremely low. However, in the model presented, not only is it assumed that crevices will initiate rapidly on all containers, but also that the propagation of these crevices will continue indefinitely since conditions will remain sufficiently oxiding for repassivation to be avoided. The mathematical development of the model is described in detail. A simple ramped distribution is used to describe the failures due to the presence of initial defects. For crevice corrosion the propagation rates are assumed to be normally distributed and to be determined predominantly by temperature. The temperature dependence of the crevice propagation rate is determined from the calculated cooling profiles for the containers and an experimentally determined Arrhenius relationship for crevice propagation rates. The cooling profiles are approximated by double or single step functions, depending on the location of the container within the vault. The experimental data upon which this model is based is extensively reviewed. This review includes descriptions of the available data to describe and quantify the processes of general corrosion, crevice corrosion and hydrogen-induced cracking. For crevice corrosion and hydrogen-induced cracking the results of studies on both Grades-2 and -12 are presented. Also, the effects of impurities in the Grade-2 material are discussed. Special attention is

  4. 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

  5. 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

  6. The structural integrity of high level waste containers for deep disposal

    International Nuclear Information System (INIS)

    Most countries with a nuclear power program are developing plans to dispose of high level waste in deep geological repositories. These facilities are typically in the range 500-1000m below ground. Although long term safety analyses mainly rely on the isolation function of the geological barrier, for the medium term (between 500 and 1000 years) a barrier such as a container (overpack) may play an important role. This paper addresses the mechanical/structural behavior of these structures under extreme geological pressures. The work described in the paper was conducted within the COMPAS project (Container Mechanical Performance Assessment) funded by the Commission of the European Communities and the United Kingdom Department of the Environment. The work was aimed at predicting the modes of failure and failure pressures which characterize the heavy, thick walled mild steel containers which might be considered for the disposal of vitrified waste. The work involved a considerable amount of analytical work, using 3-D non-linear finite element techniques, coupled with a large parallel program of experimental work. The experimental work consisted of a number of scale model tests in which the response of the containers was examined under external pressures as high as 120MPa. Extensive strain-gauge instrumentation was used to record the behavior of the models as they were driven to collapse. A number of comparative computer calculations were carried out by organizations from various European countries. Correlations were established between experimental and analytical data and guidelines regarding the choice of suitable software were established. The work concluded with a full 3-D simulation of the behavior of a container under long-term disposal conditions. In this analysis, non-linearities due to geological effects and material/geometry effects in the container were properly accounted for. 6 refs., 9 figs., 4 tabs

  7. 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

  8. 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)

  9. The perspectives for the disposal of tritium-containing wastes of nuclear power plants into deep geological formations

    International Nuclear Information System (INIS)

    Basic principles of the application of the method of the injection of liquid nuclear wastes into deep geological formations to the disposal of tritium-containing water are discussed. It is proposed that tritium-containing water should be removed together with other salt-containing waste waters of the Kalinin NPP. A scheme of the conditioning of these waters before their injection into collector layers is proposed. The prevention of harmful effects of nuclear wastes on people and environment is the necessary condition of any technology of nuclear waste management. The solution of the problem of a long-term isolation of nuclear wastes consists in the development and the application of the methods of their disposal into deep geological formations

  10. 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)

  11. 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.

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

    International Nuclear Information System (INIS)

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

  13. 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

  14. 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

  15. Instrumented measurements on radioactive waste disposal containers during experimental drop testing - 59142

    International Nuclear Information System (INIS)

    In context with disposal container safety assessment of containers for radioactive waste the German Federal Institute for Materials Research and Testing (BAM) performed numerous drop tests in the last years. The tests were accompanied by extensive and various measurement techniques especially by instrumented measurements with strain gages and accelerometers. The instrumentation of a specimen is an important tool to evaluate its mechanical behavior during impact. Test results as deceleration-time and strain-time functions constitute a main basis for the validation of assumptions in the safety analysis and for the evaluation of calculations based on finite-element methods. Strain gauges are useful to determine the time dependent magnitude of any deformation and the associated stresses. Accelerometers are widely used for the measuring of motion i.e. speed or the displacement of the rigid cask body, vibration and shock events. In addition high-speed video technique can be used to visualize and analyze the kinematical impact scenario by motion analysis. The paper describes some selected aspects on instrumented measurements and motion analysis in context with low level radioactive waste (LLW) container drop testing. (authors)

  16. 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

  17. Radiation crosslinking of styrene–butadiene rubber containing waste tire rubber and polyfunctional monomers

    International Nuclear Information System (INIS)

    The objective of this study was to investigate the influence of polyfunctional monomers (PFMs) and absorbed dose on the final characteristics of styrene–butadiene rubber (SBR) mixed with waste tire rubber (WTR). A series of SBR/WTR blends were prepared by varying the ratios of WTR in the presence of PFMs, namely trimethylolpropane trimethacrylate (TMPTMA) and trimethylolpropane triacrylate (TMPTA) and crosslinked using gamma rays. The physicochemical characteristics of the prepared blends were investigated. It was observed that tensile strength, hardness and gel content of the blends increased with absorbed dose while the blends containing TMPTA showed higher tensile strength, gel content and thermal stability as compared to the blends containing TMPTMA. Higher thermal stability was observed in the blends which were crosslinked by radiation as compared to the blends crosslinked by sulfur. These blends exhibited higher rate of swelling in organic solvents, whereas negligible swelling was observed in acidic and basic environment. - Highlights: • The effect of γ radiation on SBR blended with waste tire rubber was studied. • Two polyfunctional monomers were used to increase the crosslinking. • Mechanical properties of the blends were increased with absorbed dose. • TMPTA containing blend showed higher tensile strength and thermal stability. • Comparison with sulfur crosslinking was also done

  18. 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)

  19. Decay calculations on medium-level and actinide-containing wastes from the LWR fuel cycle. Pt. 1

    International Nuclear Information System (INIS)

    A number of basic data on medium-level and actinide-containing waste streams from the LWR fuel cycle were evaluated and the activity and thermal decay power were calculated for the nuclide inventories of cladding hulls and fuel assembly structural materials, for feed clarification sludge, medium-level aqueous process waste, low-level solid transuranium waste and for medium-level reactor operating waste. The activity as a function of decay time of the medium-level wastes decreases within 500 to 600 years by 1 to 3 orders of magnitude and is at the same time about 1 to 2 orders of magnitude lower than the activity of the high-level waste. The thermal decay power of the medium-level wastes decreases after 10 to 100 years by about 3 orders of magnitude and is about a factor of 10 to 100 less than that of high-level waste. In the very long term the residual activity (and thermal power) decreases only slowly due to the long halflives of the dominant actinides. The activity after more that 1000 years is about 1 to 2 orders of magnitude lower than that of high-level waste, the low-level transuranium waste by a factor 10 to 4, respectively. The activity per unit volume of the packaged waste of the medium-level and actinide-containing wastes because of the bigger volume of the conditioned wastes is lower by 2 to 4 orders of magnitude up to about 500 years. After more than 1000 years the activities per unit volume are lower by a factor of 20 to 200 than that of high-level waste. (orig.)

  20. 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.

  1. 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

  2. Galvanic corrosion evaluation of high activity nuclear waste container metals components

    International Nuclear Information System (INIS)

    The conceptual design for the feasibility study of an underground repository in Argentina for the final disposal of high activity nuclear waste was set. The container for such reprocessed and vitrified wastes shall have three metallic layers: a stainless steel inner layer, an external one of a metal to be selected and a thick lead intermediate layer (100 mm wide) preselected due to its good radiological protection and corrosion resistance. Surrounding the container, a bentonite and sand mixture will be used as backfilling of the rock hole. In the probable event that a small break in the metal of the container external layer would cause the simultaneous exposition of the intermediate lead layer and the external metal to the media, the corrosion behaviour of lead-metal galvanic couples in the repository simulated media (groundwater and 10% bentonite suspensions) was studied. In order to select the metal of the container external layer, 60 days galvanic couples tests of lead-titanium, lead-AISI 304 stainless steels and lead-SAE 1010 and SAE 1020 carbon steels were performed at 75 deg. C. Those test results showed the convenience of using carbon steel as an external container layer. Finally, the lead-carbon steel galvanic couples behaviour in groundwater was studied considering the following variables: Lead-carbon steel galvanic couple rate area from 1:10 to 1:40; Carbon content effect in steel corrosion kinetic; Decreasing temperature effect on lead-carbon steel galvanic couple polarity inversion in groundwater; Sea water different concentrations effect in lead-carbon steel galvanic couple corrosion kinetic at 75 deg. C. (author). 6 refs, 34 figs, 8 tabs

  3. 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

  4. Assessment of the Biodegradability of Containers for Low and Intermediate Level Nuclear Waste

    International Nuclear Information System (INIS)

    Concrete and reinforced concrete are widely used as engineered barriers (containers) for radioactive waste disposal facilities due to their isolating ability, mechanical stability and low cost. Several types of protective reinforced concrete containers for low and intermediate level waste have been designed in Ukraine. Evaluation of these containers for microbial stability is required according to NRC of Ukraine Regulation No.306.608-96. The research was therefore aimed at studying the degradation of the cement material due to microbiological interaction and the possibility of biodegraded cement as an ideal environment for the growth of other microorganisms under waste disposal conditions to satisfy the regulatory requirements. Results from this study indicated that Aspergillus niger induced gluconic and oxalic acids that dissolve portlandite (with a low leaching of calcium) after one year of contact time. This resulted in an increase in porosity, loss in tensile strength biomechanically deteriorated and cracking. XRD analysis identified crystalline precipitates within the biomass on the concrete surface as calcium oxalate dehydrate (weddellite) and calcium oxalate monohydrate (whewellite). The mechanism regarding of the microbiological interaction on the concrete surface can be summarized as follows: Phase 1: Fungi accumulate on the surface of the concrete, thereby degrading the concrete surface by biochemical and biomechanical interactions. When this effect is in the presence of air with available carbon dioxide, the micro fungi reduces the pH of the concrete from >13 to 8.5. During this phase no accumulation were observed in sections where granite aggregates are present. Phase 2: After reducing the pH of the concrete paste during phase 1, and provided that sufficient nutrients, moisture and oxygen are present sulphur oxidizing bacteria start to accumulate on the concrete surface. The result form this study therefore concluded that fungal biogeochemical activity

  5. Crevice corrosion and pitting of high-level waste containers: Integration of deterministic and probabilistic models

    International Nuclear Information System (INIS)

    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-layer canister. In this particular design, the inner barrier is made of a corrosion resistant material (CRM) such as Alloy 625 or C-22, while the outer barrier is made of a corrosion-allowance material (CAM) such as carbon steel or Alloy 400. An integrated predictive model is being developed to account for the effects of localized environmental conditions in the CRM-CAM crevice on the initiation and propagation of pits through the CRM

  6. CREVICE CORROSION and PITTING OF HIGH-LEVEL WASTE CONTAINERS: INTEGRATION OF DETERMINISTIC and PROBABILISTIC MODELS

    International Nuclear Information System (INIS)

    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-layer canister. In this particular design, the inner barrier is made of a corrosion resistant material (CRM) such as Alloy 625 or C-22, while the outer barrier is made of a corrosion-allowance material (CAM) such as carbon steel or Monel 400. An integrated predictive model is being developed to account for the effects of localized environmental conditions in the CRM-CAM crevice on the initiation and propagation of pits through the CRM

  7. Crevice corrosion ampersand pitting of high-level waste containers: integration of deterministic ampersand probabilistic models

    International Nuclear Information System (INIS)

    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-layer canister. In this particular design, the inner barrier is made of a corrosion resistant material (CRM) such as Alloy 625 or C-22, while the outer barrier is made of a corrosion-allowance material (CAM) such as carbon steel or Monel 400. An integrated predictive model is being developed to account for the effects of localized environmental conditions in the CRM-CAM crevice on the initiation and propagation of pits through the CRM

  8. Production and Energy Partition of Lactating Dairy Goats Fed Rations Containing Date Fruit Waste

    OpenAIRE

    E. Yuniarti; D. Evvyernie; D. A. Astuti

    2016-01-01

    Dates fruit waste (DFW) is a by-product of dates juice industry that contains high energy. So, it is suitable for an energy source in dairy goat ration. This study was conducted to observe the effect of DFW utilization in the ration on energy partition and productivity of lactating dairy goats. The experimental design was randomized block design using 9 primiparous lactating dairy goats. There were three types of ration as treatments used in this study, i.e. R0= 35% forage + 65% concentrate, ...

  9. Creep properties of welded joints in copper canisters for nuclear waste containment

    International Nuclear Information System (INIS)

    Copper canisters for nuclear waste containment can be expected to be exposed to temperatures up to 1000C. Since the material is pure copper, creep properties must be taken into account in particular for the welded joints in the canisters. In the paper creep rupture properties of parent metal, weld metal, and simulated heat affected zones are presented for 1100C. About ten times shorter rupture times were found for the weld metal in comparison to the parent metal. Cross weld specimens showed even shorter rupture times

  10. 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

  11. New system for the container conditioning of liquid waste in the German future finale repository 'Schacht Konrad'

    International Nuclear Information System (INIS)

    The full text of publication follows. On-site the NPP Gundremmingen liquid radioactive waste from the NPP water treatment plant is stored in resin or concentrate collecting tanks. These liquid wastes are cemented in containers in order to temporarily store them in the Bavarian interim storage Mitterteich until they are transported into final repository in 'Schacht Konrad'. With this new system liquid radioactive waste is for the first time conditioned directly into containers destined for final repository in 'Schacht Konrad'. Thus, a very secure and sustainable procedure was developed which also provides high profitability. The conditioning plant for resins and concentrate extracts the liquid waste from the respective collecting tank and transports the waste to the separation tank. This separation tank is dimensioned to ensure complete filling of a Konrad container with only one batch. Within the tank there is the option to adjust the suspensions solids content by either extracting supernatant water or by adding de-ionised water. The specific activity is analysed and after the radiologic data and the solids content are available, the containers are cemented. The required amount of cement is based on the solids content and is automatically added. In the mixer, cement and primary waste suspension are mixed. This mixture is filled into the Konrad container via the allocator. The allocator is a funnel-shaped inlet equipped with a movable tube which makes sure the mixture is evenly spread and also ensures optimal filling of the Konrad container. While filling is ongoing, the container is covered by a lowerable splash guard to avoid contamination. The room situation in Gundremmingen and the specific activities of the primary waste suspension make it necessary to disperse the plant to several rooms. Main components such as separation tanks and pumps are installed in shielded rooms. All activities are conducted remotely controlled and are supervised from the central

  12. 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)

  13. 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)

  14. 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

  15. Efficient decomposition of liquid waste containing EDTA by advanced oxidation nanotechnology

    International Nuclear Information System (INIS)

    Degradation of ethylenediaminetetraacetic acid (EDTA) present in the liquid waste was demonstrated by photocatalytic oxidation route by using nanoparticles of anatase titania. Nano sized titania photocatalyst was synthesized using sol-gel method coupled with ultrasonication mode and characterized using X-ray diffraction, transmission electron microscope, BET, Fourier transform infrared spectroscopy and TG-DTA. A cylindrical photoreactor was employed for the degradation studies. Five milligram of the nano anatase TiO2 + 0.5 ml of 30% H2O2 were employed as catalysts for the degradation studies of 1,000 mg/L EDTA. EDTA degradation was followed by a complexometric titration method. Complete degradation of 1,000 mg/L EDTA could be achieved in 90 min and the photocatalytic efficiency of the synthesized titania photocatalyst was higher than that of P-25 TiO2 for EDTA degradation. The influence of pH on the degradation of EDTA follow the order acidic > neutral > alkaline. More than ten fold increases in the decontamination factors were obtained for the chemical precipitation step for the liquid waste containing degraded EDTA compared to liquid waste without EDTA degradation. (author)

  16. The kinetics of pitting corrosion of carbon steel applied to evaluating containers for nuclear waste disposal

    International Nuclear Information System (INIS)

    This is the final summary report on a project, funded by SKB, investigating the pitting corrosion of carbon steel containers for high level nuclear waste or spent reactor fuel under granite disposal conditions. The study has covered a statistically based experimental programme to establish the pit growth kinetics, and a modelling study to determine the maximum pitting period subsequent to repository closure. It is shown that the rate of pit propagation is slower than that suggested by earlier work and that the maximum pitting period is only a small fraction of the target container life of 1000 years. An illustrative example of the methodology for estimating the corrosion allowance needed to prevent pit penetration is given. This could be applied to specific repository conditions as defined by SKB. Finally some limited recommendations are made for further studies to test and validate the methodology. (au)

  17. Design and testing of Spec 7A containers for packaging radioactive wastes

    International Nuclear Information System (INIS)

    For a variety of reasons, the containers that have or currently are being used for packaging radioactive waste have drawbacks which has motivated LLNL to investigate, design and destructively test different Type A containers. The result of this work is manifested in the TX-4, which is comparatively lightweight, increases the net payload, and the simplicity of the design and ease in handling have proved to be timesaving. The TX-4 is readily available, relatively inexpensive and practical to use. It easily meets Type A packaging specifications with a gross payload of 7000 pounds. Although no tests were performed at a higher weight, we feel that the TX-4 could pass the tests at higher gross weights if the need arises. 20 figures

  18. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Strum, M.J.; Weiss, H.; Farmer, J.C. (Lawrence Livermore National Lab., CA (USA)); Bullen, D.B. (Science and Engineering Associates, Inc., Pleasanton, CA (USA))

    1988-06-01

    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.

  20. 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%. PMID:26141285

  1. 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

  2. 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

  3. Design features of shipping containers for low and intermediate level wastes

    International Nuclear Information System (INIS)

    Immobilization of Intermediate Level Waste and Low Level Waste concentrates in cement matrices has been accepted as a conditioning process and is widely used in India since long. To keep pace with the increasing throughputs from recent solidification plants, shipping casks with improved designs are being adopted to transport single and multiple product drums from plant to storage/disposal sites. A salient feature of the vertical bottom loading shipping containers is the adoption of modular concept incorporating the loading/unloading platforms, the transportation unit and the lifting system as separable modules. Emphasis has been mainly on making the transportation units compact and devoid of any drives, controls and sensors which are prone to damages during transit which have been provided on the stationary loading/unloading platforms. The product entry into the cask is through a single bottom door in place of the split door design practised earlier thereby avoiding bulky compensatory shielding around the opening. THe lifting system accommodates the lifting tackle for the cask as well as the hoisting system for the product drum. The casks are designed to employ a fail safe and load positive pneumatic grapple where pneumatic actuation is utilized for releasing the grapple once the load is stably supported. While the intermediate level shipping container is meant for a single drum, the low level shipping container has been designed to accommodate multiple drums inside the housing with provision of a rugged mechanized conveyor segment for handling the drums within the cask. Limitation of capacity of the design has been only availability of space and handling systems at the existing facilities. The cask designs have been evaluated analytically to determine their compliance with the applicable regulations governing containers in which radioactive materials are transported. (author). 2 refs

  4. 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

  5. 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.

  6. Development of a pneumatic stowing and chocking system for packages containing radioactive waste

    International Nuclear Information System (INIS)

    Since that goods are transported, their chocking and stowing is very often done by improvisation, successfully or disastrously. When the disaster appears in comics it is always a source of an enormous amusement, when it appears in road or maritime accidents it is most of the time a source of death or severe damages. Even if transport of radioactive materials could be considered as the exception where chains and tie-down systems are used abundantly, their strength relies always on the weakness of their components. Special attention has been paid to the transport of type A or type B packages, but obviously there was a lack of interest for the transport of low level radioactive waste, even knowing that the quantities of this waste are a hunderfold or a thousandfold of the first ones. On the subject of stowing and chocking systems for radioactive waste packages, TRANSNUBEL together with the CEA-France performed under the sponsorship of the Commission of the European Communities between 1980 and 1985 a study which clearly showed that during a road accident, in case of a front end impact, the stowing system must be able to absorb entirely the kinetic energy generated by the package deceleration, which is proportional to the package mass. The chocks must be able to absorb a deceleration energy generated by the package of about 30 g at a speed of about 50 km/h. This energy of course decreases at the same time as the speed. These conclusions served as basic principles for the development by TRANSNUBEL of a pneumatic stowing and chocking system for packagings containing radioactive waste

  7. 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)

  8. 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

  9. Design features of shipping containers for low and intermediate level wastes

    International Nuclear Information System (INIS)

    Immobilisation of Intermediate Level Waste and Low Level Waste concentrates in cement matrices has been accepted as a conditioning process and has been widely used in India for some time. To keep pace with the increasingly throughputs from recent solidification plants, shipping casks with improved designs are being adopted to transport single and multiple product drums from plant to storage/disposal sites. A salient feature of the vertical bottom loading shipping containers is the adoption of a modular concept incorporating the loading/unloading platforms, the transport unit and the lifting system as separable modules. Emphasis has been mainly on making the transport units compact and devoid of any drives, controls and sensors which are prone to damage during transit and which have been provided on the stationary loading/unloading platforms. The product entry into the cask is through a single bottom door in place of the split door design practised earlier, thereby avoiding bulky compensatory shielding around the opening. The cask designs have been evaluated analytically to determine their compliance with the applicable regulations governing containers in which radioactive materials are transported. (author)

  10. 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

  11. Corrosion of iron-base waste package container materials in salt environments

    International Nuclear Information System (INIS)

    Low-carbon ferrous materials are being considered for waste package container materials in high-level nuclear waste salt repositories. The short-term corrosion rates of ASTM Type A216 Grade WCA steel have been determined under both brine-only and moist-salt conditions at 1500C for time ranging from 1 to 12 months. Tests run in moist salt with low Mg content brine yielded relatively low corrosion rates, below an adjusted value of 0.032 mm (1.3 mils) per year at 1500C. Corrosion rates in brine-only and moist-salt environments containing high concentrations of Mg were found to be a factor of 20 to 50 higher over the same experimental test times, depending on the steel's heat treatment and the specific test conditions. Austenitizing treatment reduced the corrosion resistance of the material. In the case of the as-cast steel, the measured average corrosion rates decreased with time by more than a factor of two during the 12-month testing program. Post-test examinations have shown that the corrosion product is a complex Fe-Mg hydroxide of amakinite structure, as opposed to the Fe3O4 observed in the low-Mg brines. The Mg content of the environment is believed to be a major factor leading to the higher corrosion rates and studies to understand the operative corrosion mechanisms are in progress. 1 ref., 4 figs., 2 tabs

  12. 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.)

  13. The performance of polymer containers used for the storage of radioactive waste

    International Nuclear Information System (INIS)

    An evaluation of the performance of polymeric materials after exposure to radiation and acidic aqueous solutions provides a basis for the evaluation of failure mechanisms affecting these materials. The work evaluated the importance of the combined effects of aqueous solution diffusion, radiation exposure, and temperature on the mechanical performance, diffusion profile and molecular structure of polymeric materials. This work demonstrated that the dose rate is an extremely important factor since low dose rates have been shown to result in an increase in stress at yield (15 - 20%) over the times studied, whereas higher dose rates reduced stress at yield as discussed above. Irradiation of both Nylon 6,6 and Semi-Aromatic Nylon 6,6 at dose rates of 37 and 56 kGy/hr resulted in an initial decrease in the stress at yield and subsequent recovery. Irradiation at 20 kGy/hr resulted in an initial increase in stress at yield and a continued increase throughout the aging time. It is suggested that polyamide 6,6 may be considered an acceptable material for the fabrication of storage containers for Low Level Radioactive Waste. Similarly, semi-aromatic polyamide 6,6, with its greater resistance to the combined effects of solution diffusion and radiation exposure, may be considered an acceptable material for the fabrication of containers for the storage of Intermediate Level Radioactive Waste. Finally, these results provide further explanation of the results obtained for materials such as polycarbonate, which has been previously determined to be viable candidates for the storage of High Level Radioactive Waste. (author)

  14. Developments in radiography and tomography of waste containers at the Idaho National Engineering Laboratory

    International Nuclear Information System (INIS)

    The Idaho National Engineering Laboratory (IN-F-L) has been inspecting containers (boxes and drums) of nuclear waste materials using real-time radiography (RTR) for the past ten years. Requirements governing characterization of containerized waste for short-term storage, treatment, transportation, and disposal have become more stringent. These new requirements, and the need to reduce inspection times to increase throughput, necessitate improvements in the information obtained by radiographic methods. RTR provides a qualitative view of container contents, whereas quantitative information is often required. Two projects at the INEL are converting the present qualitative radiographic inspection to the more quantitative digital radiography (DR) and computed tomography (CT) methods, while retaining the RTR function. The first project is modifying, the RTR hardware at the Radioactive Waste Management Complex (RWMC) to allow rapid processing of analog RTR images. The digital RTR (DRTR) system described here can digitize, process, and redisplay RTR images at video frame rates allowing for real-time image improvement features such as edge detection, contrast enhancement, frame subtraction, frame averaging, and a variety of digital filtering options. The second project is developing a complete radiographic and tomographic capability that allows for greater sophistication in data acquisition and processing as the operator and/or requirements demand. The approach involves modification of an industrial CT scanner with the capability to acquire radiographic and tomographic data in several modes, including conventional RTR, DR, and CT with a linear detector for high spatial resolution, and DR and CT with an area detector for high throughput. Improvements in image quality and quantitative digital radiographic capabilities of the DRTR system are shown. Status and plans for the modified CT scanner (presently under development) are also presented

  15. Pyrolysis behavior of different type of materials contained in the rejects of packaging waste sorting plants

    International Nuclear Information System (INIS)

    Highlights: ► Study of the influence of materials in the pyrolysis of real plastic waste samples. ► Inorganic compounds remain unaltered. ► Cellulosic components give rise to an increase in char formation. ► Cellulosic components promote the production of aqueous phase. ► Cellulosic components increase CO and CO2 contents in the gases. - Abstract: In this paper rejected streams coming from a waste packaging material recovery facility have been characterized and separated into families of products of similar nature in order to determine the influence of different types of ingredients in the products obtained in the pyrolysis process. The pyrolysis experiments have been carried out in a non-stirred batch 3.5 dm3 reactor, swept with 1 L min−1 N2, at 500 °C for 30 min. Pyrolysis liquids are composed of an organic phase and an aqueous phase. The aqueous phase is greater as higher is the cellulosic material content in the sample. The organic phase contains valuable chemicals as styrene, ethylbenzene and toluene, and has high heating value (HHV) (33–40 MJ kg−1). Therefore they could be used as alternative fuels for heat and power generation and as a source of valuable chemicals. Pyrolysis gases are mainly composed of hydrocarbons but contain high amounts of CO and CO2; their HHV is in the range of 18–46 MJ kg−1. The amount of CO-CO2 increases, and consequently HHV decreases as higher is the cellulosic content of the waste. Pyrolysis solids are mainly composed of inorganics and char formed in the process. The cellulosic materials lower the quality of the pyrolysis liquids and gases, and increase the production of char.

  16. 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.

  17. Technetium diffusion in clay-based materials under oxic and anoxic conditions

    International Nuclear Information System (INIS)

    Diffusion coefficients were determined for Tc in compacted clay-based materials in both anoxic and oxic environments. The soils were saturated with a synthetic groundwater solution; the principle ions in solution were CA2+, Na+ and Cl-. Anoxic conditions were established by conducting the experiments in low-O2 glove box and by mixing 0.5 wt% powdered Fe with the soils. Under anoxic conditions, apparent diffusion coefficients, Da, were 2/s for Tc in compacted backfill material (a 1:3 mix by dry mass of Lake Agassiz clay and crushed granite aggregate). Distribution coefficients, Kd, for Tc on Lake Agassiz clay and backfill material in anoxic environments were back-calculated from Da values. Based on the Kd values, Tc strongly sorbs on Lake Agassiz clay and backfill under anoxic conditions. Effective diffusion coefficients De, for Tc of 10, 16 and 110 μm2/s were measured in oxic Avonlea bentonite, Lake Agassiz clay and illite-smectite, respectively, at a clay dry bulk density of ∼ 1.2 Mg/m3; the corresponding Da values were 55, 48 and 75 μm2/s. Since anoxic conditions are expected in a disposal vault excavated deep in granitic rock in the Canadian Shield, the results suggest the migration of Tc through the backfill will be relatively slow. (author) 23 refs., 9 tabs., 4 figs

  18. Effects of a potential drop of a shipping cask, a waste container, and a bare fuel assembly during waste-handling operations; Yucca Mountain Site Characterization Project

    Energy Technology Data Exchange (ETDEWEB)

    Wu, C.L.; Lee, J.; Lu, D.L.; Jardine, L.J. [Bechtel National, Inc., San Francisco, CA (United States)

    1991-12-01

    This study investigates the effects of potential drops of a typical shipping cask, waste container, and bare fuel assembly during waste-handling operations at the prospective Yucca Mountain Repository. The waste-handling process (one stage, no consolidation configuration) is examined to estimate the maximum loads imposed on typical casks and containers as they are handled by various pieces of equipment during waste-handling operations. Maximum potential drop heights for casks and containers are also evaluated for different operations. A nonlinear finite-element model is employed to represent a hybrid spent fuel container subject to drop heights of up to 30 ft onto a reinforced concrete floor. The impact stress, strain, and deformation are calculated, and compared to the failure criteria to estimate the limiting (maximum permissible) drop height for the waste container. A typical Westinghouse 17 {times} 17 PWR fuel assembly is analyzed by a simplified model to estimate the energy absorption by various parts of the fuel assembly during a 30 ft drop, and to determine the amount of kinetic energy in a fuel pin at impact. A nonlinear finite-element analysis of an individual fuel pin is also performed to estimate the amount of fuel pellet fracture due to impact. This work was completed on May 1990.

  19. Effects of a potential drop of a shipping cask, a waste container, and a bare fuel assembly during waste-handling operations

    International Nuclear Information System (INIS)

    This study investigates the effects of potential drops of a typical shipping cask, waste container, and bare fuel assembly during waste-handling operations at the prospective Yucca Mountain Repository. The waste-handling process (one stage, no consolidation configuration) is examined to estimate the maximum loads imposed on typical casks and containers as they are handled by various pieces of equipment during waste-handling operations. Maximum potential drop heights for casks and containers are also evaluated for different operations. A nonlinear finite-element model is employed to represent a hybrid spent fuel container subject to drop heights of up to 30 ft onto a reinforced concrete floor. The impact stress, strain, and deformation are calculated, and compared to the failure criteria to estimate the limiting (maximum permissible) drop height for the waste container. A typical Westinghouse 17 x 17 PWR fuel assembly is analyzed by a simplified model to estimate the energy absorption by various parts of the fuel assembly during a 30 ft drop, and to determine the amount of kinetic energy in a fuel pin at impact. A nonlinear finite-element analysis of an individual fuel pin is also performed to estimate the amount of fuel pellet fracture due to impact. This work was completed on May 1990

  20. 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

  1. Stainless steel waste containers: an assessment of the probability of stress corrosion cracking

    International Nuclear Information System (INIS)

    The paper summarises information obtained from the literature and discussions held with corrosion experts from universities and industry, relevant to the possibility that stainless steel radioactive waste containers containing low level and intermediate level radioactive waste (LLW and ILW) could, when buried in concrete, suffer one or more of the forms of stress corrosion cracking (SCC). Stress corrosion cracking is caused by the simultaneous and synergistic action of a corrosive environment and stress. The initiation and propagation of SCC depend on a number of factors being present, namely a certain level of stress, an environment which will cause cracking and a susceptible metal or alloy. Generally the susceptibility of a metal or alloy to SCC increases as its strength level increases. The susceptibility in a specific environment will depend on: solution concentration, pH, temperature, and electrochemical potential of the metal/alloy. It is concluded that alkaline stress corrosion cracking is unlikely to occur under even the worst case conditions, that chloride stress corrosion cracking is a distinct possibility at the higher end of the temperature range (25-80oC) and that stress corrosion related to sensitization of the steel will not be a problem for the majority of container material which is less than 5 mm in cross section. Thicker section material could become sensitized leading to a local problem in these areas. Contact with metals that are electrochemically more negative in corrosion potential is likely to reduce the incidence of SCC, at least locally. Measurement of repassivation potentials and rest potentials in solutions of relevant composition would provide a firmer prediction of the extent to which a high pH could reduce the likelihood of SCC caused by chlorides. (author)

  2. 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

  3. 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.

  4. 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

  5. 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)

  6. 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

  7. Roadmapping the Resolution of Gas Generation Issues in Packages Containing Radioactive Waste/Materials

    Energy Technology Data Exchange (ETDEWEB)

    Luke, Dale Elden; Rogers, Adam Zachary; Hamp, S.

    2001-03-01

    Gas generation issues, particularly hydrogen, have been an area of concern for the transport and storage of radioactive materials and waste in the Department of Energy (DOE) complex. Potentially combustible gases can be generated through a variety of reactions, including chemical reactions and radiolytic decomposition of hydrogen-containing materials. Transportation regulations prohibit shipment of explosives and radioactive materials together. This paper discusses the major gas generation issues within the DOE Complex and the research that has been and is being conducted by the transuranic (TRU) waste, nuclear materials (NM), and spent nuclear fuels (SNF) programs within DOE’s Environmental Management (EM) organization to address gas generation concerns. This paper presents a "program level" roadmap that links technology development to program needs and identifies the probability of success in an effort to understand the programmatic risk associated with the issue of gas generation. This "program level" roadmapping involves linking technology development (and deployment) efforts to the programs’ needs and requirements for dispositioning the material/waste that generates combustible gas through radiolysis and chemical decomposition. The roadmapping effort focused on needed technical & programmatic support to the baselines (and to alternatives to the baselines) where the probability of success is low (i.e., high uncertainty) and the consequences of failure are relatively high (i.e., high programmatic risk). A second purpose for roadmapping was to provide the basis for coordinating sharing of "lessons learned" from research and development (R&D) efforts across DOE programs to increase efficiency and effectiveness in addressing gas generation issues.

  8. Photo-catalytic and photochemical degradation of liquid waste containing EDTA - 59144

    International Nuclear Information System (INIS)

    The decontamination factor of liquid waste containing 60Co is generally weak. This is due to the presence of complexant molecules. For instance, complexation of EDTA with 60Co decreases efficiency of radioactive waste treatment. The aim of this study was to degrade EDTA in H2O and CO2 and to concentrate free 60Co in order to increase decontamination factor. A first test of radioactive waste treatment by photo-catalysis was allowed to increase decontamination factor (60Co) from 16 to 196 with a device requiring to be improved. The present work concerns the first step of the degradation process development with a more powerful device. These first experiments were leaded to follow the only EDTA oxidation. EDTA degradation was carried out by the following Advanced Oxidation Processes (AOP): UV/H2O2 (photochemistry); UV/TiO2 (photo-catalysis); UV/TiO2/H2O2. A specific reactor was achieved for this study. The wavelength used was 254 nm (UVC). The photo-catalytic degradation of EDTA was carried out with Degussa P-25 titanium dioxide (TiO2), which is a semiconductor photo-catalyst. The degradation degree of EDTA and the intermediate products were monitored by TOC and ionic chromatography methods. The effects of various parameters such as pH and the quantity of H2O2 were studied. This allows us to conclude that basic pH slows down EDTA degradation. The study showed that UV/H2O2 process was the most effective treatment process under acid conditions. The rate of EDTA degradation was very high and reached 95% in 120 minutes. The presence of glyoxylic, oxalic, glycolic and formic acids was detected as degradation products. Among the intermediates produced by photochemistry, NO3- ions presence informed of the amine degradations. These results highlighted faster EDTA degradation by photochemistry than photo-catalysis. (authors)

  9. The oxidation of acid azo dye AY 36 by a manganese oxide containing mine waste

    International Nuclear Information System (INIS)

    Highlights: ► This study looks at the oxidative breakdown of the amine containing dye acid yellow 36 by a Mn oxide containing mine waste. ► The oxidation proceeds by successive one electron transfers between the dye molecule and the Mn oxide minerals. ► The initial decolorization of the dye is rapid, but does not involve the cleavage of the azo bond. -- Abstract: The oxidative breakdown of acid azo dye acid yellow 36 (AY 36) by a Mn oxide containing mine tailings is demonstrated. The oxidation reaction is pH dependent with the rate of decolorization increasing with decreasing pH. The oxidation reaction mechanism is initiated at the amino moiety and proceeds via successive, one electron transfers from the dye to the Mn oxide minerals. The reaction pathway involves the formation of a number of colorless intermediate products, some of which hydrolyze in a Mn oxide-independent step. Decolorization of the dye is rapid and is observed before the cleavage of the azo-bond, which is a slower process. The terminal oxidation products were observed to be p-benzoquinone and 3-hydroxybenzenesulfonate. The reaction order of the initial decolorization was determined to be pseudo fractional order with respect to pH and pseudo first order with respect to dye concentration and Mn tailings’ surface area

  10. High polymer-based composite containers for the disposal/storage of high radioactive waste

    International Nuclear Information System (INIS)

    Spent fuel disposal is one of the hottest topics in nuclear news, getting considerable amount of media coverage around the world. Canada as well as many other countries with nuclear electric generation plants has therefore been pushed to develop policy on this issue. One of the proposed and most widely supported strategies is to dispose of this so-called waste permanently in deep underground vaults. Through the use of engineered barriers including vault seals, vault composition, backfill and sophisticated containers this radioactive matter is isolated from the natural environment. According to a design developed by Atomic Energy of Canada, the seclusion must be maintained for approximately 500 years, which is a representative length of time it takes for the radioactive elements to decay to natural background levels. The purpose of the current study is to determine the feasibility of using poly(ether ether ketone), an advanced polymer, and continuous carbon fibre in a consolidated composite as a principal container component. Feasibility was determined by simulating the ultimate radioactive environment that the containers will be exposed to by exposing test specimens to neutron and gamma radiation fields at various temperatures (20oC - 75oC) for a variety of time intervals. (author)

  11. 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

  12. 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.

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

    International Nuclear Information System (INIS)

    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 drums containing 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. (authors)

  14. 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

  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. PMID:26250546

  16. Prokaryotic complex of newly formed soils on nepheline-containing industrial waste

    Science.gov (United States)

    Evdokimova, G. A.; Kalmykova, V. V.

    2010-06-01

    The characteristics are given of the prokaryotic complex participating in the processes of the primary soil formation on nepheline-containing waste and depending on the time of the waste disposal and degree of reclamation. The total population density of the bacteria determined with the method of fluorescent microscopy in “pure” sand ranged within 0.34—0.60 billion CFU/g soil; in the reclaimed sand under different vegatation communities, from 2.6 to 7.2 billion CFU/g soil. Gram-positive bacteria dominate in the prokaryotic complex of the nepheline sands, whereas the Grarrmegative ones dominate in the zonal soils. The bacteria predominating in the nepheline sands were classified on the basis of the comparative analysis of the nucleotide sequences in the 16S rRNA genes within the Actinobacteria class (Arthrobacter boritolerans, A. ramosus, Rhodococcusfascians, Micrococcus luteus, and Streptomyces spp.). The evolution of the microbial community in the nepheline sands in the course of their reclamation and in the course of their overgrowing by plants proceeds in way toward the microbial communities of the zonal soils on moraine deposits.

  17. Natural radioactivity content and radionuclides leachability of bricks containing industrial waste

    International Nuclear Information System (INIS)

    A study have been carried out using gamma-ray spectrometric system to determine the natural radioactivity level in bricks made from industrial waste and their associated radiation hazard. Brick-1 and brick-2 contained waste from coal power plant and granite industry, respectively. The leachability of radionuclides from these bricks was also investigated. The activity concentration values of 226Ra, 228Ra, 232Th, and 40K are 64.25, 63.15, 67.9 and 254.19 Bq kg-1, respectively in brick-1, and 193, 164.48, 164.63 and 1348.75 Bq kg-1, respectively in brick-2. The radiation hazard indexes such as radium equivalent activities (Raeq), representative level index (Iγr), external hazard index (Hex) and internal hazard index (Hin) were calculated and compared with the internationally approved values. Results indicate that brick-1 showed less radiological hazard than brick-2. This suggested that brick-1 could be used in building construction without exceeding the proposed criterion level. The leachability of 226Ra for bricks showed the activity concentration slightly exceeded the limit generally used for industrial wastewater for example 1 BqL-1. (author)

  18. Air-lift reactor system for the treatment of waste-gas-containing monochlorobenzene.

    Science.gov (United States)

    Joshi, Pradnya R; Deshmukh, Sharvari C; Morone, Amruta P; Kanade, Gajanan; Pandey, R A

    2013-01-01

    An air-lift bioreactor (ALR) system, applied for the treatment of waste-gas-containing monochlorobenzene (MCB) was seeded with pure culture of Acinetobacter calcoaceticus, isolated from soil as a starter seed. It was found that MCB was biologically converted to chloride as chloride was mineralized in the ALR. After the built up of the biomass in the ALR, the reactor parameters which have major influence on the removal efficiency and elimination capacity were studied using response surface methodology. The data generated by running the reactor for 150 days at varying conditions were fed to the model with a target to obtain the removal efficiency above 95% and the elimination capacity greater than 60%. The data analysis indicated that inlet loading was the major parameter affecting the elimination capacity and removal efficiency of >95%. The reactor when operated at optimized conditions resulted in enhanced performance of the reactor. PMID:24617061

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

    International Nuclear Information System (INIS)

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

  20. 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.

  1. 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.

  2. 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.

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

    International Nuclear Information System (INIS)

    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

  4. Selection of a mineral binder for the stabilization - solidification of waste containing aluminum metal

    International Nuclear Information System (INIS)

    The dismantling of nuclear facilities produces radioactive waste materials, some of which may contain aluminum metal. In a strongly alkaline medium, such as that encountered in conventional cementitious materials based on Portland cement, aluminum metal becomes corroded, with a continued production of dihydrogen. In order to develop a mineral matrix having enhanced compatibility with aluminum, a literature review was first undertaken to identify binders capable of reducing the pore solution pH compared with Portland cement. An experimental study was then carried out to measure the hydrogen production resulting from corrosion of aluminum metal rods encapsulated in the different selected cement pastes. The best results were achieved with magnesium phosphate cement, which released very little hydrogen over the duration of the study. This production could be reduced further by adding a corrosion inhibitor (lithium nitrate) to the mixing solution

  5. 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

  6. Stress corrosion cracking tests on high-level-waste container materials in simulated tuff repository environments

    International Nuclear Information System (INIS)

    Types 304L, 316L, and 321 austenitic stainless steel and Incoloy 825 are being considered as candidate container materials for emplacing high-level waste in a tuff repository. The stress corrosion cracking susceptibility of these materials under simulated tuff repository conditions was evaluated by using the notched C-ring method. The tests were conducted in boiling synthetic groundwater as well as in the steam/air phase above the boiling solutions. All specimens were in contact with crushed Topopah Spring tuff. The investigation showed that microcracks are frequently observed after testing as a result of stress corrosion cracking or intergranular attack. Results showing changes in water chemistry during test are also presented

  7. Study on the design and manufacturing requirements of container for low level radioactive solid waste form KRR decommissioning

    International Nuclear Information System (INIS)

    The design requirement and manufacturing criteria have been proposed on the container for the 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 ISO 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 test 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 yield stress without any structural deformation. From the results of lifting tests, it was verified that the container was safe

  8. 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.

  9. 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.

  10. Puncture phenomena in low velocity impact of radioactive waste shipping containers

    International Nuclear Information System (INIS)

    Punch penetration phenomena of a radioactive waste shipping container when subjected to low velocity impact are studied both analytically and experimentally. Model cask materials in the form of circular test samples of stainless steel, backed with high density material (lead or uranium) are dropped on a steel punch with impact velocities of 220 in/sec or less. The experimental program involves 25 static tests and 67 dynamic tests in which plate thicknesses, backing material and punch diameters are varied. A detailed finite element analysis including the effects of plastic-work hardening in the materials and plate laminate separation was made on two geometries; ballasted with an impact velocity of 66.7 in/sec and unballasted with an impact velocity of 200 in/sec. An alternate analytical program was developed based upon a system frequency spectra analysis. A Rayleigh-Ritz approach was selected, including the laminated plates, deformable punch and elastic ballast in the potential energy, and distributed mass in the kinetic energy. A model analysis using the first five frequencies was performed and stress calculations included the effects of plastic deformation. Excellent agreement was obtained between the Rayleigh-Ritz and the finite element analyses. In most cases, it was observed that initial impact was not sufficient to cause onset of punch yield or plate puncture but the plate vibration effects would initiate failure at a later time. The Rayleigh-Ritz program was used to analyze the different experimental configurations and then compared to experimental data. This research indicates that low velocity impact studies of radioactive waste shipping containers must consider the phenomena of structural vibrations. The analytical approach developed offers an accurate, quick and inexpensive method of examining low velocity impact

  11. 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.

  12. 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

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

    International Nuclear Information System (INIS)

    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

  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. Design against brittle or elastic-plastic fracture of nuclear waste container

    International Nuclear Information System (INIS)

    Design against brittle or elastic plastic fracture of nuclear waste container is discussed based on three different concepts: (i) reserve of ductility defined by means of reference temperature, (ii) deterministic design using linear or elasto-plastic fracture mechanics associated with reserve factors, and (iii) probabilistic design associated with RCCMR failure assessment diagram. Cast ferritic steel predetermined for containers of spent nuclear fuel has been used in experimental part of the study. Fracture toughness characteristics necessary for considerations have been obtained by standard 1T three point bend specimens tested statically at different temperatures. Pre-cracked Charpy type specimen has been also employed for the investigations tested statically and dynamically. Material properties necessary for the concept presented are corresponding Master Curve and Weibull distribution of fracture toughness. Special attention has been paid to dynamic loading. Large scatter in reserve factor was found depending on the selected failure assessment method for fracture toughness characteristics changing the value from 1.44 to 4.55. (author)

  16. 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

  17. MODELING THE CORROSION OF HIGH-LEVEL WASTE CONTAINERS CAM-CRM INTERFACE

    International Nuclear Information System (INIS)

    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-layer canister. In this particular design, the inner barrier is made of a corrosion resistant material (CRM) such as Alloy 625 or C-22, while the outer barrier is made of a corrosion-allowance material (CAM) such as carbon steel or Monel400. Initially, the containers will be hot and dry due to the heat generated by radioactive decay. However, the temperature will eventually drop to levels where both humid air and aqueous phase corrosion will be possible. As the outer barrier is penetrated, uniform corrosion of the CRM will be possible in exfoliated areas. The possibility for crevice formation between the CAM and CRM will also exist. In the case of either Alloy 625 or C-22, a crevice will have to form before significant penetration of the CRM can occur. Crevice corrosion of the CRMs has been well documented. Lillard and Scully have induced crevice corrosion in Alloy 625 during exposure to artificial sea water. Jones and Wilde have prepared simulated crevice solutions of FeCl2, NiCl2 and CrCl3, and measured substantial pH suppression. Asphahani measured the dissolution rates of Alloys 625 and C-22 in such artificial crevice solutions at various temperatures. Others have observed no significant localized attack in less severe environments

  18. 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

  19. 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

  20. 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

  1. 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

  2. Gamma radiolysis effects on leaching behavior of ceramic materials for nuclear fuel waste immobilization containers

    International Nuclear Information System (INIS)

    The leaching behavior of ceramic materials for nuclear fuel waste immobilization containers, under the influence of a moderate gamma dose rate (4 Gy/h), has been investigated. Samples of Al/sub 2/O/sub 3/, stabilized ZrO/sub 2/, TiO/sub 2/, cermet (70% Al/sub 2/O-30% TiC), porcelain (with high Al/sub 2/O/sub 3/ content), and concrete (with sulfate-resisting portland cement plus silica fume) have been leached in Standard Canadian Shield Saline Solution (SCSSS), and SCSSS plus clay and sand (components of the disposal system), at 1000 and 1500C for 231 and 987 days, respectively. Leaching solutions were analyzed and the surfaces of the leached samples were investigated by scanning electron microscopy in conjunction with energy dispersive X-ray spectroscopy and secondary ion mass spectrometry. Radiolysis did not appear to enhance the leaching, with or without bentonite and sand in the system. Analysis of the gas phase from sealed capsules showed O/sub 2/ depletion and production of CO/sub 2/ in all experiments containing bentonite. The decrease in O/sub 2/ is attributed to the leaching from the clay of Fe(II) species, which can participate in redox reactions with radicals generated by radiolysis. The CO/sub 2/ is produced from either the organic or inorganic fraction in the bentonite

  3. Corrosion of container materials for disposal of high-level radioactive wastes

    Energy Technology Data Exchange (ETDEWEB)

    Chun, K.S.; Park, H.S.; Yeon, J.W.; Ha, Y.K. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-01-01

    In the corrosion aspect of container for the deep geological disposal of high-level radioactive waste, disposal concepts and the related container materials, which have been developed by advanced countries, have been reviewed. The disposal circumstances could be divided into the saturated and the unsaturated zones. The candidate materials in the countries, which consider the disposal in the unsaturated zone, are the corrosion resistant materials such as supper alloys and stainless steels, but those in the saturated zone is cupper, one of the corrosion allowable materials. By the results of the pitting corrosion test of sensitized stainless steels (such as 304, 304L, 316 and 316L), pitting potential is decreased with the degree of sensitization and the pitting corrosion resistance of 316L is higher than others. And so, the long-term corrosion experiment with 316L stainless steel specimens, sebsitized and non-sensitized, under the compacted bentonite and synthetic granitic groundwater has been being carried out. The results from the experiment for 12 months indicate that no evidence of pitting corrosion of the specimens has been observed but the crevice corrosion has occurred on the sensitized specimens even for 3 months. (author). 33 refs., 19 figs., 10 tabs.

  4. Report on the performance monitoring system for the interim waste containment at the Niagara Falls Storage Site, Lewiston, New York

    International Nuclear Information System (INIS)

    The Niagara Falls Storage Site (NFSS) is an interim storage site for low-level radioactive waste, established by the US Department of Energy (DOE) at Lewiston, New York. The waste containment structure for encapsulating low-level radioactive waste at the NFSS has been designed to minimize infiltration of rainfall, prevent pollution of groundwater, preclude formation of leachate, and prevent radon emanation. Accurately determining the performance of the main engineered elements of the containment structure will be important in establishing confidence in the ability of the structure to retain the wastes. For this purpose, a waste containment performance monitoring system has been developed to verify that these elements are functioning as intended. The key objective of the performance monitoring system is the early detection of trends that could be indicative of weaknesses developing in the containment structure so that corrective action can be taken before the integrity of the structure is compromised. Consequently, subsurface as well as surface monitoring techniques will be used. After evaluating several types of subsurface instrumentation, it was determined that vibrating wire pressure transducers, in combination with surface monitoring techniques, would satisfactorily monitor the parameters of concern, such as water accumulation inside the containment facility, waste settlement, and shrinkage of the clay cover. Surface monitoring will consist of topographic surveys based on predetermined gridlines, walkover surveys, and aerial photography to detect vegetative stress or other changes not evident at ground level. This report details the objectives of the performance monitoring system, identifies the elements of the containment design whose performance will be monitored, describes the monitoring system recommended, and outlines the costs associated with the monitoring system. 5 refs., 4 figs., 3 tabs

  5. Container Approval for the Disposal of Radioactive Waste with Negligible Heat Generation in the German Konrad Repository - 12148

    International Nuclear Information System (INIS)

    Since the license for the Konrad repository was finally confirmed by legal decision in 2007, the Federal Institute for Radiation Protection (BfS) has been performing further planning and preparation work to prepare the repository for operation. Waste conditioning and packaging has been continued by different waste producers as the nuclear industry and federal research institutes on the basis of the official disposal requirements. The necessary prerequisites for this are approved containers as well as certified waste conditioning and packaging procedures. The Federal Institute for Materials Research and Testing (BAM) is responsible for container design testing and evaluation of quality assurance measures on behalf of BfS under consideration of the Konrad disposal requirements. Besides assessing the container handling stability (stacking tests, handling loads), design testing procedures are performed that include fire tests (800 deg. C, 1 hour) and drop tests from different heights and drop orientations. This paper presents the current state of BAM design testing experiences about relevant container types (box shaped, cylindrical) made of steel sheets, ductile cast iron or concrete. It explains usual testing and evaluation methods which range from experimental testing to analytical and numerical calculations. Another focus has been laid on already existing containers and packages. The question arises as to how they can be evaluated properly especially with respect to lack of completeness of safety assessment and fabrication documentation. At present BAM works on numerous applications for container design testing for the Konrad repository. Some licensing procedures were successfully finished in the past and BfS certified several container types like steel sheet, concrete until cast iron containers which are now available for waste packaging for final disposal. However, large quantities of radioactive wastes had been placed into interim storage using containers which

  6. Interaction between clay-based sealing components and crystalline host rock

    International Nuclear Information System (INIS)

    The results of hydraulic-mechanical (H-M) numerical simulation of a shaft seal installed at a fracture zone (FZ) in a crystalline host rock using the finite element method are presented. The primary function of a shaft seal is to limit short-circuiting of the groundwater flow regime via the shaft in a deep geological repository. Two different stages of system evolution were considered in this numerical modelling. Stage 1 simulates the groundwater flow into an open shaft, prior to seal installation. Stage 2 simulates the groundwater flow into the shaft seal after seal installation. Four different cases were completed to: (i) evaluate H-M response due to the interaction between clay-based sealing material and crystalline host rock in the shaft seal structure; (ii) quantify the effect of the different times between the completion of the shaft excavation and the completion of shaft seal installation on the H-M response; and (iii) define the potential effects of different sealing material configurations. Shaft sealing materials include the bentonite-sand mixture (BSM), dense backfill (DBF), and concrete plug (CP). The BSM has greater swelling capacity and lower hydraulic conductivity (K) than the DBF. The results of these analyses show that the decrease of the pore water pressure is concentrated along the fracture zone (FZ), which has the greatest K. As the time increases, the greatest decrease in pore water pressure is found around the FZ. Following FZ isolation and the subsequent filling of the shaft with water as it floods, the pore water pressure profile tends to recover back to the initial conditions prior to shaft excavation. The majority of the fluids that ultimately saturate the centre of the shaft seal flow radially inwards from the FZ. The time between the completion of the shaft excavation and the completion of shaft seal installation has a significant effect on the saturation time. A shorter time can reduce the saturation time. Since most of the inflow comes

  7. Detection of Tritium in Storage Vaults Containing RH TRU Waste at the Idaho National Engineering and Environmental Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Pui Kuan

    2005-02-01

    Waste drums containing remote-handled (RH) transuranic (TRU) waste from the Argonne National Laboratory-East (ANL-E) are stored in sealed, underground vaults at the Radioactive Waste Management Complex (RWMC) at the Idaho National Engineering and Environmental Laboratory (INEEL). The waste consists of laboratory debris from the destructive examination of fuel elements irradiated mostly in the Experimental Breeder Reactor II (EBR-II). In 2004, air samples were obtained from some of these vaults and analyzed for radioactivity. Some of the samples show that the vaults contained several DAC's (derived air concentrations) of tritium, which are considered as non-negligible by Environmental Protection Agency (EPA) regulations. Based on Acceptable Knowledge (AK) records of the waste stored in the vaults, ORIGEN2 calculations were performed to estimate the isotopic contents in the waste drums stored in the vaults. The calculations are based on the irradiation of fuel elements that produced the waste. The absolute amounts of isotopic contents in the waste drums are normalized to Cs-137 contents derived from measured surface dose rates, mostly from the Cs-137 radiation, as documented in AK records. The amounts of tritium thus calculated (assuming no loss from those produced during fission except for decay) are compared to the measured values in the air samples from the vaults. The ratio of measured tritium in the form of tritiated water vapor to un-reduced tritium from fission is found to be from below detection levels to approximately 0.2%, but mostly in the range around 1 10-4 to 1 10-3. It appears that even the debris from cutting and grinding the fuel elements contained substantial amounts of tritium, which were subsequently released from the fuel particles during years of storage in the vaults.

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

    International Nuclear Information System (INIS)

    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 evaluation of

  9. 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.

  10. 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

  11. Characteristics on the SAP-based wasteform containing radioactive molten salt waste - 16137

    International Nuclear Information System (INIS)

    This study investigated a unique wasteform containing molten salt wastes which are generated from the pyro-process for the spent fuel treatment. Using a conventional sol-gel process, SiO2-Al2O3-P2O5 (SAP) inorganic material reactive to metal chlorides were prepared. By using this inorganic composite, a monolithic wasteform were successfully fabricated via a simple process, reaction at 650 deg. C and sintering at 1100 deg. C. This unique wasteform should be qualified if it meets the requirements for final disposal. For this reasons, this paper characterized its chemical durability, physical properties, morphology and etc. In the SAP, there are three kinds of chains, Si-O-Si as a main chain, Si-O-Al as a side chain and Al-O-P/P-O-P as a reactive chain. Alkali metal chlorides were converted into metal aluminosilicate (LixAlxSi1-xO2-x) and metal phosphate(Li3PO4 and Cs2AlP3O10) while alkali earth and rare earth chlorides were changed into only metal phosphates (Sr5(PO4)3Cl and CePO4). These reaction products were compatible to borosilicate glasses which were functioned as a chemical binder for metal aluminosilicate and a physical binder for metal phosphates. By these phenomena, the wasteform was formed homogeneously above μm scale. This would affect the leaching behaviors of each radionuclides or component of binder. The leach rates of Cs and Sr under the PCT-A test condition were about 10-3g/m2day. The physical properties (Cp, k, ρ, Hv, and etc) were very reasonable. Other leaching tests (ISO, MCC-1P) are on-going. From these results, it could be concluded that SAP can be considered as an effective stabilizer on metal chlorides and the method using SAP will give a chance to minimize the waste volume for the final disposal of salt wastes through further researches. (authors)

  12. 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.

  13. 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

  14. Physiologo-biochemical characteristics of citrate-producing yeast Yarrowia lipolytica grown on glycerol-containing waste of biodiesel industry.

    Science.gov (United States)

    Morgunov, Igor G; Kamzolova, Svetlana V

    2015-08-01

    In this study, physiologo-biochemical characteristics of citrate-producing yeast Yarrowia lipolytica grown on glycerol-containing waste of biodiesel industry were studied by an investigation of growth dynamics, the consumption of glycerol, and the fatty acid fractions from waste as well as by measuring the activities of enzymes involved in the metabolism of waste. It was shown that Y. lipolytica realizes concurrent uptake of glycerol and the fatty acid fractions during conversion of glycerol-containing waste, although glycerol was utilized at a higher rate than fatty acids. Under optimal feeding of glycerol-containing waste by portions of 20 g l(-1), the citric acid production and the ratio between citric acid and isocitric acid depended on the strain used. It was revealed that wild strain Y. lipolytica VKM Y-2373 produced citrate and isocitrate with a ratio of 1.7:1, while the mutant strain Y. lipolytica NG40/UV7 synthesized presumably citric acid (122.2 g l(-1)) with a citrate-to-isocitrate ratio of 53:1 and the yield of 0.95 g g(-1). PMID:25846335

  15. 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.

  16. 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.)

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

    International Nuclear Information System (INIS)

    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(III) in solution) performed on depth

  18. 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

  19. 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

  20. Galvanic corrosion evaluation of high activity nuclear waste container metals components

    International Nuclear Information System (INIS)

    The final disposal container for vitrified high-level waste is assumed to have three metallic layers: a stainless steel inner layer, and external one of a metal to be selected and a thick lead layer (10 cm) in the middle. As design limit, the container shall act as an engineering barrier, granting the isolation of the radionuclides for approximately 1000 years. Preliminary titanium-lead galvanic couple tests showed that titanium behaved always as a cathode in the galvanic couple, promoting the galvanic corrosion of lead. This corrosion study focused on the behaviour of lead-AISI 304 stainless steel and lead-carbon steel (SAE 1010 and 1020) galvanic couples with different area relationships, temperature and media composition. High purity lead (99,999%) and commercial lead (99,9%) were used for galvanic couples tests. Tests were performed at 75, 50, 45 and 40 deg. C. Test solution was either synthetic groundwater, a suspension of 10% bentonite in groundwater, or synthetic sea water. The synthetic sea water was used at 100, 50 and 25% concentration by dilution with distilled water. Tests with lead-304 stainless steel galvanic couples showed that lead always behaves as an anode, corroding preferentially. Very low lead corrosion rates were found in lead-carbon steel galvanic couple in 10% bentonite suspension in synthetic groundwater test at 75 deg. C. An increase of carbon content in steel has very little influence on steel corrosion rate. Commercial lead has a higher corrosion rate and presented a more pronounced attack than high purity lead. Its corrosion rate is at least twice when lead-carbon steel area relationship increases from 1:10 to 1:40. There are higher steel corrosion rates in sea water than in groundwater. Lead behaves as a cathode to the end of the test. 8 refs, 85 figs, 10 tabs