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Sample records for bituminized waste packages

  1. Development of a method to determine the nuclide inventory in bituminized waste packages; Entwicklung eines Verfahrens zur Bestimmung des Nuklidinventars in bituminierten Abfallgebinden

    Energy Technology Data Exchange (ETDEWEB)

    Mesalic, E.; Kortman, F.; Lierse von Gostomski, C. [Technische Univ. Muenchen, Garching (Germany). Zentrale Technisch-Wissenschaftliche Betriebseinheit Radiochemie Muenchen (RCM)

    2014-01-15

    Until the 1980s, bitumen was used as a conditioning agent for weak to medium radioactive liquid waste. Its use can be ascribed mainly to the properties that indicated that the matrix was optimal. However, fires broke out repeatedly during the conditioning process, so that the method is meanwhile no longer permitted in Germany. There are an estimated 100 waste packages held by the public authorities in Germany that require a supplementary declaration. In contrast to the common matrices, such as for example resins or sludges, there is still no standardized technology for taking samples and subsequently determining the radio-nuclide for bitumen. Aspects, such as the thermoplastic behaviour, make determining the nuclide inventory more difficult in bituminized waste packages. The development of a standardized technology to take samples with a subsequent determination of the radio-nuclide analysis is the objective of a project funded by the BMBF. Known, new methods, specially developed for the project, are examined on inactive bitumen samples and then transferred to active samples. At first non-destructive methods are used. The resulting information forms an important basis to work out and apply destructive strategy for sampling and analysis. Since the project is on-going, this report can only address the development of the sampling process. By developing a sampling system, it will be possible to take samples from an arbitrary selected location of the package across the entire matrix level and thus gain representative analysis material. The process is currently being optimized. (orig.)

  2. Long-term properties of bituminized waste products

    International Nuclear Information System (INIS)

    This study is a survey of the factors of importance for long term behaviour of bituminized ion exchange resins. Phenomena occurring in bituminized waste products affected by the treatment, storage and disposal are identified. Test methods have been developed for characterizing product properties, which are important for the long-term behaviour. The long-term properties of bituminized ion-exchange resins are studied in a repository environment with access of water equilibrated with concrete. In these circumstances the most important properties are ralated to the interactions of bituminized waste with the surrounding barriers. The most important phenomena are water uptake due to rehydration of the resins and subsequent swelling of the product. (author)

  3. Long-term properties of bituminized waste products

    International Nuclear Information System (INIS)

    This study is a survey of the factors of importance for the long term behaviour of bituminized ion exchange resins. Phenomena occurring in bituminized waste products affected by the treatment, storage and disposal are identified. Test methods have been developed for characterizing product properties, which are important for the long-term behaviour. The long-term properties of bituminized ion-exchange resins are studied in a repository environment with access of water equilibrated with concrete. In these circumstances the most important properties are related to the interactions of bituminized waste with the surrounding barriers. The most important phenomena are water uptake due to rehydration of the resins and subsequent swelling of the product

  4. Bituminous and asphaltic membranes for radioactive waste repositories on land

    International Nuclear Information System (INIS)

    The aim of the study has been to identify, within generic designs for waste facilities, areas where bituminous materials might be incorporated as a barrier. The report is presented in three sections: Part I - The properties of bitumen are described, with particular reference to the long-term behaviour of the material. The durability of bitumen is discussed, including aspects such as ageing, microbial degradation, chemical resistance and radiation resistance. Part II -The use of bituminous materials in hydraulic engineering is outlined. Much of this section of the report concentrates on the use of bituminous materials in embankment dams and reservoir construction. A review of material technology and construction techniques is presented and a detailed assessment made of the performance of the materials in service. Part III - Generic trench type radioactive waste repository designs incorporating bituminous materials are presented. Material and construction specifications for the designs are detailed, and a performance assessment presented. (author)

  5. State of the art report on bituminized waste forms of radioactive wastes

    International Nuclear Information System (INIS)

    In this report, research and development results on the bituminization of radioactive wastes are closely reviewed, especially those regarding waste treatment technologies, waste solidifying procedures and the characteristics of asphalt and solidified forms. A new concept of the bituminization method is suggested in this report which can improve the characteristics of solidified forms. Stable solid forms with high leach resistance, high thermal resistance and good compression strength were produced by the suggested bituminization method, in which spent polyethylene from agricultural farms was added. This report can help further research and development of improved bituminized forms of radioactive wastes that will maintain long term stabilities in disposal sites. (author). 59 refs., 19 tabs., 18 figs

  6. State of the art report on bituminized waste forms of radioactive wastes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Kook; Shon, Jong Sik; Kim, Kil Jeong; Lee, Kang Moo; Jung, In Ha

    1998-03-01

    In this report, research and development results on the bituminization of radioactive wastes are closely reviewed, especially those regarding waste treatment technologies, waste solidifying procedures and the characteristics of asphalt and solidified forms. A new concept of the bituminization method is suggested in this report which can improve the characteristics of solidified forms. Stable solid forms with high leach resistance, high thermal resistance and good compression strength were produced by the suggested bituminization method, in which spent polyethylene from agricultural farms was added. This report can help further research and development of improved bituminized forms of radioactive wastes that will maintain long term stabilities in disposal sites. (author). 59 refs., 19 tabs., 18 figs

  7. Investigation of activity release from bituminized intermediate-level waste forms under thermal stresses

    International Nuclear Information System (INIS)

    To determine the consequences of a fire during fabrication, intermediate storage and transport of bituminized NaNO3 waste forms, the fractions of plutonium released from the waste forms were assessed. For this purpose, laboratory tests were made with PuO2-containing specimens as well as a field test with specimens containing Eu2O3. By the evaluation of plutonium release in the laboratory and by the determination of the total sodium release and the relative Eu/Na release in the field tests the plutonium release can be deduced from full-scale specimens. The results show that for bituminized waste forms with high NaNO3 contents (approx. 36 wt%) the average plutonium release obtained in laboratory testing is 15%. In the field tests (IAEA fire test conditions) an average Eu release of 8% was found. These results justify the statement that also for waste forms in open 175 L drum inserts a maximum plutonium release of about 15% can be expected. From the time-dependence of Eu/Na release in the field tests an induction period of 15-20 minutes between the start of testing and the first Na/Eu release can be derived. The maximum differential Na/Eu release occurs after a test period of 45 to 60 minutes duration and after 90 to 105 minutes (tests K2 and K4, respectively); after that time also the highest temperatures in the products are measured. The release values were determined for products in open 175 L drum inserts which in this form are not eligible for intermediate and ultimate storage. For bituminized waste forms in concrete packages (lost concrete shieldings) a delayed increase in temperature to only 70-80 deg. C takes place (4-5 hours after extinction of the fire) if the fire lasts 45 minutes. The concrete package remains intact under test conditions. This means that activity release from bituminized waste forms packaged in this way can be ruled out in the case under consideration. (author)

  8. Bituminous reference document: synthesis of knowledge on the long time behavior of bituminous packages; Dossier de reference bitume: synthese des connaissances sur le comportement a long terme des colis bitumes

    Energy Technology Data Exchange (ETDEWEB)

    Sercombe, J.; Adenot, F.; Vistoli, P.P.; Parraud, S.; Riglet-Martial, C.; Gwinner, B.; Felines, I.; Tiffreau, C.; Libert, M

    2004-07-01

    This document is a synthesis of the knowledge acquired at the CEA on the behavior of bituminous packages. In this framework, the CEA studied bituminous packages in generic conditions of a package lifetime. The main factors, the evolution mechanisms and influential parameters have been determined and quantitative s simulation have been developed. After a description of the main initial bituminous packages characterizations, the evolutions in saturated and un-saturated environment are exposed. (A.L.B.)

  9. WASTE PACKAGE TRANSPORTER DESIGN

    Energy Technology Data Exchange (ETDEWEB)

    D.C. Weddle; R. Novotny; J. Cron

    1998-09-23

    The purpose of this Design Analysis is to develop preliminary design of the waste package transporter used for waste package (WP) transport and related functions in the subsurface repository. This analysis refines the conceptual design that was started in Phase I of the Viability Assessment. This analysis supports the development of a reliable emplacement concept and a retrieval concept for license application design. The scope of this analysis includes the following activities: (1) Assess features of the transporter design and evaluate alternative design solutions for mechanical components. (2) Develop mechanical equipment details for the transporter. (3) Prepare a preliminary structural evaluation for the transporter. (4) Identify and recommend the equipment design for waste package transport and related functions. (5) Investigate transport equipment interface tolerances. This analysis supports the development of the waste package transporter for the transport, emplacement, and retrieval of packaged radioactive waste forms in the subsurface repository. Once the waste containers are closed and accepted, the packaged radioactive waste forms are termed waste packages (WP). This terminology was finalized as this analysis neared completion; therefore, the term disposal container is used in several references (i.e., the System Description Document (SDD)) (Ref. 5.6). In this analysis and the applicable reference documents, the term ''disposal container'' is synonymous with ''waste package''.

  10. Study of thermal reactivity during bituminization of radioactive waste

    International Nuclear Information System (INIS)

    This work deals with the study of chemical reactions and phases transitions which can occur between magnesium nitrate, sodium nitrate, cobalt sulphur product, and nickel potassium ferrocyanide, when they are heated together during bituminization process of nuclear waste. The applied methodology associates a few techniques: temperature, enthalpy, and kinetics of reaction are determined by calorimetry, reaction products are characterised by chemical analyses, mass spectrometry and XRD analysis. Three fields of temperature and energy are observed in function of composition (one compound or a mixture of compounds). The study of reactions between NaNO3 and cobalt sulphur product shows that the presence of water has got an effect on reaction temperature. The study of Mg(NO3)2, 6 H2O and CoS shows an overlapping of different signals, and that the reaction rate is very slow (a few hours). (author)

  11. Leaching tests of simulated Cogema bituminized waste form

    Energy Technology Data Exchange (ETDEWEB)

    Nakayama, S.; Akimoto, T.; Iida, Y.; Nagano, T. [Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan)

    2000-07-01

    The leaching behavior of COGEMA-type bituminized radioactive waste was studied for the atmospheric and anaerobic conditions. Active and inactive laboratory-scale bitumen samples, including two major salts of NaNO{sub 3} and BaSO{sub 4}, were contacted with deionized water, an alkaline solution (0.01 mol/L Ca(OH){sub 2} or 0.03 mol/L KOH), or a saline solution (0.5 mol/L KCl). It was found that the release of salt was reduced in the Ca(OH){sub 2} solution compared with deionized water under the atmospheric conditions. No significant difference in the concentrations of {sup 237}Np in leachants contacted with the samples for 7 days was observed between the atmospheric and the anaerobic conditions. (authors)

  12. Waste disposal package

    Science.gov (United States)

    Smith, M.J.

    1985-06-19

    This is a claim for a waste disposal package including an inner or primary canister for containing hazardous and/or radioactive wastes. The primary canister is encapsulated by an outer or secondary barrier formed of a porous ceramic material to control ingress of water to the canister and the release rate of wastes upon breach on the canister. 4 figs.

  13. Waste package performance analysis

    International Nuclear Information System (INIS)

    A performance assessment model for multiple barrier packages containing unreprocessed spent fuel has been applied to several package designs. The resulting preliminary assessments were intended for use in making decisions about package development programs. A computer model called BARIER estimates the package life and subsequent rate of release of selected nuclides. The model accounts for temperature, pressure (and resulting stresses), bulk and localized corrosion, and nuclide retardation by the backfill after water intrusion into the waste form. The assessment model assumes a post-closure, flooded, geologic repository. Calculations indicated that, within the bounds of model assumptions, packages could last for several hundred years. Intact backfills of appropriate design may be capable of nuclide release delay times on the order of 107 yr for uranium, plutonium, and americium. 8 references, 6 figures, 9 tables

  14. Assessing radioactive concentrates and waste vapor condensate in solidifying radioactive wastes by bituminization

    International Nuclear Information System (INIS)

    A brief overview is presented of chemical and radiochemical methods used in the world for the analysis of the concentrate of liquid radioactive wastes from nuclear power plants destined for bituminization. Most methods are also suitable for an analysis of the condensate of waste vapors produced in bituminization. The methods of analysis of the radioactive concentrate from the V-1 nuclear power plant in Jaslovske Bohunice and of the waste vapors condensate were developed and tested in practice. Gross gamma activity was measured using a well-type Na(Tl) scintillation detector, the content of radionuclides was determined using semiconductor Ge(Li) spectrometry. The concentration of boric acid in the concentrate was determined by titration with mannite; in the condensate, using spectrophotometry with curcumine. The content of nitrates in both the concentrate and the condensate was determined spectrophotometrically using salicylic acid, the content of nitrites was determined by spectrophotometry using sulfanilic acid and α-naphthylamine. Carbonates and chlorides were determined by titration, sodium and potassium by flame photometry. The content of organic acids was measured by gravimetry of extracted methyl esters, the content of surfactants by spectrophotometry. Infrared spectrophotometry was used in determining hydrocarbons in the waste vapor condensate. The measured value range and the measurement errors are shown for each method. (A.K.)

  15. Safety assessment for the bituminization facility of the medium-level waste

    International Nuclear Information System (INIS)

    The safety assessment for the home-made bituminization facility of the medium-level waste during hot test is made. The average external dose equivalent per month is 0.136 x 10-2 Sv. The radioactive substance released into the environment is much lower than the permissible limit. The danger of burning and exploding of the bituminized product is avoided because the operation temperature is controlled strictly below 170 C degree. The report also briefly describes the structural characteristic of the facility, the main process and radiation protection and safety measures

  16. Radioactive waste package acceptance criteria

    International Nuclear Information System (INIS)

    Preliminary acceptance criteria have been developed for packages containing nuclear waste which must be stored or disposed of by the US Department of Energy. Acceptance criteria are necessary to ensure that the waste packages are compatible with all elements of the Waste Management System. The acceptance criteria are subject to revision since many of the constraints that will be imposed on the waste packages by the Waste Management System have either not been defined or are being revised. Delineation of the acceptance criteria will provide bases for handling, transporting and disposing of the commercial waste

  17. Use of waste plastic in construction of bituminous road

    OpenAIRE

    Abhijeet Jirge; Karan patil; Mrs.Vidula Swami; Suhas patil; Sushil patil; Karan salokhe

    2012-01-01

    Bottles, containers and packing strips etc. is increasing day by day. As a result amount of waste plastic also increases. This leads to various environmental problems. Many of the wastes produced today will remain in the environment for many years leading to various environmental concerns. Therefore it is necessary to utilize thewastes effectively with technical development in each field. Many by-products are being produced using the plastic wastes. Our present work is helping to take care of...

  18. Use of waste plastic in construction of bituminous road

    Directory of Open Access Journals (Sweden)

    Abhijeet Jirge

    2012-05-01

    Full Text Available Bottles, containers and packing strips etc. is increasing day by day. As a result amount of waste plastic also increases. This leads to various environmental problems. Many of the wastes produced today will remain in the environment for many years leading to various environmental concerns. Therefore it is necessary to utilize thewastes effectively with technical development in each field. Many by-products are being produced using the plastic wastes. Our present work is helping to take care of these aspects. Plastic waste, consisting of carry bags, cups and other utilized plastic can be used as a coating over aggregate and this coated stone can be used for roadconstruction. The mix polymer coated aggregate and tyre modified bitumen have shown higher strength. Use of this mix for road construction helps to use plastics waste. Once the plastic waste is separated from municipal solid waste, the organic matter can be converted into manure and used. Our paper will discuss in detail theprocess and its successful applications.

  19. Characterisation of cemented/bituminized LAW and MAW waste products

    International Nuclear Information System (INIS)

    In the context of work for characterising low and medium activity waste products, investigations were carried out to determine the release of radioactivity from binding waste in given accidents, such as mechanical and thermal loading for the operating phase of a final store. The effects of mechanical loads on MAW cement products and the effects of thermal laods on MAW cement and MAW bitumen products were examined. The release of fine dust reaching the lungs, with a particle size of ≤10 μm from a 200 litre roller seam cement binder with a maximum mechanical load of 3x105 Nm covering the accident case is about 1.5 g and therefore corresponds to ≅ 10-4% of the total radio-activity inventory for homogeneous products. With thermal loading (60 minute oil fire, 8000C) ≅ 10-3% of the radioactivity inventory is released via the release of water from the waste binder. The activity release of MAW bitumen products containing NaNO3 (175 litre drum) with thermal load is considerably higher, as due to the NaNO3 content of the products, after an induction period of about 20 minutes there is an exothermal reaction between the bitumen and the NaNO3, which leads to burning of the bitumen with considerable aerosol formation. The Na losses are about 32% and the Pu losses, derived from the results of laboratory experiments with samples containing Eu and Pu and samples containing Eu on the original size, are only 15% maximum, even with complete burn up. It was shown for all the investigations with samples of the original size that the effects of the load cases considered can be reduced or completely avoided by additional packing (concrete shielding). (orig./RB)

  20. Utilization of Waste Clay from Boron Production in Bituminous Geosynthetic Barrier (GBR-B) Production as Landfill Liner

    OpenAIRE

    Müfide Banar; Yücel Güney; Aysun Özkan; Zerrin Günkaya; Eren Bayrakcı; Derya Ulutaş

    2016-01-01

    Bituminous geomembranes, one type of geosynthetics, include a hot bituminous mixture with mineral filler and reinforcement. In this study, boron production waste clay (CW) was used as filler to produce a geosynthetic barrier with bentonite, waste tire, and bitumen. Bentonite and waste tires were used as auxiliary fillers and bitumen as the binder. CW/bitumen, CW/bentonite/bitumen, and CW/waste tire/bitumen mixtures were prepared by using a laboratory mixer at 100°C. Hot mixtures were extruded...

  1. The reduction of packaging waste

    Energy Technology Data Exchange (ETDEWEB)

    Raney, E.A.; Hogan, J.J.; McCollom, M.L.; Meyer, R.J.

    1994-04-01

    Nationwide, packaging waste comprises approximately one-third of the waste disposed in sanitary landfills. the US Department of Energy (DOE) generated close to 90,000 metric tons of sanitary waste. With roughly one-third of that being packaging waste, approximately 30,000 metric tons are generated per year. The purpose of the Reduction of Packaging Waste project was to investigate opportunities to reduce this packaging waste through source reduction and recycling. The project was divided into three areas: procurement, onsite packaging and distribution, and recycling. Waste minimization opportunities were identified and investigated within each area, several of which were chosen for further study and small-scale testing at the Hanford Site. Test results, were compiled into five ``how-to`` recipes for implementation at other sites. The subject of the recipes are as follows: (1) Vendor Participation Program; (2) Reusable Containers System; (3) Shrink-wrap System -- Plastic and Corrugated Cardboard Waste Reduction; (4) Cardboard Recycling ; and (5) Wood Recycling.

  2. Natural analogues of bitumen and bituminized radioactive waste

    International Nuclear Information System (INIS)

    Occurrences of natural bitumen provide possibilities to identify and assess materials, processes, phenomena and conditions in nature which can serve as model cases valid also with respect to a final repository. Natural bitumens meet the basic requirements for use as natural analogues. In natural sites, processes of bitumen formation and degradation often work simultaneously. The major processes are thermal alteration, evaporation, reaction with water, biodegradation, oxidation, weathering and radiation degradation. Progress in analytical organic geochemistry made during recent years has enabled a deeper understanding of the structural and compositional effects of these processes on bitumen. This is necessary in natural analogue applications. The time scales involved in degradation processes, as observed in natural occurrences, exceed the time scales dictated by the half-lives of most important fission products in low and medium level waste by orders of magnitude. Only exposure to the weather at the surface leads to a more rapid destruction of bitumen. Trace metals in bitumen arenot released until the matrix is completely destroyed. Products of radiation degradation and weathering of bitumen are able, to a certain degree, to delay migration of the metal ions released. Impregnation with bitumen may effectively reduce the microbial decomposition of easily destructable organic waste components, as can be deduced on the basis of the excellent preservation of fossils observed in bitumen. The complexity of all the natural sites described requires extensive understanding of the origin and kind of organic material in bitumen, its maturation, migration, alteration and degradation and of the geological and tectonic evolution of the site. The latter is used for establishing the time scale

  3. Waste package materials selection process

    International Nuclear Information System (INIS)

    The office of Civilian Radioactive Waste Management (OCRWM) of the United States Department of Energy (USDOE) is evaluating a site at Yucca Mountain in Southern Nevada to determine its suitability as a mined geologic disposal system (MGDS) for the disposal of high-level nuclear waste (HLW). The B ampersand W Fuel Company (BWFC), as a part of the Management and Operating (M ampersand O) team in support of the Yucca Mountain Site Characterization Project (YMP), is responsible for designing and developing the waste package for this potential repository. As part of this effort, Lawrence Livermore National Laboratory (LLNL) is responsible for testing materials and developing models for the materials to be used in the waste package. This paper is aimed at presenting the selection process for materials needed in fabricating the different components of the waste package

  4. Classification of waste packages

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, H.P.; Sauer, M.; Rojahn, T. [Versuchsatomkraftwerk GmbH, Kahl am Main (Germany)

    2001-07-01

    A barrel gamma scanning unit has been in use at the VAK for the classification of radioactive waste materials since 1998. The unit provides the facility operator with the data required for classification of waste barrels. Once these data have been entered into the AVK data processing system, the radiological status of raw waste as well as pre-treated and processed waste can be tracked from the point of origin to the point at which the waste is delivered to a final storage. Since the barrel gamma scanning unit was commissioned in 1998, approximately 900 barrels have been measured and the relevant data required for classification collected and analyzed. Based on the positive results of experience in the use of the mobile barrel gamma scanning unit, the VAK now offers the classification of barrels as a service to external users. Depending upon waste quantity accumulation, this measurement unit offers facility operators a reliable and time-saving and cost-effective means of identifying and documenting the radioactivity inventory of barrels scheduled for final storage. (orig.)

  5. Reference waste package environment report

    International Nuclear Information System (INIS)

    One of three candidate repository sites for high-level radioactive waste packages is located at Yucca Mountain, Nevada, in rhyolitic tuff 700 to 1400 ft above the static water table. Calculations indicate that the package environment will experience a maximum temperature of ∼2300C at 9 years after emplacement. For the next 300 years the rock within 1 m of the waste packages will remain dehydrated. Preliminary results suggest that the waste package radiation field will have very little effect on the mechanical properties of the rock. Radiolysis products will have a negligible effect on the rock even after rehydration. Unfractured specimens of repository rock show no change in hydrologic characteristics during repeated dehydration-rehydration cycles. Fractured samples with initially high permeabilities show a striking permeability decrease during dehydration-rehydration cycling, which may be due to fracture healing via deposition of silica. Rock-water interaction studies demonstrate low and benign levels of anions and most cations. The development of sorptive secondary phases such as zeolites and clays suggests that anticipated rock-water interaction may produce beneficial changes in the package environment

  6. Mechanical properties of hot bituminous mixes manufactured with recycled aggregate of Silestone® waste

    Directory of Open Access Journals (Sweden)

    Rubio, M. C.

    2011-03-01

    Full Text Available This article presents the results of a research project which analyzes the viability of incorporating waste material from decorative quartz solid surfacing in the manufacture of hot bituminous mixes. For this purpose, various bituminous mixes were manufactured with waste aggregate in different percentage. A set of tests were carried out that permitted the characterization of the mechanical behavior of these mixes. The results of these tests were similar to the results obtained when conventional mixes made from virgin quarry aggregate were tested. The results for moisture sensitivity as well as for wheel track rutting on mixes showed the optimal performance of this waste material even under very demanding traffic conditions. Laboratory studies showed that the use of this waste material in the manufacture of bituminous mixes is technically viable and can provide high-quality recycled aggregates at a very low cost, which can be used in the pavement of road.

    Este artículo muestra los resultados de un proyecto de investigación que tiene por objeto analizar la viabilidad de incorporación de residuos procedentes de piedra decorativa en la fabricación de mezclas bituminosas en caliente. Para ello se fabricaron mezclas con áridos reciclados de residuos procedentes de piedra decorativa en distintos porcentajes, realizando un conjunto de ensayos que posibilitaron caracterizar el comportamiento mecánico de dichas mezclas. Los resultados obtenidos fueron similares a los de las mezclas convencionales fabricadas con áridos vírgenes procedentes de cantera. Los valores de resistencia conservada ante la acción del agua y deformación en pista, pusieron de manifiesto la aptitud del residuo incluso ante las condiciones de tráfico más exigentes. Los trabajos realizados en el laboratorio indican que la utilización del residuo en la fabricación de mezclas bituminosas es técnicamente viable, pudiendo obtenerse áridos reciclados de gran calidad y

  7. Osmosis-induced swelling of Eurobitum bituminized radioactive waste in constant total stress conditions

    Science.gov (United States)

    Valcke, E.; Marien, A.; Smets, S.; Li, X.; Mokni, N.; Olivella, S.; Sillen, X.

    2010-11-01

    In geological disposal conditions, contact of Eurobitum bituminized radioactive waste, which contains high amounts of the hygroscopic and highly soluble NaNO 3, with groundwater will result in water uptake and swelling of the waste, and in subsequent leaching of the embedded NaNO 3 and radionuclides. The swelling of and the NaNO 3 leaching from non-radioactive Eurobitum samples, comprised between two stainless steel filters and in contact with 0.1 M KOH, was studied in restricted (semi-confined) swelling conditions, i.e. under a constant total stress, or counterpressure, of 2.2, 3.3, or 4.4 MPa ( i.e. oedometer conditions). Four tests were stopped after hydration times between 800 and 1500 days, and the samples were analyzed by micro-focus X-ray Computer Tomography (μCT) and by Environmental Scanning Electron Microscopy (ESEM). The complete set of data enabled a consistent interpretation of the observations and lead to an improved understanding of the phenomenology of the water uptake, swelling, and NaNO 3 leaching in restricted swelling conditions. Under the studied conditions, the bituminous matrix surrounding the NaNO 3 crystals and pores with NaNO 3 solution behaved as a highly efficient semi-permeable membrane, i.e. osmotic processes occurred. In the main part of the leached layers, a high average NaNO 3 concentration and related to this a high osmotic pressure prevailed, explaining why in the studied range the swelling was not measurably affected by the counterpressure. At the interface with the stainless steel filters, a low permeable re-compressed bitumen layer was formed, contributing to the slow release of NaNO 3 compared to the water uptake rate. A fully coupled Chemo-Hydro-Mechanical (CHM) constitutive model has been developed that integrates the key processes involved and that reproduces satisfactorily the results; this is presented in another work. Combination of the experimental and the modelling study allow to conclude that under semi

  8. Utilization of Waste Clay from Boron Production in Bituminous Geosynthetic Barrier (GBR-B Production as Landfill Liner

    Directory of Open Access Journals (Sweden)

    Müfide Banar

    2016-01-01

    Full Text Available Bituminous geomembranes, one type of geosynthetics, include a hot bituminous mixture with mineral filler and reinforcement. In this study, boron production waste clay (CW was used as filler to produce a geosynthetic barrier with bentonite, waste tire, and bitumen. Bentonite and waste tires were used as auxiliary fillers and bitumen as the binder. CW/bitumen, CW/bentonite/bitumen, and CW/waste tire/bitumen mixtures were prepared by using a laboratory mixer at 100°C. Hot mixtures were extruded into strips by using a lab-scale corotating twin screw extruder (L/D: 40 followed by die casting (2 mm × 100 mm. Glass fleece or nonwoven polyester was used as reinforcement material and while die casting, both sides of the reinforcement materials were covered with bituminous mixture. Thickness, mass per unit area, tensile strength, elongation at yield, and hydraulic conductivity were used to characterize the geomembranes. Among all geomembranes, nonwoven polyester covered with 30% bitumen-70% boron waste clay mixture (PK-BTM30CW70 was found to be the most promising in terms of structure and mechanical behaviour. After that, consequences of its exposure to distilled water (DW, municipal solid waste landfill leachate (L-MSW, and hazardous waste landfill leachate (L-HW were examined to use for an innovative impermeable liner on solid waste landfills.

  9. Safety Analysis Report for packaging (onsite) steel waste package

    International Nuclear Information System (INIS)

    The steel waste package is used primarily for the shipment of remote-handled radioactive waste from the 324 Building to the 200 Area for interim storage. The steel waste package is authorized for shipment of transuranic isotopes. The maximum allowable radioactive material that is authorized is 500,000 Ci. This exceeds the highway route controlled quantity (3,000 A2s) and is a type B packaging

  10. Waste Package Design Methodology Report

    Energy Technology Data Exchange (ETDEWEB)

    D.A. Brownson

    2001-09-28

    The objective of this report is to describe the analytical methods and processes used by the Waste Package Design Section to establish the integrity of the various waste package designs, the emplacement pallet, and the drip shield. The scope of this report shall be the methodology used in criticality, risk-informed, shielding, source term, structural, and thermal analyses. The basic features and appropriateness of the methods are illustrated, and the processes are defined whereby input values and assumptions flow through the application of those methods to obtain designs that ensure defense-in-depth as well as satisfy requirements on system performance. Such requirements include those imposed by federal regulation, from both the U.S. Department of Energy (DOE) and U.S. Nuclear Regulatory Commission (NRC), and those imposed by the Yucca Mountain Project to meet repository performance goals. The report is to be used, in part, to describe the waste package design methods and techniques to be used for producing input to the License Application Report.

  11. Comparative study of test methods for bituminized and other low- and medium-level solidified waste materials

    International Nuclear Information System (INIS)

    Various aspects of the behaviour of bituminized or cemented simulated low- or medium-level radioactive waste in contact with water or salt solutions have been investigated. The solubility (approximately 0.5%) and the diffusion coefficient (approximately 5.10-8 cm2/sec) determining transort of water in pure bitumen have been measured for Mexphalte 40/50 at room temperature. A weighing method has been used to study water uptake and swelling of bituminized sodium nitrate, sodium sulphate or cation-exchange resin. The swelling of samples in contact with water was in some cases very pronounced. In strong salt solutions the tendency to swell is much less. The particle size of the embedded waste material is an important parameter. Thermal pre-treatment of cation-exchange resin before bituminization does not seem to improve the quality of the final product. The interaction between bituminized-exchange resin and concrete barrier materials has been studied. Microbial degradation of bitumen and bituminized waste under aerobic conditions has been investigated. It is probably of minor importance as far as leaching is concerned. A method for measuring leaching from a plane surface of cemented waste has been developed. The method avoids the problem of cracks between the sample and the container. Leaching from cemented sodium nitrate or sulphate was investigated. Absorption of CO2 from the atmosphere was found to have only minor effect on Cs- and Na-leaching but gave a pronounced decrease in Ca-leaching. The use of silica-fume as an additive to cemented sodium nitrate decreased the leach rate by a factor 4. (author)

  12. Recycling of polyethylene terephthalate (PET plastic bottle wastes in bituminous asphaltic concrete

    Directory of Open Access Journals (Sweden)

    Adebayo Olatunbosun Sojobi

    2016-12-01

    Full Text Available This research sheds light on the concept of eco-friendly road construction which comprises eco-design, eco-extraction, eco-manufacturing, eco-construction, eco-rehabilitation, eco-maintenance, eco-demolition, and socioeconomic empowerment. It also revealed the challenges being faced in its adoption and the benefits derivable from its application. Furthermore, the effects of recycling PET plastic bottle wastes produced in North Central Nigeria in bituminous asphaltic concrete (BAC used in flexible pavement construction were also evaluated. The mix design consists of 60/70 penetration-grade asphaltic concrete (5%, 68% coarse aggregate, 6% fine aggregate, and 21% filler using the dry process at 170°C. The optimum bitumen content (OBC for conventional BAC was obtained as 4% by weight of total aggregates and filler. Polymer-coated aggregate (PCA-modified BAC seems preferable because it has the potential to utilize more plastic wastes with a higher optimum plastic content (OPC of 16.7% by weight of total aggregates and filler compared to that of 9% by weight of OBC achieved by PMB-BAC. For both PMB- and PCA-modified BAC, an increase in air void, void in mineral aggregate, and Marshall stability were observed. Eco-friendly road construction which recycles PET wastes should be encouraged by government considering its potential environmental and economic benefits.

  13. Bituminization plant Jaslovske Bohunice

    International Nuclear Information System (INIS)

    In this leaflet the principle of the bituminization plant for radioactive concentrate (the intermediate liquid radioactive waste generated during the NPP A1, V-1, V-2 operations) solidification used in the Bohunice Radwaste Treatment Centre (BSC RAO) is presented

  14. Waste package performance in unsaturated rock

    International Nuclear Information System (INIS)

    The unsaturated rock and near-atmospheric pressure of the potential nuclear waste repository at Yucca Mountain present new problems of predicting waste package performance. In this paper we present some illustrations of predictions of waste package performance and discuss important data needs. 11 refs., 9 figs., 1 tab

  15. Packaged low-level waste verification system

    Energy Technology Data Exchange (ETDEWEB)

    Tuite, K.; Winberg, M.R.; McIsaac, C.V. [Idaho National Engineering Lab., Idaho Falls, ID (United States)

    1995-12-31

    The Department of Energy through the National Low-Level Waste Management Program and WMG Inc. have entered into a joint development effort to design, build, and demonstrate the Packaged Low-Level Waste Verification System. Currently, states and low-level radioactive waste disposal site operators have no method to independently verify the radionuclide content of packaged low-level waste that arrives at disposal sites for disposition. At this time, the disposal site relies on the low-level waste generator shipping manifests and accompanying records to ensure that low-level waste received meets the site`s waste acceptance criteria. The subject invention provides the equipment, software, and methods to enable the independent verification of low-level waste shipping records to ensure that the site`s waste acceptance criteria are being met. The objective of the prototype system is to demonstrate a mobile system capable of independently verifying the content of packaged low-level waste.

  16. Assessing radionuclide release and migration during deep geological disposal of bituminized waste: the COLONBO model

    International Nuclear Information System (INIS)

    Full text of publication follows: Bitumen is being used for many years in France as an embedding matrix for the conditioning of low and medium level activity waste resulting from nuclear effluent treatment. Long term management options for this kind of waste include at present deep geological disposal. As a consequence, the COLONBO model has been developed by CEA to predict the long term release of chemical species out of bitumen waste and the subsequent radionuclide migration in the deep geological near field. The COLONBO model is divided into a 'waste package' module, coupled with a 'near field' module. In the first module, the waste package is represented as a bitumen matrix containing an homogeneous dispersion of the salts resulting from the effluent insolubilization treatment. Radionuclides initially present in the waste are thus associated with a mixture of soluble (NaNO3 and Na2SO4) and sparingly soluble (mainly BaSO4) salts. Under leaching, solubilization of water in the bitumen occur at the wasteform surface, followed by water diffusion through the bitumen matrix. Consequently, the soluble salts are progressively dissolved by water accumulation, resulting in a local swelling of the waste package. This swelling induces a local alteration of the diffusive properties of the bitumen matrix, allowing an outward diffusion of the dissolved salts and radionuclides. The resulting migration of chemical species through the storage facility near field is then described in a 'near field' module. In this module, transport is accounted by a simplified reactive diffusion model through porous media, in both saturated or unsaturated conditions. Up to three different zones of the near field can be included, which are described by their respective porosity, hydraulic conductivity, diffusion and retention (Kd) parameters. In the whole system (waste package + near field), water concentration profiles, the position of salt dissolution fronts and chemical species migration are

  17. Behaviour of bituminized waste under gamma irradiation. Effect of STE3 decontamination process components

    International Nuclear Information System (INIS)

    Liquid wastes of light and medium activity are treated by chemical co-precipitation and sludge are mixed with bitumen. Irradiation is responsible of gas production and potential swelling of the embedded. It prevails on limitation of filling of storage containers and activity to 140 Ci. The aim of this work is the study of influence of the components of the decontamination process on the behaviour of bituminous wastes, in order to control swelling and to state radiolysis mechanisms, both for production and storage of wastes. For pure bitumen, mechanism of production of H2 and CH4 are specified. Oxygen consumption, localised on the surface of samples, leads to conversion of aromatic oils and resins to asphaltenes, by a chain reaction mechanism. CO2 et CO are reaction products of gaseous oxygen, respectively with bitumen and light hydrocarbons. The composition of bitumen is slightly modified to heavier and higher polarity products, with parallel hardening. NaNO3, Na2SO4, BaSO4, PPFNi, K2SO4, NiSO4, et diatoms DIT3R et DIC3 have strictly a dilution effect towards gas generation. CoS, above 1% embedded, strongly inhibits production of H2, CH4 and light hydrocarbons. Degradation of bitumen being reduced, a radical mechanism with both radicals H· et R· might exist. Kinetic shows that a bi-radicals mechanism (or more) is probable. In the same way, Raney's nickel induces a important decrease of production of H2, CH4 et C2, with a capacity of 7,7 ml/g. Swelling depends on dimension of sample gas production and dose rate. Solid content and particle size are not determining parameters. Low swelling is obtained for penetrability higher than 70 1/10 mm, This can be realised by addition of a solvent as TBP and by increasing temperature above 40 deg C. Rheological characterizations (oscillation and creeping mode) have not been successful to correlate swelling with a physical parameter. (author)

  18. CERAMIC WASTE FORM DATA PACKAGE

    Energy Technology Data Exchange (ETDEWEB)

    Amoroso, J.; Marra, J.

    2014-06-13

    The purpose of this data package is to provide information about simulated crystalline waste forms that can be used to select an appropriate composition for a Cold Crucible Induction Melter (CCIM) proof of principle demonstration. Melt processing, viscosity, electrical conductivity, and thermal analysis information was collected to assess the ability of two potential candidate ceramic compositions to be processed in the Idaho National Laboratory (INL) CCIM and to guide processing parameters for the CCIM operation. Given uncertainties in the CCIM capabilities to reach certain temperatures throughout the system, one waste form designated 'Fe-MP' was designed towards enabling processing and another, designated 'CAF-5%TM-MP' was designed towards optimized microstructure. Melt processing studies confirmed both compositions could be poured from a crucible at 1600{degrees}C although the CAF-5%TM-MP composition froze before pouring was complete due to rapid crystallization (upon cooling). X-ray diffraction measurements confirmed the crystalline nature and phase assemblages of the compositions. The kinetics of melting and crystallization appeared to vary significantly between the compositions. Impedance spectroscopy results indicated the electrical conductivity is acceptable with respect to processing in the CCIM. The success of processing either ceramic composition will depend on the thermal profiles throughout the CCIM. In particular, the working temperature of the pour spout relative to the bulk melter which can approach 1700{degrees}C. The Fe-MP composition is recommended to demonstrate proof of principle for crystalline simulated waste forms considering the current configuration of INL's CCIM. If proposed modifications to the CCIM can maintain a nominal temperature of 1600{degrees}C throughout the melter, drain, and pour spout, then the CAF-5%TM-MP composition should be considered for a proof of principle demonstration.

  19. Engineered waste-package-system design specification

    Energy Technology Data Exchange (ETDEWEB)

    1983-05-01

    This report documents the waste package performance requirements and geologic and waste form data bases used in developing the conceptual designs for waste packages for salt, tuff, and basalt geologies. The data base reflects the latest geotechnical information on the geologic media of interest. The parameters or characteristics specified primarily cover spent fuel, defense high-level waste, and commercial high-level waste forms. The specification documents the direction taken during the conceptual design activity. A separate design specification will be developed prior to the start of the preliminary design activity.

  20. Engineered waste-package-system design specification

    International Nuclear Information System (INIS)

    This report documents the waste package performance requirements and geologic and waste form data bases used in developing the conceptual designs for waste packages for salt, tuff, and basalt geologies. The data base reflects the latest geotechnical information on the geologic media of interest. The parameters or characteristics specified primarily cover spent fuel, defense high-level waste, and commercial high-level waste forms. The specification documents the direction taken during the conceptual design activity. A separate design specification will be developed prior to the start of the preliminary design activity

  1. Prevention policies addressing packaging and packaging waste: Some emerging trends.

    Science.gov (United States)

    Tencati, Antonio; Pogutz, Stefano; Moda, Beatrice; Brambilla, Matteo; Cacia, Claudia

    2016-10-01

    Packaging waste is a major issue in several countries. Representing in industrialized countries around 30-35% of municipal solid waste yearly generated, this waste stream has steadily grown over the years even if, especially in Europe, specific recycling and recovery targets have been fixed. Therefore, an increasing attention starts to be devoted to prevention measures and interventions. Filling a gap in the current literature, this explorative paper is a first attempt to map the increasingly important phenomenon of prevention policies in the packaging sector. Through a theoretical sampling, 11 countries/states (7 in and 4 outside Europe) have been selected and analyzed by gathering and studying primary and secondary data. Results show evidence of three specific trends in packaging waste prevention policies: fostering the adoption of measures directed at improving packaging design and production through an extensive use of the life cycle assessment; raising the awareness of final consumers by increasing the accountability of firms; promoting collaborative efforts along the packaging supply chains. PMID:27372152

  2. Prevention policies addressing packaging and packaging waste: Some emerging trends.

    Science.gov (United States)

    Tencati, Antonio; Pogutz, Stefano; Moda, Beatrice; Brambilla, Matteo; Cacia, Claudia

    2016-10-01

    Packaging waste is a major issue in several countries. Representing in industrialized countries around 30-35% of municipal solid waste yearly generated, this waste stream has steadily grown over the years even if, especially in Europe, specific recycling and recovery targets have been fixed. Therefore, an increasing attention starts to be devoted to prevention measures and interventions. Filling a gap in the current literature, this explorative paper is a first attempt to map the increasingly important phenomenon of prevention policies in the packaging sector. Through a theoretical sampling, 11 countries/states (7 in and 4 outside Europe) have been selected and analyzed by gathering and studying primary and secondary data. Results show evidence of three specific trends in packaging waste prevention policies: fostering the adoption of measures directed at improving packaging design and production through an extensive use of the life cycle assessment; raising the awareness of final consumers by increasing the accountability of firms; promoting collaborative efforts along the packaging supply chains.

  3. Waste Package Component Design Methodology Report

    Energy Technology Data Exchange (ETDEWEB)

    D.C. Mecham

    2004-07-12

    This Executive Summary provides an overview of the methodology being used by the Yucca Mountain Project (YMP) to design waste packages and ancillary components. This summary information is intended for readers with general interest, but also provides technical readers a general framework surrounding a variety of technical details provided in the main body of the report. The purpose of this report is to document and ensure appropriate design methods are used in the design of waste packages and ancillary components (the drip shields and emplacement pallets). The methodology includes identification of necessary design inputs, justification of design assumptions, and use of appropriate analysis methods, and computational tools. This design work is subject to ''Quality Assurance Requirements and Description''. The document is primarily intended for internal use and technical guidance for a variety of design activities. It is recognized that a wide audience including project management, the U.S. Department of Energy (DOE), the U.S. Nuclear Regulatory Commission, and others are interested to various levels of detail in the design methods and therefore covers a wide range of topics at varying levels of detail. Due to the preliminary nature of the design, readers can expect to encounter varied levels of detail in the body of the report. It is expected that technical information used as input to design documents will be verified and taken from the latest versions of reference sources given herein. This revision of the methodology report has evolved with changes in the waste package, drip shield, and emplacement pallet designs over many years and may be further revised as the design is finalized. Different components and analyses are at different stages of development. Some parts of the report are detailed, while other less detailed parts are likely to undergo further refinement. The design methodology is intended to provide designs that satisfy the safety

  4. Waste Package Component Design Methodology Report

    International Nuclear Information System (INIS)

    This Executive Summary provides an overview of the methodology being used by the Yucca Mountain Project (YMP) to design waste packages and ancillary components. This summary information is intended for readers with general interest, but also provides technical readers a general framework surrounding a variety of technical details provided in the main body of the report. The purpose of this report is to document and ensure appropriate design methods are used in the design of waste packages and ancillary components (the drip shields and emplacement pallets). The methodology includes identification of necessary design inputs, justification of design assumptions, and use of appropriate analysis methods, and computational tools. This design work is subject to ''Quality Assurance Requirements and Description''. The document is primarily intended for internal use and technical guidance for a variety of design activities. It is recognized that a wide audience including project management, the U.S. Department of Energy (DOE), the U.S. Nuclear Regulatory Commission, and others are interested to various levels of detail in the design methods and therefore covers a wide range of topics at varying levels of detail. Due to the preliminary nature of the design, readers can expect to encounter varied levels of detail in the body of the report. It is expected that technical information used as input to design documents will be verified and taken from the latest versions of reference sources given herein. This revision of the methodology report has evolved with changes in the waste package, drip shield, and emplacement pallet designs over many years and may be further revised as the design is finalized. Different components and analyses are at different stages of development. Some parts of the report are detailed, while other less detailed parts are likely to undergo further refinement. The design methodology is intended to provide designs that satisfy the safety and operational

  5. Aqueous Corrosion Rates for Waste Package Materials

    International Nuclear Information System (INIS)

    The purpose of this analysis, as directed by ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]), is to compile applicable corrosion data from the literature (journal articles, engineering documents, materials handbooks, or standards, and national laboratory reports), evaluate the quality of these data, and use these to perform statistical analyses and distributions for aqueous corrosion rates of waste package materials. The purpose of this report is not to describe the performance of engineered barriers for the TSPA-LA. Instead, the analysis provides simple statistics on aqueous corrosion rates of steels and alloys. These rates are limited by various aqueous parameters such as temperature (up to 100 C), water type (i.e., fresh versus saline), and pH. Corrosion data of materials at pH extremes (below 4 and above 9) are not included in this analysis, as materials commonly display different corrosion behaviors under these conditions. The exception is highly corrosion-resistant materials (Inconel Alloys) for which rate data from corrosion tests at a pH of approximately 3 were included. The waste package materials investigated are those from the long and short 5-DHLW waste packages, 2-MCO/2-DHLW waste package, and the 21-PWR commercial waste package. This analysis also contains rate data for some of the materials present inside the fuel canisters for the following fuel types: U-Mo (Fermi U-10%Mo), MOX (FFTF), Thorium Carbide and Th/U Carbide (Fort Saint Vrain [FSVR]), Th/U Oxide (Shippingport LWBR), U-metal (N Reactor), Intact U-Oxide (Shippingport PWR, Commercial), aluminum-based, and U-Zr-H (TRIGA). Analysis of corrosion rates for Alloy 22, spent nuclear fuel, defense high level waste (DHLW) glass, and Titanium Grade 7 can be found in other analysis or model reports

  6. Aqueous Corrosion Rates for Waste Package Materials

    Energy Technology Data Exchange (ETDEWEB)

    S. Arthur

    2004-10-08

    The purpose of this analysis, as directed by ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]), is to compile applicable corrosion data from the literature (journal articles, engineering documents, materials handbooks, or standards, and national laboratory reports), evaluate the quality of these data, and use these to perform statistical analyses and distributions for aqueous corrosion rates of waste package materials. The purpose of this report is not to describe the performance of engineered barriers for the TSPA-LA. Instead, the analysis provides simple statistics on aqueous corrosion rates of steels and alloys. These rates are limited by various aqueous parameters such as temperature (up to 100 C), water type (i.e., fresh versus saline), and pH. Corrosion data of materials at pH extremes (below 4 and above 9) are not included in this analysis, as materials commonly display different corrosion behaviors under these conditions. The exception is highly corrosion-resistant materials (Inconel Alloys) for which rate data from corrosion tests at a pH of approximately 3 were included. The waste package materials investigated are those from the long and short 5-DHLW waste packages, 2-MCO/2-DHLW waste package, and the 21-PWR commercial waste package. This analysis also contains rate data for some of the materials present inside the fuel canisters for the following fuel types: U-Mo (Fermi U-10%Mo), MOX (FFTF), Thorium Carbide and Th/U Carbide (Fort Saint Vrain [FSVR]), Th/U Oxide (Shippingport LWBR), U-metal (N Reactor), Intact U-Oxide (Shippingport PWR, Commercial), aluminum-based, and U-Zr-H (TRIGA). Analysis of corrosion rates for Alloy 22, spent nuclear fuel, defense high level waste (DHLW) glass, and Titanium Grade 7 can be found in other analysis or model reports.

  7. Method of packaging radioactive wastes

    International Nuclear Information System (INIS)

    Purpose: To decrease the leaching of radioactive waste in marine environment. Method: Fillers are placed between a drum can and an inner cage for charging radioactive wastes in order to prevent the leakage of the radioactive wastes from the drum can. Leaching inhibitors for radioactive materials are mixed with the fillers made of organic substance such as asphalts and plastics. The leaching inhibitors are made of materials in the similar chemical form to that of the radioactive materials in the wastes and mixed into the fillers to the saturation limit of dissolution. For the radioactive wastes containing spent adsorbents for iodine, the inhibitors are made of silver nitrates. (Ikeda, J.)

  8. 44-BWR WASTE PACKAGE LOADING CURVE EVALUATION

    International Nuclear Information System (INIS)

    The objective of this calculation is to evaluate the required minimum burnup as a function of initial boiling water reactor (BWR) assembly enrichment that would permit loading of spent nuclear fuel into the 44 BWR waste package configuration as provided in Attachment IV. This calculation is an application of the methodology presented in ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2003). The scope of this calculation covers a range of enrichments from 0 through 5.0 weight percent (wt%) U-235, and a burnup range of 0 through 40 GWd/MTU. This activity supports the validation of the use of burnup credit for commercial spent nuclear fuel applications. The intended use of these results will be in establishing BWR waste package configuration loading specifications. Limitations of this evaluation are as follows: (1) The results are based on burnup credit for actinides and selected fission products as proposed in YMP (2003, Table 3-1) and referred to as the ''Principal Isotopes''. Any change to the isotope listing will have a direct impact on the results of this report. (2) The results of 100 percent of the current BWR projected waste stream being able to be disposed of in the 44-BWR waste package with Ni-Gd Alloy absorber plates is contingent upon the referenced waste stream being sufficiently similar to the waste stream received for disposal. (3) The results are based on 1.5 wt% Gd in the Ni-Gd Alloy material and having no tuff inside the waste package. If the Gd loading is reduced or a process to introduce tuff inside the waste package is defined, then this report would need to be reevaluated based on the alternative materials

  9. Simulated waste package test in salt

    International Nuclear Information System (INIS)

    The Salt Repository Site Characterization Project Office (SRPO), of the U.S. Department of Energy (DOE) Office of the Civilian Radioactive Waste Management (OCRWM), in cooperation with Federal Republic of Germany (FRG), simulated waste package test at Asse Salt Mine (Asse). The purpose of this test was to determine the effect of heat produced of the decay of High-Level Radioactive Waste (HLW) on: (1) Migration of brine moisture; (2) Thermomechanical response of the salt; (3) Geomechanical response of the room mined in salt; (4) Corrosion on potential HLW waste package container materials; and (5) Generation of gases. This paper describes the test performed, results obtained, and the performance of instruments and data acquisition system deployed

  10. WASTE PACKAGE REMEDIATION SYSTEM DESCRIPTION DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    N.D. Sudan

    2000-06-22

    The Waste Package Remediation System remediates waste packages (WPs) and disposal containers (DCs) in one of two ways: preparation of rejected DC closure welds for repair or opening of the DC/WP. DCs are brought to the Waste Package Remediation System for preparation of rejected closure welds if testing of the closure weld by the Disposal Container Handling System indicates an unacceptable, but repairable, welding flaw. DC preparation of rejected closure welds will require removal of the weld in such a way that the Disposal Container Handling System may resume and complete the closure welding process. DCs/WPs are brought to the Waste Package Remediation System for opening if the Disposal Container Handling System testing of the DC closure weld indicates an unrepairable welding flaw, or if a WP is recovered from the subsurface repository because suspected damage to the WP or failure of the WP has occurred. DC/WP opening will require cutting of the DC/WP such that a temporary seal may be installed and the waste inside the DC/WP removed by another system. The system operates in a Waste Package Remediation System hot cell located in the Waste Handling Building that has direct access to the Disposal Container Handling System. One DC/WP at a time can be handled in the hot cell. The DC/WP arrives on a transfer cart, is positioned within the cell for system operations, and exits the cell without being removed from the cart. The system includes a wide variety of remotely operated components including a manipulator with hoist and/or jib crane, viewing systems, machine tools for opening WPs, and equipment used to perform pressure and gas composition sampling. Remotely operated equipment is designed to facilitate DC/WP decontamination and hot cell equipment maintenance, and interchangeable components are provided where appropriate. The Waste Package Remediation System interfaces with the Disposal Container Handling System for the receipt and transport of WPs and DCs. The Waste

  11. Nuclear waste package fabricated from concrete

    International Nuclear Information System (INIS)

    After the United States enacted the Nuclear Waste Policy Act in 1983, the Department of Energy must design, site, build and operate permanent geologic repositories for high-level nuclear waste. The Department of Energy has recently selected three sites, one being the Hanford Site in the state of Washington. At this particular site, the repository will be located in basalt at a depth of approximately 3000 feet deep. The main concern of this site, is contamination of the groundwater by release of radionuclides from the waste package. The waste package basically has three components: the containment barrier (metal or concrete container, in this study concrete will be considered), the waste form, and other materials (such as packing material, emplacement hole liners, etc.). The containment barriers are the primary waste container structural materials and are intended to provide containment of the nuclear waste up to a thousand years after emplacement. After the containment barriers are breached by groundwater, the packing material (expanding sodium bentonite clay) is expected to provide the primary control of release of radionuclide into the immediate repository environment. The loading conditions on the concrete container (from emplacement to approximately 1000 years), will be twofold; (1) internal heat of the high-level waste which could be up to 4000C; (2) external hydrostatic pressure up to 1300 psi after the seepage of groundwater has occurred in the emplacement tunnel. A suggested container is a hollow plain concrete cylinder with both ends capped. 7 refs

  12. Second Generation Waste Package Design Study

    International Nuclear Information System (INIS)

    The following describes the objectives of Project Activity 023 ''Second Generation Waste Package Design Study'' under DOE Cooperative Agreement DE-FC28-04RW12232. The objectives of this activity are: to review the current YMP baseline environment and establish corrosion test environments representative of the range of dry to intermittently wet conditions expected in the drifts as a function of time; to demonstrate the oxidation and corrosion resistance of A588 weathering steel and reference Alloy 22 samples in the representative dry to intermittently dry conditions; and to evaluate backfill and design features to improve the thermal performance analyses of the proposed second-generation waste packages using existing models developed at the University of Nevada, Reno(UNR). The work plan for this project activity consists of three major tasks: Task 1. Definition of expected worst-case environments (humidity, liquid composition and temperature) at waste package outer surfaces as a function of time, and comparison with environments defined in the YMP baseline; Task 2. Oxidation and corrosion tests of proposed second-generation outer container material; and Task 3. Second Generation waste package thermal analyses. Full funding was not provided for this project activity

  13. Physical properties of solid fuel briquettes from bituminous coal waste and biomass

    Institute of Scientific and Technical Information of China (English)

    ZARRINGHALAM-MOGHADDAM A; GHOLIPOUR-ZANJANI N; DOROSTIS; VAEZ M

    2011-01-01

    Biomass and bituminous coal fines from four different coalfields were used to produce fuel briquettes.Two physical properties of briquettes,water resistance index and compressive strength were analyzed.The influence of type and quantity of biomass on physical properties was also studied.The results reveal that depending on the mineral content of the coal,the physical properties of the briquettes differ noticeably.The comparison of briquettes with and without biomass showed that the presence of the beet pulp increased CS in all types of coal samples.Samples containing beet pulp had better physical properties than sawdust.Mezino Ⅱ coal briquettes had highest CS and WRI than the other ones.Calorific value of biomass/Mezino Ⅱ coal briquettes was lessened in comparison with raw coal,but it remained in an acceptable range.

  14. Hydrogen generation in tru waste transportation packages

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, B; Sheaffer, M K; Fischer, L E

    2000-03-27

    This document addresses hydrogen generation in TRU waste transportation packages. The potential sources of hydrogen generation are summarized with a special emphasis on radiolysis. After defining various TRU wastes according to groupings of material types, bounding radiolytic G-values are established for each waste type. Analytical methodologies are developed for prediction of hydrogen gas concentrations for various packaging configurations in which hydrogen generation is due to radiolysis. Representative examples are presented to illustrate how analytical procedures can be used to estimate the hydrogen concentration as a function of time. Methodologies and examples are also provided to show how the time to reach a flammable hydrogen concentration in the innermost confinement layer can be estimated. Finally, general guidelines for limiting the hydrogen generation in the payload and hydrogen accumulation in the innermost confinement layer are described.

  15. Drift emplaced waste package thermal response

    International Nuclear Information System (INIS)

    Thermal calculations of the effects of radioactive waste decay heat on the potential repository at Yucca Mountain, Nevada, have been conducted by the Yucca Mountain Site Characterization Project (YMP) at Lawrence Livermore National Lab. (LLNL) in conjunction with the B ampersand W Fuel Co. For a number of waste package spacings, these 3D transient calculations use the TOPAZ3D code to predict drift wall temperatures to 10,000 years following emplacement. Systematic temperature variation occurs as a function of fuel age at emplacement and Areal Mass Loading (AML) during the first few centuries after emplacement. After about 1000 years, emplacement age is not a strong driver on rock temperature; AML has a larger impact. High AMLs occur when large waste packages are emplaced end-to-end in drifts. Drift emplacement of equivalent packages results in lower rock temperatures than borehole emplacement. For an emplacement scheme with 50% of the drift length occupied by packages, an AML of 138 MTU/acre is about three times higher than the Site Characterization Plan-Conceptual Design (SCP-CD) value. With this higher AML (requiring only 1/3 of the SCP-CD repository footprint), peak drift wall temperatures do not exceed 160 degrees C, but rock temperatures exceed the boiling point of water for about 3000 years. These TOPAZ3D results have been compared with reasonable agreement with two other computer codes

  16. Development of new non destructive methods for bituminized radioactive waste drums characterization; Developpement de nouvelles methodes de caracterisation non destructive pour des dechets radioactifs enrobes dans du bitume

    Energy Technology Data Exchange (ETDEWEB)

    Pin, P

    2004-10-15

    Radioactive waste constitute a major issue for the nuclear industry. One of the key points is their characterization to optimize their management: treatment and packaging, orientation towards the suited disposal. This thesis proposes an evaluation method of the low-energy photon attenuation, based on the gamma-ray spectra Compton continuum. Effectively, the {sup 241}Am measurement by gamma-ray spectrometry is difficult due to the low energy of its main gamma-ray (59.5 keV). The photon attenuation strongly depends on the bituminous mix composition, which includes very absorbing elements. As the Compton continuum also depends on this absorption, it is possible to link the 59.5 keV line attenuation to the Compton level. Another technique is proposed to characterize uranium thanks to its fluorescence X-rays induced by the gamma emitters already present in the waste. The uranium present in the drums disturbs the neutron measurements and its measurement by self-induced X-ray fluorescence allows to correct this interference. Due to various causes of error, the total uncertainty is around 50 % on the activity of the radioisotope {sup 241}Am, corrected by the peak to Compton technique. The same uncertainty is announced on the uranium mass measured by self induced X-ray fluorescence. As a consequence of these promising results, the two methods were included in the industrial project of the 'Marcoule Sorting Unit'. One major advantage is that they do not imply any additional material because they use information already present in the gamma-ray spectra. (author)

  17. Waste to Wealth; The Utilization of Scrap Tyre as Aggregate in Bituminous Mixes for Road Construction

    OpenAIRE

    Oba, A. L.,; Onungwe, I.,

    2015-01-01

    The problem associated with solid waste management is on the increase both in the industries, urban cities and in the rural areas. In the United States of America, Asia and Europe, there are over hundreds of waste to wealth combustion plants from where solid wastes are incinerated. In Nigeria, amidst the increasing importation of vehicle tyre such plants are scarcely in existence to enhance generation of revenue from waste through the extraction of raw material for the production ...

  18. Applied Technological Direction of Power Plant Ash and Slag Waste Management when Kuznetsk Bituminous Coal is Burned

    Directory of Open Access Journals (Sweden)

    Lihach Snejana A.

    2016-01-01

    Full Text Available Currently a lot of power plants have a problem with storage of coal combustion solid by-products (ash and slag. Holding capacity of existing power plants available ash dumps were enlarged and modernized repeatedly. Many plants have two or even three of them. Today new ash dump construction is economically inconvenient due to need to assign new plots of land and their inconveniently big distance from a plant, which increase ash and slag transportation expenses. The goal of our research work is to find promising directions for ash and slag waste mass utilization based on Kuznetsk bituminous coals experimental data on ultimate composition and properties. The experimental research of ash, slag and their mixture samples from ash dumps brought us to conclusion that the most promising direction for these materials application in large quantities is construction industry including road construction. Be-sides, we lined up some other directions for ash, slag, and ash and slag mixture possible application. These directions might not provide mass utilization but they are promising from a point of view of the researched waste properties.

  19. Mixed waste chemical compatibility with packaging components

    International Nuclear Information System (INIS)

    In this paper, a chemical compatibility testing program for packaging of mixed wastes at will be described. We will discuss the choice of four y-radiation doses, four time durations, four temperatures and four waste solutions to simulate the hazardous waste components of mixed wastes for testing materials compatibility of polymers. The selected simulant wastes are (1) an aqueous alkaline mixture of sodium nitrate and sodium nitrite; (2) a chlorinated hydrocarbon mixture; (3) a simulant liquid scintillation fluid; and (4) a mixture of ketones. A selection of 10 polymers with anticipated high resistance to one or more of these types of environments are proposed for testing as potential liner or seal materials. These polymers are butadiene acrylonitrile copolymer, cross-linked polyethylene, epichlorhyarin, ethylene-propylene rubber, fluorocarbon, glass-filled tetrafluoroethylene, high-density poly-ethylene, isobutylene-isoprene copolymer, polypropylene, and styrene-butadiene rubber. We will describe the elements of the testing plan along with a metric for establishing time resistance of the packaging materials to radiation and chemicals

  20. Nuclear-waste-package materials degradation modes and accelerated testing

    International Nuclear Information System (INIS)

    This report reviews the materials degradation modes that may affect the long-term behavior of waste packages for the containment of nuclear waste. It recommends an approach to accelerated testing that can lead to the qualification of waste package materials in specific repository environments in times that are short relative to the time period over which the waste package is expected to provide containment. This report is not a testing plan but rather discusses the direction for research that might be considered in developing plans for accelerated testing of waste package materials and waste forms

  1. Industrial Waste Landfill IV upgrade package

    International Nuclear Information System (INIS)

    The Y-12 Plant, K-25 Site, and ORNL are managed by DOE's Operating Contractor (OC), Martin Marietta Energy Systems, Inc. (Energy Systems) for DOE. Operation associated with the facilities by the Operating Contractor and subcontractors, DOE contractors and the DOE Federal Building result in the generation of industrial solid wastes as well as construction/demolition wastes. Due to the waste streams mentioned, the Y-12 Industrial Waste Landfill IV (IWLF-IV) was developed for the disposal of solid industrial waste in accordance to Rule 1200-1-7, Regulations Governing Solid Waste Processing and Disposal in Tennessee. This revised operating document is a part of a request for modification to the existing Y-12 IWLF-IV to comply with revised regulation (Rule Chapters 1200-1-7-.01 through 1200-1-7-.08) in order to provide future disposal space for the ORR, Subcontractors, and the DOE Federal Building. This revised operating manual also reflects approved modifications that have been made over the years since the original landfill permit approval. The drawings referred to in this manual are included in Drawings section of the package. IWLF-IV is a Tennessee Department of Environmental and Conservation/Division of Solid Waste Management (TDEC/DSWM) Class 11 disposal unit

  2. Industrial Waste Landfill IV upgrade package

    Energy Technology Data Exchange (ETDEWEB)

    1994-03-29

    The Y-12 Plant, K-25 Site, and ORNL are managed by DOE`s Operating Contractor (OC), Martin Marietta Energy Systems, Inc. (Energy Systems) for DOE. Operation associated with the facilities by the Operating Contractor and subcontractors, DOE contractors and the DOE Federal Building result in the generation of industrial solid wastes as well as construction/demolition wastes. Due to the waste streams mentioned, the Y-12 Industrial Waste Landfill IV (IWLF-IV) was developed for the disposal of solid industrial waste in accordance to Rule 1200-1-7, Regulations Governing Solid Waste Processing and Disposal in Tennessee. This revised operating document is a part of a request for modification to the existing Y-12 IWLF-IV to comply with revised regulation (Rule Chapters 1200-1-7-.01 through 1200-1-7-.08) in order to provide future disposal space for the ORR, Subcontractors, and the DOE Federal Building. This revised operating manual also reflects approved modifications that have been made over the years since the original landfill permit approval. The drawings referred to in this manual are included in Drawings section of the package. IWLF-IV is a Tennessee Department of Environmental and Conservation/Division of Solid Waste Management (TDEC/DSWM) Class 11 disposal unit.

  3. Large packages for reactor decommissioning waste

    International Nuclear Information System (INIS)

    This study was carried out jointly by the Atomic Energy Establishment at Winfrith (now called the Winfrith Technology Centre), Windscale Laboratory and Ove Arup and Partners. The work involved the investigation of the design of large transport containers for intermediate level reactor decommissioning waste, ie waste which requires shielding, and is aimed at European requirements (ie for both LWR and gas cooled reactors). It proposes a design methodology for such containers covering the whole lifetime of a waste disposal package. The design methodology presented takes account of various relevant constraints. Both large self shielded and returnable shielded concepts were developed. The work was generic, rather than specific; the results obtained, and the lessons learned, remain to be applied in practice

  4. Qualification test of packages for transporting radioactive materials and wastes

    International Nuclear Information System (INIS)

    Since 1979 the Waste Treatment Division of Nuclear Tecnology Development Center has been developed and tested packagings for transporting radioactive materials and wastes. The Division has designed facilities for testing Type A packages in accordance with the adopted regulations. The Division has tested several packages for universities, research centers, industries, INB, FURNAS, etc. (author)

  5. Waste to Wealth; The Utilization of Scrap Tyre as Aggregate in Bituminous Mixes for Road Construction

    Directory of Open Access Journals (Sweden)

    Oba, A. L.,

    2015-11-01

    Full Text Available The problem associated with solid waste management is on the increase both in the industries, urban cities and in the rural areas. In the United States of America, Asia and Europe, there are over hundreds of waste to wealth combustion plants from where solid wastes are incinerated. In Nigeria, amidst the increasing importation of vehicle tyre such plants are scarcely in existence to enhance generation of revenue from waste through the extraction of raw material for the production of light weight aggregates, printing ink, paints, shoe polish, dry cell and battery heads. This research paper seeks to utilize vehicle scrap tyre (VST as aggregates in asphaltic mixture by adopting the dry process to evaluate the effect of rubber-bitumen interaction on asphaltic concrete properties; laboratory investigation using 4.75mm, 2.36mm and 0.600mm chunk tyre particle size modified asphalt mixture containing 2%, 4%, 6%, 8% and 10% scrap tyre and 0% tyre content as control mixture. The mixtures were subjected to Marshall Tests where the stability, flow, percentage air void, unit weight, void mineral aggregate, height of specimen and specific gravity were determined. The results obtained shows that as tyre percentage increase the stability, unit weight and specific gravity value decreases. On the other hand, as the tyre content increases, the flow and height of specimen increases while as the tyre content increases the percentage air void and VMA increased for 4.75mm Tyre Particle Size (TPS and 2.36mm TPS while for 0.600mm TPS, reverse is the case. In summary and in comparism with standard specification for road construction material, the Marshall tests conducted on the tyre modified specimens remained intact and by interpretation; material possessing such property indicates good impact resistance when use as surface course in flexible pavement. Conclusively, the use of 10% 4.75mm, 4% 2.36mm or 4% 0.600mm TPS by weight of aggregate in asphaltic concrete is

  6. Reasons for household food waste with special attention to packaging

    OpenAIRE

    Williams, Helén; Wikström, Fredrik; Otterbring, Tobias; Martin LÖFGREN; Gustafsson, Anders

    2012-01-01

    The amount of food waste needs to be reduced in order to sustain the world’s limited resources and secure enough food to all humans. Packaging plays an important role in reducing food waste. The knowledge about how packaging affects food waste in households, however, is scarce. This exploratory study examines reasons for food waste in household and especially how and to what extent packaging influences the amount of food waste. Sixty-one families measured their amount of food waste during sev...

  7. Survey on product properties of bituminized waste concentrates from reprocessing, nuclear installations and nuclear power plants

    International Nuclear Information System (INIS)

    This report contains a brief survey on the product properties after solidification with bitumen of waste concentrates from reprocessing plants, nuclear research installations and nuclear power plants. The leach behaviour of bitumen products in water and saturated salt solutions is described as well as the radiolytic and thermal stability of the products. The reports also contain some data on sedimentation of salts fixed in bitumen. (orig.)

  8. Waste forms, packages, and seals working group summary

    Energy Technology Data Exchange (ETDEWEB)

    Sridhar, N. [Center Antonio, TX (United States); McNeil, M.B. [Nuclear Regulatory Commission, Washington, DC (United States)

    1995-09-01

    This article is a summary of the proceedings of a group discussion which took place at the Workshop on the Role of Natural Analogs in Geologic Disposal of High-Level Nuclear Waste in San Antonio, Texas on July 22-25, 1991. The working group concentrated on the subject of radioactive waste forms and packaging. Also included is a description of the use of natural analogs in waste packaging, container materials and waste forms.

  9. Waste package/repository impact study: Final report

    Energy Technology Data Exchange (ETDEWEB)

    1985-09-01

    The Waste Package/Repository Impact Study was conducted to evaluate the feasibility of using the current reference salt waste package in the salt repository conceptual design. All elements of the repository that may impact waste package parameters, i.e., (size, weight, heat load) were evaluated. The repository elements considered included waste hoist feasibility, transporter and emplacement machine feasibility, subsurface entry dimensions, feasibility of emplacement configuration, and temperature limits. The evaluations are discussed in detail with supplemental technical data included in Appendices to this report, as appropriate. Results and conclusions of the evaluations are discussed in light of the acceptability of the current reference waste package as the basis for salt conceptual design. Finally, recommendations are made relative to the salt project position on the application of the reference waste package as a basis for future design activities. 31 refs., 11 figs., 11 tabs.

  10. Waste package/repository impact study: Final report

    International Nuclear Information System (INIS)

    The Waste Package/Repository Impact Study was conducted to evaluate the feasibility of using the current reference salt waste package in the salt repository conceptual design. All elements of the repository that may impact waste package parameters, i.e., (size, weight, heat load) were evaluated. The repository elements considered included waste hoist feasibility, transporter and emplacement machine feasibility, subsurface entry dimensions, feasibility of emplacement configuration, and temperature limits. The evaluations are discussed in detail with supplemental technical data included in Appendices to this report, as appropriate. Results and conclusions of the evaluations are discussed in light of the acceptability of the current reference waste package as the basis for salt conceptual design. Finally, recommendations are made relative to the salt project position on the application of the reference waste package as a basis for future design activities. 31 refs., 11 figs., 11 tabs

  11. Release rates from waste packages in a salt repository

    International Nuclear Information System (INIS)

    In this report we present estimates of radionuclide release rates from waste packages into salt. This conservative and bounding analysis shows that release rates from waste packages in salt are well below the US Nuclear Regulatory Commission's performance objectives for the engineered barrier system. 2 refs., 2 figs

  12. Initial hydrothermal waste package release experiments using spent fuel with waste package components

    International Nuclear Information System (INIS)

    The Waste/Barrier/Rock interactions program of the Basalt Waste Isolation Project (BWIP) is conducting hydrothermal experiments with fully radioactive waste forms and the other components comprising a designed waste package for a basalt repository. These additional materials include container material (steel) and packing material (crushed basalt and bentoonite). The initial 11 experiments with spent reactor fuel have been completed, showing that, after six months, the fuel begins to react only slightly with water at 2000C. Solid reaction products identified thus far consist of uranylsilicate in spent fuel-water experiments, an iron hydroxide in steel-spent fuel experiments, and a smectite clay in basalt-bearing experiments. Solution samples taken during the experiments show that concentrations of many radionuclides, including Cs, Sr, and the actinides, are lower with basalt present than without. With or without basalt, these key species occur in solution at concentrations well below those required to meet federal release criteria

  13. Mechanical Assessment of the Waste Package Subject to Vibratory Motion

    Energy Technology Data Exchange (ETDEWEB)

    M. Gross

    2004-10-14

    The purpose of this document is to provide an integrated overview of the calculation reports that define the response of the waste package and its internals to vibratory ground motion. The calculation reports for waste package response to vibratory ground motion are identified in Table 1-1. Three key calculation reports describe the potential for mechanical damage to the waste package, fuel assemblies, and cladding from a seismic event. Three supporting documents have also been published to investigate sensitivity of damage to various assumptions for the calculations. While these individual reports present information on a specific aspect of waste package and cladding response, they do not describe the interrelationship between the various calculations and the relationship of this information to the seismic scenario class for Total System Performance Assessment-License Application (TSPA-LA). This report is designed to fill this gap by providing an overview of the waste package structural response calculations.

  14. Challenges in packaging waste management in the fast food industry

    Energy Technology Data Exchange (ETDEWEB)

    Aarnio, Teija [Digita Oy, P.O. Box 135, FI-00521 Helsinki (Finland); Haemaelaeinen, Anne [Department of Energy and Environmental Technology, Lappeenranta University of Technology, P.O. Box 20, FI-53851 Lappeenranta (Finland)

    2008-02-15

    The recovery of solid waste is required by waste legislation, and also by the public. In some industries, however, waste is mostly disposed of in landfills despite of its high recoverability. Practical experiences show that the fast food industry is one example of these industries. A majority of the solid waste generated in the fast food industry is packaging waste, which is highly recoverable. The main research problem of this study was to find out the means of promoting the recovery of packaging waste generated in the fast food industry. Additionally, the goal of this article was to widen academic understanding on packaging waste management in the fast food industry, as the subject has not gained large academic interest previously. The study showed that the theoretical recovery rate of packaging waste in the fast food industry is high, 93% of the total annual amount, while the actual recovery rate is only 29% of the total annual amount. The total recovery potential of packaging waste is 64% of the total annual amount. The achievable recovery potential, 33% of the total annual amount, could be recovered, but is not mainly because of non-working waste management practices. The theoretical recovery potential of 31% of the total annual amount of packaging waste cannot be recovered by the existing solid waste infrastructure because of the obscure status of commercial waste, the improper operation of producer organisations, and the municipal autonomy. The research indicated that it is possible to reach the achievable recovery potential in the existing solid waste infrastructure through new waste management practices, which are designed and operated according to waste producers' needs and demands. The theoretical recovery potential can be reached by increasing the consistency of the solid waste infrastructure through governmental action. (author)

  15. General Corrosion and Localized Corrosion of Waste Package Outer Barrier

    Energy Technology Data Exchange (ETDEWEB)

    K.G. Mon

    2004-10-01

    The waste package design for the License Application is a double-wall waste package underneath a protective drip shield (BSC 2004 [DIRS 168489]; BSC 2004 [DIRS 169480]). The purpose and scope of this model report is to document models for general and localized corrosion of the waste package outer barrier (WPOB) to be used in evaluating waste package performance. The WPOB is constructed of Alloy 22 (UNS N06022), a highly corrosion-resistant nickel-based alloy. The inner vessel of the waste package is constructed of Stainless Steel Type 316 (UNS S31600). Before it fails, the Alloy 22 WPOB protects the Stainless Steel Type 316 inner vessel from exposure to the external environment and any significant degradation. The Stainless Steel Type 316 inner vessel provides structural stability to the thinner Alloy 22 WPOB. Although the waste package inner vessel would also provide some performance for waste containment and potentially decrease the rate of radionuclide transport after WPOB breach before it fails, the potential performance of the inner vessel is far less than that of the more corrosion-resistant Alloy 22 WPOB. For this reason, the corrosion performance of the waste package inner vessel is conservatively ignored in this report and the total system performance assessment for the license application (TSPA-LA). Treatment of seismic and igneous events and their consequences on waste package outer barrier performance are not specifically discussed in this report, although the general and localized corrosion models developed in this report are suitable for use in these scenarios. The localized corrosion processes considered in this report are pitting corrosion and crevice corrosion. Stress corrosion cracking is discussed in ''Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural Material'' (BSC 2004 [DIRS 169985]).

  16. The behaviour of radioactive waste packages under fire accident conditions

    International Nuclear Information System (INIS)

    An experimental study has been made of the behaviour of packaged Intermediate Level Wastes (ILW) subjected to heat. The conditions used represented fire accidents in the transport of the ILW to the repository in shielded transport containers and in the handling of the packages at the repository. The behaviour of four waste materials immobilised in cement and organic resin were studied. Each waste used had features which allowed the results to be applied to a wide range of other waste streams. Samples of these materials have been heated under controlled and well instrumented conditions in furnaces and pool fires. Inactive simulant wastes were used in small and full scale experiments. Fully active waste materials were used in small scale experiments only. Data are presented on the temperature profiles through the packaged ILW and on the release of volatile and particulate materials as a function of time and temperature. (orig.)

  17. Packaging and transport of low and intermediate level radioactive waste

    International Nuclear Information System (INIS)

    The paper presents an overview of Nirex proposals for the packaging and transport of low and intermediate-level radioactive waste, as well as the regulatory requirements which must be met in such operations. (author)

  18. chemo-Hydro-mechanical modelling of in-situ disposal of a bituminized radioactive waste in boom clay

    International Nuclear Information System (INIS)

    Document available in extended abstract form only. The current reference solution of the Belgian Agency for the Management of Radioactive Waste and Fissile Materials (ONDRAF/NIRAS) envisages underground disposal of Eurobitum Bituminized radioactive Waste (BW) in a geologically stable clay formation. In Belgium, the Boom Clay, which is a 30 to 35 million years old and ∼100 m thick marine sediment is being studied as a potential host formation because of its favorable properties to limit and delay the migration of the leached radionuclides to the biosphere over extended periods of time. The current disposal concept foresees that several drums (220 litres) of Eurobitum would be grouped in thick-walled cement-based secondary containers, which in turn would be placed in concrete-lined disposal galleries that are excavated at mid-depth in the clay layer. Only 80-90 % of the total volume of the drum is filled with Eurobitum.The remaining voids between the containers would be backfilled with a cement-based material. The interaction between the BW and the host clay formation is a very complicated chemo-hydro-mechanical process and depends not only on the hydro-mechanical behaviour of the Boom Clay itself, but also on that of the BW. In fact, the osmosis-induced uptake of water by the dehydrated hygroscopic salts embedded in the waste induces a geo-mechanical perturbation of the host formation, caused by the swelling and the increase of the pressure in and around the waste. The objectives of the Chemo-Hydro-Chemical (CHM) analysis presented in this work are (i) to get insights on the kinetics of water uptake by BW, dissolution of the embedded NaNO3 crystals, solute leaching, and maximum generated pressure under disposal conditions and (ii) to study the stress redistribution due to the recompression of the clay around a gallery caused by the swelling pressure of the bitumen and the admissible swelling pressure for Boom clay. Firstly, a CHM formulation of chemically and

  19. 10 CFR 60.135 - Criteria for the waste package and its components.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Criteria for the waste package and its components. 60.135... for the waste package and its components. (a) High-level-waste package design in general. (1) Packages... package and its interactions with the emplacement environment do not compromise the function of the...

  20. STUDY ON PACKAGING WASTE PREVENTION IN ROMANIA

    Directory of Open Access Journals (Sweden)

    Scortar Lucia-Monica

    2013-07-01

    It is very important to mention that individuals and businesses can often save a significant amount of money through waste prevention: waste that never gets created doesn't have management costs (handling, transporting, treating and disposing of waste. The rule is simple: the best waste is that which is not produced.

  1. CH Packaging Operations for High Wattage Waste at LANL

    International Nuclear Information System (INIS)

    This procedure provides instructions for assembling the following contact-handled (CH) packaging payloads: - Drum payload assembly - Standard Waste Box (SWB) assembly - Ten-Drum Overpack (TDOP) In addition, this procedure also provides operating instructions for the TRUPACT-II CH waste packaging. This document also provides instructions for performing ICV and OCV preshipment leakage rate tests on the following packaging seals, using a nondestructive helium (He) leak test: - ICV upper main O-ring seal - ICV outer vent port plug O-ring seal - OCV upper main O-ring seal - OCV vent port plug O-ring seal

  2. CH Packaging Operations for High Wattage Waste at LANL

    International Nuclear Information System (INIS)

    This procedure provides instructions for assembling the following contact-handled (CH) packaging payloads: - Drum payload assembly - Standard Waste Box (SWB) assembly - Ten-Drum Overpack (TDOP) In addition, this procedure provides operating instructions for the TRUPACT-II CH waste packaging. This document also provides instructions for performing ICV and OCV preshipment leakage rate tests on the following packaging seals, using a nondestructive helium (He) leak test: - ICV upper main O-ring seal - ICV outer vent port plug O-ring seal - OCV upper main O-ring seal - OCV vent port plug O-ring seal

  3. Waste package for a repository located in salt

    International Nuclear Information System (INIS)

    This paper describes the current status of the waste package designs for salt repositories. The status of the supporting studies of environment definition, corrosion of containment materials, and leaching of waste forms is also presented. Emphasis is on the results obtained in FY 83 and the planned effort in FY 84. 8 references, 3 figures, 1 table

  4. Conceptual waste packaging options for deep borehole disposal

    Energy Technology Data Exchange (ETDEWEB)

    Su, Jiann -Cherng [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Hardin, Ernest L. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

    2015-07-01

    This report presents four concepts for packaging of radioactive waste for disposal in deep boreholes. Two of these are reference-size packages (11 inch outer diameter) and two are smaller (5 inch) for disposal of Cs/Sr capsules. All four have an assumed length of approximately 18.5 feet, which allows the internal length of the waste volume to be 16.4 feet. However, package length and volume can be scaled by changing the length of the middle, tubular section. The materials proposed for use are low-alloy steels, commonly used in the oil-and-gas industry. Threaded connections between packages, and internal threads used to seal the waste cavity, are common oilfield types. Two types of fill ports are proposed: flask-type and internal-flush. All four package design concepts would withstand hydrostatic pressure of 9,600 psi, with factor safety 2.0. The combined loading condition includes axial tension and compression from the weight of a string or stack of packages in the disposal borehole, either during lower and emplacement of a string, or after stacking of multiple packages emplaced singly. Combined loading also includes bending that may occur during emplacement, particularly for a string of packages threaded together. Flask-type packages would be fabricated and heat-treated, if necessary, before loading waste. The fill port would be narrower than the waste cavity inner diameter, so the flask type is suitable for directly loading bulk granular waste, or loading slim waste canisters (e.g., containing Cs/Sr capsules) that fit through the port. The fill port would be sealed with a tapered, threaded plug, with a welded cover plate (welded after loading). Threaded connections between packages and between packages and a drill string, would be standard drill pipe threads. The internal flush packaging concepts would use semi-flush oilfield tubing, which is internally flush but has a slight external upset at the joints. This type of tubing can be obtained with premium, low

  5. 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

  6. Effects of mixed waste simulants on transportation packaging plastic components

    International Nuclear Information System (INIS)

    The purpose of hazardous and radioactive materials packaging is to, enable these materials to be transported without posing a threat to the health or property of the general public. To achieve this aim, regulations have been written establishing general design requirements for such packagings. While no regulations have been written specifically for mixed waste packaging, regulations for the constituents of mixed wastes, i.e., hazardous and radioactive substances, have been codified. The design requirements for both hazardous and radioactive materials packaging specify packaging compatibility, i.e., that the materials of the packaging and any contents be chemically compatible with each other. Furthermore, Type A and Type B packaging design requirements stipulate that there be no significant chemical, galvanic, or other reaction between the materials and contents of the package. Based on these requirements, a Chemical Compatibility Testing Program was developed in the Transportation Systems Department at Sandia National Laboratories (SNL). The program, supported by the US Department of Energy's (DOE) Transportation Management Division, EM-261 provides the means to assure any regulatory body that the issue of packaging material compatibility towards hazardous and radioactive materials has been addressed. In this paper, we describe the general elements of the testing program and the experimental results of the screening tests. The implications of the results of this testing are discussed in the general context of packaging development. Additionally, we present the results of the first phase of this experimental program. This phase involved the screening of five candidate liner and six seal materials against four simulant mixed wastes

  7. Review of DOE waste package program. Subtask 1.1. National waste package program, April-September 1983. Volume 5

    International Nuclear Information System (INIS)

    The current effort is part of an ongoing task to review the national high-level waste package effort. It includes evaluations of reference waste form, container, and packing material components with respect to determining how they may contribute to the containment and controlled release of radionuclides after waste packages have been emplaced in salt, basalt, and tuff repositories. In the current Biannual Report a section on carbon steel container corrosion has been included to complement prior work on TiCode-12 and Type 304 stainless steel. The use of crushed tuff as a packing material is discussed and waste package component interaction test data are included. Licensing data requirements to estimate the degree of compliance with NRC performance objectives are specified. 41 figures, 24 tables

  8. Review of DOE waste package program. Subtask 1.1 - National Waste Package Program, October 1983-March 1984. Volume 6

    International Nuclear Information System (INIS)

    The present effort is part of an ongoing task to review the national high-level waste package effort. It includes evaluation of reference waste form, container, and packing material components with respect to determining how they may contribute to the containment and controlled release of radionuclides after waste packages have been emplaced in salt, basalt, tuff, and granite repositories. In the current Biannual Report a review of progress in the new crystalline repository (granite) program is described. Other foreign data for this host rock have also been outlined where relevant. The use of crushed salt, and bentonite- and zeolite-containing packing materials is discussed. The effects of temperature and gamma irradiation are shown to be important with respect to defining the localized environmental conditions around a waste package and the long-term integrity of the packing

  9. Waste Package and Material Testing for the Proposed Yucca Mountain High Level Waste Repository

    International Nuclear Information System (INIS)

    Over the repository lifetime, the waste package containment barriers will perform various functions that will change with time. During the operational period, the barriers will function as vessels for handling, emplacement, and waste retrieval (if necessary). During the years following repository closure, the containment barriers will be relied upon to provide substantially complete containment, through 10,000 years and beyond. Following the substantially complete containment phase, the barriers and the waste package internal structures help minimize release of radionuclides by aqueous- and gaseous-phase transport. These requirements have lead to a defense-in-depth design philosophy. A multi-barrier design will result in a lower breach rate distributed over a longer period of time, thereby ensuring the regulatory requirements are met. The design of the Engineered Barrier System (EBS) has evolved. The initial waste package design was a thin walled package, 3/8 inch of stainless steel 304, that had very limited capacity, (3 PWR and 4 BWR assemblies) and performance characteristics, 300 to 1,000 years. This design required over 35,000 waste packages compared to today's design of just over 10,000 waste packages. The waste package designs are now based on a defense-in-depth/multi-barrier philosophy and have a capacity similar to the standard storage and rail transported spent nuclear fuel casks. Concurrent with the development of the design of the waste packages, a comprehensive waste package materials testing program has been undertaken to support the selection of containment barrier materials and to develop predictive models for the long-term behavior of these materials under expected repository conditions. The testing program includes both long-term and short-term tests and the results from these tests combination with the data published in the open literature are being used to develop models for predicting performance of the waste packages

  10. Salt Repository Project Waste Package Program Plan: Draft

    International Nuclear Information System (INIS)

    Under the direction of the Office of Civilian Radioactive Waste Management (OCRWM) created within the DOE by direction of the Nuclear Waste Policy Act of 1982 (NWPA), the mission of the Salt Repository Project (SRP) is to provide for the development of a candidate salt repository for disposal of high-level radioactive waste (HLW) and spent reactor fuel in a manner that fully protects the health and safety of the public and the quality of the environment. In consideration of the program needs and requirements discussed above, the SRP has decided to develop and issue this SRP Waste Package Program Plan. This document is intended to outline how the SRP plans to develop the waste package design and to show, with reasonable assurance, that the developed design will satisfy applicable requirements/performance objectives. 44 refs., 16 figs., 16 tabs

  11. Packaging waste recycling in Europe: is the industry paying for it?

    OpenAIRE

    Nuno F. da Cruz; Ferreira, Sandra; Cabral, Marta; Simões, Pedro; Marques, Rui Cunha

    2014-01-01

    This paper describes and examines the schemes established in five EU countries for the recycling of packaging waste. The changes in packaging waste management were mainly implemented since the Directive 94/62/EC on packaging and packaging waste entered into force. The analysis of the five systems allowed the authors to identify very different approaches to cope with the same problem: meet the recovery and recycling targets imposed by EU law. Packaging waste is a responsibility of the industry...

  12. 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

  13. TRU waste form and package criteria meeting

    Energy Technology Data Exchange (ETDEWEB)

    None

    1977-08-01

    The broad subject of the meeting is the overall ERDA TRU waste management program, although the discussions also cover performance criteria for the Waste Isolation Pilot Plant and their implications for the overall TRU program. Separate abstracts were prepared for all ten presentations. (DLC)

  14. Evaluation and compilation of DOE waste package test data

    International Nuclear Information System (INIS)

    This report summarizes evaluations by the National Institute of Standards and Technology (NIST) of Department of Energy (DOE) activities on waste packages designed for containment of radioactive high-level nuclear waste (HLW) for the six-month period August 1988 through January 1989. Included are reviews of related materials research and plans, activities for the DOE Materials Characterization Center, information on the Yucca Mountain Project, and other information regarding supporting research and special assistance. NIST comments are given on the Yucca Mountain Consultation Draft Site Characterization Plan (CDSCP) and on the Waste Compliance Plan for the West Valley Demonstration Project (WVDP) High-Level Waste (HLW) Form. 3 figs

  15. Waste package reference conceptual designs for a repository in salt

    International Nuclear Information System (INIS)

    This report provides the reference conceptual waste package designs for the Office of Nuclear Waste Isolation to baseline these designs, thereby establishing the configuration and interface controls necessary, within the Civilian Radioactive Waste Management Program, formerly the National Waste Terminal Storage Program, to proceed in an orderly manner with preliminary design. Included are designs for the current reference defense high-level waste form from the Savannah River Plant, an optimized commercial high-level waste form, and spent fuel which has been disassembled and compacted into a circular bundle containing either 12 pressurized-water reactor or 30 boiling-water reactor assemblies. For compacted spent fuel, it appears economically attractive to standardize the waste package diameter for all fuel types. The reference waste packages consist of the containerized waste form, a low carbon steel overpack, and, after emplacement, a cover of salt. The overpack is a hollow cylinder with a flat head welded to each end. Its design thickness is the sum of the structural thickness required to resist the 15.4-MPa lithostatic pressure plus the corrosion allowance necessary to assure the required structural thickness will exist through the 1000-year containment period. Based on available data and completed analyses, the reference concepts described in this report satisfy all requirements of the US Department of Energy and the US Nuclear Regulatory Commission with reasonable assurance. In addition, sufficient design maturity exists to form a basis for preliminary design; these concepts can be brought under configuration control to serve as reference package designs. Development programs are identified that will be required to support these designs during the licensing process. 19 refs., 37 figs., 31 tabs

  16. Nuclear waste package materials testing report: basaltic and tuffaceous environments

    Energy Technology Data Exchange (ETDEWEB)

    Bradley, D.J.; Coles, D.G.; Hodges, F.N.; McVay, G.L.; Westerman, R.E.

    1983-03-01

    The disposal of high-level nuclear wastes in underground repositories in the continental United States requires the development of a waste package that will contain radionuclides for a time period commensurate with performance criteria, which may be up to 1000 years. This report addresses materials testing in support of a waste package for a basalt (Hanford, Washington) or a tuff (Nevada Test Site) repository. The materials investigated in this testing effort were: sodium and calcium bentonites and mixtures with sand or basalt as a backfill; iron and titanium-based alloys as structural barriers; and borosilicate waste glass PNL 76-68 as a waste form. The testing also incorporated site-specific rock media and ground waters: Reference Umtanum Entablature-1 basalt and reference basalt ground water, Bullfrog tuff and NTS J-13 well water. The results of the testing are discussed in four major categories: Backfill Materials: emphasizing water migration, radionuclide migration, physical property and long-term stability studies. Structural Barriers: emphasizing uniform corrosion, irradiation-corrosion, and environmental-mechanical testing. Waste Form Release Characteristics: emphasizing ground water, sample surface area/solution volume ratio, and gamma radiolysis effects. Component Compatibility: emphasizing solution/rock, glass/rock, glass/structural barrier, and glass/backfill interaction tests. This area also includes sensitivity testing to determine primary parameters to be studied, and the results of systems tests where more than two waste package components were combined during a single test.

  17. Nuclear waste package materials testing report: basaltic and tuffaceous environments

    International Nuclear Information System (INIS)

    The disposal of high-level nuclear wastes in underground repositories in the continental United States requires the development of a waste package that will contain radionuclides for a time period commensurate with performance criteria, which may be up to 1000 years. This report addresses materials testing in support of a waste package for a basalt (Hanford, Washington) or a tuff (Nevada Test Site) repository. The materials investigated in this testing effort were: sodium and calcium bentonites and mixtures with sand or basalt as a backfill; iron and titanium-based alloys as structural barriers; and borosilicate waste glass PNL 76-68 as a waste form. The testing also incorporated site-specific rock media and ground waters: Reference Umtanum Entablature-1 basalt and reference basalt ground water, Bullfrog tuff and NTS J-13 well water. The results of the testing are discussed in four major categories: Backfill Materials: emphasizing water migration, radionuclide migration, physical property and long-term stability studies. Structural Barriers: emphasizing uniform corrosion, irradiation-corrosion, and environmental-mechanical testing. Waste Form Release Characteristics: emphasizing ground water, sample surface area/solution volume ratio, and gamma radiolysis effects. Component Compatibility: emphasizing solution/rock, glass/rock, glass/structural barrier, and glass/backfill interaction tests. This area also includes sensitivity testing to determine primary parameters to be studied, and the results of systems tests where more than two waste package components were combined during a single test

  18. Waste-package release rates for site suitability studies

    International Nuclear Information System (INIS)

    Performance-assessment calculations in support of the site- suitability effort for the Yucca Mountain Project will address radionuclide transport arising from various disruptive scenarios. Here we present release rates of radionuclides from individual waste packages for scenarios involving various postulated forms of water intrusion, including increased infiltration rate as well as rock immediately surrounding an individual waste package becoming saturated with ground water. We examine: (1) effect of increased water infiltration rate on release rates; increases in radionuclide release rates resulting from water filling the annulus between the waste container and the surrounding rock, as well as water saturating the pores and fractures in the rock surrounding the waste package; (3) the effect of flow in fractures in the saturated rock on release rate; and (4) release of radionuclides to the mountain surface resulting from an exploratory borehole shaft intersecting a waste package. The radionuclides considered are Tc-99; I-129; Cs-135; Np- 237; Pu-239,240,242; and Am-241,243. Release rates are calculated for both the wet-drip bathtub and the wet-continuous water-contact modes, as described in the Working Group 2 report, applying equations as published by Sadeghi, et al., [1990] and as extended in the present report

  19. WASTE PACKAGE OPERATIONS FY-99 CLOSURE METHODS REPORT

    International Nuclear Information System (INIS)

    The waste package (WP) closure weld development task is part of a larger engineering development program to develop waste package designs. The purpose of the larger waste package engineering development program is to develop nuclear waste package fabrication and closure methods that the Nuclear Regulatory Commission will find acceptable and will license for disposal of spent nuclear fuel (SNF), non-fuel components, and vitrified high-level waste within a Monitored Geologic Repository (MGR). Within the WP closure development program are several major development tasks, which, in turn, are divided into subtasks. The major tasks include: WP fabrication development, WP closure weld development, nondestructive examination (NDE) development, and remote in-service inspection development. The purpose of this report is to present the objectives, technical information, and work scope relating to the WP closure weld development.and NDE tasks and subtasks and to report results of the closure weld and NDE development programs for fiscal year 1999 (FY-99). The objective of the FY-99 WP closure weld development task was to develop requirements for closure weld surface and volumetric NDE performance demonstrations, investigate alternative NDE inspection techniques, and develop specifications for welding, NDE, and handling system integration. In addition, objectives included fabricating several flat plate mock-ups that could be used for NDE development, stress relief peening, corrosion testing, and residual stress testing

  20. A comprehensive waste collection cost model applied to post-consumer plastic packaging waste

    NARCIS (Netherlands)

    Groot, J.J.; Bing, X.; Bos-Brouwers, H.E.J.; Bloemhof, J.M.

    2014-01-01

    Post-consumer plastic packaging waste (PPW) can be collected for recycling via source separation or post-separation. In source separation, households separate plastics from other waste before collection, whereas in post-separation waste is separated at a treatment centre after collection. There are

  1. 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

  2. DESIGN ANALYSIS FOR THE NAVAL SNF WASTE PACKAGE

    Energy Technology Data Exchange (ETDEWEB)

    T.L. Mitchell

    2000-05-31

    The purpose of this analysis is to demonstrate the design of the naval spent nuclear fuel (SNF) waste package (WP) using the Waste Package Department's (WPD) design methodologies and processes described in the ''Waste Package Design Methodology Report'' (CRWMS M&O [Civilian Radioactive Waste Management System Management and Operating Contractor] 2000b). The calculations that support the design of the naval SNF WP will be discussed; however, only a sub-set of such analyses will be presented and shall be limited to those identified in the ''Waste Package Design Sensitivity Report'' (CRWMS M&O 2000c). The objective of this analysis is to describe the naval SNF WP design method and to show that the design of the naval SNF WP complies with the ''Naval Spent Nuclear Fuel Disposal Container System Description Document'' (CRWMS M&O 1999a) and Interface Control Document (ICD) criteria for Site Recommendation. Additional criteria for the design of the naval SNF WP have been outlined in Section 6.2 of the ''Waste Package Design Sensitivity Report'' (CRWMS M&O 2000c). The scope of this analysis is restricted to the design of the naval long WP containing one naval long SNF canister. This WP is representative of the WPs that will contain both naval short SNF and naval long SNF canisters. The following items are included in the scope of this analysis: (1) Providing a general description of the applicable design criteria; (2) Describing the design methodology to be used; (3) Presenting the design of the naval SNF waste package; and (4) Showing compliance with all applicable design criteria. The intended use of this analysis is to support Site Recommendation reports and assist in the development of WPD drawings. Activities described in this analysis were conducted in accordance with the technical product development plan (TPDP) ''Design Analysis for the Naval SNF Waste Package (CRWMS M

  3. Management of radioactive waste packaging drum

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Il-Sik; Shon, Kang J. S.; Lee, Y. H.; Song, I. T.; Lee, B. C.; Kim, K. J. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-02-01

    Radioactive waste drums of dismantled equipment and contaminated soil waste generated from KAERI research reactor, Seoul in 1988 were transferred to KAERI in Taejeon and are stored at low and medium level storage building until now. As a result of relatively long term storage, a corrosion on the surface of the drums was progressed and a remedial action was conducted to take a measure for the enhancement of safety. Thus, we examined the corrosion status on the surface of waste drums and studied mostly applicable methods. Resultingly, the seriously corroded drums were repackaged and the slightly corroded drums were painted after removing a corroded part on the surface. 12 refs., 14 figs., 12 tabs. (Author)

  4. WAPDEG Analysis of Waste Package and Drip shield Degradation

    Energy Technology Data Exchange (ETDEWEB)

    K. Mon

    2004-09-29

    As directed by ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]), an analysis of the degradation of the engineered barrier system (EBS) drip shields and waste packages at the Yucca Mountain repository is developed. The purpose of this activity is to provide the TSPA with inputs and methodologies used to evaluate waste package and drip shield degradation as a function of exposure time under exposure conditions anticipated in the repository. This analysis provides information useful to satisfy ''Yucca Mountain Review Plan, Final Report'' (NRC 2003 [DIRS 163274]) requirements. Several features, events, and processes (FEPs) are also discussed (Section 6.2, Table 15). The previous revision of this report was prepared as a model report in accordance with AP-SIII.10Q, Models. Due to changes in the role of this report since the site recommendation, it no longer contains model development. This revision is prepared as a scientific analysis in accordance with AP-SIII.9Q, ''Scientific Analyses'' and uses models previously validated in (1) ''Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural Material'' (BSC 2004 [DIRS 169985]); (2) ''General Corrosion and Localized Corrosion of Waste Package Outer Barrier'' (BSC 2004 [DIRS 169984]); and (3) ''General Corrosion and Localized Corrosion of Drip Shield'' (BSC 2004 [DIRS 169845]). The integrated waste package degradation (IWPD) analysis presented in this report treats several implementation-related issues, such as defining the number and size of patches per waste package that undergo stress corrosion cracking; recasting the weld flaw analysis in a form as implemented in the Closure Weld Defects (CWD) software; and, general corrosion rate manipulations (e.g., change of

  5. Post emplacement environment of waste packages

    International Nuclear Information System (INIS)

    Experiments have been conducted as part of the Nevada Nuclear Waste Storage Investigations Project to determine the changes in water chemistry due to reaction of the Topopah Spring tuff with natural groundwater at temperatures up to 1500C. The reaction extent has been investigated as a function of rock-to-water ratio, temperature, reaction time, physical state of the samples, and geographic location of the samples within the tuff unit. Results of these experiments will be used to provide information on the water chemistry to be expected if a high-level waste repository were to be constructed in the Topopah Spring tuff. 6 references, 5 figures, 1 table

  6. Tomography of actinides by photofission in bulky radioactive waste packages

    International Nuclear Information System (INIS)

    Quantifying actinides using non-destructive methods, in radioactive waste packages, is a great stake to turn packages towards appropriate storage facility. But the nature of radiations emitted by actinides (alpha radiations) makes the detection of those very difficult for large volume packages characterization. Indeed, the emitted radiation is too weak, either to be detected by emission tomography or to reach required sensitivities. Therefore, it is necessary to turn to an external probing source. Tomography based on detection of delayed neutrons induced by photofission, allows to probe bulky packages. We demonstrate the suitability of this method to an industrial stage. Firstly, we determine and qualify projection matrix which connects measures at reconstructed activity of tomographic picture. Thus, during measurements on a model and a real package, we carry out convincing tomographic reconstructions with real acquisition conditions. More, we prove that it is possible to take all disruptive chemical element into account, for tomographic reconstructions, in order to obtain the best image of activity. So, we propose a finalised tomographic device, integrating a shielding cell, and checking all the activity and distribution activity criterions fixed for acceptance of radioactive waste packages in superficial storage facility. (author)

  7. Modeling of radiation effects on nuclear waste package materials

    International Nuclear Information System (INIS)

    A methodology is developed for the assessment of radiation effects on nuclear waste package materials. An assessment of the current status of understanding with regard to waste package materials and their behavior in radiation environments is presented. The methodology is used to make prediction as to the chemically induced changes in the groundwater surrounding nuclear waste packages in a repository in tuff. The predictions indicate that mechanisms not currently being pursued by the Department of Energy may be a factor in the long-term performance of nuclear waste packages. The methodology embodies a physical model of the effects of radiation on aqueous solutions. Coupled to the physical model is a method for analyzing the complex nature of the physical model using adjoint sensitivity analysis. The sensitivity aid in both the physical understanding of the processes involved as well as aiding in eliminating portions of the model that have no bearing on the desired results. A computer implementation of the methodology is provided. 128 refs

  8. Generic Degraded Configuration Probability Analysis for the Codisposal Waste Package

    International Nuclear Information System (INIS)

    In accordance with the technical work plan, ''Technical Work Plan For: Department of Energy Spent Nuclear Fuel Work Packages'' (CRWMS M and O 2000c), this Analysis/Model Report (AMR) is developed for the purpose of screening out degraded configurations for U.S. Department of Energy (DOE) spent nuclear fuel (SNF) types. It performs the degraded configuration parameter and probability evaluations of the overall methodology specified in the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2000, Section 3) to qualifying configurations. Degradation analyses are performed to assess realizable parameter ranges and physical regimes for configurations. Probability calculations are then performed for configurations characterized by keff in excess of the Critical Limit (CL). The scope of this document is to develop a generic set of screening criteria or models to screen out degraded configurations having potential for exceeding a criticality limit. The developed screening criteria include arguments based on physical/chemical processes and probability calculations and apply to DOE SNF types when codisposed with the high-level waste (HLW) glass inside a waste package. The degradation takes place inside the waste package and is long after repository licensing has expired. The emphasis of this AMR is on degraded configuration screening and the probability analysis is one of the approaches used for screening. The intended use of the model is to apply the developed screening criteria to each DOE SNF type following the completion of the degraded mode criticality analysis internal to the waste package

  9. Secondary Waste Form Down Selection Data Package – Ceramicrete

    Energy Technology Data Exchange (ETDEWEB)

    Cantrell, Kirk J.; Westsik, Joseph H.

    2011-08-31

    As part of high-level waste pretreatment and immobilized low activity waste processing, liquid secondary wastes will be generated that will be transferred to the Effluent Treatment Facility on the Hanford Site for further treatment. These liquid secondary wastes will be converted to stable solid waste forms that will be disposed in the Integrated Disposal Facility. Currently, four waste forms are being considered for stabilization and solidification of the liquid secondary wastes. These waste forms are Cast Stone, Ceramicrete, DuraLith, and Fluidized Bed Steam Reformer. The preferred alternative will be down selected from these four waste forms. Pacific Northwest National Laboratory is developing data packages to support the down selection process. The objective of the data packages is to identify, evaluate, and summarize the existing information on the four waste forms being considered for stabilization and solidification of the liquid secondary wastes. The information included will be based on information available in the open literature and from data obtained from testing currently underway. This data package is for the Ceramicrete waste form. Ceramicrete is a relatively new engineering material developed at Argonne National Laboratory to treat radioactive and hazardous waste streams (e.g., Wagh 2004; Wagh et al. 1999a, 2003; Singh et al. 2000). This cement-like waste form can be used to treat solids, liquids, and sludges by chemical immobilization, microencapsulation, and/or macroencapsulation. The Ceramicrete technology is based on chemical reaction between phosphate anions and metal cations to form a strong, dense, durable, low porosity matrix that immobilizes hazardous and radioactive contaminants as insoluble phosphates and microencapsulates insoluble radioactive components and other constituents that do not form phosphates. Ceramicrete is a type of phosphate-bonded ceramic, which are also known as chemically bonded phosphate ceramics. The Ceramicrete

  10. INITIAL WASTE PACKAGE PROBABILISTIC CRITICALITY ANALYSIS: UNCANISTERED FUEL (TBV)

    Energy Technology Data Exchange (ETDEWEB)

    J.R. Massari

    1995-10-06

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide an assessment of the present waste package design from a criticality risk standpoint, The specific objectives of this initial analysis are to: (1) Establish a process for determining the probability of waste package criticality as a function of time (in terms of a cumulative distribution function, probability distribution function, or expected number of criticalities in a specified time interval) for various waste package concepts; (2) Demonstrate the established process by estimating the probability of criticality as a function of time since emplacement for an intact uncanistered fuel waste package (UCF-WP) configuration; and (3) Identify the dominant sequences leading to waste package criticality for subsequent detailed analysis. The purpose of this analysis is to document and demonstrate the developed process as it has been applied to the UCF-WP. This revision is performed to correct deficiencies in the previous revision and provide further detail on the calculations performed. Due to the current lack of knowledge in a number of areas, every attempt has been made to ensure that the all calculations and assumptions were conservative. This analysis is preliminary in nature, and is intended to be superseded by at least two more versions prior to license application. The information and assumptions used to generate this analysis are unverified and have been globally assigned TBV identifier TBV-059-WPD. Future versions of this analysis will update these results, possibly replacing the global TBV with a small number of TBV's on individual items, with the goal of removing all TBV designations by license application submittal. The final output of this document, the probability of UCF-WP criticality as a function of time, is therefore, also TBV. This document is intended to deal only with the risk of internal criticality with unaltered fuel

  11. Data Packages for the Hanford Immobilized Low-Activity Tank Waste Performance Assessment: 2001 Version

    International Nuclear Information System (INIS)

    Data package supporting the 2001 Immobilized Low-Activity Waste Performance Analysis. Geology, hydrology, geochemistry, facility, waste form, and dosimetry data based on recent investigation are provided. Verification and benchmarking packages for selected software codes are provided

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

    Energy Technology Data Exchange (ETDEWEB)

    Sing-Teniere, Ch.

    1998-02-01

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

  13. Preclosure analysis of conceptual waste package designs for a nuclear waste repository in tuff

    International Nuclear Information System (INIS)

    This report discusses the selection and analysis of conceptual waste package developed by the Nevada Nuclear Waste Storage Investigations (NNWSI) project for possible disposal of high-level nuclear waste at a candidate site at Yucca Mountain, Nevada. The design requirements that the waste package must conform to are listed, as are several desirable design considerations. Illustrations of the reference and alternative designs are shown. Four austenitic stainless steels (316L SS, 321 SS, 304L SS and Incoloy 825 high nickel alloy) have been selected for candidate canister/overpack materials, and 1020 carbon steel has been selected as the reference metal for the borehole liners. A summary of the results of technical and ecnonmic analyses supporting the selection of the conceptual waste package designs is included. Postclosure containment and release rates are not analyzed in this report

  14. Number of Waste Package Hit by Igneous Intrusion

    Energy Technology Data Exchange (ETDEWEB)

    M. Wallace

    2004-10-13

    The purpose of this scientific analysis report is to document calculations of the number of waste packages that could be damaged in a potential future igneous event through a repository at Yucca Mountain. The analyses include disruption from an intrusive igneous event and from an extrusive volcanic event. This analysis supports the evaluation of the potential consequences of future igneous activity as part of the total system performance assessment for the license application (TSPA-LA) for the Yucca Mountain Project (YMP). Igneous activity is a disruptive event that is included in the TSPA-LA analyses. Two igneous activity scenarios are considered: (1) The igneous intrusion groundwater release scenario (also called the igneous intrusion scenario) considers the in situ damage to waste packages or failure of waste packages that occurs if they are engulfed or otherwise affected by magma as a result of an igneous intrusion. (2) The volcanic eruption scenario depicts the direct release of radioactive waste due to an intrusion that intersects the repository followed by a volcanic eruption at the surface. An igneous intrusion is defined as the ascent of a basaltic dike or dike system (i.e., a set or swarm of multiple dikes comprising a single intrusive event) to repository level, where it intersects drifts. Magma that does reach the surface from igneous activity is an eruption (or extrusive activity) (Jackson 1997 [DIRS 109119], pp. 224, 333). The objective of this analysis is to develop a probabilistic measure of the number of waste packages that could be affected by each of the two scenarios.

  15. Oxidation and waste-to-energy output of aluminium waste packaging during incineration: A laboratory study.

    Science.gov (United States)

    López, Félix A; Román, Carlos Pérez; García-Díaz, Irene; Alguacil, Francisco J

    2015-09-01

    This work reports the oxidation behaviour and waste-to-energy output of different semi-rigid and flexible aluminium packagings when incinerated at 850°C in an air atmosphere enriched with 6% oxygen, in the laboratory setting. The physical properties of the different packagings were determined, including their metallic aluminium contents. The ash contents of their combustion products were determined according to standard BS ISO 1171:2010. The net calorific value, the required energy, and the calorific gain associated with each packaging type were determined following standard BS EN 13431:2004. Packagings with an aluminium lamina thickness of >50μm did not fully oxidise. During incineration, the weight-for-weight waste-to-energy output of the packagings with thick aluminium lamina was lower than that of packagings with thin lamina. The calorific gain depended on the degree of oxidation of the metallic aluminium, but was greater than zero for all the packagings studied. Waste aluminium may therefore be said to act as an energy source in municipal solid waste incineration systems.

  16. Microbial Effects on Nuclear Waste Packaging Materials

    International Nuclear Information System (INIS)

    Microorganisms may enhance corrosion of components of planned engineered barriers within the proposed nuclear waste repository at Yucca Mountain (YM). Corrosion could occur either directly, through processes collectively known as Microbiologically Influenced Corrosion (MIC), or indirectly, by adversely affecting the composition of water or brines that come into direct contact with engineered barrier surfaces. Microorganisms of potential concern (bacteria, archea, and fungi) include both those indigenous to Yucca Mountain and those that infiltrate during repository construction and after waste emplacement. Specific aims of the experimental program to evaluate the potential of microorganisms to affect damage to engineered barrier materials include the following: Indirect Effects--(1) Determine the limiting factors to microbial growth and activity presently in the YM environment. (2) Assess these limiting factors to aid in determining the conditions and time during repository evolution when MIC might become operant. (3) Evaluate present bacterial densities, the composition of the YM microbial community, and determining bacterial densities if limiting factors are overcome. During a major portion of the regulatory period, environmental conditions that are presently extant become reestablished. Therefore, these studies ascertain whether biomass is sufficient to cause MIC during this period and provide a baseline for determining the types of bacterial activities that may be expected. (4) Assess biogenic environmental effects, including pH, alterations to nitrate concentration in groundwater, the generation of organic acids, and metal dissolution. These factors have been shown to be those most relevant to corrosion of engineered barriers. Direct Effects--(1) Characterize and quantify microbiological effects on candidate containment materials. These studies were carried out in a number of different approaches, using whole YM microbiological communities, a subset of YM

  17. Microbial Effects on Nuclear Waste Packaging Materials

    Energy Technology Data Exchange (ETDEWEB)

    Horn, J; Martin, S; Carrillo, C; Lian, T

    2005-07-22

    Microorganisms may enhance corrosion of components of planned engineered barriers within the proposed nuclear waste repository at Yucca Mountain (YM). Corrosion could occur either directly, through processes collectively known as Microbiologically Influenced Corrosion (MIC), or indirectly, by adversely affecting the composition of water or brines that come into direct contact with engineered barrier surfaces. Microorganisms of potential concern (bacteria, archea, and fungi) include both those indigenous to Yucca Mountain and those that infiltrate during repository construction and after waste emplacement. Specific aims of the experimental program to evaluate the potential of microorganisms to affect damage to engineered barrier materials include the following: Indirect Effects--(1) Determine the limiting factors to microbial growth and activity presently in the YM environment. (2) Assess these limiting factors to aid in determining the conditions and time during repository evolution when MIC might become operant. (3) Evaluate present bacterial densities, the composition of the YM microbial community, and determining bacterial densities if limiting factors are overcome. During a major portion of the regulatory period, environmental conditions that are presently extant become reestablished. Therefore, these studies ascertain whether biomass is sufficient to cause MIC during this period and provide a baseline for determining the types of bacterial activities that may be expected. (4) Assess biogenic environmental effects, including pH, alterations to nitrate concentration in groundwater, the generation of organic acids, and metal dissolution. These factors have been shown to be those most relevant to corrosion of engineered barriers. Direct Effects--(1) Characterize and quantify microbiological effects on candidate containment materials. These studies were carried out in a number of different approaches, using whole YM microbiological communities, a subset of YM

  18. Evaluation and compilation of DOE waste package test data

    International Nuclear Information System (INIS)

    This report summarizes evaluations by the National Institute of Standards and Technology (NIST) of some of the Department of Energy (DOE) activities on waste packages designed for containment of radioactive high-level nuclear waste (HLW) for the six-month period, August 1989--January 1990. This includes reviews of related materials research and plans, information on the Yucca Mountain, Nevada disposal site activities, and other information regarding supporting research and special assistance. Short discussions are given relating to the publications reviewed and complete reviews and evaluations are included. Reports of other work are included in the Appendices

  19. REPOSITORY LAYOUT SUPPORTING DESIGN FEATURE #13- WASTE PACKAGE SELF SHIELDING

    Energy Technology Data Exchange (ETDEWEB)

    J. Owen

    1999-04-09

    The objective of this analysis is to develop a repository layout, for Feature No. 13, that will accommodate self-shielding waste packages (WP) with an areal mass loading of 25 metric tons of uranium per acre (MTU/acre). The scope of this analysis includes determination of the number of emplacement drifts, amount of emplacement drift excavation required, and a preliminary layout for illustrative purposes.

  20. Generic Degraded Congiguration Probability Analysis for DOE Codisposal Waste Package

    Energy Technology Data Exchange (ETDEWEB)

    S.F.A. Deng; M. Saglam; L.J. Gratton

    2001-05-23

    In accordance with the technical work plan, ''Technical Work Plan For: Department of Energy Spent Nuclear Fuel Work Packages'' (CRWMS M&O 2000c), this Analysis/Model Report (AMR) is developed for the purpose of screening out degraded configurations for U.S. Department of Energy (DOE) spent nuclear fuel (SNF) types. It performs the degraded configuration parameter and probability evaluations of the overall methodology specified in the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2000, Section 3) to qualifying configurations. Degradation analyses are performed to assess realizable parameter ranges and physical regimes for configurations. Probability calculations are then performed for configurations characterized by k{sub eff} in excess of the Critical Limit (CL). The scope of this document is to develop a generic set of screening criteria or models to screen out degraded configurations having potential for exceeding a criticality limit. The developed screening criteria include arguments based on physical/chemical processes and probability calculations and apply to DOE SNF types when codisposed with the high-level waste (HLW) glass inside a waste package. The degradation takes place inside the waste package and is long after repository licensing has expired. The emphasis of this AMR is on degraded configuration screening and the probability analysis is one of the approaches used for screening. The intended use of the model is to apply the developed screening criteria to each DOE SNF type following the completion of the degraded mode criticality analysis internal to the waste package.

  1. Behaviour face to packaging waste and drugs out of use

    OpenAIRE

    Nascimento, Luís; Taboada, Xavier; Cardoso, Marisa; Figueiredo, Laura; Lopes, Ivo; Torres, Rui

    2014-01-01

    According to Directive No. 2004/12/EC of 11 February, up to the present calendar year (2011 ), Portugal should meet established with respect to the recycling of packaging waste and discarded drug targets . For this, it is essential that the population has acquired over the past few years, the necessary information. So, for that it is important the active participation of everyone in this delivery, in places due to the effect . The objectives of this research consisted in knowing what they...

  2. 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.

  3. The Role of Packaging in Solid Waste Management 1966 to 1976.

    Science.gov (United States)

    Darnay, Arsen; Franklin, William E.

    The goals of waste processors and packagers obviously differ: the packaging industry seeks durable container material that will be unimpaired by external factors. Until recently, no systematic analysis of the relationship between packaging and solid waste disposal had been undertaken. This three-part document defines these interactions, and the…

  4. Shielding analysis for industrial package: focusing on dry active waste

    International Nuclear Information System (INIS)

    In this study, maximum exposure rate at DAW(Dry Active Waste) drum surface which is satisfying regulation limit was calculated for conceptual design of IP(Industrial Package). DAW can be classified as combustible and non-combustible waste and the calculation was conducted for single and mixed radionuclide for each type of waste. In case of combustible waste that mixed radionuclide is uniformly distributed, the maximum exposure rates at drum surface were 3.60E-01, 8.85E-01 and 1.27E+01 mSv/hr for IP Type 1, 2-a and 2-b, respectively. and 3.60E-01, 8.85E-01, 1.27E+01 mSv/hr for single radionuclide(Co-60). In case of non-combustible waste that mixed radionuclide is uniformly distributed, the maximum exposure rates at drum surface were 7.14E-01, 1.83E+00, 2.69E+01 mSv/hr for IP Type 1, 2-a and 2-b, respectively. and 7.13E-01, 1.81E-01, 2.62E+01 mSv/hr for single radionuclide(Co-60), Through this study, the maximum amount of DAW can be transported by IP was suggested as maximum exposure rate at drum surface and the calculation for the other types of waste will be conducted

  5. BWR ASSEMBLY SOURCE TERMS FOR WASTE PACKAGE DESIGN

    Energy Technology Data Exchange (ETDEWEB)

    T.L. Lotz

    1997-02-15

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide boiling water reactor (BWR) assembly radiation source term data for use during Waste Package (WP) design. The BWR assembly radiation source terms are to be used for evaluation of radiolysis effects at the WP surface, and for personnel shielding requirements during assembly or WP handling operations. The objectives of this evaluation are to generate BWR assembly radiation source terms that bound selected groupings of BWR assemblies, with regard to assembly average burnup and cooling time, which comprise the anticipated MGDS BWR commercial spent nuclear fuel (SNF) waste stream. The source term data is to be provided in a form which can easily be utilized in subsequent shielding/radiation dose calculations. Since these calculations may also be used for Total System Performance Assessment (TSPA), with appropriate justification provided by TSPA, or radionuclide release rate analysis, the grams of each element and additional cooling times out to 25 years will also be calculated and the data included in the output files.

  6. Solvent extraction as additional purification method for postconsumer plastic packaging waste

    NARCIS (Netherlands)

    Thoden van Velzen, E.U.; Jansen, M.

    2011-01-01

    An existing solvent extraction process currently used to convert lightly polluted post-industrial packaging waste into high quality re-granulates was tested under laboratory conditions with highly polluted post-consumer packaging waste originating from municipal solid refuse waste. The objective was

  7. EVALUATION OF WASTE PACKAGE EXTERNAL ENVIRONMENTAL CONDITION STUDY

    International Nuclear Information System (INIS)

    The U. S. Department of Energy (DOE) is studying Yucca Mountain as the possible site for a permanent underground repository for disposal of spent nuclear fuel (SNF) and other high-level waste (HLW). The emplacement of high-level radioactive waste in Yucca Mountain will release a large amount of heat into the rock above and below the repository. Due to this heat, the rock temperature will rise, and then decrease when the production of decay heat falls below the rate at which heat escapes from the hot zone. In addition to raising the rock temperature, the heat will vaporize water, which will condense in cooler regions. The condensate water may drain back toward the emplacement drifts or it may ''shed'' through the pillars between emplacement drifts. Other effects, such as coupled chemical and mechanical processes, may influence the movement of water above, within, and below the emplacement drifts. This study examined near field environmental parameters that could have an effect on the waste package, including temperature, humidity, seepage rate, pH of seepage, chemistry (dissolved salts/minerals) of seepage, composition of drift atmosphere, colloids, and biota. This report is a Type I analysis performed in support of the development of System Description Documents (SDDs). A Type I analysis is a quantitative or qualitative analysis that may fulfill any of a variety of purposes associated with the Monitored Geologic Repository (MGR), other than providing direct analytical support for design output documents. A Type I analysis may establish design input, as defined in the ''Quality Assurance Requirements and Description'' (QARD) (DOE 1998). This study establishes a technical basis for emplacement drift (i.e. at the waste package surface) environment criteria to be considered in the development of the waste package design. The information will support development of several SDDs and resolve emplacement drift external environment questions in the criteria of those

  8. On the oxidation of the dissolved organic matter in Boom clay by NaNO3 and NaNO2 from disposed Eurobitum bituminized waste

    International Nuclear Information System (INIS)

    Document available in extended abstract form only. In Belgium, Boom Clay is studied as a potential host clay formation for the final disposal of EUROBITUM bituminized waste, which consists of 60 wt% hard bitumen (Mexphalt R85/40) and 40 wt% waste. The main salts that are present in the bituminized waste are NaNO3, 20-30 wt%, and CaSO4, 4-6 wt%. After disposal of the waste in the clay, an uptake of pore water by the embedded, dehydrated and hygroscopic salts will lead to a swelling of the waste and to a release of the salts into the Boom Clay. A possible consequence of the salt release is the oxidation of the clay by nitrate and, possibly, nitrite, resulting in a lower reducing capacity of the clay towards redox sensitive radionuclides, which in turn could have an impact on the migration behaviour of these radionuclides. The extent of oxidation of authigenic Boom Clay redox sensitive components, like organic matter and pyrite is studied at the SCK.CEN. As a first step in the study of the influence of nitrate and nitrite on the redoxactive Boom Clay components, we performed batch tests with dissolved organic matter (DOM). DOM was exposed to different concentrations of nitrate and nitrite for more than one year in both biotic and abiotic conditions. This paper will discuss the results obtained by exposing DOM to nitrate and nitrite and comparing two methods for the determination of its redox capacity. NaNO3 or NaNO2, previously stored under inert atmosphere to remove all oxygen gas, was added to Boom Clay water collected from a piezometer to obtain final salt concentrations of 0.1 and 0.005 M NaNO3, or 0.05 and 0.005 M NaNO2. Sodium azide, also stored under inert atmosphere, was added (0.2 wt. %) to inhibit the microbial activity in the tests, creating abiotic conditions. All solutions were prepared in an anaerobic glove box. The nitrate and nitrite reduction by DOM was followed by analysing the concentrations of nitrate, nitrite and ammonium in the solutions and of

  9. CLASSIFICATION OF THE MGR WASTE PACKAGE REMEDIATION SYSTEM

    Energy Technology Data Exchange (ETDEWEB)

    J.A. Ziegler

    2000-11-02

    The purpose of this analysis is to document the Quality Assurance (QA) classification of the Monitored Geologic Repository (MGR) waste package remediation system structures, systems, and components (SSCs) performed by the MGR Preclosure Safety and Systems Engineering Section. This analysis also provides the basis for revision of YMP/90-55Q, Q-List (YMP 2000). The Q-List identifies those MGR SSCs subject to the requirements of DOE/RW-0333P7 ''Quality Assurance Requirements and Description'' (QARD) (DOE 2000).

  10. Performance assessment model of a single waste package

    International Nuclear Information System (INIS)

    PANDORA-1.1 is a system model for the mobilization and release of radionuclides from a spent nuclear fuel disposal package. Earlier processes affecting release are represented by input tables. Several groundwater contact alternatives and spent fuel constituents lead to different release-rate behaviors and controlling parameters. Rate control is provided by a product of parameters from hydrology, design, and/or geochemistry/waste form interaction parameters. The program is designed to accommodate evolving requirements such as a wider range of hydrological input values. A computerized configuration management system automates much of the change control process

  11. Corrosion of Metal Inclusions In Bulk Vitrification Waste Packages Erratum

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Gary L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-09-06

    This report refers to or contains Kg values for glasses LAWA44, LAWB45 and LAWC22 affected by calculations errors as identified by Papathanassiu et al. (2011). The corrected Kg values are reported in an erratum included in the revised version of the original report. The revised report can be referenced as follows: Pierce E. M. et al. (2004) Waste Form Release Data Package for the 2005 Integrated Disposal Facility Performance Assessment. PNNL-14805 Rev. 0 Erratum. Pacific Northwest National Laboratory, Richland, WA, USA.

  12. Evaluation and compilation of DOE waste package test data

    International Nuclear Information System (INIS)

    This report summarizes evaluations by the National Institute of Standards and Technology (NIST) of Department of Energy (DOE) activities on waste packages designed for containment of radioactive high-level nuclear waste (HLW) for the six-month period, February through July 1989. This includes reviews of related materials research and plans, information on the Yucca Mountain, Nevada disposal site activities, and other information regarding supporting research and special assistance. Outlines for planned interpretative reports on the topics of aqueous corrosion of copper, mechanisms of stress corrosion cracking and internal failure modes of Zircaloy cladding are included. For the publications reviewed during this reporting period, short discussions are given to supplement the completed reviews and evaluations. Included in this report is an overall review of a 1984 report on glass leaching mechanisms, as well as reviews for each of the seven chapters of this report

  13. Behaviour Test with the Leaching of a Waste package

    International Nuclear Information System (INIS)

    bibliographic data.With the whole coefficients it was made a prediction about the time involved until the total release of the radionuclides. This work is being developed by the Radioactive Waste Management Division of Cnea and it has been included in a contract with the IAEA, which also studies the changes on the mechanical resistance of the waste package,so as the release of gases from organic wastes and the container corrosion

  14. Analysis of Ecodesign Implementation and Solutions for Packaging Waste System by Using System Dynamics Modeling

    Science.gov (United States)

    Berzina, Alise; Dace, Elina; Bazbauers, Gatis

    2010-01-01

    This paper discusses the findings of a research project which explored the packaging waste management system in Latvia. The paper focuses on identifying how the policy mechanisms can promote ecodesign implementation and material efficiency improvement and therefore reduce the rate of packaging waste accumulation in landfill. The method used for analyzing the packaging waste management policies is system dynamics modeling. The main conclusion is that the existing legislative instruments can be used to create an effective policy for ecodesign implementation but substantially higher tax rates on packaging materials and waste disposal than the existing have to be applied.

  15. HORIZONTAL DROP OF THE NAVAL SNF LONG WASTE PACKAGE ON UNYIELDING SURFACE

    Energy Technology Data Exchange (ETDEWEB)

    T. Schmitt

    2000-05-23

    The objective of this calculation is to determine the structural response of a Naval Spent Nuclear Fuel (SNF) Long Waste Package (WP) subjected to a 2.4-m horizontal drop on an unyielding surface (US). The scope of this document is limited to reporting the calculation results in terms of maximum stress intensities. This calculation is associated with the waste package design and was performed by the Waste Package Design section in accordance with the development plan for ''Horizontal Drop of the Naval SNF Long Waste Package on Unyielding Surface''.

  16. Production patterns of packaging waste categories generated at typical Mediterranean residential building worksites.

    Science.gov (United States)

    González Pericot, N; Villoria Sáez, P; Del Río Merino, M; Liébana Carrasco, O

    2014-11-01

    The construction sector is responsible for around 28% of the total waste volume generated in Europe, which exceeds the amount of household waste. This has led to an increase of different research studies focusing on construction waste quantification. However, within the research studies made, packaging waste has been analyzed to a limited extent. This article focuses on the packaging waste stream generated in the construction sector. To this purpose current on-site waste packaging management has been assessed by monitoring ten Mediterranean residential building works. The findings of the experimental data collection revealed that the incentive measures implemented by the construction company to improve on-site waste sorting failed to achieve the intended purpose, showing low segregation ratios. Subsequently, through an analytical study the generation patterns for packaging waste are established, leading to the identification of the prevailing kinds of packaging and the products responsible for their generation. Results indicate that plastic waste generation maintains a constant trend throughout the whole construction process, while cardboard becomes predominant towards the end of the construction works with switches and sockets from the electricity stage. Understanding the production patterns of packaging waste will be beneficial for adapting waste management strategies to the identified patterns for the specific nature of packaging waste within the context of construction worksites. PMID:25081852

  17. Commercial waste and spent fuel packaging program. Annual report

    International Nuclear Information System (INIS)

    This document is a report of activities performed by Westinghouse Advanced Energy Systems Division - Nevada Operations in meeting subtask objectives described in the Nevada Nuclear Waste Storage Investigations (NNWSI) Project Plan and revised planning documentation for Fiscal Year (FY) 1981. Major activities included: completion of the first fuel exchange in the Spent Fuel Test - Climax program; plasma arc welder development; modification and qualification of a canister cutter; installation, and activation of a remote area monitor, constant air monitor and an alpha/beta/gamma counting system; qualification of grapples required to handle pressurized water reactor or boiling water reactor fuel and high level waste (HLW) logs; data acquisition from the 3 kilowatt soil temperature test, 2 kw fuel temperature test, and 2 kw drywell test; calorimetry of the fuel assembly used in the fuel temperature test; evaluation of moisture accumulation in the drywells and recommendations for proposed changes; revision of safety assessment document to include HLW log operations; preparation of quality assurance plan and procedures; development and qualification of all equipment and procedures to receive, handle and encapsulate both the HLW log and spent fuel for the basalt waste isolation program/near surface test facility program; preliminary studies of both the requirements to perform waste packaging for the test and evaluation facility and a cask storage program for the DOE Interim Spent Fuel Management program; and remote handling operations on radioactive source calibration in support of other contractors

  18. Thermal response of the waste package/MPC and EBS

    International Nuclear Information System (INIS)

    Thermal evaluations of the multi-Purpose Canister in repository emplacement have been performed by the Yucca Mountain Site Characterization Project Managing and Operating Contractor (YMP-M ampersand O). Thermal effects have been a major study area of the Mined Geologic Disposal System (MGDS). In the past the project has concentrated primarily on borehole-emplaced waste packages, but the Multi-Purpose Canister (MPC) with a capacity of twenty-one pressurized water reactor (PWR) fuel assemblies (or more) will likely be drift emplaced with a multi-barrier disposal container or overpack. This study investigates the thermal behavior of the waste package/MPC and its effect on the repository near-field. Results indicate that peak internal temperatures occur one to five years post emplacement, and the timing of the peak is highly dependent on the choice of design basis fuel and thermal loading. The maximum acceptable capacity of the MPC, with ten year old fuel, was determined to be twenty-one PWRs based on the current peak temperature goal of 350 degrees C; however, higher capacities can be achieved if older fuel is substituted

  19. Hanford high-level waste melter system evaluation data packages

    International Nuclear Information System (INIS)

    The Tank Waste Remediation System is selecting a reference melter system for the Hanford High-Level Waste vitrification plant. A melter evaluation was conducted in FY 1994 to narrow down the long list of potential melter technologies to a few for testing. A formal evaluation was performed by a Melter Selection Working Group (MSWG), which met in June and August 1994. At the June meeting, MSWG evaluated 15 technologies and selected six for more thorough evaluation at the Aug. meeting. All 6 were variations of joule-heated or induction-heated melters. Between the June and August meetings, Hanford site staff and consultants compiled data packages for each of the six melter technologies as well as variants of the baseline technologies. Information was solicited from melter candidate vendors to supplement existing information. This document contains the data packages compiled to provide background information to MSWG in support of the evaluation of the six technologies. (A separate evaluation was performed by Fluor Daniel, Inc. to identify balance of plant impacts if a given melter system was selected.)

  20. Solvent extraction as additional purification method for postconsumer plastic packaging waste

    OpenAIRE

    Thoden Van Velzen, E.U.; M. Jansen

    2011-01-01

    An existing solvent extraction process currently used to convert lightly polluted post-industrial packaging waste into high quality re-granulates was tested under laboratory conditions with highly polluted post-consumer packaging waste originating from municipal solid refuse waste. The objective was to study the technical feasibility of using this extraction technology and to study the quality of the produced cleaned plastic flakes. Two types of dirty plastic flakes from household waste were ...

  1. Data package format for certified transuranic waste for the Waste Isolation Pilot Plant: Revision 2

    International Nuclear Information System (INIS)

    These instructions have been prepared as a reference guide for those personnel responsible for the transmission of the data package to the WIPP. For those sites having automated data processing systems available for use, it should be understood that the shipper's computer system will be used to place the information into the specified format for transmittal. The method of input will, of course, depend upon the particular system being used. Prior to shipment, the shipper's computer system must retrieve the required information for all the packages in that shipment and write the information to an IBM or IBM-compatible personal computer in the data package format. The data package will then be transmitted in ASCII format over the specified communications system to the WIPP Waste Information System (WWIS) using RLINK, which will be furnished by WIPP. Therefore, these instructions are primarily for use by data processing personnel to aid them in programming the system to provide the transmittal information in the data package format. The method to input the data into the shipper's computer system should be determined through a joint effort between the waste generator/shipper and the data processing personnel

  2. PNGMDR 2013-2015. 2015 report by the 'optimisation of the distribution of wastes among management sectors' work-group

    International Nuclear Information System (INIS)

    PNGMDR is the French national plan for the management of radioactive materials and wastes. After a discussion of the stakes associated with the optimisation of the different sectors involved in radioactive waste management (production, sorting, processing, packaging, warehousing, storage), this reports proposes a description of wastes for which a sector optimisation study has been performed (graphite wastes, Marcoule bitumen packages, common solid residues, cemented solid wastes in concrete containers from Areva La Hague), describes the various studies optimisation ways and alternate scenarios for these different wastes. It analyses and discusses the progress status of the studies optimisation ways: improvement of the characterisation of graphite and bituminous wastes, studies and works on the storage of low-level-long-lived wastes, studies on the processing of graphite wastes, of bituminous packages and common solid residues. Other issues are evoked: the case of wastes produced by dismantling operations, and a technical-economic approach. Perspectives and future works are discussed for the different considered wastes

  3. Production patterns of packaging waste categories generated at typical Mediterranean residential building worksites

    Energy Technology Data Exchange (ETDEWEB)

    González Pericot, N., E-mail: natalia.gpericot@upm.es [Escuela Técnica Superior de Edificación, Universidad Politécnica de Madrid, Calle Juan de Herrera n°6, 28040 Madrid (Spain); Villoria Sáez, P., E-mail: paola.villoria@upm.es [Escuela Técnica Superior de Edificación, Universidad Politécnica de Madrid, Calle Juan de Herrera n°6, 28040 Madrid (Spain); Del Río Merino, M., E-mail: mercedes.delrio@upm.es [Escuela Técnica Superior de Edificación, Universidad Politécnica de Madrid, Calle Juan de Herrera n°6, 28040 Madrid (Spain); Liébana Carrasco, O., E-mail: oscar.liebana@uem.es [Escuela de Arquitectura, Universidad Europea de Madrid, Calle Tajo s/n, 28670 Villaviciosa de Odón (Spain)

    2014-11-15

    Highlights: • On-site segregation level: 1.80%; training and motivation strategies were not effective. • 70% Cardboard waste: from switches and sockets during the building services stage. • 40% Plastic waste: generated during structures and partition works due to palletizing. • >50% Wood packaging waste, basically pallets, generated during the envelope works. - Abstract: The construction sector is responsible for around 28% of the total waste volume generated in Europe, which exceeds the amount of household waste. This has led to an increase of different research studies focusing on construction waste quantification. However, within the research studies made, packaging waste has been analyzed to a limited extent. This article focuses on the packaging waste stream generated in the construction sector. To this purpose current on-site waste packaging management has been assessed by monitoring ten Mediterranean residential building works. The findings of the experimental data collection revealed that the incentive measures implemented by the construction company to improve on-site waste sorting failed to achieve the intended purpose, showing low segregation ratios. Subsequently, through an analytical study the generation patterns for packaging waste are established, leading to the identification of the prevailing kinds of packaging and the products responsible for their generation. Results indicate that plastic waste generation maintains a constant trend throughout the whole construction process, while cardboard becomes predominant towards the end of the construction works with switches and sockets from the electricity stage. Understanding the production patterns of packaging waste will be beneficial for adapting waste management strategies to the identified patterns for the specific nature of packaging waste within the context of construction worksites.

  4. Production patterns of packaging waste categories generated at typical Mediterranean residential building worksites

    International Nuclear Information System (INIS)

    Highlights: • On-site segregation level: 1.80%; training and motivation strategies were not effective. • 70% Cardboard waste: from switches and sockets during the building services stage. • 40% Plastic waste: generated during structures and partition works due to palletizing. • >50% Wood packaging waste, basically pallets, generated during the envelope works. - Abstract: The construction sector is responsible for around 28% of the total waste volume generated in Europe, which exceeds the amount of household waste. This has led to an increase of different research studies focusing on construction waste quantification. However, within the research studies made, packaging waste has been analyzed to a limited extent. This article focuses on the packaging waste stream generated in the construction sector. To this purpose current on-site waste packaging management has been assessed by monitoring ten Mediterranean residential building works. The findings of the experimental data collection revealed that the incentive measures implemented by the construction company to improve on-site waste sorting failed to achieve the intended purpose, showing low segregation ratios. Subsequently, through an analytical study the generation patterns for packaging waste are established, leading to the identification of the prevailing kinds of packaging and the products responsible for their generation. Results indicate that plastic waste generation maintains a constant trend throughout the whole construction process, while cardboard becomes predominant towards the end of the construction works with switches and sockets from the electricity stage. Understanding the production patterns of packaging waste will be beneficial for adapting waste management strategies to the identified patterns for the specific nature of packaging waste within the context of construction worksites

  5. Cleanup Verification Package for the 118-C-1, 105-C Solid Waste Burial Ground

    International Nuclear Information System (INIS)

    This cleanup verification package documents completion of remedial action for the 118-C-1, 105-C Solid Waste Burial Ground. This waste site was the primary burial ground for general wastes from the operation of the 105-C Reactor and received process tubes, aluminum fuel spacers, control rods, reactor hardware, spent nuclear fuel and soft wastes

  6. Cleanup Verification Package for the 118-C-1, 105-C Solid Waste Burial Ground

    Energy Technology Data Exchange (ETDEWEB)

    M. J. Appel and J. M. Capron

    2007-07-25

    This cleanup verification package documents completion of remedial action for the 118-C-1, 105-C Solid Waste Burial Ground. This waste site was the primary burial ground for general wastes from the operation of the 105-C Reactor and received process tubes, aluminum fuel spacers, control rods, reactor hardware, spent nuclear fuel and soft wastes.

  7. Addendum to the Safety Analysis Report for the Steel Waste Packaging. Revision 1

    International Nuclear Information System (INIS)

    The Battelle Pacific Northwest National Laboratory Safety Analysis Report (SAR) for the Steel Waste Package requires additional analyses to support the shipment of remote-handled radioactive waste and special-case waste from the 324 building hot cells to PUREX for interim storage. This addendum provides the analyses required to show that this waste can be safely shipped onsite in the configuration shown

  8. Addendum to the Safety Analysis Report for the Steel Waste Packaging. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Crow, S R

    1996-02-15

    The Battelle Pacific Northwest National Laboratory Safety Analysis Report (SAR) for the Steel Waste Package requires additional analyses to support the shipment of remote-handled radioactive waste and special-case waste from the 324 building hot cells to PUREX for interim storage. This addendum provides the analyses required to show that this waste can be safely shipped onsite in the configuration shown.

  9. SCOPING EVALUATION TO EXPLORE - ROCK FALL ACCIDENT CONDITION ANALYSIS ON MULTI-PURPOSE CANISTER WASTE PACKAGES CORRELATED FROM INTERLOCKING BASKET WASTE PACKAGE DESIGN ANALYSIS (SCPB: N/A)

    Energy Technology Data Exchange (ETDEWEB)

    Z, Ceylan

    1995-12-08

    The objective of this analysis is to correlate the results of a rock fall analysis performed for the 12 Pressurized Water Reactor (PWR) Fuel Assembly Interlocking Basket waste package (WP) in order to determine the size of rock that can strike the Multi-Purpose Canister (MPC) waste packages without breaching the containment barriers. The purpose of this analysis is to document the models and methods used in the calculations.

  10. Cold bituminous mixes

    OpenAIRE

    Delak, Boštjan

    2012-01-01

    Asphalt mix (short also asphalt) can be obtained from nature (natural asphalt) or produced technically (artificial asphalt), i. e. as a mix of bituminous binder, stone grain, and other additives that improve the asphalt properties. The beginnings of the use of asphalt can be traced from 6,000 to 4,000 BC since people back then knew some materials that contained bitumen as a binder. A more frequent use of the natural asphalt, however, starts around 3,000 BC. Different peoples used it primar...

  11. Nuclear waste management technical support in the development of nuclear waste form criteria for the NRC. Task 1. Waste package overview

    International Nuclear Information System (INIS)

    In this report the current state of waste package development for high level waste, transuranic waste, and spent fuel in the US and abroad has been assessed. Specifically, reviewed are recent and on-going research on various waste forms, container materials and backfills and tentatively identified those which are likely to perform most satisfactorily in the repository environment. Radiation effects on the waste package components have been reviewed and the magnitude of these effects has been identified. Areas requiring further research have been identified. The important variables affecting radionuclide release from the waste package have been described and an evaluation of regulatory criteria for high level waste and spent fuel is presented. Finally, for spent fuel, high level, and TRU waste, components which could be used to construct a waste package having potential to meet NRC performance requirements have been described and identified

  12. Nuclear waste management technical support in the development of nuclear waste form criteria for the NRC. Task 1. Waste package overview

    Energy Technology Data Exchange (ETDEWEB)

    Dayal, R.; Lee, B.S.; Wilke, R.J.; Swyler, K.J.; Soo, P.; Ahn, T.M.; McIntyre, N.S.; Veakis, E.

    1982-02-01

    In this report the current state of waste package development for high level waste, transuranic waste, and spent fuel in the US and abroad has been assessed. Specifically, reviewed are recent and on-going research on various waste forms, container materials and backfills and tentatively identified those which are likely to perform most satisfactorily in the repository environment. Radiation effects on the waste package components have been reviewed and the magnitude of these effects has been identified. Areas requiring further research have been identified. The important variables affecting radionuclide release from the waste package have been described and an evaluation of regulatory criteria for high level waste and spent fuel is presented. Finally, for spent fuel, high level, and TRU waste, components which could be used to construct a waste package having potential to meet NRC performance requirements have been described and identified.

  13. Leaching characteristics of paraffin waste package with pinhole

    International Nuclear Information System (INIS)

    An Effect of pinhole(perforation or pit penetration) that might be formed outside the package on the nuclide leaching from paraffin waste form was investigated. In case of single pinhole, the leached mass and cumulative fraction leached (CFL) increased with the larger diameter of pinhole, but they were not in direct proportion to the size or area of pinhole. If the total area of multiple pinholes was fixed, the leached mass showed a tendency to increase as each size was smaller and the number was more. It was also found that the leached mass was not in direct proportion to the number of pinhole in case of constant size. In order to analyze the test results, the shrinking core model(SCM) was derived from the diffusion-controlled dissolution reaction and compared with previous diffusion model

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

    Energy Technology Data Exchange (ETDEWEB)

    Colleen Shelton-Davis

    2005-11-01

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

  15. Waste Form Release Data Package for the 2001 Immobilized Low-Activity Waste Performance Assessment

    International Nuclear Information System (INIS)

    This data package documents the experimentally derived input data on the representative waste glasses LAWABP1 and HLP-31 that will be used for simulations of the immobilized lowactivity waste disposal system with the Subsurface Transport Over Reactive Multiphases (STORM) code. The STORM code will be used to provide the near-field radionuclide release source term for a performance assessment to be issued in March of 2001. Documented in this data package are data related to (1) kinetic rate law parameters for glass dissolution, (2) alkali-H ion exchange rate, (3) chemical reaction network of secondary phases that form in accelerated weathering tests, and (4) thermodynamic equilibrium constants assigned to these secondary phases. The kinetic rate law and Na+-H+ ion exchange rate were determined from single-pass flow-through experiments. Pressurized unsaturated flow and vapor hydration experiments were used for accelerated weathering or aging of the glasses. The majority of the thermodynamic data were extracted from the thermodynamic database package shipped with the geochemical code EQ3/6. However, several secondary reaction products identified from laboratory tests with prototypical LAW glasses were not included in this database, nor are the thermodynamic data available in the open literature. One of these phases, herschelite, was determined to have a potentially significant impact on the release calculations and so a solubility product was estimated using a polymer structure model developed for zeolites. Although this data package is relatively complete, final selection of ILAW glass compositions has not been done by the waste treatment plant contractor. Consequently, revisions to this data package to address new ILAW glass formulations are to be regularly expected.

  16. 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)

  17. Data Packages for the Hanford Immobilized Low Activity Tank Waste Performance Assessment 2001 Version [SEC 1 THRU 5

    Energy Technology Data Exchange (ETDEWEB)

    MANN, F.M.

    2000-03-02

    Data package supporting the 2001 Immobilized Low-Activity Waste Performance Analysis. Geology, hydrology, geochemistry, facility, waste form, and dosimetry data based on recent investigation are provided. Verification and benchmarking packages for selected software codes are provided.

  18. European experience with asphalt packaging of radioactive wastes

    International Nuclear Information System (INIS)

    In Europe, the use of asphalt is synonymous with volume reduction of low and medium level radioactive waste. It started at Marcoule, France in the early 1960's and soon was adopted by Karlsruhe in West Germany, Eurochemic in Belgium, and in other countries. The use of an asphalt (or bitumen) binder, or immobilizing agent, in the VRS (Volume Reduction and Solidification) process proved to be beneficial in many ways. At Karlsruhe, for example, the VRS asphalt system replaced a cement system (non-volume reducing) and resulted in two drums of solidified waste versus ten with cement. This process reduced the transportation problem considerably. Asphalt is an inert, waterproof material and provides significantly improved package integrity under all conditions of on-site storage, transportation, and burial. The asphalt VRS system provides considerable cost savings, particularly for the recurring items such as binder, containers, transportation, and burial. These annual savings may approach $500,000 annually for a 1000 MWe nuclear plant. End product advantages include higher resistance to leaching and other environmental impacts, as well as less internal corrosion of the drums

  19. Ceramic package fabrication for YMP nuclear waste disposal

    Energy Technology Data Exchange (ETDEWEB)

    Wilfinger, K.

    1994-08-01

    The purpose of this work is to develop alternate materials/design concepts to metal barriers for the Nevada Nuclear Waste Storage Investigations Project. There is some potential that site conditions may prove to be too aggressive for successful employment of the metal alloys under current consideration or that performance assessment models will predict metal container degradation rates that are inconsistent with the goal of substantially complete containment included in the NRC regulations. In the event that the anticipated lifetimes of metal containers are considered inadequate, alternate materials (i.e. ceramics or ceramic/metal composites) will be chosen due to superior corrosion resistance. This document was prepared using information taken from the open literature, conversations and correspondence with vendors, news releases and data presented at conferences to determine what form such a package might take. This discussion presents some ceramic material selection criteria, alternatives for the materials which might be used and alternatives for potential fabrication routes. This includes {open_quotes}stand alone{close_quotes} ceramic components and ceramic coatings/linings for metallic structures. A list of companies providing verbal or written information concerning the production of ceramic or ceramic lined waste containers appears at the end of this discussion.

  20. Aging and Phase Stability of Waste Package Outer Barrier

    Energy Technology Data Exchange (ETDEWEB)

    F. Wong

    2004-09-28

    This report was prepared in accordance with ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]). This report provides information on the phase stability of Alloy 22, the current waste package outer barrier material. The goal of this model is to determine whether the single-phase solid solution is stable under repository conditions and, if not, how fast other phases may precipitate. The aging and phase stability model, which is based on fundamental thermodynamic and kinetic concepts and principles, will be used to provide predictive insight into the long-term metallurgical stability of Alloy 22 under relevant repository conditions. The results of this model are used by ''General Corrosion and Localized Corrosion of Waste Package Outer Barrier'' as reference-only information. These phase stability studies are currently divided into three general areas: Tetrahedrally close-packed (TCP) phase and carbide precipitation in the base metal; TCP and carbide precipitation in welded samples; and Long-range ordering reactions. TCP-phase and carbide precipitates that form in Alloy 22 are generally rich in chromium (Cr) and/or molybdenum (Mo) (Raghavan et al. 1984 [DIRS 154707]). Because these elements are responsible for the high corrosion resistance of Alloy 22, precipitation of TCP phases and carbides, especially at grain boundaries, can lead to an increased susceptibility to localized corrosion in the alloy. These phases are brittle and also tend to embrittle the alloy (Summers et al. 1999 [DIRS 146915]). They are known to form in Alloy 22 at temperatures greater than approximately 600 C. Whether these phases also form at the lower temperatures expected in the repository during the 10,000-year regulatory period must be determined. The kinetics of this precipitation will be determined for both the base metal and the weld heat-affected zone (HAZ). The TCP

  1. Aging and Phase Stability of Waste Package Outer Barrier

    International Nuclear Information System (INIS)

    This report was prepared in accordance with ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]). This report provides information on the phase stability of Alloy 22, the current waste package outer barrier material. The goal of this model is to determine whether the single-phase solid solution is stable under repository conditions and, if not, how fast other phases may precipitate. The aging and phase stability model, which is based on fundamental thermodynamic and kinetic concepts and principles, will be used to provide predictive insight into the long-term metallurgical stability of Alloy 22 under relevant repository conditions. The results of this model are used by ''General Corrosion and Localized Corrosion of Waste Package Outer Barrier'' as reference-only information. These phase stability studies are currently divided into three general areas: Tetrahedrally close-packed (TCP) phase and carbide precipitation in the base metal; TCP and carbide precipitation in welded samples; and Long-range ordering reactions. TCP-phase and carbide precipitates that form in Alloy 22 are generally rich in chromium (Cr) and/or molybdenum (Mo) (Raghavan et al. 1984 [DIRS 154707]). Because these elements are responsible for the high corrosion resistance of Alloy 22, precipitation of TCP phases and carbides, especially at grain boundaries, can lead to an increased susceptibility to localized corrosion in the alloy. These phases are brittle and also tend to embrittle the alloy (Summers et al. 1999 [DIRS 146915]). They are known to form in Alloy 22 at temperatures greater than approximately 600 C. Whether these phases also form at the lower temperatures expected in the repository during the 10,000-year regulatory period must be determined. The kinetics of this precipitation will be determined for both the base metal and the weld heat-affected zone (HAZ). The TCP phases (P, μ, and σ) are present in

  2. NWTS waste package program plan. Volume I. Program strategy, description, and schedule

    International Nuclear Information System (INIS)

    This document describes the work planned for developing the technology to design, test and produce packages used for the long-term isolation of nuclear waste in deep geologic repositories. Waste forms considered include spent fuel and high-level waste. The testing and selection effort for barrier materials for radionuclide containment is described. The NWTS waste package program is a design-driven effort; waste package conceptual designs are used as input for preliminary designs, which are upgraded to a final design as materials and testing data become available. Performance assessment models are developed and validated. Milestones and a detailed schedule are given for the waste package development effort. Program logic networks defining work flow, interfaces among the NWTS Projects, and interrelationships of specific activities are presented. Detailed work elements are provided for the Waste Package Program Plan subtasks - design and development, waste form, barrier materials, and performance evaluation - for salt and basalt, host rocks for which the state of waste package knowledge and the corresponding data base are advanced

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

    International Nuclear Information System (INIS)

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

  4. Characterisation of plastic packaging waste for recycling: problems related to current approaches

    DEFF Research Database (Denmark)

    Götze, Ramona; Astrup, Thomas Fruergaard

    2013-01-01

    criteria of recycling processes. A lack of information in current waste characterisation practise on polymer resin composition, black coloured material content and the influence of surface adherent material on physico-chemical characteristics of plastic packaging waste were identified. These shortcomings......Informed decisions regarding new recycling schemes require waste characterisation studies which provide in addition to data on waste amounts and the share of recyclable fractions, accurate data on physico-chemical characteristics of the waste materials considering the material specific input...

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

    Energy Technology Data Exchange (ETDEWEB)

    SULLIVAN, T.

    2004-06-30

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

  6. Safety evaluation for packaging transportation of equipment for tank 241-C-106 waste sluicing system

    Energy Technology Data Exchange (ETDEWEB)

    Calmus, D.B.

    1994-08-25

    A Waste Sluicing System (WSS) is scheduled for installation in nd waste storage tank 241-C-106 (106-C). The WSS will transfer high rating sludge from single shell tank 106-C to double shell waste tank 241-AY-102 (102-AY). Prior to installation of the WSS, a heel pump and a transfer pump will be removed from tank 106-C and an agitator pump will be removed from tank 102-AY. Special flexible receivers will be used to contain the pumps during removal from the tanks. After equipment removal, the flexible receivers will be placed in separate containers (packagings). The packaging and contents (packages) will be transferred from the Tank Farms to the Central Waste Complex (CWC) for interim storage and then to T Plant for evaluation and processing for final disposition. Two sizes of packagings will be provided for transferring the equipment from the Tank Farms to the interim storage facility. The packagings will be designated as the WSSP-1 and WSSP-2 packagings throughout the remainder of this Safety Evaluation for Packaging (SEP). The WSSP-1 packagings will transport the heel and transfer pumps from 106-C and the WSSP-2 packaging will transport the agitator pump from 102-AY. The WSSP-1 and WSSP-2 packagings are similar except for the length.

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

    Energy Technology Data Exchange (ETDEWEB)

    J.R. Massari

    1995-10-06

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

  8. Long-Term Waste Package Degradation Studies at the Yucca Mountain Potential High-Level Nuclear Waste Repository

    Energy Technology Data Exchange (ETDEWEB)

    Mon, K. G.; Bullard, B. E.; Longsine, D. E.; Mehta, S.; Lee, J. H.; Monib, A. M.

    2002-02-26

    The Site Recommendation (SR) process for the potential repository for spent nuclear fuel (SNF) and high-level nuclear waste (HLW) at Yucca Mountain, Nevada is underway. Fulfillment of the requirements for substantially complete containment of the radioactive waste emplaced in the potential repository and subsequent slow release of radionuclides from the Engineered Barrier System (EBS) into the geosphere will rely on a robust waste container design, among other EBS components. Part of the SR process involves sensitivity studies aimed at elucidating which model parameters contribute most to the drip shield and waste package degradation characteristics. The model parameters identified included (a) general corrosion rate model parameters (temperature-dependence and uncertainty treatment), and (b) stress corrosion cracking (SCC) model parameters (uncertainty treatment of stress and stress intensity factor profiles in the Alloy 22 waste package outer barrier closure weld regions, the SCC initiation stress threshold, and the fraction of manufacturing flaws oriented favorably for through-wall penetration by SCC). These model parameters were reevaluated and new distributions were generated. Also, early waste package failures due to improper heat treatment were added to the waste package degradation model. The results of these investigations indicate that the waste package failure profiles are governed by the manufacturing flaw orientation model parameters and models used.

  9. Long-Term Waste Package Degradation Studies at the Yucca Mountain Potential High-Level Nuclear Waste Repository

    International Nuclear Information System (INIS)

    The Site Recommendation (SR) process for the potential repository for spent nuclear fuel (SNF) and high-level nuclear waste (HLW) at Yucca Mountain, Nevada is underway. Fulfillment of the requirements for substantially complete containment of the radioactive waste emplaced in the potential repository and subsequent slow release of radionuclides from the Engineered Barrier System (EBS) into the geosphere will rely on a robust waste container design, among other EBS components. Part of the SR process involves sensitivity studies aimed at elucidating which model parameters contribute most to the drip shield and waste package degradation characteristics. The model parameters identified included (a) general corrosion rate model parameters (temperature-dependence and uncertainty treatment), and (b) stress corrosion cracking (SCC) model parameters (uncertainty treatment of stress and stress intensity factor profiles in the Alloy 22 waste package outer barrier closure weld regions, the SCC initiation stress threshold, and the fraction of manufacturing flaws oriented favorably for through-wall penetration by SCC). These model parameters were reevaluated and new distributions were generated. Also, early waste package failures due to improper heat treatment were added to the waste package degradation model. The results of these investigations indicate that the waste package failure profiles are governed by the manufacturing flaw orientation model parameters and models used

  10. The treatment and packaging of waste plutonium and waste actinides for disposal

    International Nuclear Information System (INIS)

    The objectives of this work have been to review the current state of knowledge on the treatment and packaging of unusable or surplus plutonium and other waste actinides for disposal and to identify any gaps in data essential for the development of a preferred route. The exercise was based on published data which said the quantity currently to be disposed of was 50 tonnes in oxide form. A literature review over the period 1978 to 1988 was carried out and a computerised database specific to the exercise was created. From this it is concluded that there are no insuperable problems to the formulation of a disposal route although there is none currently proven. The preferred wasteform would be a glass or synthetic rock. The major complication lies in the fissile nature of plutonium which dictates limits to the package size and places restrictions on the production and disposal routes. Additional work necessary to permit a final decision is listed. (author)

  11. FABRICATION AND DEPLOYMENT OF THE 9979 TYPE AF RADIOACTIVE WASTE PACKAGING FOR THE DEPARTMENT OF ENERGY

    Energy Technology Data Exchange (ETDEWEB)

    Blanton, P.; Eberl, K.

    2013-10-10

    This paper summarizes the development, testing, and certification of the 9979 Type A Fissile Packaging that replaces the UN1A2 Specification Shipping Package eliminated from Department of Transportation (DOT) 49 CFR 173. The DOT Specification Package was used for many decades by the U.S. nuclear industry as a fissile waste container until its removal as an authorized container by DOT. This paper will discuss stream lining procurement of high volume radioactive material packaging manufacturing, such as the 9979, to minimize packaging production costs without sacrificing Quality Assurance. The authorized content envelope (combustible and non-combustible) as well as planned content envelope expansion will be discussed.

  12. Radioactive waste package assay facility. Final report - V. A

    International Nuclear Information System (INIS)

    This report provides a summary of research work carried out in support of the development of an integrated assay system for the quality checking of Intermediate Level Waste encapsulated in cement. Four non-destructive techniques were originally identified as being viable methods for obtaining radiometric inventory data from a cemented 500 litre ILW package. The major part of the programme was devoted to the development of two interrogation techniques; active neutron for measuring the total fissile content and active gamma for measuring the total actinide content. An electron linear accelerator was used to supply the interrogating beam for these two methods. In addition the linear accelerator beam could be used for high energy radiography. The results of this work are described and the performances and limitations of the non-destructive methods are summarised. The main engineering and operational features which influence the design of an integrated assay facility are outlined and a conceptual layout for a facility to inspect 750 ILW drums a year is described. Details of the detection methods, data processing and potential application of the assay facility are given in three associated HMIP reports. (Author)

  13. Long-term durability experiments with concrete-based waste packages in simulated repository conditions

    International Nuclear Information System (INIS)

    Two extensive experiments on long-term durability of waste packages in simulated repository conditions are described. The first one is a 'half-scale experiment' comprising radioactive waste product and half-scale concrete containers in site specific groundwater conditions. The second one is 'full-scale experiment' including simulated inactive waste product and full-scale concrete container stored in slowly flowing fresh water. The scope of the experiments is to demonstrate long-term behaviour of the designed waste packages in contact with moderately concrete aggressive groundwater, and to evaluate the possible interactions between the waste product, concrete container and ground water. As the waste packages are made of high-quality concrete, provisions have been made to continue the experiments for several years

  14. Corrosion of Metal Inclusions In Bulk Vitrification Waste Packages

    Energy Technology Data Exchange (ETDEWEB)

    Bacon, Diana H.; Pierce, Eric M.; Wellman, Dawn M.; Strachan, Denis M.; Josephson, Gary B.

    2006-07-31

    The primary purpose of the work reported here is to analyze the potential effect of the release of technetium (Tc) from metal inclusions in bulk vitrification waste packages once they are placed in the Integrated Disposal Facility (IDF). As part of the strategy for immobilizing waste from the underground tanks at Hanford, selected wastes will be immobilized using bulk vitrification. During analyses of the glass produced in engineering-scale tests, metal inclusions were found in the glass product. This report contains the results from experiments designed to quantify the corrosion rates of metal inclusions found in the glass product from AMEC Test ES-32B and simulations designed to compare the rate of Tc release from the metal inclusions to the release of Tc from glass produced with the bulk vitrification process. In the simulations, the Tc in the metal inclusions was assumed to be released congruently during metal corrosion as soluble TcO4-. The experimental results and modeling calculations show that the metal corrosion rate will, under all conceivable conditions at the IDF, be dominated by the presence of the passivating layer and corrosion products on the metal particles. As a result, the release of Tc from the metal particles at the surfaces of fractures in the glass releases at a rate similar to the Tc present as a soluble salt. The release of the remaining Tc in the metal is controlled by the dissolution of the glass matrix. To summarize, the release of 99Tc from the BV glass within precipitated Fe is directly proportional to the diameter of the Fe particles and to the amount of precipitated Fe. However, the main contribution to the Tc release from the iron particles is over the same time period as the release of the soluble Tc salt. For the base case used in this study (0.48 mass% of 0.5 mm diameter metal particles homogeneously distributed in the BV glass), the release of 99Tc from the metal is approximately the same as the release from 0.3 mass% soluble Tc

  15. Halomonas desiderata as a bacterial model to predict the possible biological nitrate reduction in concrete cells of nuclear waste disposals

    OpenAIRE

    Alquier, Marjorie; Kassim, Caroline; Bertron, Alexandra; Sablayrolles, Caroline; Rafrafi, Yan; Albrecht, Achim; Erable, Benjamin

    2014-01-01

    After closure of a waste disposal cell in a repository for radioactive waste, resaturation is likely to cause the release of soluble species contained in cement and bituminous matrices, such as ionic species (nitrates, sulfates, calcium and alkaline ions, etc.), organic matter (mainly organic acids), or gases (from steel containers and reinforced concrete structures as well as from radiolysis within the waste packages). However, in the presence of nitrates in the near-field of waste, the wast...

  16. Assessing microbiologically induced corrosion of waste package materials in the Yucca Mountain repository

    Energy Technology Data Exchange (ETDEWEB)

    Horn, J. M., LLNL

    1998-01-01

    The contribution of bacterial activities to corrosion of nuclear waste package materials must be determined to predict the adequacy of containment for a potential nuclear waste repository at Yucca Mountain (YM), NV. The program to evaluate potential microbially induced corrosion (MIC) of candidate waste container materials includes characterization of bacteria in the post-construction YM environment, determination of their required growth conditions and growth rates, quantitative assessment of the biochemical contribution to metal corrosion, and evaluation of overall MIC rates on candidate waste package materials.

  17. Natural additives and agricultural wastes in biopolymer formulations for food packaging

    OpenAIRE

    Valdés, Arantzazu; Mellinas, Ana Cristina; Ramos, Marina; Garrigós, María Carmen; Jiménez, Alfonso

    2014-01-01

    The main directions in food packaging research are targeted toward improvements in food quality and food safety. For this purpose, food packaging providing longer product shelf-life, as well as the monitoring of safety and quality based upon international standards, is desirable. New active packaging strategies represent a key area of development in new multifunctional materials where the use of natural additives and/or agricultural wastes is getting increasing interest. The development of ne...

  18. Yucca Mountain Project waste package design for MRS [Monitored Retrievable Storage] system studies

    International Nuclear Information System (INIS)

    This report, prepared by the Yucca Mountain Project, is the report for Task E of the MRS System Study. A number of assumptions were necessary prior to initiation of this system study. These assumptions have been defined in Section 2 for the packaging scenarios, the waste forms, and the waste package concepts and materials. Existing concepts were utilized because of schedule constraints. Section 3 provides a discussion of sensitivity considerations regarding the impact of different assumptions on the overall result of the system study. With the exception of rod consolidation considerations, the system study should not be sensitive to the parameters assumed for the waste package. The current reference waste package materials and concepts are presented in Section 4. Although stainless steel is assumed for this study, a container material has not yet been selected for Advanced Conceptual Design (ACD) from the six candidates currently under study. Section 5 discusses the current thinking for possible alternate waste package materials and concepts. These concepts are being considered in the event that the waste package emplacement environment is more severe than is currently anticipated. Task E also provides a concept in Section 6 for an MRS canister to contain consolidated fuel for storage at the MRS and eventual shipment to the repository. 5 refs., 14 figs., 10 tabs

  19. An overview of the waste handling and packaging plant, a major processing facility for remote-handled transuranic waste

    International Nuclear Information System (INIS)

    The Waste Handling and Packaging Plant (WHPP) is a FY 1991 line item project proposed for construction at the Oak Ridge National Laboratory (ORNL). The purpose of the facility is to receive, package, certify and ship remote-handled (RH) and special case (SC) transuranic (TRU) waste to the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico. The scope of the facility includes the mobilization of liquids and sludges from the Melton Valley Storage Tanks, transport of these liquids and sludges to the WHPP, solidification to a certifiable waste form, and final packaging and shipment to WIPP. Various solid hot cell wastes will be received at the WHPP from storage at ORNL and from other Department of Energy (DOE) sites. The solid wastes will be removed from the storage or shipping container, examined, processed as required, certified and packaged for shipment to WIPP. All packages coming from the processing cell will be in 55 gallon drums, and the facility will have the capability to load these directly into a shielded drum shipping cask, or to load these into the RH TRU canister for remote welding and shipment to WIPP using the RH TRU canister cask. 4 figs

  20. Preliminary evaluation of waste package releases using drift-scale thermo-hydrologic analyses

    International Nuclear Information System (INIS)

    In a 1993 performance assessment, the uncertainty associated with using panel-scale thermo-hydrologic analyses to define the near-field environment in the vicinity of the waste packages was identified as one of the major factors impacting the predicted performance. This was because of the impact of the thermo-hydrologic regime on the initiation of aqueous corrosion as well as the rate of corrosion, rate of waste form dissolution (due to the uncertainty in the percent of the waste form surface covered by water film), solubility limits, and the effective diffusion through the waste package and engineered barrier. This document presents an initial attempt to incorporate, in a more representative fashion, the anticipated thermo-hydrologic response in the vicinity of in-drift emplaced waste packages into the post-closure performance assessment. It illustrates some of the issues which must be resolved prior to justifying the inclusion of these representations into the assessment

  1. Scenarios study on post-consumer plastic packaging waste recycling

    OpenAIRE

    Thoden van Velzen, E.U.; Bos-Brouwers, H.E.J.; de Groot, J J; Bing Xiaoyun, Xiaoyun; Jansen, M; Luijsterburg, B.

    2013-01-01

    We all use plastics on a daily basis. Plastics come in many shapes, sizes and compositions and are used in a wide variety of products. Almost all of the currently used plastic packaging are made from fossil resources, which are finite. The production of plastic packages causes environmental impacts, whereas the correct use of these packages will reduce product losses and hence reduce the much more negative environmental impacts associated with product losses. Wrongly discarded plastic objects...

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1983-06-01

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

  4. Technical Basis Document No. 6: Waste Package and Drip Shield Corrosion

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, J; Pasupathi, V; Nair, P; Gordon, G; McCright, D; Gdowski, G; Carroll, S; Steinborn, T; Summers, T; Wong, F; Rebak, R; Lian, T; Ilevbare, G; Lee, J; Hua, F; Payer, J

    2003-08-01

    The waste package and drip shield will experience a wide range of interactive environmental conditions and degradation modes that will determine the overall performance of the waste package and repository. The operable modes of degradation are determined by the temperature regime of operation (region), and are summarized here. Dry-Out Region (T {ge} 120 C; 50 to 400 Years): During the pre-closure period, the waste package will be kept dry by ventilation air. During the thermal pulse, heat generated by radioactive decay will eventually increase the temperature of the waste package, drip shield and drift wall to a level above the boiling point, where the probability of seepage into drifts will become insignificant. Further heating will push the waste package surface temperature above the deliquescence point of expected salt mixtures, thereby preventing the formation of deliquescence brines from dust deposits and humid air. Phase and time-temperature-transformation diagrams predicted for Alloy 22, and validated with experimental data, indicates no significant phase instabilities (LRO and TCP precipitation) at temperatures below 300 C for 10,000 years. Neither will dry oxidation at these elevated temperatures limit waste package life. After the peak temperature is reached, the waste package will begin to cool, eventually reaching a point where deliquescence brine formation may occur. However, corrosion testing of Alloy 22 underneath such films has shown no evidence of life-limiting localized corrosion. Transition Region (120 C {ge} T {ge} 100 C; 400 to 1,000 Years): During continued cooling, the temperature of the drift wall will drop to a level close to the boiling point of the seepage brine, thus permitting the onset of seepage. Corrosion in a concentrated, possibly aggressive, liquid-phase brine, evolved through evaporative concentration, is possible while in this region. However, based upon chemical divide theory, most ({ge} 99%) of the seepage water entering the

  5. Scenarios study on post-consumer plastic packaging waste recycling

    NARCIS (Netherlands)

    Thoden van Velzen, E.U.; Bos-Brouwers, H.E.J.; Groot, J.J.; Bing Xiaoyun, Xiaoyun; Jansen, M.; Luijsterburg, B.

    2013-01-01

    We all use plastics on a daily basis. Plastics come in many shapes, sizes and compositions and are used in a wide variety of products. Almost all of the currently used plastic packaging are made from fossil resources, which are finite. The production of plastic packages causes environmental impacts,

  6. Tritium Permeation Estimate from APT and CLWR-TEF Waste Packages

    International Nuclear Information System (INIS)

    The amount of tritium permeating out of waste containers has been estimated for the Accelerator Production of Tritium project (APT) and for the Commercial Light Water Reactor - Tritium Extraction Facility project (CLWR-TEF). The waste packages analyzed include the Aluminum, Window, Tungsten, Lead, and Steel packages for the APT project, and the overpack of extracted Tritium Producing Burnable Absorber Rods (TPBARs) for the CLWR-TEF project. All of the tritium contained in the waste was assumed to be available as a gas in the free volume inside the waste container at the beginning of disposal, and to then permeate the stainless steel waste container. From estimates of the tritium content of each waste form, the void or free volume of the package, disposal temperature and container geometry, the amount of tritium exiting the waste container by permeation was calculated. Two tritium permeation paths were considered separately: through the entire wall surface area and through the weld area only, the weld area having reduced thickness and significantly less surface area compared to the wall area. Permeation out of the five APT waste containers at 50 degrees Celsius is mainly through the welds, and at 100 degrees Celsius is through the permeation out of the entire wall surface area. The largest maximum offgas rate from an APT waste stream at 50 degrees Celsius (estimated disposal temperature) was 1.8E-6 Ci/year from the weld of the Window waste package, and the smallest maximum offgas rate was 3.7E-5 Ci/year from the weld of the Lead waste package. Permeation from the CLWR-TEF overpack at 40 degrees Celsius is mainly through the entire wall surface area, with a maximum offgas rate of 1.3E-5 Ci/year

  7. Structural analysis methods and models used for the evaluation of a radioactive waste transport packaging

    International Nuclear Information System (INIS)

    This paper describes the structural analysis methods and models developed to analyze a packaging which was designed to transport transuranic radioactive waste from waste generating sites to a final repository. The packaging is a rectangular box consisting of two inner containers surrounded by polyurethane foam and an outer protective structure. The methods and models described are those developed to evaluate the effects on the packaging of the 30 ft. drop hypothetical accident as defined in Title 10 of the Code of Federal Regulation, 10CFR71. The packaging is a complex composite structure so several different models were required. The ANSYS finite element computer program was used extensively in the structural analysis. The approach selected was to use an overall system model to perform dynamic analyses to determine the packaging response for different drop orientations so that the worst orientation for the 30 ft drop could be determined

  8. Cleanup Verification Package for the 118-B-6, 108-B Solid Waste Burial Ground

    Energy Technology Data Exchange (ETDEWEB)

    M. L. Proctor

    2006-06-13

    This cleanup verification package documents completion of remedial action for the 118-B-6, 108-B Solid Waste Burial Ground. The 118-B-6 site consisted of 2 concrete pipes buried vertically in the ground and capped by a concrete pad with steel lids. The site was used for the disposal of wastes from the "metal line" of the P-10 Tritium Separation Project.

  9. Equations for predicting release rates for waste packages in unsaturated tuff

    International Nuclear Information System (INIS)

    Nuclear waste will be placed in the potential repository at Yucca Mountain in waste packages. Spent fuel assemblies or consolidated fuel rods and borosilicate glass in steel pour canisters will be enclosed in sealed containers. The waste package consists of the waste form, the cladding on spent fuel or the defense-waste pour canister, and the outside container. Current design calls for the waste packages to be surrounded by an air gap. Although the waste package is generally not seen as the primary barrier for nuclear waste isolation it must in fact meet specific regulatory requirements: substantially complete requirement and release-rate from the engineered barrier system [USNRC 1983]. This report gives derivations of equations for predicting releases rates. We consider the release of three types of species: solubility-limited species, species released congruent with solid-solid alteration of spent-fuel matrix or borosilicate glass, and readily soluble species from the fuel-cladding gap, gas plenum, and readily accessible grain boundaries. We develop analytic expressions for the release rates of individual constituents from each of these mechanisms. For a given species and for given parameters, the mechanism that results in the lowest predicted release rate is to be adopted as the rate-controlling mechanism for that species. Some of the equations are newly derived for this report, others are restated from earlier work. Release rates have been calculated for key radionuclides in a companion report. 11 refs., 7 figs

  10. Life cycle assessment of a packaging waste recycling system in Portugal

    International Nuclear Information System (INIS)

    Highlights: • We modeled a real packaging waste recycling system. • The analysis was performed using the life cycle assessment methodology. • The 2010 situation was compared with scenarios where the materials were not recycled. • The “Baseline” scenario seems to be more beneficial to the environment. - Abstract: Life Cycle Assessment (LCA) has been used to assess the environmental impacts associated with an activity or product life cycle. It has also been applied to assess the environmental performance related to waste management activities. This study analyses the packaging waste management system of a local public authority in Portugal. The operations of selective and refuse collection, sorting, recycling, landfilling and incineration of packaging waste were considered. The packaging waste management system in operation in 2010, which we called “Baseline” scenario, was compared with two hypothetical scenarios where all the packaging waste that was selectively collected in 2010 would undergo the refuse collection system and would be sent directly to incineration (called “Incineration” scenario) or to landfill (“Landfill” scenario). Overall, the results show that the “Baseline” scenario is more environmentally sound than the hypothetical scenarios

  11. Mixed waste chemical compatibility: A testing program for plastic packaging components

    International Nuclear Information System (INIS)

    The purpose of hazardous and radioactive materials packaging is to enable these materials to be transported without posing a threat to the health or property of the general public. To achieve this aim, regulations in the United States have been written establishing general design requirements for such packagings. While no regulations have been written specifically for mixed waste packaging, regulations for the constituents of mixed wastes, i.e., hazardous and radioactive substances, have been codified by the US Department of Transportation (DOT, 49 CFR 173) and the US Nuclear Regulatory Commission (NRC, 10 CFR 71). The design requirements for both hazardous [49 CFR 173.24 (e)(1)] and radioactive [49 CFR 173.412 (g)] materials packaging specify packaging compatibility, i.e., that the materials of the packaging at sign d any contents be chemically compatible with each other. Furthermore, Type A [49 CFR 173.412 (g)] and Type B (10 CFR 71.43) packaging design requirements stipulate that there be no significant chemical, galvanic, or other reaction between the materials and contents of the package. Based on these requirements, a Chemical Compatibility Testing Program was developed in the Transportation Systems Department at Sandia National Laboratories (SNL). The program attempts to assure any regulatory body that the issue of packaging material compatibility towards hazardous and radioactive materials has been addressed. This program has been described in considerable detail in an internal SNL document, the Chemical Compatibility Test Plan ampersand Procedure Report (Nigrey 1993)

  12. Synthesis of knowledge on the long-term behaviour of concretes. Applications to cemented waste packages

    International Nuclear Information System (INIS)

    As stipulated in the former law of December 91 relating to 'concrete waste package', a progress report (phenomenological reference document) was first provided in 1999. The objective was to make an assessment of the knowledge acquired on the long-term behaviour of cement-based waste packages in the context of deep disposal and/or interim storage. The present document is an updated summary report. It takes into account a new knowledge assessment, considers coupled mechanisms and should contribute to the first performance studies (operational calculations). Handling and radio-nuclides (RN) confinement are the two major functional properties requested from the concrete used for the waste packages. In unsaturated environment (interim storage/disposal prior to closing), the main problem is the generation of cracks in the material. This aspect is a key parameter from the mechanical point of view (retrievability). It can have a major impact on the disposal phase (confinement). In saturated environment (disposal post-closing phase), the main concern is the chemical degradation of the waste package concrete submitted to underground waters leaching. In this context, the major thema are: the durability of the concretes under water (chemical degradation) and in unsaturated medium (corrosion of reinforcement), matter transport, RN retention, chemistry / transport / mechanical couplings. On the other hand, laboratory data on the behaviour of concretes are used to evaluate the RN source term of waste packages in function of time (concrete waste package OPerational Model, i.e. 'Concrete MOP'). The 'MOP' provides the physico-chemical description of the RN release in relationship with the waste package degradation itself. This description is based on simplified phenomenology for which only dimensioning mechanisms are taken into account. The use of Diffu-Ca code (basic module for the MOP) on the CASTEM numerical plate-form, already allows operational predictions. (authors)

  13. Cumulative releases of radionuclides from uncontained waste packages

    International Nuclear Information System (INIS)

    This report describes mathematical predictions for the migration of radionuclides from an emplaced radioactive waste container. The model assumes a spherical-equivalent waste solid surrounded by backfill but neglects the effect of decay heat. 7 refs., 2 tabs

  14. Tabulation of thermodynamic data for chemical reactions involving 58 elements common to radioactive waste package systems

    International Nuclear Information System (INIS)

    The rate of release and migration of radionuclides from a nuclear waste repository to the biosphere is dependent on chemical interactions between groundwater, the geologic host rock, and the radioactive waste package. For the purpose of this report, the waste package includes the wasteform, canister, overpack, and repository backfill. Chemical processes of interest include sorption (ion exchange), dissolution, complexation, and precipitation. Thermochemical data for complexation and precipitation calculations for 58 elements common to the radioactive waste package are presented. Standard free energies of formation of free ions, complexes, and solids are listed. Common logarithms of equilibrium constants (log K's) for speciation and precipitation reactions are listed. Unless noted otherwise, all data are for 298.150K and one atmosphere

  15. Guidelines for the development and testing of NWTS waste-package materials

    International Nuclear Information System (INIS)

    The purpose of this document is to provide guidelines to the NWTS projects for the testing of waste package materials. The information contained herein is provided in detail to describe the required development and testing to qualify materials for use in the waste package. These materials include the waste form, structural and corrosion-resistant barriers, and backfills to be placed around the canister and overpack. The guidelines include a description of methods, procedures, and test conditions. Each potential geologic site will use the guidelines to aid in selecting specific tests to qualify materials for that site. Thus, each NWTS project must develop specific test programs to meet its requirements. The guidelines are provided as a documented description of the test methods and procedures that are available to qualify and select materials for waste packages for a variety of geologic settings and host rocks

  16. Repository environmental parameters relevant to assessing the performance of high-level waste packages

    International Nuclear Information System (INIS)

    This document provides specifications for a model/methodology and approach that could be employed in determining postclosure repository environmental parameters relevant to high-level waste package performance for the Basalt Waste Isolation Project (BWIP). Guidance is provided on (1) the identity of the relevant repository environmental parameters (groundwater characteristics, temperature, radiation, and pressure), (2) the models/methodologies employed to determine the parameters, and (3) the input data base for the model/methodologies. Supporting studies included are (1) an analysis of potential waste package failure modes leading to identification of the relevant repository environmental parameters, (2) an evaluation of the credible range of the repository environmental parameters for the BWIP situation, and (3) a summary review of existing models/methodologies currently employed in determining repository environmental parameters relevant to waste package performance. 7 refs

  17. Waste package materials testing for a salt repository: 1983 status summary report

    International Nuclear Information System (INIS)

    The United States plans to safely dispose of nuclear waste in deep, stable geologic formations. As part of these plans, the US Department of Energy is sponsoring research on the designing and testing of waste packages and waste package materials. This fiscal year 1983 status report summarizes recent results of waste package materials testing in a salt environment. The results from these tests will be used by waste package designers and performance assessment experts. Release characteristics data are available on two waste forms (spent fuel and waste-containing glass) that were exposed to leaching tests at various radiation levels, temperatures, pH, glass surface area to solution volume ratios, and brine solutions simulating expected salt repository conditions. Candidate materials tested for corrosion resistance and other properties include iron alloys; TI-CODE 12, the most promising titanium alloy for containment; and nickel alloys. In component interaction testing, synergistic effects have not ruled out any candidate material. 21 refs., 37 figs., 15 tabs

  18. A Fruit of Yucca Mountain: The Remote Waste Package Closure System

    Energy Technology Data Exchange (ETDEWEB)

    Kevin Skinner; Greg Housley; Colleen Shelton-Davis

    2011-11-01

    Was the death of the Yucca Mountain repository the fate of a technical lemon or a political lemon? Without caution, this debate could lure us away from capitalizing on the fruits of the project. In March 2009, Idaho National Laboratory (INL) successfully demonstrated the Waste Package Closure System, a full-scale prototype system for closing waste packages that were to be entombed in the now abandoned Yucca Mountain repository. This article describes the system, which INL designed and built, to weld the closure lids on the waste packages, nondestructively examine the welds using four different techniques, repair the welds if necessary, mitigate crack initiating stresses in the surfaces of the welds, evacuate and backfill the packages with an inert gas, and perform all of these tasks remotely. As a nation, we now have a proven method for securely sealing nuclear waste packages for long term storage—regardless of whether or not the future destination for these packages will be an underground repository. Additionally, many of the system’s features and concepts may benefit other remote nuclear applications.

  19. Evaluation of low and intermediate level radioactive solidified waste forms and packages

    International Nuclear Information System (INIS)

    Evaluation of low and intermediate level radioactive waste forms and packages with respect to compliance with quality and safety requirements for transport, interim storage and disposal has become a very important part of the radioactive waste management strategy in many countries. The evaluation of waste forms and packages provides precise basic data for regulatory bodies to establish safety requirements, and implement quality control and quality assurance procedures for radioactive waste management programmes. The requirements depend very much upon the disposal option selected, treatment technology used, waste form characteristics, package quality and other factors. The regulatory requirements can also influence the methodology of waste form/package evaluation together with selection and analysis of data for quality control and safety assurance. A coordinated research programme started at the end of 1985 and brought together 12 participants from 11 countries. The results of the programme and each particular project were discussed at three Research Coordination Meetings held in Cairo, Egypt, in May, 1986; in Beijing, China, in April, 1998; and at Harwell Laboratory, United Kingdom, in November, 1989. This document summarises the salient features and results achieved during the four year investigation and a recommendation for future work in this area. Refs, figs and tabs

  20. Determination of activation energy of pyrolysis of carton packaging wastes and its pure components using thermogravimetry.

    Science.gov (United States)

    Alvarenga, Larissa M; Xavier, Thiago P; Barrozo, Marcos Antonio S; Bacelos, Marcelo S; Lira, Taisa S

    2016-07-01

    Many processes have been used for recycling of carton packaging wastes. The pyrolysis highlights as a promising technology to be used for recovering the aluminum from polyethylene and generating products with high heating value. In this paper, a study on pyrolysis reactions of carton packaging wastes and its pure components was performed in order to estimate the kinetic parameters of these reactions. For this, dynamic thermogravimetric analyses were carried out and two different kinds of kinetic models were used: the isoconversional and Independent Parallel Reactions. Isoconversional models allowed to calculate the overall activation energy of the pyrolysis reaction, in according to their conversions. The IPR model, in turn, allowed the calculation of kinetic parameters of each one of the carton packaging and paperboard subcomponents. The carton packaging pyrolysis follows three separated stages of devolatilization. The first step is moisture loss. The second stage is perfectly correlated to devolatilization of cardboard. The third step is correlated to devolatilization of polyethylene. PMID:27156364

  1. Packaging waste recycling in Europe: is the industry paying for it?

    Science.gov (United States)

    da Cruz, Nuno Ferreira; Ferreira, Sandra; Cabral, Marta; Simões, Pedro; Marques, Rui Cunha

    2014-02-01

    This paper describes and examines the schemes established in five EU countries for the recycling of packaging waste. The changes in packaging waste management were mainly implemented since the Directive 94/62/EC on packaging and packaging waste entered into force. The analysis of the five systems allowed the authors to identify very different approaches to cope with the same problem: meet the recovery and recycling targets imposed by EU law. Packaging waste is a responsibility of the industry. However, local governments are generally in charge of waste management, particularly in countries with Green Dot schemes or similar extended producer responsibility systems. This leads to the need of establishing a system of financial transfers between the industry and the local governments (particularly regarding the extra costs involved with selective collection and sorting). Using the same methodological approach, the authors also compare the costs and benefits of recycling from the perspective of local public authorities for France, Portugal and Romania. Since the purpose of the current paper is to take note of who is paying for the incremental costs of recycling and whether the industry (i.e. the consumer) is paying for the net financial costs of packaging waste management, environmental impacts are not included in the analysis. The work carried out in this paper highlights some aspects that are prone to be improved and raises several questions that will require further research. In the three countries analyzed more closely in this paper the industry is not paying the net financial cost of packaging waste management. In fact, if the savings attained by diverting packaging waste from other treatment (e.g. landfilling) and the public subsidies to the investment on the "recycling system" are not considered, it seems that the industry should increase the financial support to local authorities (by 125% in France, 50% in Portugal and 170% in Romania). However, in France and

  2. Packaging waste recycling in Europe: Is the industry paying for it?

    International Nuclear Information System (INIS)

    Highlights: • We study the recycling schemes of France, Germany, Portugal, Romania and the UK. • The costs and benefits of recycling are compared for France, Portugal and Romania. • The balance of costs and benefits depend on the perspective (strictly financial/economic). • Financial supports to local authorities ought to promote cost-efficiency. - Abstract: This paper describes and examines the schemes established in five EU countries for the recycling of packaging waste. The changes in packaging waste management were mainly implemented since the Directive 94/62/EC on packaging and packaging waste entered into force. The analysis of the five systems allowed the authors to identify very different approaches to cope with the same problem: meet the recovery and recycling targets imposed by EU law. Packaging waste is a responsibility of the industry. However, local governments are generally in charge of waste management, particularly in countries with Green Dot schemes or similar extended producer responsibility systems. This leads to the need of establishing a system of financial transfers between the industry and the local governments (particularly regarding the extra costs involved with selective collection and sorting). Using the same methodological approach, the authors also compare the costs and benefits of recycling from the perspective of local public authorities for France, Portugal and Romania. Since the purpose of the current paper is to take note of who is paying for the incremental costs of recycling and whether the industry (i.e. the consumer) is paying for the net financial costs of packaging waste management, environmental impacts are not included in the analysis. The work carried out in this paper highlights some aspects that are prone to be improved and raises several questions that will require further research. In the three countries analyzed more closely in this paper the industry is not paying the net financial cost of packaging waste

  3. Packaging waste recycling in Europe: is the industry paying for it?

    Science.gov (United States)

    da Cruz, Nuno Ferreira; Ferreira, Sandra; Cabral, Marta; Simões, Pedro; Marques, Rui Cunha

    2014-02-01

    This paper describes and examines the schemes established in five EU countries for the recycling of packaging waste. The changes in packaging waste management were mainly implemented since the Directive 94/62/EC on packaging and packaging waste entered into force. The analysis of the five systems allowed the authors to identify very different approaches to cope with the same problem: meet the recovery and recycling targets imposed by EU law. Packaging waste is a responsibility of the industry. However, local governments are generally in charge of waste management, particularly in countries with Green Dot schemes or similar extended producer responsibility systems. This leads to the need of establishing a system of financial transfers between the industry and the local governments (particularly regarding the extra costs involved with selective collection and sorting). Using the same methodological approach, the authors also compare the costs and benefits of recycling from the perspective of local public authorities for France, Portugal and Romania. Since the purpose of the current paper is to take note of who is paying for the incremental costs of recycling and whether the industry (i.e. the consumer) is paying for the net financial costs of packaging waste management, environmental impacts are not included in the analysis. The work carried out in this paper highlights some aspects that are prone to be improved and raises several questions that will require further research. In the three countries analyzed more closely in this paper the industry is not paying the net financial cost of packaging waste management. In fact, if the savings attained by diverting packaging waste from other treatment (e.g. landfilling) and the public subsidies to the investment on the "recycling system" are not considered, it seems that the industry should increase the financial support to local authorities (by 125% in France, 50% in Portugal and 170% in Romania). However, in France and

  4. Survey of waste package designs for disposal of high-level waste/spent fuel in selected foreign countries

    International Nuclear Information System (INIS)

    This report presents the results of a survey of the waste package strategies for seven western countries with active nuclear power programs that are pursuing disposal of spent nuclear fuel or high-level wastes in deep geologic rock formations. Information, current as of January 1989, is given on the leading waste package concepts for Belgium, Canada, France, Federal Republic of Germany, Sweden, Switzerland, and the United Kingdom. All but two of the countries surveyed (France and the UK) have developed design concepts for their repositories, but none of the countries has developed its final waste repository or package concept. Waste package concepts are under study in all the countries surveyed, except the UK. Most of the countries have not yet developed a reference concept and are considering several concepts. Most of the information presented in this report is for the current reference or leading concepts. All canisters for the wastes are cylindrical, and are made of metal (stainless steel, mild steel, titanium, or copper). The canister concepts have relatively thin walls, except those for spent fuel in Sweden and Germany. Diagrams are presented for the reference or leading concepts for canisters for the countries surveyed. The expected lifetimes of the conceptual canisters in their respective disposal environment are typically 500 to 1,000 years, with Sweden's copper canister expected to last as long as one million years. Overpack containers that would contain the canisters are being considered in some of the countries. All of the countries surveyed, except one (Germany) are currently planning to utilize a buffer material (typically bentonite) surrounding the disposal package in the repository. Most of the countries surveyed plan to limit the maximum temperature in the buffer material to about 100 degree C. 52 refs., 9 figs

  5. Survey of waste package designs for disposal of high-level waste/spent fuel in selected foreign countries

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, K.J.; Lakey, L.T.; Silviera, D.J.

    1989-09-01

    This report presents the results of a survey of the waste package strategies for seven western countries with active nuclear power programs that are pursuing disposal of spent nuclear fuel or high-level wastes in deep geologic rock formations. Information, current as of January 1989, is given on the leading waste package concepts for Belgium, Canada, France, Federal Republic of Germany, Sweden, Switzerland, and the United Kingdom. All but two of the countries surveyed (France and the UK) have developed design concepts for their repositories, but none of the countries has developed its final waste repository or package concept. Waste package concepts are under study in all the countries surveyed, except the UK. Most of the countries have not yet developed a reference concept and are considering several concepts. Most of the information presented in this report is for the current reference or leading concepts. All canisters for the wastes are cylindrical, and are made of metal (stainless steel, mild steel, titanium, or copper). The canister concepts have relatively thin walls, except those for spent fuel in Sweden and Germany. Diagrams are presented for the reference or leading concepts for canisters for the countries surveyed. The expected lifetimes of the conceptual canisters in their respective disposal environment are typically 500 to 1,000 years, with Sweden's copper canister expected to last as long as one million years. Overpack containers that would contain the canisters are being considered in some of the countries. All of the countries surveyed, except one (Germany) are currently planning to utilize a buffer material (typically bentonite) surrounding the disposal package in the repository. Most of the countries surveyed plan to limit the maximum temperature in the buffer material to about 100{degree}C. 52 refs., 9 figs.

  6. Scale-up considerations relevant to experimental studies of nuclear waste-package behavior

    International Nuclear Information System (INIS)

    Results from a study that investigated whether testing large-scale nuclear waste-package assemblages was technically warranted are reported. It was recognized that the majority of the investigations for predicting waste-package performance to date have relied primarily on laboratory-scale experimentation. However, methods for the successful extrapolation of the results from such experiments, both geometrically and over time, to actual repository conditions have not been well defined. Because a well-developed scaling technology exists in the chemical-engineering discipline, it was presupposed that much of this technology could be applicable to the prediction of waste-package performance. A review of existing literature documented numerous examples where a consideration of scaling technology was important. It was concluded that much of the existing scale-up technology is applicable to the prediction of waste-package performance for both size and time extrapolations and that conducting scale-up studies may be technically merited. However, the applicability for investigating the complex chemical interactions needs further development. It was recognized that the complexity of the system, and the long time periods involved, renders a completely theoretical approach to performance prediction almost hopeless. However, a theoretical and experimental study was defined for investigating heat and fluid flow. It was concluded that conducting scale-up modeling and experimentation for waste-package performance predictions is possible using existing technology. A sequential series of scaling studies, both theoretical and experimental, will be required to formulate size and time extrapolations of waste-package performance

  7. Scale-up considerations relevant to experimental studies of nuclear waste-package behavior

    Energy Technology Data Exchange (ETDEWEB)

    Coles, D.G.; Peters, R.D.

    1986-04-01

    Results from a study that investigated whether testing large-scale nuclear waste-package assemblages was technically warranted are reported. It was recognized that the majority of the investigations for predicting waste-package performance to date have relied primarily on laboratory-scale experimentation. However, methods for the successful extrapolation of the results from such experiments, both geometrically and over time, to actual repository conditions have not been well defined. Because a well-developed scaling technology exists in the chemical-engineering discipline, it was presupposed that much of this technology could be applicable to the prediction of waste-package performance. A review of existing literature documented numerous examples where a consideration of scaling technology was important. It was concluded that much of the existing scale-up technology is applicable to the prediction of waste-package performance for both size and time extrapolations and that conducting scale-up studies may be technically merited. However, the applicability for investigating the complex chemical interactions needs further development. It was recognized that the complexity of the system, and the long time periods involved, renders a completely theoretical approach to performance prediction almost hopeless. However, a theoretical and experimental study was defined for investigating heat and fluid flow. It was concluded that conducting scale-up modeling and experimentation for waste-package performance predictions is possible using existing technology. A sequential series of scaling studies, both theoretical and experimental, will be required to formulate size and time extrapolations of waste-package performance.

  8. Long term governance for radioactive waste management. Final report of Cowan2 - work package 4

    International Nuclear Information System (INIS)

    This report aims at identifying key features for the long term governance of radioactive waste. It is proposed by the COWAN2 Work Package 4 the purpose of which was to identify, discuss and analyse the institutional, ethical, economic and legal considerations raised by long term radioactive waste storage or disposal on the three interrelated issues of: responsibility and ownership of radioactive waste on the long term, continuity of local dialogue between stakeholders and monitoring of radioactive waste management facilities, and compensation and sustainable development. The aim is also to propose guidelines in order to better address long term issues in decision-making processes and start long term governance

  9. Nuclear waste package corrosion behavior in the proposed Yucca Mountain repository

    International Nuclear Information System (INIS)

    The corrosion performance of spent nuclear fuel waste packages is becoming increasingly important in establishing the viability of the proposed Yucca Mountain repository system. Current package concepts propose the use of a 2 cm thick nickel-base superalloy (Alloy 22) shell as the main barrier to prevent corrosion penetration over many thousands of years. The expected package service conditions, as well as their variability and uncertainty, are discussed. The electrochemical conditions known to be responsible for passive behavior and its breakdown in Alloy 22 and similar alloys are examined in the light of the predicted repository environment. Durability prediction approaches and their conclusions are considered. Efforts to determine the relative impact of localized modes of failure and uniform passive dissolution on package durability are reviewed, along with open issues in need of resolution and alternative package designs. The basic question of the validity of extrapolating corrosion behavior over many times the duration of the present base of experience is addressed

  10. An initial appraisal of destructive testing of radioactive waste packages

    International Nuclear Information System (INIS)

    An initial appraisal of destructive testing techniques has been undertaken, as part of the Department of the Environment initiative to establish an independent Quality Checking Facility for LLW and ILW packages. The reasons for destructive testing are discussed, and examples of chemical separation and analyses techniques are given. (author)

  11. Safety Test Planning for a Type B Packaging of Radioactive Waste

    International Nuclear Information System (INIS)

    KHNP-NETEC has been developing a new Type B packaging for on-site transport of radioactive waste. The packaging is designed to meet the requirements of IAEA and Korean regulations. Demonstration of compliance with the performance standards required in the regulations must be accomplished by analyses or safety tests or by a combination thereof. Analyses on the packaging for normal transport and accident conditions were completed using computer programs. In order to verify the analysis results and demonstrate the compliance with the regulatory requirements, a safety test planning was established and the safety tests on a prototype model will be carried out

  12. SECOND WASTE PACKAGE PROBABILISTIC CRITICALITY ANALYSIS: GENERATION AND EVALUATION OF INTERNAL CRITICIALITY CONFIGURATIONS

    Energy Technology Data Exchange (ETDEWEB)

    P. Gottlieb, J.R. Massari, J.K. McCoy

    1996-03-27

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development (WPD) department to provide an evaluation of the criticality potential within a waste package having sonic or all of its contents degraded by corrosion and removal of neutron absorbers. This analysis is also intended to provide an estimate of the consequences of any internal criticality, particularly in terms of any increase in radionuclide inventory. These consequence estimates will be used as part of the WPD input to the Total System Performance Assessment. The ultimate objective of this analysis is to augment the information gained from the Initial Waste Package Probabilistic Criticality Analyses (Ref. 5.8 and 5.9, hereafter referred to as IPA) to a degree which will support preliminary waste package design recommendations intended to reduce the risk of waste package criticality and the risk to total repository system performance posed by the consequences of any criticality. The IPA evaluated the criticality potential under the assumption that the waste package basket retained its structural integrity, so that the assemblies retained their initial separation, even when the neutron absorbers had been leached from the basket. This analysis is based on the more realistic condition that removal of the neutron absorbers is a consequence of the corrosion of the steel in which they are contained, which has the additional consequence of reducing the structural support between assemblies. The result is a set of more reactive configurations having a smaller spacing between assemblies, or no inter-assembly spacing at all. Another difference from the IPA is the minimal attention to probabilistic evaluation given in this study. Although the IPA covered a time horizon to 100,000 years, the lack of consideration of basket degradation modes made it primarily applicable to the first 10,000 years. In contrast, this study, by focusing on the degraded modes of the basket, is primarily

  13. Post-closure performance assessment of waste packages for the Yucca Mountain Project

    International Nuclear Information System (INIS)

    This report details a system model of some core features of the performance of waste packages for the permanent disposal of spent nuclear fuel at the Yucca Mountain Site. The model is realized in the prototype computer program PANDORA-1.1. The PANDORA system model links processes leading to possible release of radionuclides from the waste package. The PANDORA submodels are being developed for processes and conditions specific to this potential repository site, notably the comparatively dry location in an arid area and well above the groundwater table, and the rock medium of porous partially welded tuff

  14. Probablistic Analyses of Waste Package Quantities Impacted by Potential Igneous Disruption at Yucca Mountain

    Science.gov (United States)

    Wallace, M. G.; Iuzzolina, H.

    2005-12-01

    A probabilistic analysis was conducted to estimate ranges for the numbers of waste packages that could be damaged in a potential future igneous event through a repository at Yucca Mountain. The analysis includes disruption from an intrusive igneous event and from an extrusive volcanic event. This analysis supports the evaluation of the potential consequences of future igneous activity as part of the total system performance assessment for the license application for the Yucca Mountain Project (YMP). The first scenario, igneous intrusion, investigated the case where one or more igneous dikes intersect the repository. A swarm of dikes was characterized by distributions of length, width, azimuth, and number of dikes and the spacings between them. Through the use in part of a latin hypercube simulator and a modified video game engine, mathematical relationships were built between those parameters and the number of waste packages hit. Corresponding cumulative distribution function curves (CDFs) for the number of waste packages hit under several different scenarios were calculated. Variations in dike thickness ranges, as well as in repository magma bulkhead positions were examined through sensitivity studies. It was assumed that all waste packages in an emplacement drift would be impacted if that drift was intersected by a dike. Over 10,000 individual simulations were performed. Based on these calculations, out of a total of over 11,000 planned waste packages distributed over an area of approximately 5.5 km2 , the median number of waste packages impacted was roughly 1/10 of the total. Individual cases ranged from 0 waste packages to the entire inventory being impacted. The igneous intrusion analysis involved an explicit characterization of dike-drift intersections, built upon various distributions that reflect the uncertainties associated with the inputs. The second igneous scenario, volcanic eruption (eruptive conduits), considered the effects of conduits formed in

  15. Calculation of the Naval Long and Short Waste Package Three-Dimensional Thermal Interface Temperatures

    International Nuclear Information System (INIS)

    The purpose of this calculation is to evaluate the thermal performance of the Naval Long and Naval Short spent nuclear fuel (SNF) waste packages (WP) in the repository emplacement drift. The scope of this calculation is limited to the determination of the temperature profiles upon the surfaces of the Naval Long and Short SNF waste package for up to 10,000 years of emplacement. The temperatures on the top of the outside surface of the naval canister are the thermal interfaces for the Naval Nuclear Propulsion Program (NNPP). The results of this calculation are intended to support Licensing Application design activities

  16. Natural additives and agricultural wastes in biopolymer formulations for food packaging

    Science.gov (United States)

    Valdés, Arantzazu; Mellinas, Ana Cristina; Ramos, Marina; Garrigós, María Carmen; Jiménez, Alfonso

    2014-02-01

    The main directions in food packaging research are targeted towards improvements in food quality and food safety. For this purpose, food packaging providing longer product shelf-life, as well as the monitoring of safety and quality based upon international standards, is desirable. New active packaging strategies represent a key area of development in new multifunctional materials where the use of natural additives and/or agricultural wastes is getting increasing interest. The development of new materials, and particularly innovative biopolymer formulations, can help to address these requirements and also with other packaging functions such as: food protection and preservation, marketing and smart communication to consumers. The use of biocomposites for active food packaging is one of the most studied approaches in the last years on materials in contact with food. Applications of these innovative biocomposites could help to provide new food packaging materials with improved mechanical, barrier, antioxidant and antimicrobial properties. From the food industry standpoint, concerns such as the safety and risk associated with these new additives, migration properties and possible human ingestion and regulations need to be considered. The latest innovations in the use of these innovative formulations to obtain biocomposites are reported in this review. Legislative issues related to the use of natural additives and agricultural wastes in food packaging systems are also discussed.

  17. Consumption and recovery of packaging waste in Germany in 2009; Aufkommen und Verwertung von Verpackungsabfaellen in Deutschland im Jahr 2009

    Energy Technology Data Exchange (ETDEWEB)

    Schueler, Kurt [GVM Gesellschaft fuer Verpackungsmarktforschung mbH, Mainz (Germany)

    2012-04-15

    Pursuant to EU Directive 94/62/EC on packaging and packaging waste dated 20.12.1994 in connection with Directive 2004/12/EC, EU Member States are obliged to report annually on the consumption and recovery of packaging. This report shall be prepared on the basis of the Commission's decision of 22.03.2005 on establishing mandatory table formats (2005/270/EC). The study determines the quantity of packaging (packaging consumption) for the material groups of glass, plastics, paper, aluminium, tin plate, composites, other steel, wood and other packaging materials placed on the market in Germany. In addition to the quantity of packaging used in Germany, filled exports and imports were also ascertained in order to calculate the consumption rate. The quantity of packaging waste of waste relevance in Germany was calculated on the basis of the quantity of packaging placed on the market as e.g. reusable and durable packaging will only be discarded at some point in the future. All existing data from associations, the waste disposal industry and environmental statistics were compiled and documented systematically in order to determine the recovery quantities and recovery paths. The quantities incinerated at waste incineration plants with energy recovery could only be calculated as the difference between the total quantity to be discarded and quantities actually recovered. In 2008, 15.05 million tons of packaging were consumed and became waste. Compared to the reference year 2008, packaging consumption decreased by 6.2 %. A total of 12.73 million tons was recovered in terms of material or energy, of which a total of 2.45 million tons outside Germany. In addition, 1.42 million tons of imported packaging waste were recovered in Germany. In 2009, 1.55 million tons were incinerated at waste incineration plants with energy recovery.

  18. Consumption and recovery of packaging waste in Germany in 2008; Aufkommen und Verwertung von Verpackungsabfaellen in Deutschland im Jahr 2008

    Energy Technology Data Exchange (ETDEWEB)

    Schueler, Kurt [Gesellschaft fuer Verpackungsmarktforschung mbH, Mainz (Germany)

    2010-12-15

    Pursuant to EU Directive 94/62/EC on packaging and packaging waste dated 20.12.1994 in connection with Directive 2004/12/EC, EU Member States are obliged to report annually on the consumption and recovery of packaging. This report shall be prepared on the basis of the Commission's decision of 22.03.2005 on establishing mandatory table formats (2005/270/EC). The study determines the quantity of packaging (packaging consumption) for the material groups of glass, plastics, paper, aluminium, tin plate, composites, other steel, wood and other packaging materials placed on the market in Germany. In addition to the quantity of packaging used in Germany, filled exports and imports were also ascertained in order to calculate the consumption rate. The quantity of packaging waste of waste relevance in Germany was calculated on the basis of the quantity of packaging placed on the market as e.g. reusable and durable packaging will only be discarded at some point in the future. All existing data from associations, the waste disposal industry and environmental statistics were compiled and documented systematically in order to determine the recovery quantities and recovery paths. The quantities incinerated at waste incineration plants with energy recovery could only be calculated as the difference between the total quantity to be discarded and quantities actually recovered. In 2008, 16.04 million tons of packaging were consumed and became waste. Compared to the reference year 2005, packaging consumption increased by 3.7 % (minus 0.4 % compared to 2007). A total of 13.10 million tons was recovered in terms of material or energy, of which a total of 2.41 million tons outside Germany. In addition, 1.40 million tons of imported packaging waste were recovered in Germany. In 2008, 2.10 million tons were incinerated at waste incineration plants with energy recovery. (orig.)

  19. Dose Rate Calucaltion for the DHL W/DOE SNF Codisposal Waste Package

    Energy Technology Data Exchange (ETDEWEB)

    G. Radulescu

    2000-02-12

    The purpose of this calculation is to determine the surface dose rates of the short codisposal waste package (WP) of defense high-level waste (DHLW) and TRIGA (Training, Research, Isotopes, General Atomics) spent nuclear fuel (SNF). The WP contains the TRIGA SNF, in a standardized 18-in. DOE (U.S. Department of Energy) SNF canister, and five 3-m-long Savannah River Site (SRS) DHLW pour glass canisters, which surround the DOE SNF canister.

  20. STRUCTURAL CALCULATIONS FOR THE CODISPOSAL OF TRIGA SPENT NUCLEAR FUEL IN A WASTE PACKAGE

    Energy Technology Data Exchange (ETDEWEB)

    S. Mastilovic

    1999-07-28

    The purpose of this analysis is to determine the structural response of a TRIGA Department of Energy (DOE) spent nuclear fuel (SNF) codisposal canister placed in a 5-Defense High Level Waste (DHLW) waste package (WP) and subjected to a tipover design basis event (DBE) dynamic load; the results will be reported in terms of displacements and stress magnitudes. This activity is associated with the WP design.

  1. Annotated bibliography for the design of waste packages for geologic disposal of spent fuel and high-level waste

    International Nuclear Information System (INIS)

    This bibliography identifies documents that are pertinent to the design of waste packages for geologic disposal of nuclear waste. The bibliography is divided into fourteen subject categories so that anyone wishing to review the subject of leaching, for example, can turn to the leaching section and review the abstracts of reports which are concerned primarily with leaching. Abstracts are also cross referenced according to secondary subject matter so that one can get a complete list of abstracts for any of the fourteen subject categories. All documents which by their title alone appear to deal with the design of waste packages for the geologic disposal of spent fuel or high-level waste were obtained and reviewed. Only those documents which truly appear to be of interest to a waste package designer were abstracted. The documents not abstracted are listed in a separate section. There was no beginning date for consideration of a document for review. About 1100 documents were reviewed and about 450 documents were abstracted

  2. 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

  3. Evaluation and compilation of DOE waste package test data: Biannual report, August 1986-January 1987

    Energy Technology Data Exchange (ETDEWEB)

    Interrante, C.; Escalante, E.; Fraker, A.; Harrison, S.; Shull, R.; Linzer, M.; Ricker, R.; Ruspi, J.

    1987-10-01

    This report summarizes results of the National Bureau of Standards (NBS) evaluations of Department of Energy (DOE) activities on waste packages designed for containment of radioactive high-level nuclear waste (HLW). The waste package is a proposed engineered barrier that is part of a permanent repository for HLW. Metal alloys are the principal barriers within the engineered system. Technical discussions are given for the corrosion of metals proposed for the canister, particularly carbon and stainless steels, and copper. In the section on tuff, the current level of understanding of several canister materials is questioned. Within the Basalt Waste Isolation Project (BWIP) section, discussions are given on problems concerning groundwater, materials for use in the metallic overpack, and diffusion through the packing. For the proposed salt site, questions are raised on the work on both ASTM A216 Steel and Ti-Code 12. NBS work related to the vitrification of HLW borosilicate glass at the West Valley Demonstration Project (WVDP) and the Defense Waste Processing Facility (DWPF) is covered. NBS reviews of selected DOE technical reports and a summary of current waste-package activities of the Materials Characterization Center (MCC) is presented. Using a database management system, a computerized database for storage and retrieval of reviews and evaluations of HLW data has been developed and is described. 17 refs., 2 figs., 2 tabs.

  4. Investigation of the behaviour of packaged radioactive waste under fire accident conditions

    International Nuclear Information System (INIS)

    A study has been made of the behaviour of packaged intermediate level waste (ILW) when exposed to fire conditions so as to provide information to support safety cases for ILW transport and disposal. The temperatures used in the study were selected to exceed those that the waste might be subject to in fire accidents during the transport and handling of ILW. Four waste materials, immobilised in cement or in organic resin, with properties representative of a wide range of waste streams were included in the study. Tests were carried out on samples of both real waste materials and non-radioactive simulants, and also on full-scale (500 litre) drums of simulant wastes. The overall release fractions were low, even for external temperatures of up to 1000oC. Examination showed that the stainless steel drums were still in good condition and on sectioning, little damage to the matrix or decrease in its strength was evident. (author)

  5. Chemical compatibility screening results of plastic packaging to mixed waste simulants

    Energy Technology Data Exchange (ETDEWEB)

    Nigrey, P.J.; Dickens, T.G.

    1995-12-01

    We have developed a chemical compatibility program for evaluating transportation packaging components for transporting mixed waste forms. We have performed the first phase of this experimental program to determine the effects of simulant mixed wastes on packaging materials. This effort involved the screening of 10 plastic materials in four liquid mixed waste simulants. The testing protocol involved exposing the respective materials to {approximately}3 kGy of gamma radiation followed by 14 day exposures to the waste simulants of 60 C. The seal materials or rubbers were tested using VTR (vapor transport rate) measurements while the liner materials were tested using specific gravity as a metric. For these tests, a screening criteria of {approximately}1 g/m{sup 2}/hr for VTR and a specific gravity change of 10% was used. It was concluded that while all seal materials passed exposure to the aqueous simulant mixed waste, EPDM and SBR had the lowest VTRs. In the chlorinated hydrocarbon simulant mixed waste, only VITON passed the screening tests. In both the simulant scintillation fluid mixed waste and the ketone mixture simulant mixed waste, none of the seal materials met the screening criteria. It is anticipated that those materials with the lowest VTRs will be evaluated in the comprehensive phase of the program. For specific gravity testing of liner materials the data showed that while all materials with the exception of polypropylene passed the screening criteria, Kel-F, HDPE, and XLPE were found to offer the greatest resistance to the combination of radiation and chemicals.

  6. Installation in the A-1 plant of an experimental bituminization line by VUCHZ

    International Nuclear Information System (INIS)

    Following the termination of the experimental operation of the bituminization line at the Research Institute for Chemical Installations in Brno, the line was dismantled and transferred to the nucler power plant in Jaslovske Bohunice. The installation of the line, the layout of the assemblies are described and the results of tests with non-radioactive simulated wastes and actual radioactive wastes briefly described. An amount of 3.2 m3 of actual radioactive wastes from the V-1 nuclear power plant was processed in the tests. The results confirmed the suitability of bituminization for processing liquid radioactive wastes from WWER nuclear power plants. (Z.M.)

  7. 77 FR 50716 - Certain Food Waste Disposers and Components and Packaging Thereof; Notice of Commission...

    Science.gov (United States)

    2012-08-22

    ... COMMISSION Certain Food Waste Disposers and Components and Packaging Thereof; Notice of Commission... complainant Emerson Electric Co. of St. Louis, Missouri to amend the Notice of Investigation (``NOI'') and... investigation may be viewed on the Commission's electronic docket (EDIS) at http://edis.usitc.gov ....

  8. Advanced spent-fuel waste package fill material: Depleted uranium dioxide

    International Nuclear Information System (INIS)

    The use of depleted uranium dioxide (DUO2) particles has been investigated as fill material inside repository waste packages containing light water reactor (LWR) spent nuclear fuel (SNF). The use of DUO2 fill may eliminate repository criticality concerns, reduce radionuclide release rates from the repository, and dispose of excess depleted uranium

  9. Waste Form Release Data Package for the 2005 Integrated Disposal Facility Performance Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Pierce, Eric M.; McGrail, B. Peter; Rodriguez, Elsa A.; Schaef, Herbert T.; Saripalli, Prasad; Serne, R. Jeffrey; Krupka, Kenneth M.; Martin, P. F.; Baum, Steven R.; Geiszler, Keith N.; Reed, Lunde R.; Shaw, Wendy J.

    2004-09-01

    This data package documents the experimentally derived input data on the representative waste glasses; LAWA44, LAWB45, and LAWC22. This data will be used for Subsurface Transport Over Reactive Multi-phases (STORM) simulations of the Integrated Disposal Facility (IDF) for immobilized low-activity waste (ILAW). The STORM code will be used to provide the near-field radionuclide release source term for a performance assessment to be issued in July 2005. Documented in this data package are data related to 1) kinetic rate law parameters for glass dissolution, 2) alkali (Na+)-hydrogen (H+) ion exchange rate, 3) chemical reaction network of secondary phases that form in accelerated weathering tests, and 4) thermodynamic equilibrium constants assigned to these secondary phases. The kinetic rate law and Na+-H+ ion exchange rate were determined from single-pass flow-through experiments. Pressurized unsaturated flow (PUF) and product consistency (PCT) tests where used for accelerated weathering or aging of the glasses in order to determine a chemical reaction network of secondary phases that form. The majority of the thermodynamic data used in this data package were extracted from the thermody-namic database package shipped with the geochemical code EQ3/6, version 8.0. Because of the expected importance of 129I release from secondary waste streams being sent to IDF from various thermal treatment processes, parameter estimates for diffusional release and solubility-controlled release from cementitious waste forms were estimated from the available literature.

  10. 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

  11. Waste package performance assessment: Deterministic system model, program scope and specification

    International Nuclear Information System (INIS)

    Integrated assessments of the performance of nuclear waste package designs must be made in order to qualify waste package designs with respect to containment time and release-rate requirements. PANDORA is a computer-based model of the waste package and of the processes affecting it over the long terms, specific to conditions at the proposed Yucca Mountain, Nevada, site. The processes PANDORA models include: changes in inventories due to radioactive decay, gamma radiation dose rate in and near the package, heat transfer, mechanical behavior, groundwater contact, corrosion, waste form alteration, and radionuclide release. The model tracks the development and coupling of these processes over time. The process models are simplified ones that focus on major effects and on coupling. This report documents our conceptual model development and provides a specification for the computer program. The current model is the first in a series. Succeeding models will use guidance from results of preceding models in the PANDORA series and will incorporate results of recently completed experiments and calculations on processes affecting performance. 22 refs., 21 figs., 9 tabs

  12. Insight into economies of scale for waste packaging sorting plants

    DEFF Research Database (Denmark)

    Cimpan, Ciprian; Wenzel, Henrik; Maul, Anja;

    2015-01-01

    of economies of scale and discussed complementary relations occurring between capacity size, technology level and operational practice. Processing costs (capital and operational expenditure) per unit waste input were found to decrease from above 100 € for small plants with a basic technology level to 60...

  13. Effect of aged waste package and basalt on radioelement release

    International Nuclear Information System (INIS)

    Results of experiments are described that combine backfill, radioactive waste, and repository host rock in a single flowing groundwater stream in a manner analogous to a hydraulic breach of a waste repository. The experimental design is used to identify the chemical interactions that would occur if repository components were breached by flowing water. The results indicate that of three parameters studied, the alteration of the repository components as might occur upon aging had the most substantial influence on the migration of radioactive elements dissolved from the solid radioactive waste. The other two parameters, the metal alloy used in the apparatus and an ionizing radiation field imposed on the experimental apparatus, had little or no measurable effect on radioactive element transport by flowing water. Inasmuch as the alteration of the repository materials represent aging in an actual repository, it is concluded that changes with age may detrimentally affect the ability of a repository to isolate plutonium and neptunium, and possibly other radioactive elements in nuclear waste. 37 references, 2 figures, 2 tables

  14. Using Single-Camera 3-D Imaging to Guide Material Handling Robots in a Nuclear Waste Package Closure System

    International Nuclear Information System (INIS)

    Nuclear reactors for generating energy and conducting research have been in operation for more than 50 years, and spent nuclear fuel and associated high-level waste have accumulated in temporary storage. Preparing this spent fuel and nuclear waste for safe and permanent storage in a geological repository involves developing a robotic packaging system--a system that can accommodate waste packages of various sizes and high levels of nuclear radiation. During repository operation, commercial and government-owned spent nuclear fuel and high-level waste will be loaded into casks and shipped to the repository, where these materials will be transferred from the casks into a waste package, sealed, and placed into an underground facility. The waste packages range from 12 to 20 feet in height and four and a half to seven feet in diameter. Closure operations include sealing the waste package and all its associated functions, such as welding lids onto the container, filling the inner container with an inert gas, performing nondestructive examinations on welds, and conducting stress mitigation. The Idaho National Laboratory is designing and constructing a prototype Waste Package Closure System (WPCS). Control of the automated material handling is an important part of the overall design. Waste package lids, welding equipment, and other tools must be moved in and around the closure cell during the closure process. These objects are typically moved from tool racks to a specific position on the waste package to perform a specific function. Periodically, these objects are moved from a tool rack or the waste package to the adjacent glovebox for repair or maintenance. Locating and attaching to these objects with the remote handling system, a gantry robot, in a loosely fixtured environment is necessary for the operation of the closure cell. Reliably directing the remote handling system to pick and place the closure cell equipment within the cell is the major challenge

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

    International Nuclear Information System (INIS)

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

  16. Stochastic analysis of radioactive waste package performance using first-order reliability method

    International Nuclear Information System (INIS)

    In order to license an underground radioactive waste repository, it is important to demonstrate regulatory compliance with authoritative regulations. And it is evident from NRC's criteria for waste package performance that a stochastic analysis is necessary to provide that these criteria can be met with confidence. The first-order reliability method is an attractive approach to stochastic analysis and particularly useful when statistical information is incomplete, as is common for problems in the subsurface environment. Results from a first-order reliability analysis include an estimate of the probability of exceeding a specified performance criteria and measures of sensitivity of the stochastic solution to changes in random variables and their statistical moments. A method of stochastic analysis is illustrated by analyzing canister corrosion in a radioactive waste package

  17. Demands placed on waste package performance testing and modeling by some general results on reliability analysis

    International Nuclear Information System (INIS)

    Waste packages for a US nuclear waste repository are required to provide reasonable assurance of maintaining substantially complete containment of radionuclides for 300 to 1000 years after closure. The waiting time to failure for complex failure processes affecting engineered or manufactured systems is often found to be an exponentially-distributed random variable. Assuming that this simple distribution can be used to describe the behavior of a hypothetical single barrier waste package, calculations presented in this paper show that the mean time to failure (the only parameter needed to completely specify an exponential distribution) would have to be more than 107 years in order to provide reasonable assurance of meeting this requirement. With two independent barriers, each would need to have a mean time to failure of only 105 years to provide the same reliability. Other examples illustrate how multiple barriers can provide a strategy for not only achieving but demonstrating regulatory compliance

  18. The effects of gamma radiation on the corrosion of candidate materials for the fabrication of nuclear waste packages

    International Nuclear Information System (INIS)

    The influence of gamma radiation on the corrosion of candidate materials for the fabrication of nuclear waste packages has been comprehensively reviewed. The comparison of corrosion of the various materials was compared in three distinct environments: Environment A; Mg2+-enriched brines in which hydrolysis of the cation produces acidic environments and the Mg2+ interferes with the formation of protective films; Environment B; saline environments with a low Mg2+ content which remain neutral; Environment C; moist aerated conditions.The reference design of nuclear waste package for emplacement in the proposed waste repository in Yucca Mountain, Nevada, employs a dual wall arrangement, in which a 2 cm thick nickel alloy inner barrier is encapsulated within a 10 cm thick mild steel outer barrier. It is felt that this arrangement will give considerable containment lifetimes, since no common mode failure exists for the two barriers. The corrosion performance of this waste package will be determined by the exposure environment established within the emplacement drifts. Key features of the Yucca Mountain repository in controlling waste package degradation are expected to be the permanent availability of oxygen and the limited presence of water. When water contacts the surface of the waste package, its gamma radiolysis could produce an additional supply of corrosive agents. the gamma field will be produced by the radioactive decay of radionuclides within the waste form, and its magnitude will depend on the nature and age of the waste form as well as the material and wall thickness of the waste package

  19. Equilibrium moisture content of waste mixtures from post-consumer carton packaging.

    Science.gov (United States)

    Bacelos, M S; Freire, J T

    2012-01-01

    The manufacturing of boards and roof tiles is one of the routes to reuse waste from the recycled-carton-packaging process. Such a process requires knowledge of the hygroscopic behaviour of these carton-packaging waste mixtures in order to guarantee the quality of the final product (e.g. boards and roof tiles). Thus, with four carton-packaging waste mixtures of selected compositions (A, B, C and D), the sorption isotherms were obtained at air temperature of 20, 40 and 60 degrees C by using the static method. This permits one to investigate which model can relate the equilibrium moisture content of the mixture with that of a pure component through the mass fraction of each component in the mixtures. The results show that the experimental data can be well described by the weighted harmonic mean model. This suggests that the mean equilibrium moisture content of the carton-packaging mixture presents a non-linear relationship with each single, pure compound. PMID:22720401

  20. Effect of ionizing radiation on the waste package environment

    Energy Technology Data Exchange (ETDEWEB)

    Reed, D.T. [Argonne National Lab., IL (USA); Van Konynenburg, R.A. [Lawrence Livermore National Lab., CA (USA)

    1991-05-01

    The radiolytic production of nitrogen oxides, nitrogen acids and ammonia are discussed in relation to the expected environment in a high-level waste repository that may be constructed at the Yucca Mountain site if it is found to be suitable. Both literature data and repository-relevant data are summarized for air-water vapor systems. The limiting cases of a dry air and a pure water vapor gas phase are also discussed. Design guidelines and recommendations, based solely on the potential consequence of radiation enhancement of corrosion, are given. 13 refs., 5 figs., 1 tab.

  1. Development and evaluation of a tracer-injection hydrothermal technique for studies of waste package interactions

    International Nuclear Information System (INIS)

    A tracer-injection system has been developed for use in characterizing reactions of waste package materials under hydrothermal conditions. High-pressure liquid chromatographic instrumentation has been coupled with Dickson-type rocking autoclaves to allow injection of selected components into the hydrothermal fluid while maintaining run temperature and pressure. Hydrothermal experiments conducted using this system included the interactions of depleted uranium oxide and Zircaloy-4 metal alloy discs with trace levels of 99Tc and non-radioactive Cs and I in a simulated groundwater matrix. After waste-package components and simulated waste forms were pre-conditioned in the autoclave systems (usually 4 to 6 weeks), known quantities of tracer-doped fluids were injected into the autoclaves' gold reaction bag at run conditions. Time-sequenced sampling of the hydrothermal fluid providing kinetic data on the reactions of tracers with waste package materials. The injection system facilitates the design of experiments that will better define ''steady-state'' fluid compositions in hydrothermal reactions. The injection system will also allow for the formation of tracer-bearing solid phases in detectable quantities

  2. Prompt gamma neutron activation analysis of toxic elements in radioactive waste packages.

    Science.gov (United States)

    Ma, J-L; Carasco, C; Perot, B; Mauerhofer, E; Kettler, J; Havenith, A

    2012-07-01

    The French Alternative Energies and Atomic Energy Commission (CEA) and National Radioactive Waste Management Agency (ANDRA) are conducting an R&D program to improve the characterization of long-lived and medium activity (LL-MA) radioactive waste packages. In particular, the amount of toxic elements present in radioactive waste packages must be assessed before they can be accepted in repository facilities in order to avoid pollution of underground water reserves. To this aim, the Nuclear Measurement Laboratory of CEA-Cadarache has started to study the performances of Prompt Gamma Neutron Activation Analysis (PGNAA) for elements showing large capture cross sections such as mercury, cadmium, boron, and chromium. This paper reports a comparison between Monte Carlo calculations performed with the MCNPX computer code using the ENDF/B-VII.0 library and experimental gamma rays measured in the REGAIN PGNAA cell with small samples of nickel, lead, cadmium, arsenic, antimony, chromium, magnesium, zinc, boron, and lithium to verify the validity of a numerical model and gamma-ray production data. The measurement of a ∼20kg test sample of concrete containing toxic elements has also been performed, in collaboration with Forschungszentrum Jülich, to validate the model in view of future performance studies for dense and large LL-MA waste packages. PMID:22406218

  3. Destructive and non-destructive tests for radioactive waste packages Task 3 Characterization of radioactive waste forms. A series of final reports (1985-89) No 43

    International Nuclear Information System (INIS)

    On the basis of preliminary waste acceptance requirements quality control of radioactive waste has to be performed prior to interim storage or final disposal. The quality control can either be achieved by random tests on conditioned radioactive waste packages or by process qualification of the conditioning processes. One of the most important criteria is the activity of the radioactive waste product or packages. To get some first information on the waste package γ-spectrometric measurement is performed as non-destructive test. Besides the γ-emitting nuclides the α and β-emitting nuclides can be estimated by calculation if the waste was generated in nuclear power plants and the nuclide relations are known. If the non-destructive determination of nuclides is not sufficient or the non-radioactive content of the waste packages has to be identified sampling from the waste packages has to be performed. This can best be done by core drilling. To avoid the need of water for cooling the drill head, air cooled core drilling is investigated. As mixed wastes is not allowed for final disposal the determination of possible organic toxic materials like PCB, dioxin and furane-compounds in cemented wastes is conducted by GC-MS-investigations. For getting more knowledge in the field of process qualification concerning super compaction, instrumentation of the super compaction process is investigated and tested

  4. Evaluation on the structural soundness of the transport package for low-level radioactive waste for subsurface disposal against aircraft impact by finite element method

    International Nuclear Information System (INIS)

    The structural analysis of aircraft crush on the transport package for low-level radioactive waste was performed using the impact force which was already used for the evaluation of the high-level waste transport package by LSDYNA code. The transport package was deformed, and stresses due to the crush exceeded elastic range. However, plastic strains yieled in the package were far than the elongation of the materials and the body of the package did not contact the disposal packages due to the deformation of the package. Therefore, it was confirmed that the package keeps its integrity against aircraft crush. (author)

  5. Hanford low-level waste process chemistry testing data package

    International Nuclear Information System (INIS)

    Recently, the Tri-Party Agreement (TPA) among the State of Washington Department of Ecology, U.S. Department of Energy (DOE) and the US Environmental Protection Agency (EPA) for the cleanup of the Hanford Site was renegotiated. The revised agreement specifies vitrification as the encapsulation technology for low level waste (LLW). A demonstration, testing, and evaluation program underway at Westinghouse Hanford Company to identify the best overall melter-system technology available for vitrification of Hanford Site LLW to meet the TPA milestones. Phase I is a open-quotes proof of principleclose quotes test to demonstrate that a melter system can process a simulated highly alkaline, high nitrate/nitrite content aqueous LLW feed into a glass product of consistent quality. Seven melter vendors were selected for the Phase I evaluation: joule-heated melters from GTS Duratek, Incorporated (GDI); Envitco, Incorporated (EVI); Penberthy Electomelt, Incorporated (PEI); and Vectra Technologies, Incorporated (VTI); a gas-fired cyclone burner from Babcock ampersand Wilcox (BCW); a plasma torch-fired, cupola furnace from Westinghouse Science and Technology Center (WSTC); and an electric arc furnace with top-entering vertical carbon electrodes from the U.S. Bureau of Mines (USBM)

  6. Biologically-Induced Micropitting of Alloy 22, a Candidate Nuclear Waste Packaging Material

    International Nuclear Information System (INIS)

    The effects of potential microbiologically influenced corrosion (MIC) on candidate packaging materials for nuclear waste containment are being assessed. Coupons of Alloy 22, the outer barrier candidate for waste packaging, were exposed to a simulated, saturated repository environment (or microcosm) consisting of crushed rock (tuff) from the Yucca Mountain repository site and a continual flow of simulated groundwater for periods up to five years at room temperature and 30 C. Coupons were incubated with YM tuff under both sterile and non-sterile conditions. Surfacial analysis by scanning electron microscopy of the biotically-incubated coupons show development of both submicron-sized pinholes and pores; these features were not present on either sterile or untreated control coupons. Room temperature, biotically-incubated coupons show a wide distribution of pores covering the coupon surface, while coupons incubated at 30 C show the pores restricted to polishing ridges

  7. Evaluation and compilation of DOE [Department of Energy] waste package test data

    International Nuclear Information System (INIS)

    This report summarizes evaluations by the National Institute of Standards and Technology (NIST) of Department of Energy (DOE) activities on waste packages designed for containment of radioactive high-level nuclear waste (HLW) for the six month period February 1988 through July 1988. Activities for the DOE Materials Characterization Center are reviewed for the period January 1988 through June 1988. A summary is given of the Yucca Mountain, Nevada disposal site activities. Short discussions relating to the reviewed publications are given and complete reviews and evaluations are included. 20 refs., 1 fig., 1 tab

  8. Review of DOE Waste Package Program. Semiannual report, October 1984-March 1985. Volume 8

    Energy Technology Data Exchange (ETDEWEB)

    Davis, M.S. (ed.)

    1985-12-01

    A large number of technical reports on waste package component performance were reviewed over the last year in support of the NRC`s review of the Department of Energy`s (DOE`s) Environmental Assessment reports. The intent was to assess in some detail the quantity and quality of the DOE data and their relevance to the high-level waste repository site selection process. A representative selection of the reviews is presented for the salt, basalt, and tuff repository projects. Areas for future research have been outlined. 141 refs.

  9. A radioactive waste transportation package monitoring system for normal transport and accident emergency response conditions

    International Nuclear Information System (INIS)

    This paper addresses spent fuel and high level waste transportation history and prospects, discusses accident histories of radioactive material transport, discusses emergency responder needs and provides a general description of the Transportation Intelligent Monitoring System (TRANSIMS) design. The key objectives of the monitoring system are twofold: (1) to facilitate effective emergency response to accidents involving a radioactive waste transportation package, while minimizing risk to the public and emergency first-response personnel, and (2) to allow remote monitoring of transportation vehicle and payload conditions to enable research into radioactive material transportation for normal and accident conditions. (J.P.N.)

  10. Packaging for transport and disposal of low level waste at Drigg

    International Nuclear Information System (INIS)

    Solid low level waste (LLW) disposal operations at the British Nuclear Fuels plc (BNFL) Drigg site are currently being upgraded. A major feature of this upgrade is the introduction of waste compaction, containerisation and orderly emplacement of packages in concrete lined trenches (vaults). This paper summarises the current status of the upgrade with particular emphasis on progress towards specification of a product container design that is consistent with the overall aim of achieving long term post-closure site stability and will also meet the requirements for transport to Drigg through the public domain under the conditions of the 1985 IAEA Transport Regulations. (author)

  11. Review of DOE Waste Package Program. Semiannual report, October 1984-March 1985. Volume 8

    International Nuclear Information System (INIS)

    A large number of technical reports on waste package component performance were reviewed over the last year in support of the NRC's review of the Department of Energy's (DOE's) Environmental Assessment reports. The intent was to assess in some detail the quantity and quality of the DOE data and their relevance to the high-level waste repository site selection process. A representative selection of the reviews is presented for the salt, basalt, and tuff repository projects. Areas for future research have been outlined. 141 refs

  12. Systems of Packaging Waste Recycling in the EU: Comparing Five Different Case-Studies

    OpenAIRE

    Simões, Pedro; Cruz, Nuno; Marques, Rui

    2012-01-01

    All European Union (EU) member states have to comply with the demanding recycling rates targets that were set for the recovery of packaging waste in the Directive 94/62/EC. Nevertheless, each country has its own system for accomplishing these targets. Some already had national legislation when the Directive entered into force. Others had to “start from scratch†. Indeed, many countries have experienced massive improvements in the waste management systems in the last years; these include th...

  13. 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

  14. Regulatory authority of the Rocky Mountain states for low-level radioactive waste packaging and transportation

    International Nuclear Information System (INIS)

    The newly-formed Rocky Mountain Low-Level Radioactive Waste Compact is an interstate agreement for the management of low-level radioactive waste (LLW). Eligible members of the compact are Arizona, Colorado, Nevada, New Mexico, Utah, and Wyoming. Each state must ratify the compact within its legislature for the compact to become effective in that state and to make that state a full-fledged member of the compact. By so adopting the compact, each state agrees to the terms and conditions specified therein. Among those terms and conditions are provisions requiring each member state to adopt and enforce procedures requiring low-level waste shipments originating within its borders and destined for a regional facility to conform to packaging and transportation requirements and regulations. These procedures are to include periodic inspections of packaging and shipping practices, periodic inspections of waste containers while in the custody of carriers and appropriate enforcement actions for violations. To carry out this responsibility, each state must have an adequate statutory and regulatory inspection and enforcement authority to ensure the safe transportation of low-level radioactive waste. Three states in the compact region, Arizona, Utah and Wyoming, have incorporated the Department of Transportation regulations in their entirety, and have no published rules and regulations of their own. The other states in the compact, Colorado, Nevada and New Mexico all have separate rules and regulations that incorporate the DOT regulations. A brief description of the regulatory requirements of each state is presented

  15. Evaluation and compilation of DOE waste package test data: Biannual report, August 1987--January 1988

    International Nuclear Information System (INIS)

    This report summarizes results of the National Bureau of Standards (NBS) evaluations on waste packages designed for containment of radioactive high-level nuclear waste (HLW). The waste package is a proposed engineered barrier that is part of a permanent repository for HLW. Metal alloys are the principal barriers within the engineered system. Since enactment of the Budget Reconciliation Act for Fiscal Year 1988, the Yucca Mountain, Nevada, site (in which tuff is the geologic medium) is the only site that will be characterized for use as high-level nuclear waste repository. During the reporting period of August 1987 to January 1988, five reviews were completed for tuff, and these were grouped into the categories: ferrous alloys, copper, groundwater chemistry, and glass. Two issues are identified for the Yucca Mountain site: the approach used to calculate corrosion rates for ferrous alloys, and crevice corrosion was observed in a copper-nickel alloy. Plutonium can form pseudo-colloids that may facilitate transport. NBS work related to the vitrification of HLW borosilicate glass at the West Valley Demonstration Project (WVDP) and the Defense Waste Processing Facility (DWPF) and activities of the DOE Materials Characterization Center (MCC) for the 6-month reporting period are also included. 27 refs., 3 figs

  16. Evaluation and compilation of DOE waste package test data: Biannual report, August 1987--January 1988

    Energy Technology Data Exchange (ETDEWEB)

    Interrante, C.; Escalante, E.; Fraker, A.; Ondik, H.; Plante, E.; Ricker, R.; Ruspi, J.

    1988-08-01

    This report summarizes results of the National Bureau of Standards (NBS) evaluations on waste packages designed for containment of radioactive high-level nuclear waste (HLW). The waste package is a proposed engineered barrier that is part of a permanent repository for HLW. Metal alloys are the principal barriers within the engineered system. Since enactment of the Budget Reconciliation Act for Fiscal Year 1988, the Yucca Mountain, Nevada, site (in which tuff is the geologic medium) is the only site that will be characterized for use as high-level nuclear waste repository. During the reporting period of August 1987 to January 1988, five reviews were completed for tuff, and these were grouped into the categories: ferrous alloys, copper, groundwater chemistry, and glass. Two issues are identified for the Yucca Mountain site: the approach used to calculate corrosion rates for ferrous alloys, and crevice corrosion was observed in a copper-nickel alloy. Plutonium can form pseudo-colloids that may facilitate transport. NBS work related to the vitrification of HLW borosilicate glass at the West Valley Demonstration Project (WVDP) and the Defense Waste Processing Facility (DWPF) and activities of the DOE Materials Characterization Center (MCC) for the 6-month reporting period are also included. 27 refs., 3 figs.

  17. 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

  18. 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

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

    International Nuclear Information System (INIS)

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

  20. Technical support for GEIS: radioactive waste isolation in geologic formations. Volume 2. Commercial waste forms, packaging and projections for preconceptual repository design studies

    International Nuclear Information System (INIS)

    This volume, Y/OWI/TM-36/2, ''Commercial Waste Forms, Packaging and Projections for Preconceptual Repository Design Studies,'' is one of a 23-volume series, ''Technical Support for GEIS: Radioactive Waste Isolation in Geologic Formations,'' Y/OWI/TM-36, which supplements the ''Contribution to Draft Generic Environmental Impact Statement on Commercial Waste Management: Radioactive Waste Isolation in Geologic Formations,'' Y/OWI/TM-44. The series provides a more complete technical basis for the preconceptual designs, resource requirements, and environmental source terms associated with isolating commercial LWR wastes in underground repositories in salt, granite, shale and basalt. Wastes are considered from three fuel cycles: uranium and plutonium recycling, no recycling of spent fuel and uranium-only recycling. This volume contains the data base for waste forms, packages, and projections from the commercial waste defined by the Office of Waste Isolation in ''Nuclear Waste Projections and Source Term Data for FY 1977,'' Y/OWI/TM-34. Also, as an alternative data base for repository design and analysis, waste forms, packages, and projections for commercial waste defined by Battelle Pacific Northwest Laboratory (BPNL) have been included. This data base consists of a reference case for use in the alternative design study and a definition of combustible wastes for use in mine fire and hydrogen generation analyses

  1. Review of waste package verification tests. Semiannual report, April 1985-September 1985

    International Nuclear Information System (INIS)

    Several studies were completed this period to evaluate experimental and analytical methodologies being used in the DOE waste package program. The first involves a determination of the relevance of the test conditions being used by DOE to characterize waste package component behavior in a salt repository system. Another study focuses on the testing conditions and procedures used to measure radionuclide solubility and colloid formation in repository groundwaters. An attempt was also made to evaluate the adequacy of selected waste package performance codes. However, the latter work was limited by an inability to obtain several codes from DOE. Nevertheless, it was possible to comment briefly on the structures and intents of the codes based on publications in the open literature. The final study involved an experimental program to determine the likelihood of stress-corrosion cracking of austenitic stainless steels and Incoloy 825 in simulated tuff repository environments. Tests for six-month exposure periods in water and air-steam conditions are described. 52 figs., 48 tabs

  2. Petrologic and geochemical characterization of the Topopah Spring Member of the Paintbrush Tuff: outcrop samples used in waste package experiments

    Energy Technology Data Exchange (ETDEWEB)

    Knauss, K.G.

    1984-06-01

    This report summarizes characterization studies conducted with outcrop samples of Topopah Spring Member of the Paintbrush Tuff (Tpt). In support of the Waste Package Task within the Nevada Nuclear Waste Storage Investigation (NNWSI), Tpt is being studied both as a primary object and as a constituent used to condition water that will be reacted with waste form, canister, or packing material. These studies directly or indirectly support NNWSI subtasks concerned with waste package design and geochemical modeling. To interpret the results of subtask experiments, it is necessary to know the exact nature of the starting material in terms of the intial bulk composition, mineralogy, and individual phase geochemistry. 31 figures, 5 tables.

  3. An econometric analysis of regional differences in household waste collection: The case of plastic packaging waste in Sweden

    International Nuclear Information System (INIS)

    The Swedish producer responsibility ordinance mandates producers to collect and recycle packaging materials. This paper investigates the main determinants of collection rates of household plastic packaging waste in Swedish municipalities. This is done by the use of a regression analysis based on cross-sectional data for 252 Swedish municipalities. The results suggest that local policies, geographic/demographic variables, socio-economic factors and environmental preferences all help explain inter-municipality collection rates. For instance, the collection rate appears to be positively affected by increases in the unemployment rate, the share of private houses, and the presence of immigrants (unless newly arrived) in the municipality. The impacts of distance to recycling industry, urbanization rate and population density on collection outcomes turn out, though, to be both statistically and economically insignificant. A reasonable explanation for this is that the monetary compensation from the material companies to the collection entrepreneurs vary depending on region and is typically higher in high-cost regions. This implies that the plastic packaging collection in Sweden may be cost ineffective. Finally, the analysis also shows that municipalities that employ weight-based waste management fees generally experience higher collection rates than those municipalities in which flat and/or volume-based fees are used

  4. Development of backfill material as an engineered barrier in the waste package system. Interim topical report

    International Nuclear Information System (INIS)

    A backfill barrier, emplaced between the containerized waste and the host rock, can both protect the other engineered barriers and act as a primary barrier to the release of radionuclides from the waste package. Attributes that a backfill should provide in order to carry out its required function have been identified. Primary attributes are those that have a direct effect upon the release and transport of radionuclides from the waste package. Supportive attributes do not directly affect radionuclide release but are necessary to support the primary attributes. The primary attributes, in order of importance, are: minimize (retard or exclude) the migration of ground water between the host rock and the waste canister system; retard the migration of selected chemical species (corrosive species and radionuclides) in the ground water; control the Eh and pH of the ground water within the waste-package environment. The supportive attributes are: self-seal any cracks or discontinuities in the backfill or interfacing host geology; retain performance properties at all repository temperatures; retain peformance properties during and after receiving repository levels of gamma radiation; conduct heat from the canister system to the host geology; retain mechanical properties and provide resistance to applied mechanical forces; retain morphological stability and compatibility with structural barriers and with the host geology for required period of time. Screening and selection of candidate backfill materials has resulted in a preliminary list of materials for testing. Primary emphasis has been placed on sodium and calcium bentonites and zeolites used in conjunction with quartz sand or crushed host rock. Preliminary laboratory studies have concentrated on permeability, sorption, swelling pressure, and compaction properties of candidate backfill materials

  5. Qualification tests of packages used for transport and storage of low activity radioactive wastes in INR Pitesti

    International Nuclear Information System (INIS)

    Radioactive wastes generated by the TRIGA INR research reactor are packaged according to the national and international rules and standards. The technology for packaging and treatment of radioactive wastes can also be used at the Nuclear Power Plant Cernavoda. The qualification tests for the package used for transport and storage of radioactive wastes (low activity, up to 6.07 GBq (0.164 Ci) per drum) are described. The package used is a drum manufactured from 1 mm thick mild steel with the dimensions: height 915 ± 10 mm; diameter 600 ± 5 mm; volume 220 litres. To achieve adequate safety in the transport of radioactive wastes strict precautions must be taken according to the IAEA Regulations for the Safe Transport of Radioactive Materials. The adequacy of the package design is therefore of primary importance, the design requirements being supplemented by careful construction, quality assurance and inspection procedures. Taking into consideration the above requirements, qualification tests for the prototype package were carried out. These tests include compression, penetration, free fall, leaching, safety in use (biological protection), checking of chemical and mechanical characteristics, and the effect of the product on the environment. Performance of these tests, and the results obtained, prove that our technology for treatment and packaging of radioactive waste is in accordance with international rules. (author)

  6. The Role of Capillary Barrier in Reducing Moisture Content on Waste Packages

    International Nuclear Information System (INIS)

    Assessment of the performance of engineered capillary barriers at the potential Yucca Mountain nuclear waste repository site, in which 1.67-m-diameter waste packages are to be emplaced in 5-m-diameter tunnels according to current design, brings up aspects not commonly considered in more typical applications of capillary barriers (e.g., near-surface landfills). Engineered capillary barriers typically consist of two layers of granular materials with a sloping interface, in which the contrast in capillarity between the layers keeps infiltrating water in the upper layer. One issue is the effect of thermohydrologic processes that would occur at elevated repository temperatures (and temperature gradients). For example, backfill materials may be altered from that of the as-placed material by the hydrothermal regime imposed by the emplacement of waste in the repository, changing hydrologic properties in a way that degrades the performance of the barrier. A reduction of permeability in the upper layer might diminish the capacity of the upper layer to divert incoming seepage or to cause a ''vapor lid'' whereby buoyant vapor flow would be trapped, then condense and drain onto waste packages. Other concerns are the result of highly spatially and temporally variable seepage distribution and the very limited spatial scale available for flow attenuation and diversion

  7. Waste package/engineered barrier system design concepts for Yucca Mountain repository

    International Nuclear Information System (INIS)

    The U.S. Department of Energy (DOE) is responsible for the siting, construction and operation of mined geologic disposal system (MGDS) for high level waste. The U.S. Nuclear Regulatory Commission (NRC) has the responsibility for promulgating the technical requirements necessary to license all phases of repository operation. The development of MGDS has been delegated to the DOE's Yucca Mountain Site Characterization Project Office. The B ampersand W Fuel Company, as part of the Civilian Radioactive Waste Management System Management and Operating Contractor, is responsible for designing the waste package (WP) and the engineered barrier system (EBS). The goal of the design effort is to achieve a conservative, licensable design that meets the regulatory requirements with sufficient margin for uncertainty. Attainment of this goal relies on a multibarrier approach, the unsaturated nature of the Yucca Mountain site, consideration of, technical alternatives, and sufficient resolution of technical and regulatory uncertainties

  8. 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

  9. Quality intercomparison testing of waste package assay systems on UK nuclear sites

    Energy Technology Data Exchange (ETDEWEB)

    Daish, S.R. [NNC Ltd., WQCL, Dorchester (United Kingdom); Leech, N.A. [The Environment Agency, Lancaster (United Kingdom)

    2003-07-01

    The independent monitoring of solid low-level radioactive waste (LLW) disposals in the united kingdom is undertaken by NNC limited on behalf of the environment agency and SEPA at the Waste Quality Checking Laboratory (WQCL) at Winfrith. A review of the potential for on-site checking of site operator's drum monitoring equipment was carried out at WQCL in 1998. As a result of this review, drums of simulated waste have been prepared and developed at WQCL. These standard waste packages form the basis of an on-going programme of on-site intercomparison tests on site operator's gamma assay instrumentation, which commenced in December 1999. The purpose of the programme is to provide the Agency with a check on site operator's waste drum measurements as part of the its ongoing monitoring programme. The use of reference drums containing defined radionuclides of known radioactivity allows the Agency to assess the adequacy of operator's arrangements for assaying drummed LLW destined for disposal in the BNFL Drigg repository in Cumbria. The waste assay systems tested to date are described and the results of the first eleven tests performed are used to compare and contrast two types of gamma assay system in common use on nuclear sites in the United Kingdom. (orig.)

  10. Quality intercomparison testing of waste package assay systems on UK nuclear sites

    International Nuclear Information System (INIS)

    The independent monitoring of solid low-level radioactive waste (LLW) disposals in the united kingdom is undertaken by NNC limited on behalf of the environment agency and SEPA at the Waste Quality Checking Laboratory (WQCL) at Winfrith. A review of the potential for on-site checking of site operator's drum monitoring equipment was carried out at WQCL in 1998. As a result of this review, drums of simulated waste have been prepared and developed at WQCL. These standard waste packages form the basis of an on-going programme of on-site intercomparison tests on site operator's gamma assay instrumentation, which commenced in December 1999. The purpose of the programme is to provide the Agency with a check on site operator's waste drum measurements as part of the its ongoing monitoring programme. The use of reference drums containing defined radionuclides of known radioactivity allows the Agency to assess the adequacy of operator's arrangements for assaying drummed LLW destined for disposal in the BNFL Drigg repository in Cumbria. The waste assay systems tested to date are described and the results of the first eleven tests performed are used to compare and contrast two types of gamma assay system in common use on nuclear sites in the United Kingdom. (orig.)

  11. The effects of gamma radiation on the corrosion of candidate materials for the fabrication of nuclear waste packages

    Energy Technology Data Exchange (ETDEWEB)

    Shoesmith, D.W. [Univ. of Western Ontario, Dept. of Chemistry, London, Ontario (Canada); King, F

    1999-07-01

    The influence of gamma radiation on the corrosion of candidate materials for the fabrication of nuclear waste packages has been comprehensively reviewed. The comparison of corrosion of the various materials was compared in three distinct environments: Environment A; Mg{sup 2+}-enriched brines in which hydrolysis of the cation produces acidic environments and the Mg{sup 2+} interferes with the formation of protective films; Environment B; saline environments with a low Mg{sup 2+} content which remain neutral; Environment C; moist aerated conditions.The reference design of nuclear waste package for emplacement in the proposed waste repository in Yucca Mountain, Nevada, employs a dual wall arrangement, in which a 2 cm thick nickel alloy inner barrier is encapsulated within a 10 cm thick mild steel outer barrier. It is felt that this arrangement will give considerable containment lifetimes, since no common mode failure exists for the two barriers. The corrosion performance of this waste package will be determined by the exposure environment established within the emplacement drifts. Key features of the Yucca Mountain repository in controlling waste package degradation are expected to be the permanent availability of oxygen and the limited presence of water. When water contacts the surface of the waste package, its gamma radiolysis could produce an additional supply of corrosive agents. the gamma field will be produced by the radioactive decay of radionuclides within the waste form, and its magnitude will depend on the nature and age of the waste form as well as the material and wall thickness of the waste package.

  12. TECHNICAL PEER REVIEW REPORT - YUCCA MOUNTAIN: WASTE PACKAGE CLOSURE CONTROL SYSTEM

    International Nuclear Information System (INIS)

    The objective of the Waste Package Closure System (WPCS) project is to assist in the disposal of spent nuclear fuel (SNF) and associated high-level wastes (HLW) at the Yucca Mountain site in Nevada. Materials will be transferred from the casks into a waste package (WP), sealed, and placed into the underground facility. The SNF/HLW transfer and closure operations will be performed in an aboveground facility. The objective of the Control System is to bring together major components of the entire WPCS ensuring that unit operations correctly receive, and respond to, commands and requests for data. Integrated control systems will be provided to ensure that all operations can be performed remotely. Maintenance on equipment may be done using hands-on or remote methods, depending on complexity, exposure, and ease of access. Operating parameters and nondestructive examination results will be collected and stored as permanent electronic records. Minor weld repairs must be performed within the closure cell if the welds do not meet the inspection acceptance requirements. Any WP with extensive weld defects that require lids to be removed will be moved to the remediation facility for repair

  13. Contaminant Release Data Package for Residual Waste in Single-Shell Hanford Tanks

    Energy Technology Data Exchange (ETDEWEB)

    Deutsch, William J.; Cantrell, Kirk J.; Krupka, Kenneth M.

    2007-12-01

    The Hanford Federal Facility Agreement and Consent Order requires that a Resource Conservation and Recovery Act (RCRA) Facility Investigation report be submitted to the Washington State Department of Ecology. The RCRA Facility Investigation report will provide a detailed description of the state of knowledge needed for tank farm performance assessments. This data package provides detailed technical information about contaminant release from closed single-shell tanks necessary to support the RCRA Facility Investigation report. It was prepared by Pacific Northwest National Laboratory (PNNL) for CH2M HILL Hanford Group, Inc., which is tasked by the U.S. Department of Energy (DOE) with tank closure. This data package is a compilation of contaminant release rate data for residual waste in the four Hanford single-shell tanks (SSTs) that have been tested (C-103, C-106, C-202, and C-203). The report describes the geochemical properties of the primary contaminants of interest from the perspective of long-term risk to groundwater (uranium, technetium-99, iodine-129, chromium, transuranics, and nitrate), the occurrence of these contaminants in the residual waste, release mechanisms from the solid waste to water infiltrating the tanks in the future, and the laboratory tests conducted to measure release rates.

  14. TECHNICAL PEER REVIEW REPORT - YUCCA MOUNTAIN: WASTE PACKAGE CLOSURE CONTROL SYSTEM

    Energy Technology Data Exchange (ETDEWEB)

    NA

    2005-10-25

    The objective of the Waste Package Closure System (WPCS) project is to assist in the disposal of spent nuclear fuel (SNF) and associated high-level wastes (HLW) at the Yucca Mountain site in Nevada. Materials will be transferred from the casks into a waste package (WP), sealed, and placed into the underground facility. The SNF/HLW transfer and closure operations will be performed in an aboveground facility. The objective of the Control System is to bring together major components of the entire WPCS ensuring that unit operations correctly receive, and respond to, commands and requests for data. Integrated control systems will be provided to ensure that all operations can be performed remotely. Maintenance on equipment may be done using hands-on or remote methods, depending on complexity, exposure, and ease of access. Operating parameters and nondestructive examination results will be collected and stored as permanent electronic records. Minor weld repairs must be performed within the closure cell if the welds do not meet the inspection acceptance requirements. Any WP with extensive weld defects that require lids to be removed will be moved to the remediation facility for repair.

  15. Evaluation of the Thermal Response of the 5-DHLW Waste Package-Hypothetical Fire Accident

    International Nuclear Information System (INIS)

    The purpose of this calculation is to determine the thermal response of the 5-defense high level waste (DHLW)/Department of Energy (DOE) codisposal waste package (WP) to the hypothetical fire accident. The objective is to calculate the temperature response of the DHLW glass to the hypothetical short-term fire defined in 10 CFR 71, Section 73(c)(4), Reference 1. The scope of the calculation includes evaluation of the accident with the waste package above ground, at the Yucca Mountain surface facility. The scope is intended to cover a DHLW WP. This WP is loaded with DHLW canisters containing glass from the Savannah River Site (SRS) and a DOE canister containing Training, Research, and Isotope General Atomics (TRIGA) spent nuclear fuel (SNF). The information provided by the sketches attached to this calculation is that for the potential design of the type of WP considered in this calculation. In addition to the nominal design configuration thermal load case, the effects of varying the central DOE canister and DHLW thermal loads are determined. Also, the effects of varying values of the flame and WP outer surface emissivities are evaluated

  16. Excess Weapons Plutonium Disposition: Plutonium Packaging, Storage and Transportation and Waste Treatment, Storage and Disposal Activities

    Energy Technology Data Exchange (ETDEWEB)

    Jardine, L J; Borisov, G B

    2004-07-21

    A fifth annual Excess Weapons Plutonium Disposition meeting organized by Lawrence Livermore National Laboratory (LLNL) was held February 16-18, 2004, at the State Education Center (SEC), 4 Aerodromnya Drive, St. Petersburg, Russia. The meeting discussed Excess Weapons Plutonium Disposition topics for which LLNL has the US Technical Lead Organization responsibilities. The technical areas discussed included Radioactive Waste Treatment, Storage, and Disposal, Plutonium Oxide and Plutonium Metal Packaging, Storage and Transportation and Spent Fuel Packaging, Storage and Transportation. The meeting was conducted with a conference format using technical presentations of papers with simultaneous translation into English and Russian. There were 46 Russian attendees from 14 different Russian organizations and six non-Russian attendees, four from the US and two from France. Forty technical presentations were made. The meeting agenda is given in Appendix B and the attendance list is in Appendix C.

  17. Pyrolysis of plastic packaging waste: A comparison of plastic residuals from material recovery facilities with simulated plastic waste.

    Science.gov (United States)

    Adrados, A; de Marco, I; Caballero, B M; López, A; Laresgoiti, M F; Torres, A

    2012-05-01

    Pyrolysis may be an alternative for the reclamation of rejected streams of waste from sorting plants where packing and packaging plastic waste is separated and classified. These rejected streams consist of many different materials (e.g., polyethylene (PE), polypropylene (PP), polystyrene (PS), polyvinyl chloride (PVC), polyethylene terephthalate (PET), acrylonitrile butadiene styrene (ABS), aluminum, tetra-brik, and film) for which an attempt at complete separation is not technically possible or economically viable, and they are typically sent to landfills or incinerators. For this study, a simulated plastic mixture and a real waste sample from a sorting plant were pyrolyzed using a non-stirred semi-batch reactor. Red mud, a byproduct of the aluminum industry, was used as a catalyst. Despite the fact that the samples had a similar volume of material, there were noteworthy differences in the pyrolysis yields. The real waste sample resulted, after pyrolysis, in higher gas and solid yields and consequently produced less liquid. There were also significant differences noted in the compositions of the compared pyrolysis products. PMID:21795037

  18. DOSE RATES FOR WESTINGHOUSE 17X17 MOX PWR SNF IN A WASTE PACKAGE (SCPB: N/A)

    Energy Technology Data Exchange (ETDEWEB)

    T.L. Lotz

    1997-01-29

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to estimate the dose rate on and near the surface a Multi-Purpose Canister (MPC) PWR waste package (WP) which is loaded with Westinghouse 17 x 17 mixed oxide (MOX) PWR fuel. The 21 PWR MPC WP is used to provide an upper bound for waste package designs since the 12 PWR MPC WP will have a smaller source term and an equivalent amount of shielding. the objectives of this evaluation are to calculate the requested dose rate(s) and document the calculation in a fashion to allow comparisons to other waste forms and WP designs at a future time.

  19. Expected very-near-field thermal environments for advanced spent-fuel and defense high-level waste packages

    International Nuclear Information System (INIS)

    The very-near-field thermal environments expected in a nuclear waste repository in a salt formation have been evaluated for the Westinghouse Form I advanced waste package concepts. The repository descriptions used to supplement the waste package designs in these analyses are realistic and take into account design constraints to assure conservatism. As a result, areal loadings are well below the acceptable values established for salt repositories. Predicted temperatures are generally well below any temperature limits which have been discussed for waste packages in a salt formation. These low temperatures result from the conservative repository designs. Investigations are also made of the sensitivity of these temperatures to areal loading, canister separation, and other design features

  20. Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural Material

    International Nuclear Information System (INIS)

    Stress corrosion cracking is one of the most common corrosion-related causes for premature breach of metal structural components. Stress corrosion cracking is the initiation and propagation of cracks in structural components due to three factors that must be present simultaneously: metallurgical susceptibility, critical environment, and static (or sustained) tensile stresses. This report was prepared according to ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]). The purpose of this report is to provide an evaluation of the potential for stress corrosion cracking of the engineered barrier system components (i.e., the drip shield, waste package outer barrier, and waste package stainless steel inner structural cylinder) under exposure conditions consistent with the repository during the regulatory period of 10,000 years after permanent closure. For the drip shield and waste package outer barrier, the critical environment is conservatively taken as any aqueous environment contacting the metal surfaces. Appendix B of this report describes the development of the SCC-relevant seismic crack density model (SCDM). The consequence of a stress corrosion cracking breach of the drip shield, the waste package outer barrier, or the stainless steel inner structural cylinder material is the initiation and propagation of tight, sometimes branching, cracks that might be induced by the combination of an aggressive environment and various tensile stresses that can develop in the drip shields or the waste packages. The Stainless Steel Type 316 inner structural cylinder of the waste package is excluded from the stress corrosion cracking evaluation because the Total System Performance Assessment for License Application (TSPA-LA) does not take credit for the inner cylinder. This document provides a detailed description of the process-level models that can be applied to assess the performance of Alloy 22

  1. Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural Material

    Energy Technology Data Exchange (ETDEWEB)

    G. Gordon

    2004-10-13

    Stress corrosion cracking is one of the most common corrosion-related causes for premature breach of metal structural components. Stress corrosion cracking is the initiation and propagation of cracks in structural components due to three factors that must be present simultaneously: metallurgical susceptibility, critical environment, and static (or sustained) tensile stresses. This report was prepared according to ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]). The purpose of this report is to provide an evaluation of the potential for stress corrosion cracking of the engineered barrier system components (i.e., the drip shield, waste package outer barrier, and waste package stainless steel inner structural cylinder) under exposure conditions consistent with the repository during the regulatory period of 10,000 years after permanent closure. For the drip shield and waste package outer barrier, the critical environment is conservatively taken as any aqueous environment contacting the metal surfaces. Appendix B of this report describes the development of the SCC-relevant seismic crack density model (SCDM). The consequence of a stress corrosion cracking breach of the drip shield, the waste package outer barrier, or the stainless steel inner structural cylinder material is the initiation and propagation of tight, sometimes branching, cracks that might be induced by the combination of an aggressive environment and various tensile stresses that can develop in the drip shields or the waste packages. The Stainless Steel Type 316 inner structural cylinder of the waste package is excluded from the stress corrosion cracking evaluation because the Total System Performance Assessment for License Application (TSPA-LA) does not take credit for the inner cylinder. This document provides a detailed description of the process-level models that can be applied to assess the

  2. Use of simple transport equations to estimate waste package performance requirements

    International Nuclear Information System (INIS)

    A method of developing waste package performance requirements for specific nuclides is described. The method is based on: Federal regulations concerning permissible concentrations in solution at the point of discharge to the accessible environment; a simple and conservative transport model; baseline and potential worst-case release scenarios. Use of the transport model enables calculation of maximum permissible release rates within a repository in basalt for each of the scenarios. The maximum permissible release rates correspond to performance requirements for the engineered barrier system. The repository was assumed to be constructed in a basalt layer. For the cases considered, including a well drilled into an aquifer 1750 m from the repository center, little significant advantage is obtained from a 1000-yr as opposed to a 100-yr waste package. A 1000-yr waste package is of importance only for nuclides with half-lives much less than 100 yr which travel to the accessible environment in much less than 1000 yr. Such short travel times are extremely unlikely for a mined repository. Among the actinides, the most stringent maximum permissible release rates are for 236U and 234U. A simple solubility calculation suggests, however, that these performance requirements can be readily met by the engineered barrier system. Under the reducing conditions likely to occur in a repository located in basalt, uranium would be sufficiently insoluble that no solution could contain more than about 0.01% of the maximum permissible concentration at saturation. The performance requirements derived from the one-dimensional modeling approach are conservative by at least one to two orders of magnitude. More quantitative three-dimensional modeling at specific sites should enable relaxation of the performance criteria derived in this study. 12 references, 8 figures, 8 tables

  3. Waste Form Release Data Package for the 2005 Integrated Disposal Facility Performance Assessment Erratum

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Gary L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-09-06

    This report refers to or contains Kg values for glasses LAWA44, LAWB45 and LAWC22 affected by calculations errors as identified by Papathanassiu et al. (2011). The corrected Kg values are reported in an erratum included in the revised version of the original report. The revised report can be referenced as follows: Pierce E. M. et al. (2004) Waste Form Release Data Package for the 2005 Integrated Disposal Facility Performance Assessment. PNNL-14805 Rev. 0 Erratum. Pacific Northwest National Laboratory, Richland, WA, USA.

  4. Engineered barrier system and waste package design concepts for a potential geologic repository at Yucca Mountain

    Energy Technology Data Exchange (ETDEWEB)

    Short, D.W.; Ruffner, D.J.; Jardine, L.J.

    1991-10-01

    We are using an iterative process to develop preliminary concept descriptions for the Engineered Barrier System and waste-package components for the potential geologic repository at Yucca Mountain. The process allows multiple design concepts to be developed subject to major constraints, requirements, and assumptions. Involved in the highly interactive and interdependent steps of the process are technical specialists in engineering, metallic and nonmetallic materials, chemistry, geomechanics, hydrology, and geochemistry. We have developed preliminary design concepts that satisfy both technical and nontechnical (e.g., programmatic or policy) requirements.

  5. R and D applied to the non-destructive characterization of waste packages for long term storage or deep disposal

    Energy Technology Data Exchange (ETDEWEB)

    Malvache, P.; Perot, B.; Ma, J.L.; Pettier, J.L. [CEA Cadarache, Dept. d' Etudes des Dechets, DED, 13 - Saint Paul lez Durance (France); Capdevila, J.M.; Huot, N. [CEA Saclay, Dir. du Developpement et de l' Innovation Nucleares DDIN, 91 - Gif Sur Yvette (France); Moulin, V. [CEA Grenoble, Lab. d' Electronique, de Technologie de l' Information LETI, DSYS, 38 (France)

    2001-07-01

    To ensure the quality and traceability of waste package management in the long term, knowledge on these packages is necessary so as to confirm their compliance to storage or disposal specifications. Research is focused on the management of the knowledge on these packages (fabrication means, materials contained,...) and on the acquisition, through measurement, of their characteristics. Within this context, many studies are underway at the CEA in the field of measurements so as to obtain non- destructive tools to access the parameters which allow the waste packages to be characterized. The two main R and D investigations concern: the nuclear measurement methods for the detection and quantification of radionuclides and of chemical elements considered as important for storage or disposal safety ; the measurement methods for the physical characteristics of the packages by high energy photon imaging, thus allowing pictures of the contents of large, high density and sometimes irradiating packages to be known. During the last five years, the research at the CEA focused on these two areas and resulted in a significant evolution in the non-destructive characterization means for long lived waste packages. (author)

  6. Radioactive waste package assay facility. Volume 1. Application of assay technology

    International Nuclear Information System (INIS)

    This report, in three volumes, covers the work carried out by Taylor Woodrow Construction Ltd., and two major sub-contractors: Harwell Laboratory (AEA Technology) and Siemens Plessey Controls Ltd., on the development of a radioactive waste package assay facility, for cemented 500 litre intermediate level waste drums. In volume 1, the reasons for assay are considered together with the various techniques that can be used, and the information that can be obtained. The practical problems associated with the use of the various techniques in an integrated assay facility are identified, and the key parameters defined. Engineering and operational features are examined and provisional designs proposed for facilities at three throughput levels: 15,000, 750 and 30 drums per year respectively. The capital and operating costs for such facilities have been estimated. A number of recommendations are made for further work. 16 refs., 14 figs., 13 tabs

  7. Environmental and economic benefit of recycling model of packaging waste:a case study on aluminum

    Institute of Scientific and Technical Information of China (English)

    Huang Pingsha

    2004-01-01

    In order to achieve sustainable utilization of natural resources, save energy and protect environment and ecosystem, it is important for a region or a nation to develop and implement a viable waste recycling model from both theoretical and practical point of view. Some packaging recycling models operated in developed countries are introduced in this article. Aluminium can recovery and recycling is emphasized. Cost effective, economic and environmental benefit of different models are compared and analyzed. The result shows that all recycling models have their characteristics due to the initial purpose of recovery and the situation of the implementing country. However, all the models contribute to the reduction of municipal solid waste disposal and resources conservation.

  8. GENERIC DEGRADATION SCENARIO AND CONFIGURATION ANALYSIS FOR DOE CODISPOSAL WASTE PACKAGE

    International Nuclear Information System (INIS)

    The purpose of this analysis is to develop a generic set of degradation scenarios and associated configurations for various Department of Energy (DOE) spent nuclear fuel (SNF) types when codisposed with the high-level waste (HLW) glass inside a waste package (WP). The degradation takes place inside the WP. These scenarios and configurations are developed as refinements of the standard degradation scenarios and potentially critical configuration classes given in Section 3.1 of the ''Disposal Criticality Analysis Methodology Topical Report'' (Ref. 1). Certain degradation scenarios and configurations will change when EDA II design is baselined. In accordance with AP-3.10Q, Revision 0, ICN 0, a work direction was developed, issued, and used in the preparation of this document (Ref. 13, p. 7)

  9. The Development of Concrete Packages for Geological Disposal of B- and TRU Radioactive Waste: Collaboration Between ANDRA and RWMC

    International Nuclear Information System (INIS)

    A cooperation program on long-term concrete disposal packages solutions for ILW-LL has been developed between ANDRA (French National Radioactive Waste Management Agency) and RWMC (Japanese Radioactive Waste Management Funding and Research Center). The main application of this cooperation program has been high retention ability packages (confinement and durability) for hulls and end pieces in CSDC canister. Respective package designs differ slightly. The RWMC package is based on Ultra High Performance Concrete monolithic package manufactured by continuous placing. ANDRA package designs are based on High Performance Concrete: one option is a monolithic type package; another option investigates the performance of concrete junction with small poured concrete lids. The paper focuses on respective approaches and cooperation in two specific areas: thermal crack prevention design and associated tools, and manufacturing results (full scale and small scale mock ups). Each partner has consolidated his own approach and benefits from the experience and expertise feedback of the other partner on similar problems. Cooperation is fruitful especially in knowledge and methodology consolidation, and in the package industrial feasibility performance check. (authors)

  10. 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

  11. The effect of waste package components on radionuclides released from spent fuel under hydrothermal conditions

    International Nuclear Information System (INIS)

    The Basalt Waste Isolation Project has conducted a series of hydrothermal experiments to characterize waste/barrier/rock interactions as a part of its study of the Columbia River basalts as a potential medium for a nuclear waste repository. Hydrothermal tests of 3--15 months duration were performed with light water reactor spent fuel and simulated groundwater, in combination with candidate container materials (low-carbon steel or copper) and/or basalt, in order to evaluate the effect of waste package materials on spent fuel radionuclide release behavior. In all experiments, 14C, 129I, and 137Cs occurred only as dissolved species, whereas the actinides occurred in 400 nm filtrates primarily as spent fuel particles. Actinide concentrations in 1.8 nm filtrates were below detection in steel-bearing experiments. In the system spent fuel + copper, apparent time-invariant concentrations of 14C and 137Cs were obtained, but in the spent fuel + steel system, the concentrations of 14C and 137Cs increased gradually throughout the experiments. In experiments containing basalt or steel + basalt, 137Cs concentrations decreased with time. In tests with copper + basalt, 14C and 129I concentrations attained time-variant values and 137Cs concentrations decreased. Concentrations for the actinides and fission products measured in these experiments were below those calculated from Federal regulations governing radionuclide release. 12 refs., 7 figs., 5 tabs

  12. Sampling methods and non-destructive examination techniques for large radioactive waste packages

    International Nuclear Information System (INIS)

    Progress is reported on work undertaken to evaluate quality checking methods for radioactive wastes. A sampling rig was designed, fabricated and used to develop techniques for the destructive sampling of cemented simulant waste using remotely operated equipment. An engineered system for the containment of cooling water was designed and manufactured and successfully demonstrated with the drum and coring equipment mounted in both vertical and horizontal orientations. The preferred in-cell orientation was found to be with the drum and coring machinery mounted in a horizontal position. Small powdered samples can be taken from cemented homogeneous waste cores using a hollow drill/vacuum section technique with the preferred subsampling technique being to discard the outer 10 mm layer to obtain a representative sample of the cement core. Cement blends can be dissolved using fusion techniques and the resulting solutions are stable to gelling for periods in excess of one year. Although hydrochloric acid and nitric acid are promising solvents for dissolution of cement blends, the resultant solutions tend to form silicic acid gels. An estimate of the beta-emitter content of cemented waste packages can be obtained by a combination of non-destructive and destructive techniques. The errors will probably be in excess of +/-60 % at the 95 % confidence level. Real-time X-ray video-imaging techniques have been used to analyse drums of uncompressed, hand-compressed, in-drum compacted and high-force compacted (i.e. supercompacted) simulant waste. The results have confirmed the applicability of this technique for NDT of low-level waste. 8 refs., 12 figs., 3 tabs

  13. Investigation and determination of safety and health physics specifications for packaging and conditioning of decommissioning wastes

    International Nuclear Information System (INIS)

    This report analyzed the impact of the new IAEA recommendations on future transports of radioactive materials in waste containers intended for final storage in the KONRAD facility. Through the application of the new transport regulations to the KONRAD containers, the major portion of the transports can be made in standardized industrial packaging. The transition area for the use of Type B casks was described. Also, safety and health physics aspects were examined by computer and/or by experiment. For this, loading and conditioning experiments were performed with full-scale standardized Type V containers. The loading of containers was tested with steel scrap, rubble and bulky materials in drums. The feasibility of filling in the hollow spaces in containers, even with loose rubble, was demonstrated. For this purpose, filler material with special flow characteristics was selected and tested. Post-test examinations of the containers showed that the hollow spaces had been practically completely filled in the tests. The health physics aspects of the large-scale packaging of radioactive waste from decommissioning were considered, and tips for the future application of this technique were given. (orig.)

  14. Environment on the Surfaces of the Drip Shield and Waste Package Outer Barrier

    International Nuclear Information System (INIS)

    This report provides supporting analysis of the conditions at which an aqueous solution can exist on the drip shield or waste package surfaces, including theoretical underpinning for the evolution of concentrated brines that could form by deliquescence or evaporation, and evaluation of the effects of acid-gas generation on brine composition. This analysis does not directly feed the total system performance assessment for the license application (TSPA-LA), but supports modeling and abstraction of the in-drift chemical environment (BSC 2004 [DIRS 169863]; BSC 2004 [DIRS 169860]). It also provides analyses that may support screening of features, events, and processes, and input for response to regulatory inquiries. This report emphasizes conditions of low relative humidity (RH) that, depending on temperature and chemical conditions, may be dry or may be associated with an aqueous phase containing concentrated electrolytes. Concentrated solutions at low RH may evolve by evaporative concentration of water that seeps into emplacement drifts, or by deliquescence of dust on the waste package or drip shield surfaces. The minimum RH for occurrence of aqueous conditions is calculated for various chemical systems based on current understanding of site geochemistry and equilibrium thermodynamics. The analysis makes use of known characteristics of Yucca Mountain waters and dust from existing tunnels, laboratory data, and relevant information from the technical literature and handbooks

  15. Environment on the Surfaces of the Drip Shield and Waste Package Outer Barrier

    Energy Technology Data Exchange (ETDEWEB)

    T. Wolery

    2005-02-22

    This report provides supporting analysis of the conditions at which an aqueous solution can exist on the drip shield or waste package surfaces, including theoretical underpinning for the evolution of concentrated brines that could form by deliquescence or evaporation, and evaluation of the effects of acid-gas generation on brine composition. This analysis does not directly feed the total system performance assessment for the license application (TSPA-LA), but supports modeling and abstraction of the in-drift chemical environment (BSC 2004 [DIRS 169863]; BSC 2004 [DIRS 169860]). It also provides analyses that may support screening of features, events, and processes, and input for response to regulatory inquiries. This report emphasizes conditions of low relative humidity (RH) that, depending on temperature and chemical conditions, may be dry or may be associated with an aqueous phase containing concentrated electrolytes. Concentrated solutions at low RH may evolve by evaporative concentration of water that seeps into emplacement drifts, or by deliquescence of dust on the waste package or drip shield surfaces. The minimum RH for occurrence of aqueous conditions is calculated for various chemical systems based on current understanding of site geochemistry and equilibrium thermodynamics. The analysis makes use of known characteristics of Yucca Mountain waters and dust from existing tunnels, laboratory data, and relevant information from the technical literature and handbooks.

  16. Project and design of apparatus and equipment of pilot plant experimental bituminization line PS 44 EBO

    International Nuclear Information System (INIS)

    The modifications are described made in the design of major components of the pilot plant experimental bituminization line involving: the mixing equipment, the film rotor evaporator, the calciner, and various transport mechanisms for handling drums and containers. The difficulty and complexity of the problem which Kralovopolska strojirna Brno, the end supplier of the bituminization line, will have to tackle is evident mainly from the case of bridge cranes, which feed the line, and the full-portal gantry crane which handles the containers and drums at regional sites of radioactive waste disposal. (Z.M.)

  17. Radioactive waste isolation in salt: peer review of Westinghouse Electric Corporation's report on reference conceptual designs for a repository waste package

    International Nuclear Information System (INIS)

    This report documents the findings of the peer panel constituted by Argonne National Laboratory to review Region A of Westinghouse Electric Corporation's report entitled Waste Package Reference Conceptual Designs for a Repository in Salt. The panel determined that the reviewed report does not provide reasonable assurance that US Nuclear Regulatory Commission (NRC) requirements for waste packages will be met by the proposed design. It also found that it is premature to call the design a ''reference design,'' or even a ''reference conceptual design.'' This review report provides guidance for the preparation of a more acceptable design document

  18. Repository Criticality Control with Depleted-Uranium-Dioxide Cermet Waste Packages

    International Nuclear Information System (INIS)

    It is proposed that the structural components and internal basket structures of waste packages (WPs) be constructed of depleted uranium dioxide (DUO2)-steel cermets. The cermet contains 2 DUO2 imbedded in a steel matrix. The WPs are filled with spent nuclear fuel (SNF) and placed 2 in a geological repository. The WP provides a handling container for placement of SNF in the repository and is an engineered barrier to delay SNF degradation and subsequent release of radionuclides. SNF and other fissile wastes contain enriched uranium and transuranic fissile isotopes; thus, the potential for nuclear criticality exists. Most of the transuranic fissile isotopes, such as 239Pu, will have decayed to 233U or 235U before significant fissile-isotope migration from the degraded SNF or other fissile waste forms has occurred. Consequently, post-closure repository criticality issues are primarily from the fissile isotopes of uranium. As the WP degrades, the 238U in the DUO2-steel cermet would mix with the degrading SNF and isotopically dilute 233U and 235U to levels that would ensure that post-closure criticality would not occur

  19. Evaluating laser-driven Bremsstrahlung radiation sources for imaging and analysis of nuclear waste packages.

    Science.gov (United States)

    Jones, Christopher P; Brenner, Ceri M; Stitt, Camilla A; Armstrong, Chris; Rusby, Dean R; Mirfayzi, Seyed R; Wilson, Lucy A; Alejo, Aarón; Ahmed, Hamad; Allott, Ric; Butler, Nicholas M H; Clarke, Robert J; Haddock, David; Hernandez-Gomez, Cristina; Higginson, Adam; Murphy, Christopher; Notley, Margaret; Paraskevoulakos, Charilaos; Jowsey, John; McKenna, Paul; Neely, David; Kar, Satya; Scott, Thomas B

    2016-11-15

    A small scale sample nuclear waste package, consisting of a 28mm diameter uranium penny encased in grout, was imaged by absorption contrast radiography using a single pulse exposure from an X-ray source driven by a high-power laser. The Vulcan laser was used to deliver a focused pulse of photons to a tantalum foil, in order to generate a bright burst of highly penetrating X-rays (with energy >500keV), with a source size of chips. The uranium penny was clearly resolved to sub-mm accuracy over a 30cm(2) scan area from a single shot acquisition. In addition, neutron generation was demonstrated in situ with the X-ray beam, with a single shot, thus demonstrating the potential for multi-modal criticality testing of waste materials. This feasibility study successfully demonstrated non-destructive radiography of encapsulated, high density, nuclear material. With recent developments of high-power laser systems, to 10Hz operation, a laser-driven multi-modal beamline for waste monitoring applications is envisioned. PMID:27484945

  20. Evaluating laser-driven Bremsstrahlung radiation sources for imaging and analysis of nuclear waste packages.

    Science.gov (United States)

    Jones, Christopher P; Brenner, Ceri M; Stitt, Camilla A; Armstrong, Chris; Rusby, Dean R; Mirfayzi, Seyed R; Wilson, Lucy A; Alejo, Aarón; Ahmed, Hamad; Allott, Ric; Butler, Nicholas M H; Clarke, Robert J; Haddock, David; Hernandez-Gomez, Cristina; Higginson, Adam; Murphy, Christopher; Notley, Margaret; Paraskevoulakos, Charilaos; Jowsey, John; McKenna, Paul; Neely, David; Kar, Satya; Scott, Thomas B

    2016-11-15

    A small scale sample nuclear waste package, consisting of a 28mm diameter uranium penny encased in grout, was imaged by absorption contrast radiography using a single pulse exposure from an X-ray source driven by a high-power laser. The Vulcan laser was used to deliver a focused pulse of photons to a tantalum foil, in order to generate a bright burst of highly penetrating X-rays (with energy >500keV), with a source size of <0.5mm. BAS-TR and BAS-SR image plates were used for image capture, alongside a newly developed Thalium doped Caesium Iodide scintillator-based detector coupled to CCD chips. The uranium penny was clearly resolved to sub-mm accuracy over a 30cm(2) scan area from a single shot acquisition. In addition, neutron generation was demonstrated in situ with the X-ray beam, with a single shot, thus demonstrating the potential for multi-modal criticality testing of waste materials. This feasibility study successfully demonstrated non-destructive radiography of encapsulated, high density, nuclear material. With recent developments of high-power laser systems, to 10Hz operation, a laser-driven multi-modal beamline for waste monitoring applications is envisioned.

  1. Geochemistry Model Abstraction and Sensitivity Studies for the 21 PWR CSNF Waste Package

    Energy Technology Data Exchange (ETDEWEB)

    P. Bernot; S. LeStrange; E. Thomas; K. Zarrabi; S. Arthur

    2002-10-29

    The CSNF geochemistry model abstraction, as directed by the TWP (BSC 2002b), was developed to provide regression analysis of EQ6 cases to obtain abstracted values of pH (and in some cases HCO{sub 3}{sup -} concentration) for use in the Configuration Generator Model. The pH of the system is the controlling factor over U mineralization, CSNF degradation rate, and HCO{sub 3}{sup -} concentration in solution. The abstraction encompasses a large variety of combinations for the degradation rates of materials. The ''base case'' used EQ6 simulations looking at differing steel/alloy corrosion rates, drip rates, and percent fuel exposure. Other values such as the pH/HCO{sub 3}{sup -} dependent fuel corrosion rate and the corrosion rate of A516 were kept constant. Relationships were developed for pH as a function of these differing rates to be used in the calculation of total C and subsequently, the fuel rate. An additional refinement to the abstraction was the addition of abstracted pH values for cases where there was limited O{sub 2} for waste package corrosion and a flushing fluid other than J-13, which has been used in all EQ6 calculation up to this point. These abstractions also used EQ6 simulations with varying combinations of corrosion rates of materials to abstract the pH (and HCO{sub 3}{sup -} in the case of the limiting O{sub 2} cases) as a function of WP materials corrosion rates. The goodness of fit for most of the abstracted values was above an R{sup 2} of 0.9. Those below this value occurred during the time at the very beginning of WP corrosion when large variations in the system pH are observed. However, the significance of F-statistic for all the abstractions showed that the variable relationships are significant. For the abstraction, an analysis of the minerals that may form the ''sludge'' in the waste package was also presented. This analysis indicates that a number a different iron and aluminum minerals may form in

  2. Sorption of aqueous phosphorus onto bituminous and lignitous coal ashes

    Energy Technology Data Exchange (ETDEWEB)

    Yan Jinying; Kirk, Donald W. [Department of Chemical Engineering and Applied Chemistry, University of Toronto, 200 College Street, Toronto, Ontario, M5S 3E5 (Canada); Jia, Charles Q. [Department of Chemical Engineering and Applied Chemistry, University of Toronto, 200 College Street, Toronto, Ontario, M5S 3E5 (Canada)], E-mail: cqjia@chem-eng.toronto.edu; Liu Xinan [College of Chemical Engineering, Chongqing University, Chongqing (China)

    2007-09-05

    Aiming at the development of a phosphorus removal technology for waste water, phosphate (PO{sub 4}{sup 3-}) retention behavior of bituminous and lignitous coal ashes was investigated using a batch reactor. Ash samples, including fresh and weathered fly and bottom ashes, were studied for their sorption isotherms and reversibility. Fly ashes had a much higher phosphate retention capacity (4000-30,000 mg P/kg) than bottom ashes (15-600 mg P/kg). Lignitous coal ashes were more capable of retaining phosphate than bituminous coal ashes. The retention process was largely irreversible, and the irreversibility increased with the increase in the retention capacity. Weathering enlarged the retention capacity of the bituminous bottom ash, but substantially lowered that of the fly ash, likely due to the difference in the weather-induced changes between the fly and bottom ashes. Sorption isotherms of fly ashes were found to be adequately represented by the Langmuir model while those of bottom ashes fitted better to the Freundlich model. Concentrations of Ca{sup 2+} and PO{sub 4}{sup 3-} in the aqueous phase were measured at the end of sorption and desorption experiments, and were compared with solubilities of three calcium phosphate minerals. The aqueous solutions were saturated or super-saturated with respect to tricalcium phosphate (Ca{sub 3}(PO{sub 4}){sub 2}) and hydroxyapatite (Ca{sub 5}(PO{sub 4}){sub 3}OH), and slightly under-saturated with respect to amorphous calcium phosphate. It is concluded that precipitation of calcium phosphate is the predominant mechanism for phosphate retention by coal ash under the conditions studied. There is a strong and positive correlation between alkalinity and phosphate sorption capacity. Consequently, acid neutralization capacity (ANC) can be used as an indicator of phosphate sorption capacity of coal ashes.

  3. An investigation of the usability of sound recognition for source separation of packaging wastes in reverse vending machines.

    Science.gov (United States)

    Korucu, M Kemal; Kaplan, Özgür; Büyük, Osman; Güllü, M Kemal

    2016-10-01

    In this study, we investigate the usability of sound recognition for source separation of packaging wastes in reverse vending machines (RVMs). For this purpose, an experimental setup equipped with a sound recording mechanism was prepared. Packaging waste sounds generated by three physical impacts such as free falling, pneumatic hitting and hydraulic crushing were separately recorded using two different microphones. To classify the waste types and sizes based on sound features of the wastes, a support vector machine (SVM) and a hidden Markov model (HMM) based sound classification systems were developed. In the basic experimental setup in which only free falling impact type was considered, SVM and HMM systems provided 100% classification accuracy for both microphones. In the expanded experimental setup which includes all three impact types, material type classification accuracies were 96.5% for dynamic microphone and 97.7% for condenser microphone. When both the material type and the size of the wastes were classified, the accuracy was 88.6% for the microphones. The modeling studies indicated that hydraulic crushing impact type recordings were very noisy for an effective sound recognition application. In the detailed analysis of the recognition errors, it was observed that most of the errors occurred in the hitting impact type. According to the experimental results, it can be said that the proposed novel approach for the separation of packaging wastes could provide a high classification performance for RVMs.

  4. An investigation of the usability of sound recognition for source separation of packaging wastes in reverse vending machines.

    Science.gov (United States)

    Korucu, M Kemal; Kaplan, Özgür; Büyük, Osman; Güllü, M Kemal

    2016-10-01

    In this study, we investigate the usability of sound recognition for source separation of packaging wastes in reverse vending machines (RVMs). For this purpose, an experimental setup equipped with a sound recording mechanism was prepared. Packaging waste sounds generated by three physical impacts such as free falling, pneumatic hitting and hydraulic crushing were separately recorded using two different microphones. To classify the waste types and sizes based on sound features of the wastes, a support vector machine (SVM) and a hidden Markov model (HMM) based sound classification systems were developed. In the basic experimental setup in which only free falling impact type was considered, SVM and HMM systems provided 100% classification accuracy for both microphones. In the expanded experimental setup which includes all three impact types, material type classification accuracies were 96.5% for dynamic microphone and 97.7% for condenser microphone. When both the material type and the size of the wastes were classified, the accuracy was 88.6% for the microphones. The modeling studies indicated that hydraulic crushing impact type recordings were very noisy for an effective sound recognition application. In the detailed analysis of the recognition errors, it was observed that most of the errors occurred in the hitting impact type. According to the experimental results, it can be said that the proposed novel approach for the separation of packaging wastes could provide a high classification performance for RVMs. PMID:27378630

  5. Waste Materials from Tetra Pak Packages as Reinforcement of Polymer Concrete

    Directory of Open Access Journals (Sweden)

    Miguel Martínez-López

    2015-01-01

    Full Text Available Different concentrations (from 1 to 6 wt% and sizes (0.85, 1.40, and 2.36 mm of waste Tetra Pak particles replaced partially silica sand in polymer concrete. As is well known, Tetra Pak packages are made up of three raw materials: cellulose (75%, low density polyethylene (20%, and aluminum (5%. The polymer concrete specimens were elaborated with unsaturated polyester resin (20% and silica sand (80% and irradiated by using gamma rays at 100 and 200 kGy. The obtained results have shown that compressive and flexural strength and modulus of elasticity decrease gradually, when either Tetra Pak particle concentration or particle size is increased, as regularly occurs in composite materials. Nevertheless, improvements of 14% on both compressive strength and flexural strength as well as 5% for modulus of elasticity were obtained when polymer concrete is irradiated.

  6. Effect of Components on the Performance of Asphalt Modiifed by Waste Packaging Polyethylene

    Institute of Scientific and Technical Information of China (English)

    ZHANG Maorong; FANG Changqing; ZHOU Shisheng; CHENG Youliang; YU Ruien; LIU Shaolong; LIU Xiaolong; SU Jian

    2016-01-01

    Waste packaging polyethylene (WPE) was used to modify raw asphalt by melt blending the components at 190℃ for 1 h in a simple mixer and subsequently machining them at 120℃ for 1 h in a high-speed shearing machine. The effect of modiifcation on the degree of the penetration, the softening point and the ductility of the asphalt was studied using lfuorescent microscopy, infrared spectrometry, component changes and various other techniques. The experimental results showed that no chemical reactions took place in the components themselves (saturate, aromatic, asphaltene and resin) during the modifications. The softening point and penetration of the asphalt were found to be closely related to the resulting contents of the asphaltene, saturate and resin components. In addition, aromatics were identified as having the greatest impact on the ductility of the asphalt.

  7. Preliminary Criticality Analysis of Degraded SNF Accumulations to a Waste Package (SCPB: N/A) 

    Energy Technology Data Exchange (ETDEWEB)

    J.W. Davis

    2005-12-15

    This study is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide input to a separate evaluation on the probability of criticality in the far-field environment. These calculations are performed in sufficient detail to provide conservatively bounding configurations to support separate probabilistic analyses. The objective of this evaluation is to provide input to a risk analysis which will show that criticalities involving commercial spent nuclear fuel (SNF) are not credible, or indicate additional measures that are required for the Engineered Barrier Segment (EBS) to make such events incredible. Minimum critical volumes and masses of UO{sub 2}/H{sub 2}O/tuff mixtures are determined without application of regulatory safety limits. This study does not address or demonstrate compliance with regulatory limits.

  8. Near-Field Hydrology Data Package for the Immobilized Low-Activity Waste 2001 Performance Assessment

    International Nuclear Information System (INIS)

    Lockheed Martin Hanford Company (LMHC) is designing and assessing the performance of disposal facilities to receive radioactive wastes that are currently stored in single- and double-shell tanks at the Hanford Site. The preferred method for disposing of the portion that is classified as immobilized low-activity waste (ILAW) is to vitrify the waste and place the product in new-surface, shallow land burial facilities. The LMHC project to assess the performance of these disposal facilities is the Hanford ILAW Performance Assessment (PA) Activity. The goal of this project is to provide a reasonable expectation that the disposal of the waste is protective of the general public, groundwater resources, air resources, surface water resources, and inadvertent intruders. Achieving this goal will require prediction of contaminant migration from the facilities. This migration is expected to occur primarily via the movement of water through the facilities and the consequent transport of dissolved contaminants in the pore water of the vadose zone. Pacific Northwest National Laboratory (PNNL) assists LMHC in its performance assessment activities. One of PNNL's tasks is to provide estimates of the physical, hydraulic, and transport properties of the materials comprising the disposal facilities and the disturbed region around them. These materials are referred to as the near-field materials. Their properties are expressed as parameters of constitutive models used in simulations of subsurface flow and transport. In addition to the best-estimate parameter values, information on uncertainty in the parameter values and estimates of the changes in parameter values over time are required to complete the PA. These parameter estimates and information are contained in this report, the Near-Field Hydrology Data Package

  9. Geochemical data package for the Hanford immobilized low-activity tank waste performance assessment (ILAW PA)

    International Nuclear Information System (INIS)

    Lockheed Martin Hanford Company (LMHC) is designing and assessing the performance of disposal facilities to receive radioactive wastes that are stored in single- and double-shell tanks at the Hanford Site. The preferred method of disposing of the portion that is classified as low-activity waste is to vitrify the liquid/slurry and place the solid product in near-surface, shallow-land burial facilities. The LMHC project to assess the performance of these disposal facilities is the Hanford Immobilized Low-Activity Tank Waste (ILAW) Performance Assessment (PA) activity. The goal of this project is to provide a reasonable expectation that the disposal of the waste is protective of the general public, groundwater resources, air resources, surface-water resources, and inadvertent intruders. Achieving this goal will require prediction of contaminant migration from the facilities. This migration is expected to occur primarily via the movement of water through the facilities, and the consequent transport of dissolved contaminants in the porewater of the vadose zone. Pacific Northwest National Laboratory assists LMHC in their performance assessment activities. One of the PNNL tasks is to provide estimates of the geochemical properties of the materials comprising the disposal facility, the disturbed region around the facility, and the physically undisturbed sediments below the facility (including the vadose zone sediments and the aquifer sediments in the upper unconfined aquifer). The geochemical properties are expressed as parameters that quantify the adsorption of contaminants and the solubility constraints that might apply for those contaminants that may exceed solubility constraints. The common parameters used to quantify adsorption and solubility are the distribution coefficient (Kd) and the thermodynamic solubility product (Ksp), respectively. In this data package, the authors approximate the solubility of contaminants using a more simplified construct, called the

  10. Preliminary Systems Design Study assessment report. Volume 6, Waste Isolation Pilot Plant and transportation package acceptable concepts

    Energy Technology Data Exchange (ETDEWEB)

    Mayberry, J.L.; Feizollahi, F.; Del Signore, J.C.

    1992-01-01

    The System Design Study (SDS), part of the Waste Technology Development Department at the Idaho National Engineering Laboratory (INEL), examined techniques for the remediation of hazardous and transuranic waste stored at Radioactive Waste Management Complex`s Subsurface Disposal Area at the INEL. Using specific technologies, system concepts for treating the buried waste and the surrounding contaminated soil were evaluated. Evaluation included implementability, effectiveness, and cost. The SDS resulted in the development of technology requirements including demonstration, testing, and evaluation activities needed for implementing each concept. This volume contains introduction section containing a brief SDS background and lists the general assumptions and considerations used during the development of the system concepts. The introduction section is followed by sections describing two system concepts that produce a waste form in compliance with the Waste Isolation Pilot Plant (WIPP) Waste Acceptance Criteria (WAC) and transportation package (TRAMPAC) requirements. This system concept category is referred to as Waste Form 4, ``WIPP and TRAMPAC Acceptable.`` The following two system concepts are under this category: Sort, Treat, and Repackage System (4-BE-2); Volume Reduction and Packaging System (4-BE-4).

  11. Steam and air plasma gasification of bituminous coal and petrocoke

    Directory of Open Access Journals (Sweden)

    Vladimir Messerle

    2012-12-01

    Full Text Available This paper presents a numerical analysis and experimental investigation of two very different solid fuels, low-rank bituminous coal of 40 % ash content and petrocoke of 3 % ash content, gasification under steam and air plasma conditions with an aim of producing synthesis gas. Numerical analysis was fulfilled using the software package TERRA for equilibrium computation. Using the results of the numerical simulation, experiments on plasma steam gasification of the petrocoke and air and steam gasification of the coal were conducted in an original installation. Nominal power of the plasma installation is 100 kWe and sum consumption of the reagents is up to 20 kg/h. High quality synthesis gas was produced in the experiments on solid fuels plasma gasification. It has been found that the synthesis gas content at about 97.4 vol.% can be produced. Comparison between the numerical and experimental results showed satisfactory agreement. 

  12. Technical assessment of processing plants as exemplified by the sorting of beverage cartons from lightweight packaging wastes

    NARCIS (Netherlands)

    Feil, A.; Thoden van Velzen, E.U.; Jansen, M.; Vitz, P.; Go, N.; Pretz, T.

    2016-01-01

    The recovery of beverage cartons (BC) in three lightweight packaging waste processing plants (LP) was analyzed with different input materials and input masses in the area of 21-50. Mg. The data was generated by gravimetric determination of the sorting products, sampling and sorting analysis. Sinc

  13. 77 FR 17093 - Certain Food Waste Disposers and Components and Packaging Thereof: Notice of Receipt of Complaint...

    Science.gov (United States)

    2012-03-23

    ...Notice is hereby given that the U.S. International Trade Commission has received a complaint entitled Certain Food Waste Disposers and Components and Packaging Thereof, DN 2886; the Commission is soliciting comments on any public interest issues raised by the complaint or complainant's filing under section 210.8(b) of the Commission's Rules of Practice and Procedure (19 CFR...

  14. Reduced Pressure Electron Beam Welding Evaluation Activities on a Ni-Cr-Mo Alloy for Nuclear Waste Packages

    Energy Technology Data Exchange (ETDEWEB)

    Wong, F; Punshon, C; Dorsch, T; Fielding, P; Richard, D; Yang, N; Hill, M; DeWald, A; Rebak, R; Day, S; Wong, L; Torres, S; McGregor, M; Hackel, L; Chen, H-L; Rankin, J

    2003-09-11

    The current waste package design for the proposed repository at Yucca Mountain Nevada, USA, employs gas tungsten arc welding (GTAW) in fabricating the waste packages. While GTAW is widely used in industry for many applications, it requires multiple weld passes. By comparison, single-pass welding methods inherently use lower heat input than multi-pass welding methods which results in lower levels of weld distortion and also narrower regions of residual stresses at the weld TWI Ltd. has developed a Reduced Pressure Electron Beam (RPEB) welding process which allows EB welding in a reduced pressure environment ({le} 1 mbar). As it is a single-pass welding technique, use of RPEB welding could (1) achieve a comparable or better materials performance and (2) lead to potential cost savings in the waste package manufacturing as compared to GTAW. Results will be presented on the initial evaluation of the RPEB welding on a Ni-Cr-Mo alloy (a candidate alloy for the Yucca Mountain waste packages) in the areas of (a) design and manufacturing simplifications, (b) material performance and (c) weld reliability.

  15. Radioactive waste packages stored at the Aube facility for low-intermediate activity wastes. A selective and controlled storage; Les colis de dechets radioactifs stockes au centre de stockage FMA de l'Aube. Une stockage selectif et maitrise

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-07-01

    The waste package is the first barrier designed to protect the man and the environment from the radioactivity contained in wastes. Its design is thus particularly stringent and controlled. This brochure describes the different types of packages for low to intermediate activity wastes like those received and stored at the Aube facility, and also the system implemented by the ANDRA (the French national agency of radioactive wastes) and by waste producers to safely control each step of the design and fabrication of these packages. (J.S.)

  16. 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).

  17. Role of a buffer component within an engineered barrier waste package and a preliminary evaluation of bentonite as a backfill material

    International Nuclear Information System (INIS)

    This paper deals with the functions, properties, and compositions of backfill components to be used in the geologic disposal of high-level nuclear waste in basalt. A conceptual design for a repository located in basalt is being developed by the Basalt Waste Isolation Project (BWIP) in which these backfill components are part of the waste package and the repository sealing system (rooms, tunnels, and shafts). The first part of the paper concerns the role of a buffer component which is located between the primary and secondary physical barriers of the waste package (the canister and overpack). The second part of the paper deals with the chemical and physical properties of bentonite, which is a primary candidate for a backfill material both in the outer backfill barrier of the waste package and in the rooms, tunnels, and shafts above the waste package

  18. Investigation of metallic, ceramic, and polymeric materials for engineered barrier applications in nuclear-waste packages

    Energy Technology Data Exchange (ETDEWEB)

    Westerman, R.E.

    1980-10-01

    An effort to develop licensable engineered barrier systems for the long-term (about 1000 yr) containment of nuclear wastes under conditions of deep continental geologic disposal has been underway at Pacific Northwest Laboratory since January 1979, under the auspices of the High-Level Waste Immobilization Program. In the present work, the barrier system comprises the hard or structural elements of the package: the canister, the overpack(s), and the hole sleeve. A number of candidate metallic, ceramic, and polymeric materials were put through mechanical, corrosion, and leaching screening tests to determine their potential usefulness in barrier-system applications. Materials demonstrating adequate properties in the screening tests will be subjected to more detailed property tests, and, eventually, cost/benefit analyses, to determine their ultimate applicability to barrier-system design concepts. The following materials were investigated: two titanium alloys of Grade 2 and Grade 12; 300 and 400 series stainless steels, Inconels, Hastelloy C-276, titanium, Zircoloy, copper-nickel alloys and cast irons; total of 14 ceramic materials, including two grades of alumina, plus graphite and basalt; and polymers such as polyamide-imide, polyarylene, polyimide, polyolefin, polyphenylene sulfide, polysulfone, fluoropolymer, epoxy, furan, silicone, and ethylene-propylene terpolymer (EPDM) rubber. The most promising candidates for further study and potential use in engineered barrier systems were found to be rubber, filled polyphenylene sulfide, fluoropolymer, and furan derivatives.

  19. Investigation of metallic, ceramic, and polymeric materials for engineered barrier applications in nuclear-waste packages

    International Nuclear Information System (INIS)

    An effort to develop licensable engineered barrier systems for the long-term (about 1000 yr) containment of nuclear wastes under conditions of deep continental geologic disposal has been underway at Pacific Northwest Laboratory since January 1979, under the auspices of the High-Level Waste Immobilization Program. In the present work, the barrier system comprises the hard or structural elements of the package: the canister, the overpack(s), and the hole sleeve. A number of candidate metallic, ceramic, and polymeric materials were put through mechanical, corrosion, and leaching screening tests to determine their potential usefulness in barrier-system applications. Materials demonstrating adequate properties in the screening tests will be subjected to more detailed property tests, and, eventually, cost/benefit analyses, to determine their ultimate applicability to barrier-system design concepts. The following materials were investigated: two titanium alloys of Grade 2 and Grade 12; 300 and 400 series stainless steels, Inconels, Hastelloy C-276, titanium, Zircoloy, copper-nickel alloys and cast irons; total of 14 ceramic materials, including two grades of alumina, plus graphite and basalt; and polymers such as polyamide-imide, polyarylene, polyimide, polyolefin, polyphenylene sulfide, polysulfone, fluoropolymer, epoxy, furan, silicone, and ethylene-propylene terpolymer (EPDM) rubber. The most promising candidates for further study and potential use in engineered barrier systems were found to be rubber, filled polyphenylene sulfide, fluoropolymer, and furan derivatives

  20. The MEDINA facility for the assay of the chemotoxic inventory of radioactive waste packages

    International Nuclear Information System (INIS)

    A new analytical facility called MEDINA (Multi Element Determination based on Instrumental Neutron Activation) and based on Prompt- and Delayed-Gamma-Neutron-Activation-Analysis (P and DGNAA) using a 14 MeV neutron generator is developed for the determination of non-radioactive elements and substances in 200-l radioactive waste drums. The qualitative performance of MEDINA is studied by measuring (1) the gamma-ray spectra for the empty chamber, (2) for an empty 200-l steel drum in presence or not of 60Co and 137Cs sources and (3) for a 200-l steel drum filled with concrete in order to determine the level of interferences induced by activation products of the facility components and to investigate further optimization steps to improve the measurements sensitivity. The measurement carried out for the empty drum with the radioactive sources show the possibility to characterize radioactive waste packages containing both 60Co and 137Cs with activities ranging between 20 and 80 MBq at least. (author)

  1. Geotechnical, Hydrogeologic and Vegetation Data Package for 200-UW-1 Waste Site Engineered Surface Barrier Design

    Energy Technology Data Exchange (ETDEWEB)

    Ward, Andy L.

    2007-11-26

    Fluor Hanford (FH) is designing and assessing the performance of engineered barriers for final closure of 200-UW-1 waste sites. Engineered barriers must minimize the intrusion and water, plants and animals into the underlying waste to provide protection for human health and the environment. The Pacific Northwest National Laboratory (PNNL) developed Subsurface Transport Over Multiple Phases (STOMP) simulator is being used to optimize the performance of candidate barriers. Simulating barrier performance involves computation of mass and energy transfer within a soil-atmosphere-vegetation continuum and requires a variety of input parameters, some of which are more readily available than others. Required input includes parameter values for the geotechnical, physical, hydraulic, and thermal properties of the materials comprising the barrier and the structural fill on which it will be constructed as well as parameters to allow simulation of plant effects. This report provides a data package of the required parameters as well as the technical basis, rationale and methodology used to obtain the parameter values.

  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. Assessment of collection schemes for packaging and other recyclable waste in European Union-28 Member States and capital cities.

    Science.gov (United States)

    Seyring, Nicole; Dollhofer, Marie; Weißenbacher, Jakob; Bakas, Ioannis; McKinnon, David

    2016-09-01

    The Waste Framework Directive obliged European Union Member States to set up separate collection systems to promote high quality recycling for at least paper, metal, plastic and glass by 2015. As implementation of the requirement varies across European Union Member States, the European Commission contracted BiPRO GmbH/Copenhagen Resource Institute to assess the separate collection schemes in the 28 European Union Member States, focusing on capital cities and on metal, plastic, glass (with packaging as the main source), paper/cardboard and bio-waste. The study includes an assessment of the legal framework for, and the practical implementation of, collection systems in the European Union-28 Member States and an in depth-analysis of systems applied in all capital cities. It covers collection systems that collect one or more of the five waste streams separately from residual waste/mixed municipal waste at source (including strict separation, co-mingled systems, door-to-door, bring-point collection and civic amenity sites). A scoreboard including 13 indicators is elaborated in order to measure the performance of the systems with the capture rates as key indicators to identify best performers. Best performance are by the cities of Ljubljana, Helsinki and Tallinn, leading to the key conclusion that door-to-door collection, at least for paper and bio-waste, and the implementation of pay-as-you-throw schemes results in high capture and thus high recycling rates of packaging and other municipal waste. PMID:27357560

  4. 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.

  5. “零废弃”包装设计案例探析%CASE STUDY OF "ZERO WASTE" PACKAGING DESIGN

    Institute of Scientific and Technical Information of China (English)

    刘春雷; 翟健含; 施斌杰

    2015-01-01

    目的:现阶段包装废弃物在资源污染与浪费中占有很大比例。所以设计师们在如何处理对包装废弃物的探讨设计也一直没有停歇。方法:“零废弃”包装设计作为绿色设计的一种。是在包装设计前要考虑包装生命周期的一种设计方法。使得通过“零废弃”设计后的包装在完成包装的基本功能后,可以作为一个独立物品而单独使用。结论:这是一种对包装废弃物的绿色设计探析。采用“零废弃”包装可以减少包装废弃物的增加,减轻环境的污染。%Objective: At present packaging waste occupies a large proportion in the polution and waste of resources. So the designers explores in how to deal with the design of packaging waste. Methods: "Zero waste" packaging design is a kind of green design. It is packaging design method of the life cycle in front of the packaging design. After the" zero waste "design in the completion of the basic function of packaging, the packaging can be used alone as an independent materials. Conclusion: It is a kind of green design and analysis of packaging waste. Using" zero waste "packaging can reduce the increase of packaging waste and reduce environmental polution.

  6. Problem of the NPP liquid radioactive wastes processing and disposal

    International Nuclear Information System (INIS)

    Modern methods of NPP radioactive waste processing and disposal are briefly presented, bituminization with the following disposal in the clayey soil in particular. Soviet installations of liquid waste bituminization and results of proving ground preservation of bituminic blocks are briefly described. These results indicate a possibility of bituminic material disposal with specific activity of 1 Ci/l directly in the soil without waterproofing. High safety and effectivity of waste bituminization is shown in comparison with preservation variant of liquid radioactive concentrates in capacities

  7. Recharge Data Package for the Immobilized Low-Activity Waste 2001 Performance Assessment

    Energy Technology Data Exchange (ETDEWEB)

    MJ Fayer; EM Murphy; JL Downs; FO Khan; CW Lindenmeier; BN Bjornstad

    2000-01-18

    Lockheed Martin Hanford Company (LMHC) is designing and assessing the performance of disposal facilities to receive radioactive wastes that are currently stored in single- and double-shell tanks at the Hanford Site. The preferred method of disposing of the portion that is classified as immobilized low-activity waste (ILAW) is to vitrify the waste and place the product in near-surface, shallow-land burial facilities. The LMHC project to assess the performance of these disposal facilities is known as the Hanford ILAW Performance Assessment (PA) Activity, hereafter called the ILAW PA project. The goal of this project is to provide a reasonable expectation that the disposal of the waste is protective of the general public, groundwater resources, air resources, surface-water resources, and inadvertent intruders. Achieving this goal will require predictions of contaminant migration from the facility. To make such predictions will require estimates of the fluxes of water moving through the sediments within the vadose zone around and beneath the disposal facility. These fluxes, loosely called recharge rates, are the primary mechanism for transporting contaminants to the groundwater. Pacific Northwest National Laboratory (PNNL) assists LMHC in their performance assessment activities. One of the PNNL tasks is to provide estimates of recharge rates for current conditions and long-term scenarios involving the shallow-land disposal of ILAW. Specifically, recharge estimates are needed for a filly functional surface cover; the cover sideslope, and the immediately surrounding terrain. In addition, recharge estimates are needed for degraded cover conditions. The temporal scope of the analysis is 10,000 years, but could be longer if some contaminant peaks occur after 10,000 years. The elements of this report compose the Recharge Data Package, which provides estimates of recharge rates for the scenarios being considered in the 2001 PA. Table S.1 identifies the surface features and

  8. Recharge Data Package for the Immobilized Low-Activity Waste 2001 Performance Assessment

    International Nuclear Information System (INIS)

    Lockheed Martin Hanford Company (LMHC) is designing and assessing the performance of disposal facilities to receive radioactive wastes that are currently stored in single- and double-shell tanks at the Hanford Site. The preferred method of disposing of the portion that is classified as immobilized low-activity waste (ILAW) is to vitrify the waste and place the product in near-surface, shallow-land burial facilities. The LMHC project to assess the performance of these disposal facilities is known as the Hanford ILAW Performance Assessment (PA) Activity, hereafter called the ILAW PA project. The goal of this project is to provide a reasonable expectation that the disposal of the waste is protective of the general public, groundwater resources, air resources, surface-water resources, and inadvertent intruders. Achieving this goal will require predictions of contaminant migration from the facility. To make such predictions will require estimates of the fluxes of water moving through the sediments within the vadose zone around and beneath the disposal facility. These fluxes, loosely called recharge rates, are the primary mechanism for transporting contaminants to the groundwater. Pacific Northwest National Laboratory (PNNL) assists LMHC in their performance assessment activities. One of the PNNL tasks is to provide estimates of recharge rates for current conditions and long-term scenarios involving the shallow-land disposal of ILAW. Specifically, recharge estimates are needed for a filly functional surface cover; the cover sideslope, and the immediately surrounding terrain. In addition, recharge estimates are needed for degraded cover conditions. The temporal scope of the analysis is 10,000 years, but could be longer if some contaminant peaks occur after 10,000 years. The elements of this report compose the Recharge Data Package, which provides estimates of recharge rates for the scenarios being considered in the 2001 PA. Table S.1 identifies the surface features and

  9. Geochemical data package for the Hanford immobilized low-activity tank waste performance assessment (ILAW PA)

    Energy Technology Data Exchange (ETDEWEB)

    DI Kaplan; RJ Serne

    2000-02-24

    Lockheed Martin Hanford Company (LMHC) is designing and assessing the performance of disposal facilities to receive radioactive wastes that are stored in single- and double-shell tanks at the Hanford Site. The preferred method of disposing of the portion that is classified as low-activity waste is to vitrify the liquid/slurry and place the solid product in near-surface, shallow-land burial facilities. The LMHC project to assess the performance of these disposal facilities is the Hanford Immobilized Low-Activity Tank Waste (ILAW) Performance Assessment (PA) activity. The goal of this project is to provide a reasonable expectation that the disposal of the waste is protective of the general public, groundwater resources, air resources, surface-water resources, and inadvertent intruders. Achieving this goal will require prediction of contaminant migration from the facilities. This migration is expected to occur primarily via the movement of water through the facilities, and the consequent transport of dissolved contaminants in the porewater of the vadose zone. Pacific Northwest National Laboratory assists LMHC in their performance assessment activities. One of the PNNL tasks is to provide estimates of the geochemical properties of the materials comprising the disposal facility, the disturbed region around the facility, and the physically undisturbed sediments below the facility (including the vadose zone sediments and the aquifer sediments in the upper unconfined aquifer). The geochemical properties are expressed as parameters that quantify the adsorption of contaminants and the solubility constraints that might apply for those contaminants that may exceed solubility constraints. The common parameters used to quantify adsorption and solubility are the distribution coefficient (K{sub d}) and the thermodynamic solubility product (K{sub sp}), respectively. In this data package, the authors approximate the solubility of contaminants using a more simplified construct

  10. Upgrading of recycled plastics obtained from flexible packaging waste by adding nanosilicates

    Science.gov (United States)

    Garofalo, E.; Claro, M.; Scarfato, P.; Di Maio, L.; Incarnato, L.

    2015-12-01

    Currently, the growing consumption of polymer products creates large quantities of waste materials resulting in public concern in the environment and people life. The efficient treatment of polymer wastes is still a difficult challenge and the recycling process represents the best way to manage them. Recently, many researchers have tried to develop nanotechnology for polymer recycling. The products prepared through the addition of nanoparticles to post-used plastics could offer the combination of improved properties, low weight, easy of processing and low cost which is not easily and concurrently found by other methods of plastic recycling. In this study materials, obtained by the separation and mechanical recycling of post-consumer packaging films of small size (

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

    Energy Technology Data Exchange (ETDEWEB)

    Marusich, Robert M.

    2013-08-15

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

  12. Crumb Rubber in cold recycled bituminous mixes

    DEFF Research Database (Denmark)

    Dondi, Giulio; Tataranni, Piergiorgio; Pettinari, Matteo;

    2014-01-01

    Today recycling is one of the most innovative and interesting techniques for the rehabilitation of destressed road pavements. In recent years the increased interest in this process, has led to the development of various alternative methods for the recovery and the reuse of road bituminous materials....... Cold recycling is, among the recycling techniques, certainly the most studied and developed: it allows the recovering of bituminous material from an existing pavement without the addition of heat, whilst ensuring the creation of high quality bound base layers. A wide range of materials have been tested...... together with Reclaimed Asphalt Pavement (RAP) and, consequently, there is a large number of variables that can be considered in the mix-design process of new eco-friendly Cold Recycled Mixes. In particular, the present research involves the use of Crumb Rubber within a mixture containing 100% Reclaimed...

  13. Tritium Packages and 17th RH Canister Categories of Transuranic Waste Stored Below Ground within Area G

    Energy Technology Data Exchange (ETDEWEB)

    Hargis, Kenneth Marshall [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-03-01

    A large wildfire called the Las Conchas Fire burned large areas near Los Alamos National Laboratory (LANL) in 2011 and heightened public concern and news media attention over transuranic (TRU) waste stored at LANL’s Technical Area 54 (TA-54) Area G waste management facility. The removal of TRU waste from Area G had been placed at a lower priority in budget decisions for environmental cleanup at LANL because TRU waste removal is not included in the March 2005 Compliance Order on Consent (Reference 1) that is the primary regulatory driver for environmental cleanup at LANL. The Consent Order is a settlement agreement between LANL and the New Mexico Environment Department (NMED) that contains specific requirements and schedules for cleaning up historical contamination at the LANL site. After the Las Conchas Fire, discussions were held by the U.S. Department of Energy (DOE) with the NMED on accelerating TRU waste removal from LANL and disposing it at the Waste Isolation Pilot Plant (WIPP). This report summarizes available information on the origin, configuration, and composition of the waste containers within the Tritium Packages and 17th RH Canister categories; their physical and radiological characteristics; the results of the radioassays; and potential issues in retrieval and processing of the waste containers.

  14. 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

  15. Preparation, Characterization and Hot Storage Stability of Asphalt Modified by Waste Polyethylene Packaging

    Institute of Scientific and Technical Information of China (English)

    Changqing Fang; Ying Zhang; Qian yu; Xing Zhou; Dagang Guo; Ruien Yu; Min Zhang

    2013-01-01

    Waste polyethylene packaging (WPE) was used to modify asphalt,and hot storage stability of the modified asphalt was studied in this paper.The morphological change and component loss of WPE modified asphalt were characterized by fluorescence microscopy,Fourier transform infrared spectroscopy (FT-IR),differential scanning calorimetry (DSC),thermogravimetry (TG) and isolation testing.In addition,the mechanism of the hot storage stability of WPE modified asphalt was discussed.The results showed that the modification of asphalt with WPE was a physical process.It was found that the filament or partly network-like structure formed in the modified asphalt system was beneficial to improving the hot storage stability.Moreover,the addition of WPE resulted in a decrease in both the light components volatilization and the macromolecules decomposition of asphalt.It was demonstrated that when the content of WPE in matrix asphalt was less than 10 wt%,the service performances of modified asphalt could be better.

  16. Different unit price of packaging sizes matter for food waste : A case study how companies' price strategies of different packaging sizes of dairy products affect the emissions of CO2 equivalents

    OpenAIRE

    Lövdahl, Maria; Palmér, Karin

    2015-01-01

    One third of all food that is produced ends up as waste, making it an important for society. Companies in food sector offer various packaging sizes to consumers to meet different household needs, where smaller packaging sizes often have a higher unit price than larger packaging sizes. In Sweden, 70% of households consist of 1-2 person households, which indicate a need for smaller packaging sizes. Price is a determining factor in the choice of products, which may mean that consumers choose pac...

  17. Noise Reduction Properties of an Experimental Bituminous Slurry with Crumb Rubber Incorporated by the Dry Process

    Directory of Open Access Journals (Sweden)

    Moisés Bueno

    2014-08-01

    Full Text Available Nowadays, cold technology for asphalt pavement in the field of road construction is considered as an alternative solution to conventional procedures from both an economic and environmental point of view. Among these techniques, bituminous slurry surfacing is obtaining an important role due to the properties of the obtained wearing course. The functional performance of this type of surfaces is directly related to its rough texture. Nevertheless, this parameter has a significant influence on the tire/road noise generation. To reduce this undesirable effect on the sound performance, new designs of elastic bituminous slurries have been developed. Within the FENIX project, this work presents the acoustical characterization of an experimental bituminous slurry with crumb rubber from wasted automobile tires incorporated by the dry process. The obtained results show that, under controlled operational parameters, the close proximity sound levels associated to the experimental slurry are considerably lower than those emitted by a conventional slurry wearing course. However, after one year of supporting traffic loads and different weather conditions, the evaluated bituminous slurry, although it conserves the original noise reduction properties in relation to the conventional one, noticeably increases the generated sound emission. Therefore, it is required to continue improving the design of experimental surfaces in order to enhance its long-term performance.

  18. Evaluation and compilation of DOE [Department of Energy] waste package test data; Biannual report, February 1988--July 1988

    Energy Technology Data Exchange (ETDEWEB)

    Interrante, C.; Escalante, E.; Fraker, A.; Plante, E.

    1989-10-01

    This report summarizes evaluations by the National Institute of Standards and Technology (NIST) of Department of Energy (DOE) activities on waste packages designed for containment of radioactive high-level nuclear waste (HLW) for the six month period February 1988 through July 1988. Activities for the DOE Materials Characterization Center are reviewed for the period January 1988 through June 1988. A summary is given of the Yucca Mountain, Nevada disposal site activities. Short discussions relating to the reviewed publications are given and complete reviews and evaluations are included. 20 refs., 1 fig., 1 tab.

  19. Evaluation and compilation of DOE waste package test data; Volume 8: Biannual report, August 1989--January 1990

    Energy Technology Data Exchange (ETDEWEB)

    Interrante, C.G. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of High-Level Waste Management; Fraker, A.C.; Escalante, E. [National Inst. of Standards and Technology (MSEL), Gaithersburg, MD (United States). Metallurgy Div.

    1993-06-01

    This report summarizes evaluations by the National Institute of Standards and Technology (NIST) of some of the Department of Energy (DOE) activities on waste packages designed for containment of radioactive high-level nuclear waste (HLW) for the six-month period, August 1989--January 1990. This includes reviews of related materials research and plans, information on the Yucca Mountain, Nevada disposal site activities, and other information regarding supporting research and special assistance. Short discussions are given relating to the publications reviewed and complete reviews and evaluations are included. Reports of other work are included in the Appendices.

  20. FINITE-ELEMENT ANALYSIS OF ROCK FALL ON UNCANISTERED FUEL WASTE PACKAGE DESIGNS (SCPB: N/A)

    Energy Technology Data Exchange (ETDEWEB)

    Z. Ceylan

    1996-10-18

    The objective of this analysis is to explore the Uncanistered Fuel (UCF) Tube Design waste package (WP) resistance to rock falls. This analysis will also be used to determine the size of rock that can strike the WP without causing failure in the containment barriers from a height based on the starter tunnel dimensions. The purpose of this analysis is to document the models and methods used in the calculations.

  1. Selection of candidate container materials for the conceptual waste package design for a potential high level nuclear waste repository at Yucca Mountain

    Energy Technology Data Exchange (ETDEWEB)

    Van Konynenburg, R.A.; Halsey, W.G.; McCright, R.D.; Clarke, W.L. Jr. [Lawrence Livermore National Lab., CA (United States); Gdowski, G.E. [KMI, Inc., Albuquerque, NM (United States)

    1993-02-01

    Preliminary selection criteria have been developed, peer-reviewed, and applied to a field of 41 candidate materials to choose three alloys for further consideration during the advanced conceptual design phase of waste package development for a potential high level nuclear waste repository at Yucca Mountain, Nevada. These three alloys are titanium grade 12, Alloy C-4, and Alloy 825. These selections are specific to the particular conceptual design outlined in the Site Characterization Plan. Other design concepts that may be considered in the advanced conceptual design phase may favor other materials choices.

  2. Selection of candidate container materials for the conceptual waste package design for a potential high level nuclear waste repository at Yucca Mountain

    International Nuclear Information System (INIS)

    Preliminary selection criteria have been developed, peer-reviewed, and applied to a field of 41 candidate materials to choose three alloys for further consideration during the advanced conceptual design phase of waste package development for a potential high level nuclear waste repository at Yucca Mountain, Nevada. These three alloys are titanium grade 12, Alloy C-4, and Alloy 825. These selections are specific to the particular conceptual design outlined in the Site Characterization Plan. Other design concepts that may be considered in the advanced conceptual design phase may favor other materials choices

  3. Study on Plastic Coated Overburnt Brick Aggregate as an Alternative Material for Bituminous Road Construction

    OpenAIRE

    Sarkar, Dipankar; Pal, Manish; Sarkar, Ashoke K.

    2016-01-01

    There are different places in India where natural stone aggregates are not available for constructional work. Plastic coated OBBA can solve the problem of shortage of stone aggregate to some extent. The engineers are always encouraged to use locally available materials. The present investigation is carried out to evaluate the plastic coated OBBA as an alternative material for bituminous road construction. Shredded waste plastics are mixed with OBBA in different percentages as 0.38, 0.42, 0.46...

  4. Waste Generator Instructions: Key to Successful Implementation of the US DOE's 435.1 for Transuranic Waste Packaging Instructions (LA-UR-12-24155) - 13218

    Energy Technology Data Exchange (ETDEWEB)

    French, David M. [LANL EES-12, Carlsbad, NM, Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Hayes, Timothy A. [LANL EES-12, Carlsbad, NM, Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Pope, Howard L. [Aspen Resources Ltd., Inc., P.O. Box 3038, Boulder, CO 80307 (United States); Enriquez, Alejandro E. [LANL NCO-4, Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Carson, Peter H. [LANL NPI-7, Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States)

    2013-07-01

    In times of continuing fiscal constraints, a management and operation tool that is straightforward to implement, works as advertised, and virtually ensures compliant waste packaging should be carefully considered and employed wherever practicable. In the near future, the Department of Energy (DOE) will issue the first major update to DOE Order 435.1, Radioactive Waste Management. This update will contain a requirement for sites that do not have a Waste Isolation Pilot Plant (WIPP) waste certification program to use two newly developed technical standards: Contact-Handled Defense Transuranic Waste Packaging Instructions and Remote-Handled Defense Transuranic Waste Packaging Instructions. The technical standards are being developed from the DOE O 435.1 Notice, Contact-Handled and Remote-Handled Transuranic Waste Packaging, approved August 2011. The packaging instructions will provide detailed information and instruction for packaging almost every conceivable type of transuranic (TRU) waste for disposal at WIPP. While providing specificity, the packaging instructions leave to each site's own discretion the actual mechanics of how those Instructions will be functionally implemented at the floor level. While the Technical Standards are designed to provide precise information for compliant packaging, the density of the information in the packaging instructions necessitates a type of Rosetta Stone that translates the requirements into concise, clear, easy to use and operationally practical recipes that are waste stream and facility specific for use by both first line management and hands-on operations personnel. The Waste Generator Instructions provide the operator with step-by-step instructions that will integrate the sites' various operational requirements (e.g., health and safety limits, radiological limits or dose limits) and result in a WIPP certifiable waste and package that can be transported to and emplaced at WIPP. These little known but widely

  5. Continuous using of the scaling factors for radionuclide evaluation in the packaged solid wastes originated from the Japanese Nuclear Power Plants since 2003

    International Nuclear Information System (INIS)

    The amounts and concentration of the nuclides in the waste packages are estimated by measuring some key nuclides, mostly gamma emitters, from outside of the packages and by applying the scaling factor method (using the relationship between some easy to measure key nuclides and the other difficult to measure nuclides). The solid wastes are classified into two kinds of packages: homogeneous solid wastes made from concentrated liquid wastes and spent fuels solidified with cement asphalt, or plastics and heterogeneous solid wastes made of cutting metals, compacted or fused filters solidified with mortars. Japan Nuclear Energy Safety Organization (JNES) established in 2005 is in charge of the confirmation of the inside contents with radionuclide information and compliance with formalities for safety maintenance and control. (S. Ohno)

  6. Conceptual designs for waste packages for horizontal or vertical emplacement in a repository in salt for reference in the site characterization plan

    International Nuclear Information System (INIS)

    This report includes the options of horizontal and vertical emplacement, the addition of a phased repository, an additional waste form (intact spent fuel), revised geotechnical data appropriate for the Deaf Smith County site, new corrosion data for the container, and new repository design data. The waste package consists of waste form and canister within a thick-walled, low-carbon steel container surrounded by packing. The container is a hollow cylinder with a flat head welded to each end. The design concepts for the waste container or vertical and horizontal emplacement are identical. This report discusses the results of analyses of aspects of the reference waste package concept needing changes because of new data and information believed applicable to the Deaf Smith County site. Included are waste package conceptual designs or (1) the reference defense high-level waste form from the Savannah River Plant; (2) intact spent fuel with our pressurized-water-reactor or nine boiling-water-reactor assemblies per package for emplacement during Phase 1 of repository operation; and (3) spent fuel which has been disassembled and consolidated into a segmented cylindrical canister with rods from either 12 pressurized-water-reactor or 30 boiling-water-reactor assemblies per package for emplacement during Phase 2. 30 refs., 61 figs., 30 tabs

  7. Introduction to Radcalc: A computer program to calculate the radiolytic production of hydrogen gas from radioactive wastes in packages

    International Nuclear Information System (INIS)

    A calculational technique for quantifying the concentration of hydrogen generated by radiolysis in sealed radioactive waste containers was developed in a U.S. Department of Energy (DOE) study conducted by EG ampersand G Idaho, Inc., and the Electric Power Research Institute (EPRI) TMI-2 Technology Transfer Office. The study resulted in report GEND-041, entitled open-quotes A Calculational Technique to Predict Combustible Gas Generation in Sealed Radioactive Waste Containersclose quotes. The study also resulted in a presentation to the U.S. Nuclear Regulatory Commission (NRC) which gained acceptance of the methodology for use in ensuring compliance with NRC IE Information Notice No. 84-72 (NRC 1984) concerning the generation of hydrogen within packages. NRC IE Information Notice No. 84-72: open-quotes Clarification of Conditions for Waste Shipments Subject to Hydrogen Gas Generationclose quotes applies to any package containing water and/or organic substances that could radiolytically generate combustible gases. EPRI developed a simple computer program in a spreadsheet format utilizing GEND-041 calculational methodology to predict hydrogen gas concentrations in low-level radioactive wastes containers termed Radcalc. The computer code was extensively benchmarked against TMI-2 (Three Mile Island) EPICOR II resin bed measurements. The benchmarking showed that the model developed predicted hydrogen gas concentrations within 20% of the measured concentrations. Radcalc for Windows was developed using the same calculational methodology. The code is written in Microsoft Visual C++ 2.0 and includes a Microsoft Windows compatible menu-driven front end. In addition to hydrogen gas concentration calculations, Radcalc for Windows also provides transportation and packaging information such as pressure buildup, total activity, decay heat, fissile activity, TRU activity, and transportation classifications

  8. Long term behaviour of low and intermediate level waste packages under repository conditions. Results of a co-ordinated research project 1997-2002

    International Nuclear Information System (INIS)

    The development and application of approaches and technologies that provide long term safety is an essential issue in the disposal of radioactive waste. For low and intermediate level radioactive waste, engineered barriers play an important role in the overall safety and performance of near surface repositories. Thus, developing a strong technical basis for understanding the behaviour and performance of engineered barriers is an important consideration in the development and establishment of near surface repositories for radioactive waste. In 1993, a Co-ordinated Research Project (CRP) on Performance of Engineered Barrier Materials in Near Surface Disposal Facilities for Radioactive Waste was initiated by the IAEA with the twin goals of addressing some of the gaps in the database on radionuclide isolation and long term performance of a wide variety of materials and components that constitute the engineered barriers system (IAEA-TECDOC-1255 (2001)). However, during the course of the CRP, it was realized that that the scope of the CRP did not include studies of the behaviour of waste packages over time. Given that a waste package represents an important component of the overall near surface disposal system and the fact that many Member States have active R and D programmes related to waste package testing and evaluation, a new CRP was launched, in 1997, on Long Term Behaviour of Low and Intermediate Level Waste Packages Under Repository Conditions. The CRP was intended to promote research activities on the subject area in Member States, share information on the topic among the participating countries, and contribute to advancing technologies for near surface disposal of radioactive waste. Thus, this CRP complements the afore mentioned CRP on studies of engineered barriers. With the active participation and valuable contributions from twenty scientists and engineers from Argentina, Canada, Czech Republic, Egypt, Finland, India, Republic of Korea, Norway, Romania

  9. Design of an innovative, ecological portable waste compressor for in-house recycling of paper, plastic and metal packaging waste.

    Science.gov (United States)

    Xevgenos, D; Athanasopoulos, N; Kostazos, P K; Manolakos, D E; Moustakas, K; Malamis, D; Loizidou, M

    2015-05-01

    Waste management in Greece relies heavily on unsustainable waste practices (mainly landfills and in certain cases uncontrolled dumping of untreated waste). Even though major improvements have been achieved in the recycling of municipal solid waste during recent years, there are some barriers that hinder the achievement of high recycling rates. Source separation of municipal solid waste has been recognised as a promising solution to produce high-quality recycled materials that can be easily directed to secondary materials markets. This article presents an innovative miniature waste separator/compressor that has been designed and developed for the source separation of municipal solid waste at a household level. The design of the system is in line with the Waste Framework Directive (2008/98/EC), since it allows for the separate collection (and compression) of municipal solid waste, namely: plastic (polyethylene terephthalate and high-density polyethylene), paper (cardboard and Tetrapak) and metal (aluminium and tin cans). It has been designed through the use of suitable software tools (LS-DYNA, INVENTROR and COMSOL). The results from the simulations, as well as the whole design process and philosophy, are discussed in this article.

  10. Design of an innovative, ecological portable waste compressor for in-house recycling of paper, plastic and metal packaging waste.

    Science.gov (United States)

    Xevgenos, D; Athanasopoulos, N; Kostazos, P K; Manolakos, D E; Moustakas, K; Malamis, D; Loizidou, M

    2015-05-01

    Waste management in Greece relies heavily on unsustainable waste practices (mainly landfills and in certain cases uncontrolled dumping of untreated waste). Even though major improvements have been achieved in the recycling of municipal solid waste during recent years, there are some barriers that hinder the achievement of high recycling rates. Source separation of municipal solid waste has been recognised as a promising solution to produce high-quality recycled materials that can be easily directed to secondary materials markets. This article presents an innovative miniature waste separator/compressor that has been designed and developed for the source separation of municipal solid waste at a household level. The design of the system is in line with the Waste Framework Directive (2008/98/EC), since it allows for the separate collection (and compression) of municipal solid waste, namely: plastic (polyethylene terephthalate and high-density polyethylene), paper (cardboard and Tetrapak) and metal (aluminium and tin cans). It has been designed through the use of suitable software tools (LS-DYNA, INVENTROR and COMSOL). The results from the simulations, as well as the whole design process and philosophy, are discussed in this article. PMID:25819929

  11. Innovative Soft-Sided Waste Packaging System Implementation at a Small Department of Energy Environmental Restoration/Waste Management (ER/WM) Site

    Energy Technology Data Exchange (ETDEWEB)

    Wolf, J.

    2002-02-28

    Weiss Associates (WA) performs a broad range of environmental restoration/waste management (ER/WM) activities for the U.S. Department of Energy (DOE) at the former Laboratory for Energy-Related Health Research (LEHR), University of California, Davis (UC Davis). Over the last three years, the LEHR ER/WM program transitioned from a baseline packaging system of steel, 2.7 cubic meter (3.5-cubic yard) B-25 boxes to a 7.0 cubic meter (9.1-cubic yard) soft-sided container (Lift Liner) system. The transition increased efficiencies in processing, packaging, and storage, and when combined with decreased procurement costs, achieved a $402,000 cost savings (Table I). Additional disposal costs between $128,600 and $182,600 were avoided by minimizing void space. Future cost savings by the end of fiscal year 2003 are projected between $250,640 and $1,003,360.

  12. Repository environmental parameters and models/methodologies relevant to assessing the performance of high-level waste packages in basalt, tuff, and salt

    International Nuclear Information System (INIS)

    This document provides specifications for models/methodologies that could be employed in determining postclosure repository environmental parameters relevant to the performance of high-level waste packages for the Basalt Waste Isolation Project (BWIP) at Richland, Washington, the tuff at Yucca Mountain by the Nevada Test Site, and the bedded salt in Deaf Smith County, Texas. Guidance is provided on the identify of the relevant repository environmental parameters; the models/methodologies employed to determine the parameters, and the input data base for the models/methodologies. Supporting studies included are an analysis of potential waste package failure modes leading to identification of the relevant repository environmental parameters, an evaluation of the credible range of the repository environmental parameters, and a summary of the review of existing models/methodologies currently employed in determining repository environmental parameters relevant to waste package performance. 327 refs., 26 figs., 19 tabs

  13. Report to Congress on the potential use of lead in the waste packages for a geologic repository at Yucca Mountain, Nevada

    International Nuclear Information System (INIS)

    In the Report of the Senate Committee on Appropriations accompanying the Energy and Water Appropriation Act for 1989, the Committee directed the Department of Energy (DOE) to evaluate the use of lead in the waste packages to be used in geologic repositories for spent nuclear fuel and high-level waste. The evaluation that was performed in response to this directive is presented in this report. This evaluation was based largely on a review of the technical literature on the behavior of lead, reports of work conducted in other countries, and work performed for the waste-management program being conducted by the DOE. The initial evaluation was limited to the potential use of lead in the packages to be used in the repository. Also, the focus of this report is post closure performance and not on retrievability and handling aspects of the waste package. 100 refs., 8 figs., 15 tabs

  14. Response Modelling of Bitumen, Bituminous Mastic and Mortar

    NARCIS (Netherlands)

    Woldekidan, M.F.

    2011-01-01

    This research focuses on testing and modelling the viscoelastic response of bituminous binders. The main goal is to find an appropriate response model for bituminous binders. The desired model should allow implementation into numerical environments such as ABAQUS. On the basis of such numerical envi

  15. The impact of policy interactions on the recycling of plastic packaging waste in Germany

    OpenAIRE

    Gandenberger, Carsten; Orzanna, Robert; Klingenfuß, Sara; Sartorius, Christian

    2014-01-01

    Due to the environmental challenges associated with the strong growth of plastic waste worldwide, the EU Commission recently published a green paper on a European Strategy on Plastic Waste in the Environment (COM (2013), 123 final), which highlights the challenges and opportunities that arise from improving the management of plastic waste in the EU. The European Waste Directive (2008/98/EC) which was transposed into German law through the Kreislaufwirtschaftsgesetz (KrWG) established the so-c...

  16. Climate accounting for waste management, Phase I and II. Summary: Phase 1: Glass Packaging, Metal packaging, paper, cardboard, plastic and wet organic waste. Phase 2: Wood waste and residual waste from households; Klimaregnskap for avfallshaandtering, Fase I og II. Sammendrag: Fase 1: Glassemballasje, metallemballasje, papir, papp, plastemballasje og vaatorganisk avfall. Fase 2: Treavfall og restavfall fra husholdninger

    Energy Technology Data Exchange (ETDEWEB)

    Raadal, Hanne Lerche; Modahl, Ingunn Saur; Lyng, Kari-Anne

    2009-09-15

    involves the lowest greenhouse gas load for the types of waste glass packaging, metal packaging and plastic packaging. Biological treatment (biogas production) provides the lowest GHG (greenhouse gas) impact for the treatment of wet organic waste. Energy recovery provides the lowest GHG impact for the treatment of paper, cardboard and wood waste. Disposal provides the greatest greenhouse gas load for all the analyzed types of waste, but plastic and glass containers. For waste composition has a major impact on greenhouse gas emissions for the landfill and the energy efficiency of the waste. The composition varies both with the types of waste disposed and with what kind of source separation schemes offered in the various municipalities. This in turn can vary depending on population density (urban areas / cities versus scattered buildings), and motivation of the individual citizen to source sorting. Energy recovery means the lowest greenhouse gas emissions for an 'average composite' residual waste in Norway. Analysis of residual waste should always be considered in context with the total amounts and handling of sorted out waste types, as well as total amounts and composition of residual waste. This is important to achieve a comprehensive assessment and avoid suboptimalization. Transport related greenhouse gas emissions are generally of relatively little importance in relation to the environmental benefits arising from the material and / or energy utilization. 3. The model is used to calculate the net greenhouse gas emissions resulting from disposal of a total of approximately 4.1 million tons of waste from households, industry, construction and service industries. 4. Analysis of a realistic optimal scenario for disposal of household waste show that this system can be virtually carbon-neutral. 5. The choice of which assumptions to be incorporated in this type of analysis depends on the purpose of analysis, in addition to local and geographical conditions. 6. Relevant

  17. The paradox of packaging optimization – a characterization of packaging source reduction in the Netherlands

    NARCIS (Netherlands)

    van Sluisveld, M.A.E.; Worrell, E.

    2013-01-01

    The European Council Directive 94/62/EC for Packaging and Packaging Waste requires that Member States implement packaging waste prevention measures. However, consumption and subsequently packaging waste figures are still growing annually. It suggests that policies to accomplish packaging waste preve

  18. Techno-economic assessment of central sorting at material recovery facilities - the case of lightweight packaging waste

    DEFF Research Database (Denmark)

    Cimpan, Ciprian; Maul, Anja; Wenzel, Henrik;

    2016-01-01

    by documenting typical steps taken in a techno-economic assessment of MRFs, using the specific example of lightweight packaging waste (LWP) sorting in Germany. Thus, the study followed the steps of dimensioning of buildings and equipment, calculation of processing costs and projections of revenues from material...... 7 to 21 million EUR and the yearly operational expenditure grew by a factor of 2.4 from 2 to 4.7 million EUR. As a result, specific unit processing cost decreased from 110 to 70 EUR/tonne. Material sales and disposal costs summed to between a net cost of 25 EUR/tonne and net revenue of 50 EUR....../tonne. Measured as total materials recovery, the difference between optimal and typical operation was approximately 15% points. The complex nature of LWP waste combined with challenging processing conditions were identified as important factors explaining the relatively low overall recovery efficiencies achieved...

  19. Skid resistance of bituminous surfaces in Ohio

    Science.gov (United States)

    Colony, D. C.

    1984-03-01

    Statistical studies of skid number (SN) on the state highway system in Ohio reveal apparently systematic variations in distributions of SN among the twelve districts in the state. These variations appear to be significantly related to traffic and to geological characteristics of a given district. Regression calculations using traffic index, a function of ADT and age of the pavement surface in months, show a high correlation with SN, although standard errors of estimate were too large to make the equations useful for predicting SN from traffic variables. Similar results were obtained by elating SN to physiographic features in the state. Pavement surface type (bituminous or pcc) and coarse aggregate type (limestone, gravel or slag) are significantly related to SN.

  20. Radon concentration measurements in bituminous coal mines

    International Nuclear Information System (INIS)

    Radon measurements were carried out in Kozlu, Karadon and Uezuelmez underground coal mines of Zonguldak bituminous coal basin in Turkey. Passive-time integrating method, which is the most widely used technique for the measurement of radon concentration in air, was applied by using nuclear etched track detectors (CR-39) in the study area. The radon concentration measurements were performed on a total of 42 points in those three mines. The annual exposure, the annual effective dose and lifetime fatality risk, which are the important parameters for the health of workers, were estimated based on chronic occupational exposure to the radon gas, which is calculated using UNCEAR-2000 and ICRP-65 models. The radon concentrations at several coal production faces are higher than the action level of 1000 Bq m-3. It is suggested that the ventilation rates should be rearranged to reduce the radon concentration. (authors)

  1. Radon concentration measurements in bituminous coal mines.

    Science.gov (United States)

    Fisne, Abdullah; Okten, Gündüz; Celebi, Nilgün

    2005-01-01

    Radon measurements were carried out in Kozlu, Karadon and Uzülmez underground coal mines of Zonguldak bituminous coal basin in Turkey. Passive-time integrating method, which is the most widely used technique for the measurement of radon concentration in air, was applied by using nuclear etched track detectors (CR-39) in the study area. The radon concentration measurements were performed on a total of 42 points in those three mines. The annual exposure, the annual effective dose and lifetime fatality risk, which are the important parameters for the health of workers, were estimated based on chronic occupational exposure to the radon gas, which is calculated using UNCEAR-2000 and ICRP-65 models. The radon concentrations at several coal production faces are higher than the action level of 1000 Bq m(-3). It is suggested that the ventilation rates should be rearranged to reduce the radon concentration.

  2. Data Package for Secondary Waste Form Down-Selection—Cast Stone

    Energy Technology Data Exchange (ETDEWEB)

    Serne, R. Jeffrey; Westsik, Joseph H.

    2011-09-05

    Available literature on Cast Stone and Saltstone was reviewed with an emphasis on determining how Cast Stone and related grout waste forms performed in relationship to various criteria that will be used to decide whether a specific type of waste form meets acceptance criteria for disposal in the Integrated Disposal Facility (IDF) at Hanford. After the critical review of the Cast Stone/Saltstone literature, we conclude that Cast Stone is a good candidate waste form for further consideration. Cast stone meets the target IDF acceptance criteria for compressive strength, no free liquids, TCLP leachate are below the UTS permissible concentrations and leach rates for Na and Tc-99 are suiteably low. The cost of starting ingredients and equipment necessary to generate Cast Stone waste forms with secondary waste streams are low and the Cast Stone dry blend formulation can be tailored to accommodate variations in liquid waste stream compositions. The database for Cast Stone short-term performance is quite extensive compared to the other three candidate waste solidification processes. The solidification of liquid wastes in Cast Stone is a mature process in comparison to the other three candidates. Successful production of Cast Stone or Saltstone has been demonstrated from lab-scale monoliths with volumes of cm3 through m3 sized blocks to 210-liter sized drums all the way to the large pours into vaults at Savannah River. To date over 9 million gallons of low activity liquid waste has been solidified and disposed in concrete vaults at Savannah River.

  3. Study on safety evaluation of monolithic cement packages of radioactive wastes under deep-sea conditions

    International Nuclear Information System (INIS)

    For sea disposal of the low-level radioactive wastes, the safety of monolithic cement solidified products to be the main solidified waste for dumping was evaluated. Safety evaluation covers the results of integrity test under deep-sea conditions, development of nondestructive inspection and leaching test of nuclides of the above solidified waste. It is concluded that previous evaluation for the sea disposal of radioactive wastes should be more conservative than the real situation, because the cement solidified products have appreciable retardation effect for dispersion of radionuclides and thus the effect of containment is enhanced. (author)

  4. Waste package degradation from thermal and chemical processes in performance assessments for the Yucca Mountain disposal system for spent nuclear fuel and high-level radioactive waste

    International Nuclear Information System (INIS)

    This paper summarizes modeling of waste container degradation in performance assessments conducted between 1984 and 2008 to evaluate feasibility, viability, and assess compliance of a repository for spent nuclear fuel and high-level radioactive waste at Yucca Mountain, Nevada. As understanding of the Yucca Mountain disposal system increased, modeling of container degradation evolved from a component of the source term in 1984 to a separate module describing both container and drip shield degradation in 2008. A thermal module for evaluating the influence of higher heat loads from more closely packed, large waste packages was also introduced. In addition, a module for evaluating drift chemistry was added in later PAs to evaluate the potential for localized corrosion of the outer barrier of the waste container composed of Alloy 22, a highly corrosion-resistant nickel–chromium–tungsten–molybdenum alloy. The uncertainty of parameters related to container degradation contributed significantly to the estimated uncertainty of performance measures (cumulative release in assessments prior to 1995 and individual dose, thereafter). - Highlights: • Progression of container modeling from an assumption of complete failure before 1993 to a stochastic description after 1995 is described. • Introduction of thermal-hydrologic model to examine interaction between the engineered and natural barrier, and container degradation is described. • The introduction of a water chemistry module to evaluate the possibility of localized corrosion of the container is described. • The addition of seismic damage models to evaluate container damage from drift degradation and container movement is described

  5. Study on Plastic Coated Overburnt Brick Aggregate as an Alternative Material for Bituminous Road Construction

    Directory of Open Access Journals (Sweden)

    Dipankar Sarkar

    2016-01-01

    Full Text Available There are different places in India where natural stone aggregates are not available for constructional work. Plastic coated OBBA can solve the problem of shortage of stone aggregate to some extent. The engineers are always encouraged to use locally available materials. The present investigation is carried out to evaluate the plastic coated OBBA as an alternative material for bituminous road construction. Shredded waste plastics are mixed with OBBA in different percentages as 0.38, 0.42, 0.46, 0.50, 0.54, and 0.60 of the weight of brick aggregates. Marshall Method of mix design is carried out to find the optimum bitumen content of such bituminous concrete mix prepared by plastic coated OBBA. Bulk density, Marshall Stability, flow, Marshall Quotient, ITS, TSR, stripping, fatigue life, and deformations have been determined accordingly. Marshall Stability value of 0.54 percent of plastic mix is comparatively higher than the other mixes except 0.60 percent of plastic mix. Test results are within the prescribed limit for 0.54 percent of plastic mix. There is a significant reduction in rutting characteristics of the same plastic mix. The fatigue life of the mix is also significantly higher. Thus plastic coated OBBA is found suitable in construction of bituminous concrete road.

  6. Some aspects regarding the qualifications tests of packages used for transport and storage of radioactive waste (low activity) in INR Pitesti

    International Nuclear Information System (INIS)

    Radioactive wastes generated by TRIGA INR research reactor are packaged according to the national and international standards and to the IAEA Regulations for the Safe Transport of Radioactive Materials and Advisory Material for the Application of the IAEA Transport Regulations. The technology for packaging and treatment of Radioactive wastes used in our institute can be applied, in perspective, also at the Nuclear Power Plant Cernavoda, after commissioning. This paper describes the qualifications tests (type tests) for packages used for transport and storage (for a long period of about 30 years) of radioactive wastes (low activity, up to 0.5068x1010 Bq/drum, 0.164 Ci/drum, respectively). The package used is a drum manufactured by Romanian industry (according to the national standard 7683-79) of 1 mm thick mild steel with the following dimensions: height: 915 ± 10 mm, diameter: 600 ± 5 mm, volume: 220 liters, approximately. There are presented the type tests carried-out, e.g. compression, penetration, free fall, lixiviation, safety in utilizing (biological protection), checking of chemical and mechanical characterization and the effect of the product on the environment, the results, interpretation and conclusions. The performing of the above mentioned tests and other additional ones, the results obtained, prove that our technology for treatment and packaging of radioactive wastes is in accordance with IAEA Regulations in the field. (author). 6 refs, 6 figs

  7. Vendor Assessment for the Waste Package Closure System (Yucca Mountain Project)

    International Nuclear Information System (INIS)

    The Idaho National Engineering and Environmental Laboratory (INEEL) has been tasked with developing, designing, constructing, and operating a full-scale prototype of the work package closure system. As a precursor to developing the conceptual design, all commercially available equipment was assessed to identify any existing technology gaps. This report presents the results of that assessment for all major equipment

  8. Geology Data Package for the Single-Shell Tank Waste Management Areas at the Hanford Site

    Energy Technology Data Exchange (ETDEWEB)

    Reidel, Stephen P.; Chamness, Mickie A.

    2007-12-14

    This data package discusses the geology of the single-shell tank (SST) farms and the geologic history of the area. The purpose of this report is to provide the most recent geologic information available for the SST farms. This report builds upon previous reports on the tank farm geology and Integrated Disposal Facility geology with information available after those reports were published.

  9. Geology Data Package for the Single-Shell Tank Waste Management Areas at the Hanford Site

    Energy Technology Data Exchange (ETDEWEB)

    Reidel, Steve P.; Chamness, Mickie A.

    2007-01-01

    This data package discusses the geology of the single-shell tank (SST) farms and the geologic history of the area. The focus of this report is to provide the most recent geologic information available for the SST farms. This report builds upon previous reports on the tank farm geology and Integrated Disposal Facility geology with information available after those reports were published.

  10. Vendor Assessment for the Waste Package Closure System (Yucca Mtn. Project)

    Energy Technology Data Exchange (ETDEWEB)

    Colleen Shelton-Davis

    2003-09-01

    The Idaho National Engineering and Environmental Laboratory (INEEL) has been tasked with developing, designing, constructing, and operating a full-scale prototype of the work package closure system. As a precursor to developing the conceptual design, all commercially available equipment was assessed to identify any existing technology gaps. This report presents the results of that assessment for all major equipment.

  11. Vendor Assessment for the Waste Package Closure System (Yucca Mountain Project)

    Energy Technology Data Exchange (ETDEWEB)

    Shelton-Davis, C.V.

    2003-09-26

    The Idaho National Engineering and Environmental Laboratory (INEEL) has been tasked with developing, designing, constructing, and operating a full-scale prototype of the work package closure system. As a precursor to developing the conceptual design, all commercially available equipment was assessed to identify any existing technology gaps. This report presents the results of that assessment for all major equipment.

  12. Specific transport and storage solutions: Waste management facing current and future stakes of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    development process; - Current solutions proposed by COGEMA LOGISTICS; - Transport of Low Level and Very Low Level waste (LLW and VLLW); - Shipment of large volumes; - Non-reusable transport packaging; - Packaging transport permit the storage in final disposal area; - A steep increase of transport flows in the next 10 years; - Future challenges, waste from dismantling operations; - Transport packaging of IP 2 type and IP 2 package; - Some examples of standardized packages; - DV78 Package; - CC102 over pack; - A solution for liquid waste from NPPs, TNTMCIEL; - Transport of Intermediate Level Waste (ILW) and High Level Waste (HLW); - Alpha waste (or TRU waste); - TNTMGEMINI; - RD 26 package; - Bituminized waste (ILW): the TNTM833; - Technological waste in CBFC'2 (ILW): the TNTM837; - Universal Canisters for Vitrified waste (HLW) and Compacted waste (ILW); - Vitrified waste (HLW); - Transport solutions for CSD-V canisters: the TNTM28 VT; - Dual purpose solutions for CSD-V canisters, the TS 28; - CSD-V programme of return; - Compacted waste (ILW): the TNTM843; - Some future stakes and new developments. The following conclusions completes the paper. COGEMA LOGISTICS experience in designing packages and transporting nuclear materials is part of the COGEMA / AREVA waste management policy, which aims to waste volume reduction through waste sorting and packaging in universal canisters or standard waste containers. For the design and manufacture of special shipping casks, as for nuclear transportation and storage, COGEMA LOGISTICS tailors its solutions to its international customers' requirements. Personnel safety, transportation safety and environmental protection are put first. These reasons have lead COGEMA LOGISTICS to become the world's leading designer of nuclear fuel packages, with unique expertise in designing packages for back-end transportation and storage of nuclear materials. Beyond being a transport company and packaging designer, COGEMA LOGISTICS is a partner implementing

  13. 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

  14. The radiation characteristics of the transport packages with vitrified high-level waste

    Science.gov (United States)

    Bogatov, S. A.; Mitenkova, E. F.; Novikov, N. V.

    2015-12-01

    The calculation method of neutron yield in the (α, n) reaction for a homogeneous material of arbitrary composition is represented. It is shown that the use of the ORIGEN 2 code excluding the real elemental composition of vitrified high-level waste leads to significant underestimation of the neutron yield in the (α, n) reaction. For vitrified high-level waste and spent nuclear fuel from VVER, the neutron fluxes are analyzed. The thickness of the protective materials for a transfer cask and a shipping cask with vitrified highlevel waste are estimated.

  15. The radiation characteristics of the transport packages with vitrified high-level waste

    Energy Technology Data Exchange (ETDEWEB)

    Bogatov, S. A. [JSC VNIPIpromtechnologii (Russian Federation); Mitenkova, E. F., E-mail: mit@ibrae.ac.ru; Novikov, N. V. [Russian Academy of Sciences, Nuclear Safety Institute (Russian Federation)

    2015-12-15

    The calculation method of neutron yield in the (α, n) reaction for a homogeneous material of arbitrary composition is represented. It is shown that the use of the ORIGEN 2 code excluding the real elemental composition of vitrified high-level waste leads to significant underestimation of the neutron yield in the (α, n) reaction. For vitrified high-level waste and spent nuclear fuel from VVER, the neutron fluxes are analyzed. The thickness of the protective materials for a transfer cask and a shipping cask with vitrified highlevel waste are estimated.

  16. Facilities for processing, packaging and storage of intermediate level radioactive waste at the Winfrith Technology Center

    International Nuclear Information System (INIS)

    The UK strategy for the management of radioactive waste is the responsibility of the Secretary of State for the Environment. The implementation of that strategy is a function of Her Majesty's Inspectorate of Pollution (HMIP), a department within the Department of the Environment. Close liaison between HMIP and Her Majesty's Nuclear Installation Inspectorate (HMNII) takes place ensuring that site licenses are only granted for proposals which are consistent with the national waste management strategy. The task of developing national low level solid waste (LLSW) and intermediate level waste (ILW) disposal facilities rests with the UK Nirex, Ltd., a company formed by major national nuclear utilities with independent and government representation. This company is currently investigating the suitability of Sellafield, Cumbria and Dounreay, Caithness for a national repository. Operation of such a facility is unlikely however before the year 2005. 6 refs., 2 figs., 2 tabs

  17. Deep geological repositories in the UK. Current underground planning and competency requirements for waste packaging systems

    International Nuclear Information System (INIS)

    Nuclear fission has been used for commercial electricity generation in the UK since 1956. This early (and world first) entry into the nuclear energy industry has created a huge challenge in the form of waste disposal and decommissioning, as many of the original facilities have left behind types of waste that by today's standards are both abnormal and also poorly documented. The UK has already agreed on the concept of using deep geological format ions for the permanent storage of this waste, and the material yet to be produced. However, the choice of site is still undecided. This paper outlines the main issues at stake and discusses the previous approach and future proposals for the construction of a permanent waste repository in the United Kingdom.

  18. Packaging for logistical support

    Science.gov (United States)

    Twede, Diana; Hughes, Harold

    Logistical packaging is conducted to furnish protection, utility, and communication for elements of a logistical system. Once the functional requirements of space logistical support packaging have been identified, decision-makers have a reasonable basis on which to compare package alternatives. Flexible packages may be found, for example, to provide adequate protection and superior utility to that of rigid packages requiring greater storage and postuse waste volumes.

  19. Tank vapor sampling and analysis data package for tank 241-C-106 waste retrieval sluicing system process test phase III

    Energy Technology Data Exchange (ETDEWEB)

    LOCKREM, L.L.

    1999-08-13

    This data package presents sampling data and analytical results from the March 28, 1999, vapor sampling of Hanford Site single-shell tank 241-C-106 during active sluicing. Samples were obtained from the 296-C-006 ventilation system stack and ambient air at several locations. Characterization Project Operations (CPO) was responsible for the collection of all SUMMATM canister samples. The Special Analytical Support (SAS) vapor team was responsible for the collection of all triple sorbent trap (TST), sorbent tube train (STT), polyurethane foam (PUF), and particulate filter samples collected at the 296-C-006 stack. The SAS vapor team used the non-electrical vapor sampling (NEVS) system to collect samples of the air, gases, and vapors from the 296-C-006 stack. The SAS vapor team collected and analyzed these samples for Lockheed Martin Hanford Corporation (LMHC) and Tank Waste Remediation System (TWRS) in accordance with the sampling and analytical requirements specified in the Waste Retrieval Sluicing System Vapor Sampling and Analysis Plan (SAP) for Evaluation of Organic Emissions, Process Test Phase III, HNF-4212, Rev. 0-A, (LMHC, 1999). All samples were stored in a secured Radioactive Materials Area (RMA) until the samples were radiologically released and received by SAS for analysis. The Waste Sampling and Characterization Facility (WSCF) performed the radiological analyses. The samples were received on April 5, 1999.

  20. Food waste: The role of date labels, package size, and product category

    OpenAIRE

    Wilson, Norbert L.W.; Rickard, Bradley J.; Saputo, Rachel; Ho, Shuay-Tsyr

    2015-01-01

    The presence of food waste, and ways to reduce food waste, has generated significant debate among industry stakeholders, policy makers, and consumer groups in the United States and elsewhere. Many have argued that the variety of date labels used by food manufacturers leads to confusion about food quality and food safety among consumers. Here we develop a laboratory experiment with treatments that expose subjects to different date labels (Sell by, Best by, Use by, and Fresh by) for six food pr...

  1. Development of non-standard equipment of bituminization line

    International Nuclear Information System (INIS)

    A brief description is presented of some components of the bituminization line developed at the Research Institute of Chemical Installations in Brno. The film rotor evaporator has an evaporation area of 2 m2. The mixing tank is the main technological assembly for processing loose materials, i.e., sorbents and fly ash which it mixes with melted bitumen. The screw conveyor with a three-way valve is designed for batching the dried radioactive sorbent and fly ash into mixing tanks. Attention is also devoted to certain systems of measurement and control of the bituminization line, namely to measurement of the level of the bituminization product, the flow rate of bituminization emulsion and the water-sorbent mixture and to the diagnostics of failure and accident conditions of the film rotor evaporator. (Z.M.)

  2. EXPERIMENTAL STUDY ON HORIZONTAL COMBUSTION TECHNIQUE FOR BITUMINOUS COAL BRIQUET

    Institute of Scientific and Technical Information of China (English)

    路春美; 程世庆; 邵延玲; 张晔

    1997-01-01

    Through a lot of experiments, a new kind of stove using horizontal combustion technique for bituminous coal briquet has been developed. Making use of this stove, studies have been made on burning process of bituminous coal briquet, distribution of temperature field in the stove, the regularities of evolution and combustion of volatile matter, the burning rate and efficiency of bituminous coal briquet, characteristics of fire-sealing and sulfur-retention. The results show that, with the technique, some achievements can be obtained in combustion of bituminous coal briquet, such as lower pollution that the flue gas black degree is below 0.5R and dust concentration is below 90mg/m3 . The stove's combustion efficiency reaches 90%, sulfur fixing efficiency is 60%, and CO concentration is decreased by 40% compared with other traditional stoves. With so many advantages, the stove can be used extensively in civil stoves and smaller industrial boilers.

  3. Effects of bituminous layer as backfill material on mechanical behavior in tunnel model

    OpenAIRE

    Moriyoshi, Akihiro; Takano, Shin-ei; Urata, Hiroyuki; Suzuki, Tetsuya; Yoshida, Takaki

    2001-01-01

    This paper describes the effects of bituminous material as a backfill material on mechanical behavior in model tunnel in laboratory. It is known that load spreading and relaxation of bituminous material are good properties. Then if we use bituminous material as a backfill material of tunnel, the tunnel will have waterproof, good load spreading property. We used new bituminous material (Aquaphalt) which can solidify in water. We conducted relaxation test in tension for new bituminous mat...

  4. Dual Use Packaging Project

    Data.gov (United States)

    National Aeronautics and Space Administration — NASA calculation that over a kg of packaging waste are generated per day for a 6 member crew. This represents over 1.5 metric tons of waste during a Mars mission....

  5. The bituminous sands : a Canadian mirage?

    Energy Technology Data Exchange (ETDEWEB)

    Rousse, D.R. [Quebec Univ., Chicoutimi, PQ (Canada). Dept. des Sciences Appliquees; Nasr, G.J. [Lebanese Univ., Roumieh (Lebanon). Faculty of Engineering; Turcotte, S.F. [Quebec Univ., Montreal, PQ (Canada). Centre d' Etudes Internationales et Mondialisation; Salah, N.B. [Ecole Superieure des Sciences et Techniques De Tunis, Tunis (Tunisia). LMMP

    2009-07-01

    This paper examined the controversy about the potential role of a significant increase in Canadian oil sands production in order to bridge the upcoming gap between the world's increasing energy demand and the total recoverable oil supply. The paper presented the actual potential of different scenarios and considered the prediction cost forecasts. A brief overview of environmental impacts and the real return on investments were also provided. Environmental impacts that were considered included land degradation; water contamination; ecosystem damage; and air pollution. Nuclear energy was also presented as a possible solution. The paper demonstrated that even in a very optimistic scenario, Canada's oil sands accelerated production has a negligible effect on the aforementioned gap, has a considerable impact on environment that has yet to be accounted for. Energy ratios that were presented included energy return on energy investment; energy available on energy used; and energy payback. It was concluded that enhanced recovery techniques are clearly needed for future sustainable exploitation of these bituminous sands. 32 refs., 1 fig.

  6. Conditioning CANDU reactor wastes for disposal

    International Nuclear Information System (INIS)

    A Waste Treatment Centre (WTC) is being constructed at the Chalk River Nuclear Laboratories to develop and demonstrate processes for converting reactor wastes to a form suitable for disposal. The WTC contains a starved air incinerator for reducing the volume of combustible solid wastes, a reverse osmosis section for reducing the volume of liquid wastes and an immobilization section for incorporating the conditioned wastes in bitumen. The incinerator is commissioned on inactive waste: approximately 16.5 Mg of waste packaged in polyethylene bags has been incinerated in 17 burns. Average weight and volume reductions of 8.4:1 and 32:1, respectively, have been achieved. Construction of the reverse osmosis section of WTC is complete and inactive commissioning will begin in 1982 January. The reverse osmosis section was designed to process 30,000 m3/a of dilute radioactive waste. The incinerator ash and concentrated aqueous waste will be immobiblized in bitumen using a horizontal mixer and wiped-film evaporator. Results obtained during inactive commissioning of the incinerator are described along with recent results of laboratory programs directed at demonstrating the reverse osmosis and bituminization processes

  7. Corrosion testing of selected packaging materials for disposal of high-level waste glass in rock salt formations

    International Nuclear Information System (INIS)

    In previous corrosion studies performed in salt brines, unalloyed steels, Ti 99.8-Pd and Hastelloy C4 have proved to be the most promising materials for long-term resistant packagings to be used in heat-generating waste (vitrified HLW, spent fuel) disposal in rock-salt formations. To characterise the corrosion behaviour of these materials in more detail, further in-depth laboratory-scale and in-situ corrosion studies have been performed in the present study. Besides the above-mentioned materials, also some in-situ investigations of the iron-base materials Ni-Resist D2 and D4, cast iron and Si-cast iron have been carried out in order to complete the results available to date. (orig.)

  8. Viability Assessment of a Repository at Yucca Mountain. Volume 2: Preliminary Design Concept for the Repository and Waste Package

    Energy Technology Data Exchange (ETDEWEB)

    None

    1998-12-01

    This volume describes the major design features of the Monitored Geologic Repository. This document is not intended to provide an exhaustive, detailed description of the repository design. Rather, this document summarizes the major systems and primary elements of the design that are radiologically significant, and references the specific technical documents and design analyses wherein the details can be found. Not all portions of the design are at the same level of completeness. Highest priority has been given to assigning resources to advance the design of the Monitored Geologic Repository features that are important to radiological safety and/or waste isolation and for which there is no NRC licensing precedent. Those features that are important to radiological safety and/or waste isolation, but for which there is an NRC precedent, receive second priority. Systems and features that have no impact on radiological safety or waste isolation receive the lowest priority. This prioritization process, referred to as binning, is discussed in more detail in Section 2.3. Not every subject discussed in this volume is given equal treatment with regard to the level of detail provided. For example, less detail is provided for the surface facility design than for the subsurface and waste package designs. This different level of detail is intentional. Greater detail is provided for those functions, structures, systems, and components that play key roles with regard to protecting radiological health and safety and that are not common to existing nuclear facilities already licensed by NRC. A number of radiological subjects are not addressed in the VA, (e.g., environmental qualification of equipment). Environmental qualification of equipment and other radiological safety considerations will be addressed in the LA. Non-radiological safety considerations such as silica dust control and other occupational safety considerations are considered equally important but are not addressed in

  9. Single Shell Tank Waste Characterization Project for Tank B-110, Core 9 - data package and PNL validation summary report

    International Nuclear Information System (INIS)

    This Data Package contains results obtained by Pacific Northwest Laboratory (PNL) staff in the characterization and analyses of Core 9 segments taken from the Single-Shell Tank (SST) 110B. The characterization and analysis of Core 9 segments are outlined in the Waste Characterization Plan for Hanford Site Single-Shell Tanks and in the Pacific Northwest Laboratory (PNL) Single-Shell Tank Waste Characterization Support FY 89/90 Statement of Work (SOW), Rev. 1 dated March, 1990. Specific analyses for each sub-sample taken from a segment are delineated in Test Instructions prepared by the PNL Single-Shell Tank Waste Characterization Project Management Office (SST Project) in accordance with procedures contained in the SST Waste Characterization Procedure Compendium (PNL-MA-599). Analytical procedures used in the characterization activities are also included in PNL-MA-599. Core 9 included five segments although segment 1 did not have sufficient material for characterization. The five samplers were received from Westinghouse Hanford Company (WHC) on 11/21-22/89. Each segment was contained in a sampler and was enclosed in a shipping cask. The shipping cask was butted up to the 325-A hot cell and the sampler moved into the hot cell. The material in the sampler (i.e., the segment) was extruded from the sampler, limited physical characteristics assessed, and photographed. At this point samples were taken for particle size and volatile organic analyses. Each segment was then homogenized. Sub-samples were taken for required analyses as delineated in the appropriate Test Instruction. Table 1 includes sample numbers assigned to Core 9 segment materials being transferred from 325-A Hot Cell. Sample numbers 90-0298, 90-0299, 90-0302, and 90-0303 were included in Table 1 although no analyses were requested for these samples. Table 2 lists Core 9 sub-sample numbers per sample preparation method

  10. Experimental study on the detection of free fluids and gases in waste packages by acoustic methods

    International Nuclear Information System (INIS)

    The objective of the project was to evaluate the potential and the limits of various nondestructive methods for testing the contents of 200-litre drums filled with radioactive waste. The following test problems were to be studied: 1. Detection of free water on the surface of the waste matrix (concrete); 2. Determination of the waste matrix level; 3. Determination of internal gas pressure. The following methods were found to be suitable: For Test problem 1: Measurement of Lamb wave attenuation, Acoustic impedance measurement (AIM) and Analysis of swash sound; For Test problem 2: Acoustic impedance measurement (AIM) and Measurement of Lamb wave attenuation; For Test problem 3: A method of pressure compensation and Analysis of cover resonances after striking the cover. It was not possible, however, to detect the concrete level by localisation of friction points using acoustic emission methods. 53 figs

  11. THE PROCESS OF WASTE MANAGEMENT IN POST-CONSUMER PACKAGING: CASE STUDY MCDONALD'S

    Directory of Open Access Journals (Sweden)

    Robson dos Santos

    2013-09-01

    Full Text Available This research considers the increasing concern of society in general environmental issues, shows the importance of an Environmental Management System to improve the image of a company towards society in which it is embedded. Shows that proper waste management can result in financial and environmental benefits for companies that practice. To address the practical issues of the theme, was chosen the company McDonald's, as a service company fast food, that have a quantity of waste, and creates conditions for application of the techniques of environmental management in this sector. Thus, this article aims to demonstrate through case study and descriptive research, the commitment that this large network of fast-food has with the preservation of the environment through its waste management and investments in economic, social and environmental the country.

  12. Experience in treatment of components, waste treatment, packaging and shipping related to decommissioning

    International Nuclear Information System (INIS)

    Common practice of the treatment of large amounts of activated core components, contaminated components and waste treatment is described. Single tasks of actual repair and backfitting work comprise up to 800 tons of material to be removed, packed, shipped and decontaminated. Decontamination for unrestricted release has already been successfully performed. Special container systems for shipment and storage of every kind of radioactive waste are presented. Containers resulting in maximum shipping, storage and disposal capacity as well as shipping and storage casks for canned and uncanned fuel are described. The experience gained with nuclear facilities in Europe is correlated to the decommissioning tasks to be performed in the near future. 7 figures

  13. Development of the NIREX generic transport safety assessment to assist in the provision of waste packaging advice

    International Nuclear Information System (INIS)

    and a comparison of the risks with the Nirex dose and risk targets. In addition, this paper also describes how the methodology of the GTSA has been developed into an innovative software tool, TranSAT, which is routinely used as part of the packaging waste advice service offered by Nirex. (author)

  14. Remaining Sites Verification Package for the 100-D-2 Lead Sheeting Waste Site, Waste Site Reclassification Form 2007-030

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2008-03-19

    The 100-D-2 Lead Sheeting waste site was located approximately 50 m southwest of the 185-D Building and approximately 16 m north of the east/west oriented road. The site consisted of a lead sheet covering a concrete pad. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  15. Degradation mode survey candidate titanium-base alloys for Yucca Mountain project waste package materials. Revision 1

    International Nuclear Information System (INIS)

    The Yucca Mountain Site Characterization Project (YMP) is evaluating materials from which to fabricate high-level nuclear waste containers (hereafter called waste packages) for the potential repository at Yucca Mountain, Nevada. Because of their very good corrosion resistance in aqueous environments titanium alloys are considered for container materials. Consideration of titanium alloys is understandable since about one-third (in 1978) of all titanium produced is used in applications where corrosion resistance is of primary importance. Consequently, there is a considerable amount of data which demonstrates that titanium alloys, in general, but particularly the commercial purity and dilute α grades, are highly corrosion resistant. This report will discuss the corrosion characteristics of Ti Gr 2, 7, 12, and 16. The more highly alloyed titanium alloys which were developed by adding a small Pd content to higher strength Ti alloys in order to give them better corrosion resistance will not be considered in this report. These alloys are all two phase (α and β) alloys. The palladium addition while making these alloys more corrosion resistant does not give them the corrosion resistance of the single phase α and near-α (Ti Gr 12) alloys

  16. Degradation mode survey candidate titanium-base alloys for Yucca Mountain project waste package materials. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Gdowski, G.E.

    1997-12-01

    The Yucca Mountain Site Characterization Project (YMP) is evaluating materials from which to fabricate high-level nuclear waste containers (hereafter called waste packages) for the potential repository at Yucca Mountain, Nevada. Because of their very good corrosion resistance in aqueous environments titanium alloys are considered for container materials. Consideration of titanium alloys is understandable since about one-third (in 1978) of all titanium produced is used in applications where corrosion resistance is of primary importance. Consequently, there is a considerable amount of data which demonstrates that titanium alloys, in general, but particularly the commercial purity and dilute {alpha} grades, are highly corrosion resistant. This report will discuss the corrosion characteristics of Ti Gr 2, 7, 12, and 16. The more highly alloyed titanium alloys which were developed by adding a small Pd content to higher strength Ti alloys in order to give them better corrosion resistance will not be considered in this report. These alloys are all two phase ({alpha} and {beta}) alloys. The palladium addition while making these alloys more corrosion resistant does not give them the corrosion resistance of the single phase {alpha} and near-{alpha} (Ti Gr 12) alloys.

  17. Substitution potentials of recycled HDPE and wood particles from post-consumer packaging waste in Wood-Plastic Composites.

    Science.gov (United States)

    Sommerhuber, Philipp F; Welling, Johannes; Krause, Andreas

    2015-12-01

    The market share of Wood-Plastic Composites (WPC) is small but expected to grow sharply in Europe. This raises some concerns about suitable wood particles needed in the wood-based panels industry in Europe. Concerns are stimulated by the competition between the promotion of wooden products through the European Bioeconomy Strategy and wood as an energy carrier through the Renewable Energy Directive. Cascade use of resources and valorisation of waste are potential strategies to overcome resource scarcity. Under experimental design conditions, WPC made from post-consumer recycled wood and plastic (HDPE) were compared to WPC made from virgin resources. Wood content in the polymer matrix was raised in two steps from 0% to 30% and 60%. Mechanical and physical properties and colour differences were characterized. The feasibility of using cascaded resources for WPC is discussed. Results indicate the technical and economic feasibility of using recycled HDPE from packaging waste for WPC. Based on technical properties, 30% recycled wood content for WPC is feasible, but economic and political barriers of efficient cascading of biomass need to be overcome. PMID:26376122

  18. The FAME drum measuring system as a tool for data capture for waste package documentation; Die Fassmessanlage FAME als Instrument der Datenerfassung fuer die Abfallgebindedokumentation

    Energy Technology Data Exchange (ETDEWEB)

    Risch, Ralph; Doppler, Thomas [Wissenschaftlich-Technische Ingenieurberatung GmbH (WTI), Juelich (Germany)

    2010-08-15

    Incipient commissioning of the Konrad repository requires large volumes of radioactive waste to be prepared for final storage and to be declared. Performing these duties within an acceptable period of time needs manpower resources and optimized technical procedures such that standardized data capture allows speedy qualification and documentation of the waste. The FAME drum measuring system ensures complete capture of data relevant to documentation, the performance of quality-assured dose rate measurements and gamma spectrometric measurements, respectively, and export of the data by defined export functions. Capturing documentation-related data about the drums also allows the use of standardized computation methods approved by the authorities. This clearly reduces the expense involved in compiling waste package documentations and their examination by an expert of the Federal Office of Radiation Protection. After a positive outcome of the examination by the expert, the drums so documented can be put into licensed repository packages and prepared for storage in the Konrad repository. (orig.)

  19. Remaining Sites Verification Package for the 120-F-1 Glass Dump Waste Site, Waste Site Reclassification Form 2008-028

    Energy Technology Data Exchange (ETDEWEB)

    J. M. Capron

    2008-06-27

    The 120-F-1 waste site consisted of two dumping areas located 660 m southeast of the 105-F Reactor containing laboratory equipment and bottles, demolition debris, light bulbs and tubes, small batteries, small drums, and pesticide contaminated soil. It is probable that 108-F was the source of the debris but the material may have come from other locations within the 100-F Area. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  20. Exploring sustainable market opportunities for Local Generation's mixed plastic packaging waste

    OpenAIRE

    Jayamani, Saravana; Goel, Aastha; Radhakrishnan, Manish Krishnan

    2011-01-01

    European Union’s 2020 target by 2020 is driving all the member states towards low carbon green economy and UK is in forefront of the change. UK government is using the transition to green economy to tackle economic instability and deficit. To ease the transition, various government departments are working with businesses and supporting them through many grants and funds. Anaerobic digestion (AD) is one such green business that generates energy from food waste that would have been otherwise se...

  1. The bituminization in Marcoule and la Hague plants. Description of the processes and the fire risk prevention

    International Nuclear Information System (INIS)

    Chemical precipitation is an efficient process to decontaminate liquid effluents and has been selected as early as 1966 for the liquid waste treatment of MARCOULE. The process has been designed by the French CEA and improved along the time. The first implementation in LA HAGUE was for the STE 2 facility (Station de Traitement d'Effluents no 2). A new facility, STE 3, has been commissioned in 1989 in LA HAGUE, benefiting from these previous experience and therefore contributing into a drastic reduction of liquid discharges. The sludges coming from the treatment in MARCOULE and STE 3 are bituminized into continuous extruders. The resulting waste is kept in interim storage, waiting for final disposal. The industrial production, up to now, is close to 10000 drums (55 gallons) in LA HAGUE and 60000 drums in MARCOULE. A retrieval operation of the MARCOULE drums from the present storage has been launched to send them into a recently built interim storage operating since the beginning of 2000. At the present time, in LA HAGUE, evaporation has been preferred to chemical precipitation for the treatment of liquid effluents. After concentration, the effluents go to the vitrification and the volume of resulting waste is much smaller. However, bituminization is still the preferred option for the 10000 m3 of sludges resulting from the STE 2 operations which are being stored in tanks. A complete reevaluation of the operation safety measures of bituminization has been made lately in both plants following recent incidents on other installations. The paper describes the chemical precipitation and the bituminization processes and focuses on safety measures which are implemented for prevention, detection and fight. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-07-01

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

  3. Bituminization of low- and intermediate-level radioactive concentrates

    International Nuclear Information System (INIS)

    The results and experiences are summed up from the research and development of the technology of bituminization of low- and intermediate-level radioactive concentrates generated in the operation of nuclear power plants with WWER-440 reactors. The experiments took place on a pilot plant bituminization line with various model solutions at an evaporator capacity of 70, 100 and 140 litres per hour. The composition of the solutions changed in dependence on the knowledge of the composition of actual concentrates in the V-1 nuclear power plant at Jaslovske Bohunice. The following factors were studied in the concentrate: the effect of its pH, its borate content and content of metal carbonates, the content of organic acid salts, the content of detergents, etc., on the process of bituminization. Physico-chemical conditions are described under which the operation of the evaporator was fail-safe and filling of the bituminization product with salts homogeneous. A low water content of up to 1% was achieved. The properties of the bituminization product were negatively affected by a high level of heavy metal oxides, surfactants, oxalates and citrates. In order to improve the properties of the product it will be necessary to replace bitumen emulsion Silembit-60 used as reinforcement matrix, by a different type of bitumen. (Z.M.)

  4. Waste specification and quantity structure. Basis for phase-out of nuclear power use (July 2011). Report on work package no. 3. Preliminary safety analysis for the site Gorleben

    International Nuclear Information System (INIS)

    The report on work package 3 of the preliminary safety analysis of the site Gorleben in the frame of nuclear phase-out in Germany characterizing the waste quantity structure covers the following issues: Boundary conditions and procedural method; heat generating radioactive waste from power reactors (direct final disposal of spent fuel elements and radioactive waste from reprocessing in other countries and the reprocessing plant in Karlsruhe); heat generating radioactive waste from research and prototype reactors; waste with negligible heat generation.

  5. Effect of ionizing radiation on the waste package environment; Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Reed, D.T. [Argonne National Lab., IL (USA); Van Konynenburg, R.A. [Lawrence Livermore National Lab., CA (USA)

    1991-05-01

    The radiolytic production of nitrogen oxides, nitrogen acids and ammonia are discussed in relation to the expected environment in a high-level waste repository that may be constructed at the Yucca Mountain site if it is found to be suitable. Both literature data and repository-relevant data are summarized for air-water vapor systems. The limiting cases of a dry air and a pure water vapor gas phase are also discussed. Design guidelines and recommendations, based solely on the potential consequence of radiation enhancement of corrosion, are given. 13 refs., 5 figs., 1 tab.

  6. Radioactive waste package assay facility. Volume 2. Investigation of active neutron and active gamma interrogation

    International Nuclear Information System (INIS)

    Volume 2 of this report describes the theoretical and experimental work carried out at Harwell on active neutron and active gamma interrogation of 500 litre cemented intermediate level waste drums. The design of a suitable neutron generating target in conjunction with a LINAC was established. Following theoretical predictions of likely neutron responses, an experimental assay assembly was built. Responses were measured for simulated drums of ILW, based on CAGR, Magnox and PCM wastes. Good correlations were established between quantities of 235-U, nat-U and D2O contained in the drums, and the neutron signals. Expected sensitivities are -1g of fissile actinide and -100g of total actinide. A measure of spatial distribution is obtainable. The neutron time spectra obtained during neutron interrogation were more complex than expected, and more analysis is needed. Another area of discrepancy is the difference between predicted and measured thermal neutron flux in the drum. Clusters of small 3He proportional counters were found to be much superior for fast neutron detection than larger diameter counters. It is necessary to ensure constancy of electron beam position relative to target(s) and drum, and prudent to measure the target neutron or gamma output as appropriate. 59 refs., 77 figs., 11 tabs

  7. Operation of the radioactive waste treatment facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kil Jeong; Ahn, Seom Jin; Lee, Kang Moo; Lee, Young Hee; Sohn, Jong Sik; Bae, Sang Min; Kang, Kwon Ho; Lim, Kil Sung; Sohn, Young Joon; Kim, Tae Kook; Jeong, Kyung Hwan; Wi, Geum San; Park, Seung Chul; Park, Young Woong; Yoon, Bong Keun

    1996-12-01

    The radioactive wasted generated at Korea Atomic Energy Research Institute (KAERI) in 1996 are about 118m{sup 3} of liquid waste and 204 drums of solid waste. Liquid waste were treated by the evaporation process, the bituminization process, and the solar evaporation process. In 1996, 100.5m{sup 3} of liquid waste was treated. (author). 84 tabs., 103 figs.

  8. Operation of the radioactive waste treatment facility

    International Nuclear Information System (INIS)

    The radioactive wasted generated at Korea Atomic Energy Research Institute (KAERI) in 1996 are about 118m3 of liquid waste and 204 drums of solid waste. Liquid waste were treated by the evaporation process, the bituminization process, and the solar evaporation process. In 1996, 100.5m3 of liquid waste was treated. (author). 84 tabs., 103 figs

  9. The Innovations, Technology and Waste Management Approaches to Safely Package and Transport the World's First Radioactive Fusion Research Reactor for Burial

    Energy Technology Data Exchange (ETDEWEB)

    Keith Rule; Erik Perry; Jim Chrzanowski; Mike Viola; Ron Strykowsky

    2003-09-15

    Original estimates stated that the amount of radioactive waste that will be generated during the dismantling of the Tokamak Fusion Test Reactor will approach two million kilograms with an associated volume of 2,500 cubic meters. The materials were activated by 14 MeV neutrons and were highly contaminated with tritium, which present unique challenges to maintain integrity during packaging and transportation. In addition, the majority of this material is stainless steel and copper structural metal that were specifically designed and manufactured for this one-of-a-kind fusion research reactor. This provided further complexity in planning and managing the waste. We will discuss the engineering concepts, innovative practices, and technologies that were utilized to size reduce, stabilize, and package the many unique and complex components of this reactor. This waste was packaged and shipped in many different configurations and methods according to the transportation regulations and disposal facility requirements. For this particular project, we were able to utilize two separate disposal facilities for burial. This paper will conclude with a complete summary of the actual results of the waste management costs, volumes, and best practices that were developed from this groundbreaking and successful project.

  10. Recharge Data Package for Hanford Single-Shell Tank Waste Management Areas

    Energy Technology Data Exchange (ETDEWEB)

    Fayer, Michael J.; Keller, Jason M.

    2007-09-24

    Pacific Northwest National Laboratory (PNNL) assists CH2M HILL Hanford Group, Inc., in its preparation of the Resource Conservation and Recovery Act (RCRA) Facility Investigation report. One of the PNNL tasks is to use existing information to estimate recharge rates for past and current conditions as well as future scenarios involving cleanup and closure of tank farms. The existing information includes recharge-relevant data collected during activities associated with a host of projects, including those of RCRA, the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA), the CH2M HILL Tank Farm Vadose Zone Project, and the PNNL Remediation and Closure Science Project. As new information is published, the report contents can be updated. The objective of this data package was to use published data to provide recharge estimates for the scenarios being considered in the RCRA Facility Investigation. Recharge rates were estimated for areas that remain natural and undisturbed, areas where the vegetation has been disturbed, areas where both the vegetation and the soil have been disturbed, and areas that are engineered (e.g., surface barrier). The recharge estimates supplement the estimates provided by PNNL researchers in 2006 for the Hanford Site using additional field measurements and model analysis using weather data through 2006.

  11. Study on the cause of the fire and explosion incident at Bituminization Demonstration Facility of PNC Tokai Works

    International Nuclear Information System (INIS)

    On March 11, 1997, a fire and explosion incident occurred in Bituminization Demonstration Facility (BDF) at Tokai Reprocessing Plant (TRP). The cause of the incident was thoroughly investigated and after thirty meetings of 'Investigation Committee of Science and Technology Agency (STA)'. STA reported possible cause of this incident. Continually detailed investigation was discussed at 'Follow-up Committee of STA'. Based on the results of five meetings, JNC concluded that the main cause of fire was frictional heating and viscous heating in the extruder, which provided an unusually high temperature in the bituminized product at filling, since there was no chemical analysis data suggesting the generation of chemical heat. It is considered that these excessive heat generations were due to the decreased feed rate of liquid waste to the extruder. The Follow-up Committee almost agreed to this conclusion. This paper describes general information concerning the incident and details of the cause of the incident. (author)

  12. Geologic Data Package for 2001 Immobilized Low-Activity Waste Performance Assessment

    Energy Technology Data Exchange (ETDEWEB)

    SP Reidel; DG Horton

    1999-12-21

    This database is a compilation of existing geologic data from both the existing and new immobilized low-activity waste disposal sites for use in the 2001 Performance Assessment. Data were compiled from both surface and subsurface geologic sources. Large-scale surface geologic maps, previously published, cover the entire 200-East Area and the disposal sites. Subsurface information consists of drilling and geophysical logs from nearby boreholes and stored sediment samples. Numerous published geological reports are available that describe the subsurface geology of the area. Site-specific subsurface data are summarized in tables and profiles in this document. Uncertainty in data is mainly restricted to borehole information. Variations in sampling and drilling techniques present some correlation uncertainties across the sites. A greater degree of uncertainty exists on the new site because of restricted borehole coverage. There is some uncertainty to the location and orientation of elastic dikes across the sites.

  13. Geologic Data Package for 2001 Immobilized Low-Activity Waste Performance Assessment

    International Nuclear Information System (INIS)

    This database is a compilation of existing geologic data from both the existing and new immobilized low-activity waste disposal sites for use in the 2001 Performance Assessment. Data were compiled from both surface and subsurface geologic sources. Large-scale surface geologic maps, previously published, cover the entire 200-East Area and the disposal sites. Subsurface information consists of drilling and geophysical logs from nearby boreholes and stored sediment samples. Numerous published geological reports are available that describe the subsurface geology of the area. Site-specific subsurface data are summarized in tables and profiles in this document. Uncertainty in data is mainly restricted to borehole information. Variations in sampling and drilling techniques present some correlation uncertainties across the sites. A greater degree of uncertainty exists on the new site because of restricted borehole coverage. There is some uncertainty to the location and orientation of elastic dikes across the sites

  14. Separate Collection Modes for Packaging Waste in China%我国包装废物的分类收集模式

    Institute of Scientific and Technical Information of China (English)

    刘娜; 李丽; 周炳炎; 王琪; 汪群慧; 陈德平

    2011-01-01

    Demonstrations of three different systems for sorting and recovery of typical packaging waste from plastic, paper and metal in three residential communities were established. The demonstrated systems were natural state, public education about sortiing and professional sorting based on public education. With questionnaires, site sampling and in-home interviews, waste flow data including generation amount, reclaimed amount and recovery rate of each kind of packaging waste in the different modes were statistically analyzed based on substance flow analytic tracking observation. The material flow was also analyzed from production to recycling and reuse of packaging waste. The results showed that the current recovery system of packaging waste ( natural state) had some effect on sorting collection. Different packaging waste had different key characteristics, such as paper and metal packaging waste with high recovery rates in families, and the recovery of plastic packaging waste mainly from social workers. With education,packaging waste which was of low-value and difficult to sort could be effectively recovered.%针对塑料类、纸类和金属类等典型包装废物,采用3种包装废物分类收集模式,并在3个居民小区进行示范.3种分类收集模式为:①自然状态;②在小区进行分类收集的宣传教育;③在宣传教育的同时由专业人员对小区生活垃圾进行分类.采用物质流分析的跟踪观察法,通过问卷调查、现场采样和入户采访,对不同模式下各类包装废物的产生量、回收量与回收率等废物流数据进行统计,得到包装废物从产生、回收到再利川和最终处置的物流分析图.结果表明:自然状态下包装废物也有一定的回收率,即我国现有的回收体系对包装废物的分类收集有一定的效果;不同类型包装废物的主要回收环节不同,纸类和金属类包装废物在家庭的回收率较高,而塑料类包装废物的回收主要依赖社

  15. Technical assessment of processing plants as exemplified by the sorting of beverage cartons from lightweight packaging wastes.

    Science.gov (United States)

    Feil, A; Thoden van Velzen, E U; Jansen, M; Vitz, P; Go, N; Pretz, T

    2016-02-01

    The recovery of beverage cartons (BC) in three lightweight packaging waste processing plants (LP) was analyzed with different input materials and input masses in the area of 21-50Mg. The data was generated by gravimetric determination of the sorting products, sampling and sorting analysis. Since the particle size of beverage cartons is larger than 120mm, a modified sampling plan was implemented and targeted multiple sampling (3-11 individual samplings) and a total sample size of respectively 1200l (ca. 60kg) for the BC-products and of about 2400l (ca. 120kg) for material-heterogeneous mixed plastics (MP) and sorting residue products. The results infer that the quantification of the beverage carton yield in the process, i.e., by including all product-containing material streams, can be specified only with considerable fluctuation ranges. Consequently, the total assessment, regarding all product streams, is rather qualitative than quantitative. Irregular operation conditions as well as unfavorable sampling conditions and capacity overloads are likely causes for high confidence intervals. From the results of the current study, recommendations can basically be derived for a better sampling in LP-processing plants. Despite of the suboptimal statistical results, the results indicate very clear that the plants show definite optimisation potentials with regard to the yield of beverage cartons as well as the required product purity. Due to the test character of the sorting trials the plant parameterization was not ideal for this sorting task and consequently the results should be interpreted with care.

  16. Waste Receiving and Packaging, Module 2A, Supplemental Design Requirements Document

    Energy Technology Data Exchange (ETDEWEB)

    Lamberd, D.L.; Boothe, G.F.; Hinkle, A.L.; Horgos, R.M.; LeClair, M.D.; Nash, C.R.; Ocampo, V.P.; Pauly, T.R.; Stroup, J.L.; Weingardt, K.M.

    1994-04-26

    The Supplemental Design Requirements Document (SDRD) is used to communicate plant design information from Westinghouse Hanford Company (WHC) to the US Department of Energy (DOE) and the cognizant Architect Engineer (A/E). Information in the SDRD serves two purposes: to convey design requirements that are too detailed for inclusion in a Functional Design Criteria (FDC) report; and to serve as a means of change control for design commitments in the Conceptual Design Report. The mission of WRAP 2A on the Hanford site is the treatment of contact handled low level mixed waste (MW) for final disposal. The overall systems engineering steps used to reach construction and operation of WRAP 2A are depicted in Figure 1. The WRAP 2A SDRD focuses on the requirements to address the functional analysis provided in Figure 1. This information is provided in sections 2 through 5 of this SDRD. The mission analysis and functional analysis are to be provided in a separate supporting document. The organization of sections 2 through 5 corresponds to the requirements identified in the WRAP 2A functional analysis.

  17. Waste Receiving and Packaging, Module 2A, Supplemental Design Requirements Document

    International Nuclear Information System (INIS)

    The Supplemental Design Requirements Document (SDRD) is used to communicate plant design information from Westinghouse Hanford Company (WHC) to the US Department of Energy (DOE) and the cognizant Architect Engineer (A/E). Information in the SDRD serves two purposes: to convey design requirements that are too detailed for inclusion in a Functional Design Criteria (FDC) report; and to serve as a means of change control for design commitments in the Conceptual Design Report. The mission of WRAP 2A on the Hanford site is the treatment of contact handled low level mixed waste (MW) for final disposal. The overall systems engineering steps used to reach construction and operation of WRAP 2A are depicted in Figure 1. The WRAP 2A SDRD focuses on the requirements to address the functional analysis provided in Figure 1. This information is provided in sections 2 through 5 of this SDRD. The mission analysis and functional analysis are to be provided in a separate supporting document. The organization of sections 2 through 5 corresponds to the requirements identified in the WRAP 2A functional analysis

  18. HOW THE ROCKY FLATS ENVIRONMENTAL TECHNOLOGY SITE DEVELOPED A NEW WASTE PACKAGE USING A POLYUREA COATING THAT IS SAFELY AND ECONOMICALLY ELIMINATING SIZE REDUCTION OF LARGE ITEMS

    Energy Technology Data Exchange (ETDEWEB)

    Dorr, Kent A.; Hogue, Richard S.; Kimokeo, Margaret K.

    2003-02-27

    One of the major challenges involved in closing the Rocky Flats Environmental Technology Site (RFETS) is the disposal of extremely large pieces of contaminated production equipment and building debris. Past practice has been to size reduce the equipment into pieces small enough to fit into approved, standard waste containers. Size reducing this equipment is extremely expensive, and exposes workers to high-risk tasks, including significant industrial, chemical, and radiological hazards. RFETS has developed a waste package using a Polyurea coating for shipping large contaminated objects. The cost and schedule savings have been significant.

  19. Edible packaging materials.

    Science.gov (United States)

    Janjarasskul, Theeranun; Krochta, John M

    2010-01-01

    Research groups and the food and pharmaceutical industries recognize edible packaging as a useful alternative or addition to conventional packaging to reduce waste and to create novel applications for improving product stability, quality, safety, variety, and convenience for consumers. Recent studies have explored the ability of biopolymer-based food packaging materials to carry and control-release active compounds. As diverse edible packaging materials derived from various by-products or waste from food industry are being developed, the dry thermoplastic process is advancing rapidly as a feasible commercial edible packaging manufacturing process. The employment of nanocomposite concepts to edible packaging materials promises to improve barrier and mechanical properties and facilitate effective incorporation of bioactive ingredients and other designed functions. In addition to the need for a more fundamental understanding to enable design to desired specifications, edible packaging has to overcome challenges such as regulatory requirements, consumer acceptance, and scaling-up research concepts to commercial applications. PMID:22129343

  20. Edible packaging materials.

    Science.gov (United States)

    Janjarasskul, Theeranun; Krochta, John M

    2010-01-01

    Research groups and the food and pharmaceutical industries recognize edible packaging as a useful alternative or addition to conventional packaging to reduce waste and to create novel applications for improving product stability, quality, safety, variety, and convenience for consumers. Recent studies have explored the ability of biopolymer-based food packaging materials to carry and control-release active compounds. As diverse edible packaging materials derived from various by-products or waste from food industry are being developed, the dry thermoplastic process is advancing rapidly as a feasible commercial edible packaging manufacturing process. The employment of nanocomposite concepts to edible packaging materials promises to improve barrier and mechanical properties and facilitate effective incorporation of bioactive ingredients and other designed functions. In addition to the need for a more fundamental understanding to enable design to desired specifications, edible packaging has to overcome challenges such as regulatory requirements, consumer acceptance, and scaling-up research concepts to commercial applications.

  1. Calculation Package for the Analysis of Performance of Cells 1-6, with Underdrain, of the Environmental Management Waste Management Facility Oak Ridge, Tennessee

    Energy Technology Data Exchange (ETDEWEB)

    Gonzales D.

    2010-03-30

    This calculation package presents the results of an assessment of the performance of the 6 cell design of the Environmental Management Waste Management Facility (EMWMF). The calculations show that the new cell 6 design at the EMWMF meets the current WAC requirement. QA/QC steps were taken to verify the input/output data for the risk model and data transfer from modeling output files to tables and calculation.

  2. Input data of the thermal analysis on the disposal drift and waste package pitch. Reference case of disposal pit vertical emplacement concept. Set of data

    International Nuclear Information System (INIS)

    The thermal analysis was carried out to determine disposal tunnel spacing and waste package pitch for vertical emplacement as parameters and these specification were assumed in reference case in project overview report of second progress report (H12) on research and development for the geological disposal of HLW in Japan. This report summarizes the input data of reference. The CD-ROM recording the input data is attached. (author)

  3. 放射性废物货包检测平台设计%Design of Detection Platform for Radioactive Waste Package

    Institute of Scientific and Technical Information of China (English)

    孔劲松; 孟开

    2013-01-01

    放射性废物货包检测是放射性废物管理的重要环节.根据放射性废物货包的技术要求参数,借鉴国内外先进检测技术和工艺手段,设计了一套控制精度高、连续可调范围大的货包检测平台,并进行了平台的受力分析.实践检验表明,该装置具有良好的可操作性.%Detection of radioactive waste standard package is an important part in the radioactive waste management.Considering the features of the radioactive waste package and special security requirement,the advanced detection technique and processing technology from China and international are used as the reference,and a set of package detection platform with high control precision and wide continuous range of adjustability is finally designed.The force analysis and practical application of platform indicate that the platform is with fine operability.

  4. Measurement of heat generation from simulated bituminized product

    International Nuclear Information System (INIS)

    The fire and explosion incident occurred at Bituminization Demonstration Facility of PNC Tokai Works on March 11, 1997. In order to ascertain the cause of incident, the investigation has been pushed forward. For the investigation, we prepared simulated bituminized product of measurement of heat generation in low temperature region less than 200degC. We used calvet Calorimeter MS80 for the heat generation measurement. Result of measurement, we were able to catch the feeble heat generation from bituminized product. The maximum calorific value that was able to detect it in isothermal measurement was approximately 1 mW/g in 160degC. It was approximately 2 mW/g in 200degC. And, as the another measurement, the measurement condition went heat rate by 0.01degC/minute, the highest temperature 190degC. As a result, the maximum generation of heat value that was able to detect it was approximately 0.5 mW/g. I changed simulated bituminized products and measured these. A difference of condition is salt particle size, salt content rate (45%, 60%), addition of the simulated precipitate. But there was not a difference in the generation of heat characteristic detected. (author)

  5. Reuse of steel slag in bituminous paving mixtures.

    Science.gov (United States)

    Sorlini, Sabrina; Sanzeni, Alex; Rondi, Luca

    2012-03-30

    This paper presents a comprehensive study to evaluate the mechanical properties and environmental suitability of electric arc furnace (EAF) steel slag in bituminous paving mixtures. A variety of tests were executed on samples of EAF slag to characterize the physical, geometrical, mechanical and chemical properties as required by UNI EN specifications, focusing additionally on the volumetric expansion associated with hydration of free CaO and MgO. Five bituminous mixtures of aggregates for flexible road pavement were designed containing up to 40% of EAF slag and were tested to determine Marshall stability and indirect tensile strength. The leaching behaviour of slag samples and bituminous mixtures was evaluated according to the UNI EN leaching test. The tested slag showed satisfactory physical and mechanical properties and a release of pollutants generally below the limits set by the Italian code. Tests on volume stability of fresh materials confirmed that a period of 2-3 months is necessary to reduce effects of oxides hydration. The results of tests performed on bituminous mixtures with EAF slag were comparable with the performance of mixtures containing natural aggregates and the leaching tests provided satisfactory results. PMID:22305201

  6. 30 CFR 716.4 - Special bituminous coal mines.

    Science.gov (United States)

    2010-07-01

    ... INTERIOR INITIAL PROGRAM REGULATIONS SPECIAL PERFORMANCE STANDARDS § 716.4 Special bituminous coal mines... may be used if necessary to prevent erosion. (3) Spoil piles will be graded and contoured with no more... of an amendment or revision to the State of Wyoming's regulatory program, regulations, or...

  7. IN-PACKAGE CHEMISTRY ABSTRACTION

    International Nuclear Information System (INIS)

    This report was developed in accordance with the requirements in ''Technical Work Plan for Postclosure Waste Form Modeling'' (BSC 2005 [DIRS 173246]). The purpose of the in-package chemistry model is to predict the bulk chemistry inside of a breached waste package and to provide simplified expressions of that chemistry as a function of time after breach to Total Systems Performance Assessment for the License Application (TSPA-LA). The scope of this report is to describe the development and validation of the in-package chemistry model. The in-package model is a combination of two models, a batch reactor model, which uses the EQ3/6 geochemistry-modeling tool, and a surface complexation model, which is applied to the results of the batch reactor model. The batch reactor model considers chemical interactions of water with the waste package materials, and the waste form for commercial spent nuclear fuel (CSNF) waste packages and codisposed (CDSP) waste packages containing high-level waste glass (HLWG) and DOE spent fuel. The surface complexation model includes the impact of fluid-surface interactions (i.e., surface complexation) on the resulting fluid composition. The model examines two types of water influx: (1) the condensation of water vapor diffusing into the waste package, and (2) seepage water entering the waste package as a liquid from the drift. (1) Vapor-Influx Case: The condensation of vapor onto the waste package internals is simulated as pure H2O and enters at a rate determined by the water vapor pressure for representative temperature and relative humidity conditions. (2) Liquid-Influx Case: The water entering a waste package from the drift is simulated as typical groundwater and enters at a rate determined by the amount of seepage available to flow through openings in a breached waste package

  8. IN-PACKAGE CHEMISTRY ABSTRACTION

    Energy Technology Data Exchange (ETDEWEB)

    E. Thomas

    2005-07-14

    This report was developed in accordance with the requirements in ''Technical Work Plan for Postclosure Waste Form Modeling'' (BSC 2005 [DIRS 173246]). The purpose of the in-package chemistry model is to predict the bulk chemistry inside of a breached waste package and to provide simplified expressions of that chemistry as a function of time after breach to Total Systems Performance Assessment for the License Application (TSPA-LA). The scope of this report is to describe the development and validation of the in-package chemistry model. The in-package model is a combination of two models, a batch reactor model, which uses the EQ3/6 geochemistry-modeling tool, and a surface complexation model, which is applied to the results of the batch reactor model. The batch reactor model considers chemical interactions of water with the waste package materials, and the waste form for commercial spent nuclear fuel (CSNF) waste packages and codisposed (CDSP) waste packages containing high-level waste glass (HLWG) and DOE spent fuel. The surface complexation model includes the impact of fluid-surface interactions (i.e., surface complexation) on the resulting fluid composition. The model examines two types of water influx: (1) the condensation of water vapor diffusing into the waste package, and (2) seepage water entering the waste package as a liquid from the drift. (1) Vapor-Influx Case: The condensation of vapor onto the waste package internals is simulated as pure H{sub 2}O and enters at a rate determined by the water vapor pressure for representative temperature and relative humidity conditions. (2) Liquid-Influx Case: The water entering a waste package from the drift is simulated as typical groundwater and enters at a rate determined by the amount of seepage available to flow through openings in a breached waste package.

  9. Technical assessment of processing plants as exemplified by the sorting of beverage cartons from lightweight packaging wastes.

    Science.gov (United States)

    Feil, A; Thoden van Velzen, E U; Jansen, M; Vitz, P; Go, N; Pretz, T

    2016-02-01

    The recovery of beverage cartons (BC) in three lightweight packaging waste processing plants (LP) was analyzed with different input materials and input masses in the area of 21-50Mg. The data was generated by gravimetric determination of the sorting products, sampling and sorting analysis. Since the particle size of beverage cartons is larger than 120mm, a modified sampling plan was implemented and targeted multiple sampling (3-11 individual samplings) and a total sample size of respectively 1200l (ca. 60kg) for the BC-products and of about 2400l (ca. 120kg) for material-heterogeneous mixed plastics (MP) and sorting residue products. The results infer that the quantification of the beverage carton yield in the process, i.e., by including all product-containing material streams, can be specified only with considerable fluctuation ranges. Consequently, the total assessment, regarding all product streams, is rather qualitative than quantitative. Irregular operation conditions as well as unfavorable sampling conditions and capacity overloads are likely causes for high confidence intervals. From the results of the current study, recommendations can basically be derived for a better sampling in LP-processing plants. Despite of the suboptimal statistical results, the results indicate very clear that the plants show definite optimisation potentials with regard to the yield of beverage cartons as well as the required product purity. Due to the test character of the sorting trials the plant parameterization was not ideal for this sorting task and consequently the results should be interpreted with care. PMID:26547410

  10. Feasibility study of 235U and 239Pu characterization in radioactive waste drums using neutron-induced fission delayed gamma rays

    Science.gov (United States)

    Nicol, T.; Pérot, B.; Carasco, C.; Brackx, E.; Mariani, A.; Passard, C.; Mauerhofer, E.; Collot, J.

    2016-10-01

    This paper reports a feasibility study of 235U and 239Pu characterization in 225 L bituminized waste drums or 200 L concrete waste drums, by detecting delayed fission gamma rays between the pulses of a deuterium-tritium neutron generator. The delayed gamma yields were first measured with bare samples of 235U and 239Pu in REGAIN, a facility dedicated to the assay of 118 L waste drums by Prompt Gamma Neutron Activation Analysis (PGNAA) at CEA Cadarache, France. Detectability in the waste drums is then assessed using the MCNPX model of MEDINA (Multi Element Detection based on Instrumental Neutron Activation), another PGNAA cell dedicated to 200 L drums at FZJ, Germany. For the bituminized waste drum, performances are severely hampered by the high gamma background due to 137Cs, which requires the use of collimator and shield to avoid electronics saturation, these elements being very penalizing for the detection of the weak delayed gamma signal. However, for lower activity concrete drums, detection limits range from 10 to 290 g of 235U or 239Pu, depending on the delayed gamma rays of interest. These detection limits have been determined by using MCNPX to calculate the delayed gamma useful signal, and by measuring the experimental gamma background in MEDINA with a 200 L concrete drum mock-up. The performances could be significantly improved by using a higher interrogating neutron emission and an optimized experimental setup, which would allow characterizing nuclear materials in a wide range of low and medium activity waste packages.

  11. Solid waste reclamation and recycling: Packaging and containers. (Latest citations from the NTIS bibliographic database). Published Search

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-08-01

    The bibliography contains citations concerning techniques and management of packaging and container recycling. References discuss recycling of tin and aluminum cans, reverse vending machines, reusable packaging and containers, and the future of containers. Environmental aspects, government programs, and development of recycling markets are covered. (Contains 50-250 citations and includes a subject term index and title list.) (Copyright NERAC, Inc. 1995)

  12. Amenability of Muzret bituminous coal to oil agglomeration

    Energy Technology Data Exchange (ETDEWEB)

    Sahinoglu, E.; Uslu, T. [Karadeniz Technical University, Department of Mining Engineering, 61080 Trabzon (Turkey)

    2008-12-15

    Laboratory scale agglomeration tests were undertaken to investigate the amenability of Muzret (Yusufeli-Artvin) bituminous coal to oil agglomeration. Kerosene was extensively used as oil in the tests. In addition, fuel oil, diesel oil, and hazelnut oil were also used in order to determine the effect of oil type. The effects of the parameters including coal content, kerosene content, agglomeration time, coal particle size, pH, oil type, and agitation rate, on the combustible matter recovery, ash reduction and pyritic sulphur reduction, were investigated. It was found that Muzret bituminous coal could be readily cleaned by oil agglomeration with substantial reductions in ash and pyritic sulphur content. Maximum combustible matter recovery, ash reduction and pyritic sulphur reduction were achieved to be 85.54%, 59.98%, and 85.17%, respectively. (author)

  13. Performance of PAHs emission from bituminous coal combustion

    Institute of Scientific and Technical Information of China (English)

    严建华; 尤孝方; 李晓东; 倪明江; 尹雪峰; 岑可法

    2004-01-01

    Carcinogenic and mutagenic polycyclic aromatic hydrocarbons (PAHs) generated in coal combustion have caused great environmental health concern. Seventeen PAHs (16 high priority PAHs recommended by USEPA plus Benzo[e]pyrene) present in five raw bituminous coals and released during bituminous coal combustion were studied. The effects of combustion temperature, gas atmosphere, and chlorine content of raw coal on PAHs formation were investigated. Two additives (copper and cupric oxide) were added when the coal was burned. The results indicated that significant quantities of PAHs were produced from incomplete combustion of coal pyrolysis products at high temperature, and that temperature is an important causative factor of PAHs formation. PAHs concentrations decrease with the increase of chlorine content in oxygen or in nitrogen atmosphere. Copper and cupric oxide additives can promote PAHs formation (especially the multi-ring PAHs) during coal combustion.

  14. Waste Package Misload Probability

    Energy Technology Data Exchange (ETDEWEB)

    J.K. Knudsen

    2001-11-20

    The objective of this calculation is to calculate the probability of occurrence for fuel assembly (FA) misloads (i.e., Fa placed in the wrong location) and FA damage during FA movements. The scope of this calculation is provided by the information obtained from the Framatome ANP 2001a report. The first step in this calculation is to categorize each fuel-handling events that occurred at nuclear power plants. The different categories are based on FAs being damaged or misloaded. The next step is to determine the total number of FAs involved in the event. Using the information, a probability of occurrence will be calculated for FA misload and FA damage events. This calculation is an expansion of preliminary work performed by Framatome ANP 2001a.

  15. Identification of aggregates for Tennessee bituminous surface courses

    Science.gov (United States)

    Sauter, Heather Jean

    Tennessee road construction is a major venue for federal and state spending. Tax dollars each year go to the maintenance and construction of roads. One aspect of highway construction that affects the public is the safety of its state roads. There are many factors that affect the safety of a given road. One factor that was focused on in this research was the polish resistance capabilities of aggregates. Several pre-evaluation methods have been used in the laboratory to predict what will happen in a field situation. A new pre-evaluation method was invented that utilized AASHTO T 304 procedure upscaled to accommodate surface bituminous aggregates. This new method, called the Tennessee Terminal Textural Condition Method (T3CM), was approved by Tennessee Department of Transportation to be used as a pre-evaluation method on bituminous surface courses. It was proven to be operator insensitive, repeatable, and an accurate indication of particle shape and texture. Further research was needed to correlate pre-evaluation methods to the current field method, ASTM E 274-85 Locked Wheel Skid Trailer. In this research, twenty-five in-place bituminous projects and eight source evaluations were investigated. The information gathered would further validate the T3CM and find the pre-evaluation method that best predicted the field method. In addition, new sources of aggregates for bituminous surface courses were revealed. The results of this research have shown T3CM to be highly repeatable with an overall coefficient of variation of 0.26% for an eight sample repeatability test. It was the best correlated pre-evaluation method with the locked wheel skid trailer method giving an R2 value of 0.3946 and a Pearson coefficient of 0.710. Being able to predict field performance of aggregates prior to construction is a powerful tool capable of saving time, money, labor, and possibly lives.

  16. Antioxidative, hemocompatible, fluorescent carbon nanodots from an "end-of-pipe" agricultural waste: exploring its new horizon in the food-packaging domain.

    Science.gov (United States)

    Das Purkayastha, Manashi; Manhar, Ajay Kumar; Das, Vijay Kumar; Borah, Anjan; Mandal, Manabendra; Thakur, Ashim Jyoti; Mahanta, Charu Lata

    2014-05-21

    The attention of researchers is burgeoning toward oilseed press-cake valorization for its high protein content. Protein removal from oil-cakes generates large quantities of fibrous residue (oil-and-protein spent meal) as a byproduct, which currently has very limited practical utility. In the wake of increasing awareness in waste recycling, a simple environmentally benign hydrothermal carbonization process to convert this "end-of-pipe" waste (spent meal) into antioxidative, hemocompatible, fluorescent carbonaceous nanoparticles (FCDs) has been described. In the present investigation, an interesting application of FCDs in fabricating low-cost rapeseed protein-based fluorescent film, with improved antioxidant potential (17.5-19.3-fold) and thermal stability has been demonstrated. The nanocomposite film could also be used as forgery-proof packaging due to its photoluminescence property. For assessing the feasibility of antioxidative FCDs in real food systems, a comparative investigation was further undertaken to examine the effect of such nanocarbon-loaded composite film on the oxidative shelf life of rapeseed oil. Oil samples packed in nanocomposite film sachets showed significant delay in oxidative rancidity compared to those packed in pristine protein-film sachet (free fatty acids, peroxide value, and thiobarbituric acid-reactive substances reduced up to 1.4-, 2-, and 1.2-fold, respectively). The work presents a new concept of biobased fluorescent packaging and avenues for harnessing this potent waste. PMID:24784501

  17. Continuous use after 1998 of the scaling factors for radionuclide evaluation in the packaged solid wastes generated from the Fukushima Daiichi Nuclear Power Plants of Tokyo Electric Power Company

    International Nuclear Information System (INIS)

    The waste packages containing solidified radioactive wastes generated from nuclear power plants are evaluated by measuring radioactivity from outside the drums and applying the scaling factor method to estimate the inside radionuclide concentrations. Japan Nuclear Energy Safety Organization (JNES) is in charge of the confirmation of the applicability of this method and this report presents the results of the examining processes of the working committee for this purpose. The properties of the solidified waste packages, the guidelines of using the scaling factors, the concept of renewal of reactor composing materials, are explained with radionuclide information and compliance with formalities for safety maintenance and control. (S. Ohno)

  18. Synthesis of knowledge on the long-term behaviour of concretes. Applications to cemented waste packages; Synthese des connaissances sur le comportement a long terme des betons. Application aux colis cimentes

    Energy Technology Data Exchange (ETDEWEB)

    Richet, C.; Galle, C.; Le Bescop, P.; Peycelon, H.; Bejaoui, S.; Tovena, I.; Pointeau, I.; L' Hostis, V.; Levera, P

    2004-03-01

    As stipulated in the former law of December 91 relating to 'concrete waste package', a progress report (phenomenological reference document) was first provided in 1999. The objective was to make an assessment of the knowledge acquired on the long-term behaviour of cement-based waste packages in the context of deep disposal and/or interim storage. The present document is an updated summary report. It takes into account a new knowledge assessment, considers coupled mechanisms and should contribute to the first performance studies (operational calculations). Handling and radio-nuclides (RN) confinement are the two major functional properties requested from the concrete used for the waste packages. In unsaturated environment (interim storage/disposal prior to closing), the main problem is the generation of cracks in the material. This aspect is a key parameter from the mechanical point of view (retrievability). It can have a major impact on the disposal phase (confinement). In saturated environment (disposal post-closing phase), the main concern is the chemical degradation of the waste package concrete submitted to underground waters leaching. In this context, the major thema are: the durability of the concretes under water (chemical degradation) and in unsaturated medium (corrosion of reinforcement), matter transport, RN retention, chemistry / transport / mechanical couplings. On the other hand, laboratory data on the behaviour of concretes are used to evaluate the RN source term of waste packages in function of time (concrete waste package OPerational Model, i.e. 'Concrete MOP'). The 'MOP' provides the physico-chemical description of the RN release in relationship with the waste package degradation itself. This description is based on simplified phenomenology for which only dimensioning mechanisms are taken into account. The use of Diffu-Ca code (basic module for the MOP) on the CASTEM numerical plate-form, already allows operational

  19. Development of a non-destructive experimental method and a calculation procedure for the identification and quantification of gamma emitter radionuclides in packages containing low and medium-level radioactive wastes

    International Nuclear Information System (INIS)

    In this report, a non-destructive method on gamma spectrometry is investigated as a means of providing information on the radiological characteristics of low- and medium-level radioactive waste already packaged in drums. Experiments are reported using full-scale and mock-up scale homogeneous packages, in order to obtain real calibration curves and attenuation data, and heterogeneous packages in order to determine the influence of the concrete shield, and the distribution in the drum of radioactive materials, and collimation shape influence. For performing calibration curves and filters Europium-152 standard packages were employed and Cs-137, Co-60 and Mn-54 standard packages for quality control. Conclusions are presented on counting methods to provide a practically constant efficiency for any position in the drum and for any length, surface size or shape of the radioactive material. Calculation codes have been implemented to carry out the quantification of every packages studied. 37 figs., 35 tabs

  20. Background studies in support of a feasibility assessment on the use of copper-base materials for nuclear waste packages in a repository in tuff

    International Nuclear Information System (INIS)

    This report combines six work units performed in FY'85--86 by the Copper Development Association and the International Copper Research Association under contract with the University of California. The work includes literature surveys and state-of-the-art summaries on several considerations influencing the feasibility of the use of copper-base materials for fabricating high-level nuclear waste packages for the proposed repository in tuff rock at Yucca Mountain, Nevada. The general conclusion from this work was that copper-base materials are viable candidates for inclusion in the materials selection process for this application. 55 refs., 48 figs., 22 tabs

  1. A Versatile System for the In-Field Non-Destructive Characterization of Radioactive Waste Packages and of Objects in the Defense against Nuclear Threats

    International Nuclear Information System (INIS)

    In-filed non-destructive characterization of radioactive waste packages and of objects in the defense of nuclear threats requires purpose-built but similar equipment. Based on commercially available measuring devices like dose-rate and gamma detectors, X-ray and gamma-transmission sources etc. a versatile and mobile mechanical positioning system for these devices is designed, assembled and operated facilitating basic to even complex measuring procedures. Several in-field measuring campaigns resulted in its further optimization. Today an highly mobile and flexible mechanical manipulator system is available supporting nearly all types of required measuring devices thus rising to nearly all occasions. (authors)

  2. Background studies in support of a feasibility assessment on the use of copper-base materials for nuclear waste packages in a repository in tuff

    Energy Technology Data Exchange (ETDEWEB)

    Van Konynenburg, R.A. [Lawrence Livermore National Lab., CA (USA); Kundig, K.J.A.; Lyman, W.S.; Prager, M.; Meyers, J.R.; Servi, I.S. [CDA/INCRA Joint Advisory Group, Greenwich, CT (USA)

    1990-06-01

    This report combines six work units performed in FY`85--86 by the Copper Development Association and the International Copper Research Association under contract with the University of California. The work includes literature surveys and state-of-the-art summaries on several considerations influencing the feasibility of the use of copper-base materials for fabricating high-level nuclear waste packages for the proposed repository in tuff rock at Yucca Mountain, Nevada. The general conclusion from this work was that copper-base materials are viable candidates for inclusion in the materials selection process for this application. 55 refs., 48 figs., 22 tabs.

  3. Proceedings of the 6th Annual Meeting for Excess Weapons Plutonium Disposition: Plutonium Packaging, Storage and Transportation and WasteTreatment, Storage and Disposal Activities

    Energy Technology Data Exchange (ETDEWEB)

    Jardine, L J

    2005-06-30

    The sixth annual Excess Weapons Plutonium Disposition meeting organized by Lawrence Livermore National Laboratory (LLNL) was held November 15-17, 2004, at the State Education Center (SEC), 4 Aerodromnya Drive, St. Petersburg, Russia. The meeting discussed Excess Weapons Plutonium Disposition topics for which LLNL has the US Technical Lead Organization responsibilities. The technical areas discussed included Radioactive Waste Treatment, Storage, and Disposal, and Plutonium Oxide and Plutonium Metal Packaging, Storage and Transportation and Spent Fuel Packaging, Storage and Transportation. The meeting was conducted with a conference format using technical presentations of papers with simultaneous translation into English and Russian. There were 55 Russian attendees from 16 different Russian organizations and four non-Russian attendees from the US. Forty technical presentations were made. The meeting agenda is given in Appendix B and the attendance list is in Appendix C. The 16 different Russian design, industrial sites, and scientific organizations in attendance included staff from Rosatom/Minatom, Federal Nuclear and Radiation Safety Authority of Russia (GOSATOMNADZOR, NIERA/GAN), All Russian Designing & Scientific Research Institute of Complex Power Technology (VNIPIET), Khlopin Radium Institute (KRI), A. A. Bochvar All Russian Scientific Research Institute of Inorganic Materials (VNIINM), All Russian & Design Institute of Production Engineering (VNIPIPT), Ministry of Atomic Energy of Russian Federation Specialized State Designing Institute (GSPI), State Scientific Center Research Institute of Atomic Reactors (RIAR), Siberian Chemical Combine Tomsk (SCC), Mayak PO, Mining Chemical Combine (MCC K-26), Institute of Biophysics (IBPh), Sverdlosk Scientific Research Institute of Chemical Machine Building (SNIIChM), Kurchatov Institute (KI), Institute of Physical Chemistry Russian Academy of Science (IPCh RAS) and Radon PO-Moscow. The four non-Russian attendees included

  4. A similarity solution for two-phase fluid and heat flow near high-level nuclear waste packages emplaced in porous media

    International Nuclear Information System (INIS)

    The emplacement of a heat source, such as a high-level nuclear waste package, into a geologic medium causes strongly coupled thermal and hydrologic behavior. Under certain conditions, a heat pipe may develop, with significant impact on conditions at the heat source. In an infinite, homogenous, permeable medium with a constant-strength linear heat source, the partial differential equations governing fluid and heat flows in a radial geometry can be converted to ordinary differential equations by using a similarity variable, η = r/√t. These equations are numerically integrated using an iterative 'shooting' method to provide a description of temperature, pressure, saturation, heat flow, gas flow, and liquid flow conditions around a heat source such as a nuclear waste package. The similarity solution is verified by numerical finite-difference simulations. Illustrative solutions are given for a range of hydrologic and thermal parameters, and the likelihood of heat-pipe development for conditions at several proposed repository sites is discussed. (author)

  5. A user's guide to the GoldSim/BLT-MS integrated software package:a low-level radioactive waste disposal performance assessment model.

    Energy Technology Data Exchange (ETDEWEB)

    Knowlton, Robert G.; Arnold, Bill Walter; Mattie, Patrick D.

    2007-03-01

    Sandia National Laboratories (Sandia), a U.S. Department of Energy National Laboratory, has over 30 years experience in the assessment of radioactive waste disposal and at the time of this publication is providing assistance internationally in a number of areas relevant to the safety assessment of radioactive waste disposal systems. In countries with small radioactive waste programs, international technology transfer program efforts are often hampered by small budgets, schedule constraints, and a lack of experienced personnel. In an effort to surmount these difficulties, Sandia has developed a system that utilizes a combination of commercially available software codes and existing legacy codes for probabilistic safety assessment modeling that facilitates the technology transfer and maximizes limited available funding. Numerous codes developed and endorsed by the United States Nuclear Regulatory Commission (NRC) and codes developed and maintained by United States Department of Energy are generally available to foreign countries after addressing import/export control and copyright requirements. From a programmatic view, it is easier to utilize existing codes than to develop new codes. From an economic perspective, it is not possible for most countries with small radioactive waste disposal programs to maintain complex software, which meets the rigors of both domestic regulatory requirements and international peer review. Therefore, revitalization of deterministic legacy codes, as well as an adaptation of contemporary deterministic codes, provides a credible and solid computational platform for constructing probabilistic safety assessment models. This document is a reference users guide for the GoldSim/BLT-MS integrated modeling software package developed as part of a cooperative technology transfer project between Sandia National Laboratories and the Institute of Nuclear Energy Research (INER) in Taiwan for the preliminary assessment of several candidate low

  6. Bituminous solidification, disposal, transport and burial of spent ion-exchange resins. Part of a coordinated programme on treatment of spent ion exchange resins

    International Nuclear Information System (INIS)

    The project dealing with the incorporation of spent ion-exchange resins into bitumen was performed within the Agency coordinated research programme on treatment of spent ion-exchange resins. Physical and chemical properties of commercial ion-exchange resins, bitumens and bituminized resins were studied. It was shown that bitumen with low oil content and with a softening point of 60-70 deg. C are applicable for the incorporation of resins. The final waste form is allowed to contain maximum 50% resin. The comprehensive study of the biological resistance of B-30 bitumen was performed. That showed that any bacteriological attack can be regarded as generally insignificant. A continuously operating technology was realized on a semi-plant scale. The best operating conditions of this technology were determined. On the basis of the experience gained from the experiments a design of the bituminization plant of 50m3 dry resin/year treatment capacity was proposed

  7. The Garbage Project Revisited: From a 20th Century Archaeology of Food Waste to a Contemporary Study of Food Packaging Waste

    OpenAIRE

    La Vergne Lehmann

    2015-01-01

    In 1973, Dr. Bill Rathje and his students at the University of Arizona began what was to become a two decade long investigation into American consumer waste habits. An archaeologist by profession, Rathje decided to adapt traditional archaeological methods and apply them to contemporary archaeological situations. This provided a platform for improving the understanding of what was really happening with, amongst other forms of waste, food at the consumer household level. The Garbage Project was...

  8. Vegetation cover and long-term conservation of radioactive waste packages: the case study of the CSM waste disposal facility (Manche District, France).

    Science.gov (United States)

    Petit-Berghem, Yves; Lemperiere, Guy

    2012-03-01

    The CSM is the first French waste disposal facility for radioactive waste. Waste material is buried several meters deep and protected by a multi-layer cover, and equipped with a drainage system. On the surface, the plant cover is a grassland vegetation type. A scientific assessment has been carried out by the Géophen laboratory, University of Caen, in order to better characterize the plant cover (ecological groups and associated soils) and to observe its medium and long term evolution. Field assessments made on 10 plots were complemented by laboratory analyses carried out over a period of 1 year. The results indicate scenarios and alternative solutions which could arise, in order to passively ensure the long-term safety of the waste disposal system. Several proposals for a blanket solution are currently being studied and discussed, under the auspices of international research institutions in order to determine the most appropriate materials for the storage conditions. One proposal is an increased thickness of these materials associated with a geotechnical barrier since it is well adapted to the forest plants which are likely to colonize the site. The current experiments that are carried out will allow to select the best option and could provide feedback for other waste disposal facility sites already being operated in France (CSFMA waste disposal facility, Aube district) or in other countries.

  9. In-Package Chemistry Abstraction

    International Nuclear Information System (INIS)

    This report was developed in accordance with the requirements in ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]). The purpose of the in-package chemistry model is to predict the bulk chemistry inside of a breached waste package and to provide simplified expressions of that chemistry as function of time after breach to Total Systems Performance Assessment for the License Application (TSPA-LA). The scope of this report is to describe the development and validation of the in-package chemistry model. The in-package model is a combination of two models, a batch reactor model that uses the EQ3/6 geochemistry-modeling tool, and a surface complexation model that is applied to the results of the batch reactor model. The batch reactor model considers chemical interactions of water with the waste package materials and the waste form for commercial spent nuclear fuel (CSNF) waste packages and codisposed waste packages that contain both high-level waste glass (HLWG) and DOE spent fuel. The surface complexation model includes the impact of fluid-surface interactions (i.e., surface complexation) on the resulting fluid composition. The model examines two types of water influx: (1) the condensation of water vapor that diffuses into the waste package, and (2) seepage water that enters the waste package from the drift as a liquid. (1) Vapor Influx Case: The condensation of vapor onto the waste package internals is simulated as pure H2O and enters at a rate determined by the water vapor pressure for representative temperature and relative humidity conditions. (2) Water Influx Case: The water entering a waste package from the drift is simulated as typical groundwater and enters at a rate determined by the amount of seepage available to flow through openings in a breached waste package. TSPA-LA uses the vapor influx case for the nominal scenario for simulations where the waste package has been

  10. In-Package Chemistry Abstraction

    Energy Technology Data Exchange (ETDEWEB)

    E. Thomas

    2004-11-09

    This report was developed in accordance with the requirements in ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]). The purpose of the in-package chemistry model is to predict the bulk chemistry inside of a breached waste package and to provide simplified expressions of that chemistry as function of time after breach to Total Systems Performance Assessment for the License Application (TSPA-LA). The scope of this report is to describe the development and validation of the in-package chemistry model. The in-package model is a combination of two models, a batch reactor model that uses the EQ3/6 geochemistry-modeling tool, and a surface complexation model that is applied to the results of the batch reactor model. The batch reactor model considers chemical interactions of water with the waste package materials and the waste form for commercial spent nuclear fuel (CSNF) waste packages and codisposed waste packages that contain both high-level waste glass (HLWG) and DOE spent fuel. The surface complexation model includes the impact of fluid-surface interactions (i.e., surface complexation) on the resulting fluid composition. The model examines two types of water influx: (1) the condensation of water vapor that diffuses into the waste package, and (2) seepage water that enters the waste package from the drift as a liquid. (1) Vapor Influx Case: The condensation of vapor onto the waste package internals is simulated as pure H2O and enters at a rate determined by the water vapor pressure for representative temperature and relative humidity conditions. (2) Water Influx Case: The water entering a waste package from the drift is simulated as typical groundwater and enters at a rate determined by the amount of seepage available to flow through openings in a breached waste package. TSPA-LA uses the vapor influx case for the nominal scenario for simulations where the waste

  11. Chapter No.6. Radioactive waste (RAW)

    International Nuclear Information System (INIS)

    UJD SR supervises all phases of radioactive waste management at nuclear installations and final phases of institutional radioactive waste management. Total concentrates generation and minimisation trends during recent period (for NPP V1, NPP V2 Bohunice and NPP Mochovce) as well as increasing of their conditioning are presented. So nearly 8000 m3 of concentrates were stored at the end of year 2001, representing 75.5% of storage capacity at Bohunice site. Total amount of solid waste stored at VVER NPPs reached 3500 m3 at December 2001. The solid waste storage capacity of NPP V1 Bohunice is spent. NPP A1 Bohunice with a heterogeneous reactor using natural uranium, moderated by heavy water and cooled by CO2 had been in operation for 4 years. It was finally shut down in February 1977 after an accident (INES level 4). Main activities in 2001 were focused on an increasing of radwaste management safety by the waste solidification or storage improvement, especially in the case of inorganic spent fuel coolant. The total solid radwaste inventory in 2001 (including filters) represents cca 700 m3 of soft and pre-pressed not sorted waste, 700 m3 of sorted waste, more then 1 000 t of scrap and next 260 m3 of scrap in drums. The total volume of contaminated soils and concrete is more than 7200 m3. The volume of solid radwaste depends on the extent of dismantling works and following treatment, conditioning and disposal. The total volume is also increased by products of bituminization and cementation facilities in drums stored before conditioning at NPP A1 Bohunice site. The facilities focused only on treatment of radwaste from NPP A1 decommissioning are operated by the same operator as a part of this decommissioned nuclear installation. Experimental and research treatment plants are operated by VUJE. A facility for the treatment and conditioning of operational liquid radwaste from NPP Mochovce is under preparation and a start of construction is foreseen in 2004. The nuclear

  12. Packaging fluency

    DEFF Research Database (Denmark)

    Mocanu, Ana; Chrysochou, Polymeros; Bogomolova, Svetlana;

    2011-01-01

    Research on packaging stresses the need for packaging design to read easily, presuming fast and accurate processing of product-related information. In this paper we define this property of packaging as “packaging fluency”. Based on the existing marketing and cognitive psychology literature...... on packaging design and processing fluency, our aim is to define and conceptualise packaging fluency. We stress the important role of packaging fluency since it is anticipated that a fluent package would influence the evaluative judgments for a product. We conclude this paper by setting the research agenda...

  13. 基于连锁超市的包装废弃物回收模型%Recycling Model of Waste Packaging Materials for Chain Supermarkets

    Institute of Scientific and Technical Information of China (English)

    黄勇; 邱丽艳

    2013-01-01

    阐述了包装废弃物回收的意义,基于资源整合与共享的理论思想,提出连锁超市的回收模式,即基于连锁超市的配送体系,在连锁超市合理选址的基础上,依超市建立回收站点,集中回收可再利用的包装;充分利用配送中心的网络资源,建立以达到降低物流成本,实现资源再利用为目的的模型.最后用模糊综合评价法对其进行了评价.%In this paper, we introduced the significance of the recycling of waste packaging materials and on the basis of the ideas of resource integration and sharing, proposed the recycling model of the chain supermarkets, which, on the basis of the reasonable location of the supermarkets, recycled reusable packaging materials in a centralized way and made full use of the network resource of the distribution centers to reduce logistics cost and realize resource reclamation. At the end, we evaluated the model using fuzzy comprehensive evaluation method.

  14. Waste and Disposal: Research and Development

    Energy Technology Data Exchange (ETDEWEB)

    Neerdael, B.; Marivoet, J.; Put, M.; Van Iseghem, P

    2002-04-01

    This contribution to the annual report describes the main activities of the Waste and Disposal Department of the Belgian Nuclear Research Center SCK-CEN. Achievements in 2001 in three topical areas are reported on: performance assessments (PA), waste forms/packages and near- and far field studies. Performance assessment calculations were made for the geological disposal of high-level and long-lived waste in a clay formation. SCK-CEN partcipated in several PA projects supported by the European Commission. In the BENIPA project, the role of bentonite barriers in performance assessments of HLW disposal systems is evaluated. The applicability of various output variables (concentrations, fluxes) as performance and safety indicators is investigated in the SPIN project. The BORIS project investigates the chemical behaviour and the migration of radionuclides at the Borehole injection site at Krasnoyarsk-26 and Tomsk-7. SCK-CEN contributed to an impact assessment of a radium storage facility at Olen (Belgium) and conducted PA for site-specific concepts regarding surface or deep disposal of low-level waste at the nuclear zones in the Mol-Dessel region. As regards R and D on waste forms and packages, SCK continued research on the compatbility of various waste forms (bituminised waste, vitrified waste, spent fuel) with geological disposal in clay. Main emphasis in 2001 was on corrosion studies on vitrified high-level waste, the investigation of localised corrosion of candidate container and overpack materials and the study of the effect of the degradation of cellulose containing waste as well as of bituminized waste on the solubility and the sorption of Pu and Am in geological disposal conditions in clay. With regard to near- and far-field studies, percolation and diffusion experiments to determine migration parameters of key radionuclides were continued. The electromigration technique was used to study the migration of redox sensitive species like uranium. In addition to

  15. Remaining Sites Verification Package for the 128-B-3 Burn Pit Site, Waste Site Reclassification Form 2006-058

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2006-11-17

    The 128-B-3 waste site is a former burn and disposal site for the 100-B/C Area, located adjacent to the Columbia River. The 128-B-3 waste site has been remediated to meet the remedial action objectives specified in the Remaining Sites ROD. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results of sampling at upland areas of the site also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

  16. Remaining Sites Verification Package for the 116-C-3, 105-C Chemical Waste Tanks, Waste Site Reclassification Form 2008-002

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2008-01-31

    The 116-C-3 waste site consisted of two underground storage tanks designed to receive mixed waste from the 105-C Reactor Metals Examination Facility chemical dejacketing process. Confirmatory evaluation and subsequent characterization of the site determined that the southern tank contained approximately 34,000 L (9,000 gal) of dejacketing wastes, and that the northern tank was unused. In accordance with this evaluation, the verification sampling and modeling results support a reclassification of this site to Interim Closed Out. The results of verification sampling demonstrate that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also show that residual contaminant concentrations are protective of groundwater and the Columbia River.

  17. Remaining Sites Verification Package for the 100-B-1 Surface Chemical and Solid Waste Dumping Area, Waste Site Reclassification Form 2006-003

    Energy Technology Data Exchange (ETDEWEB)

    R. A. Carlson

    2006-04-24

    The 100-B-1 waste site was a dumping site that was divided into two areas. One area was used as a laydown area for construction materials, and the other area was used as a chemical dumping area. The 100-B-1 Surface Chemical and Solid Waste Dumping Area site meets the remedial action objectives specified in the Remaining Sites ROD. The results demonstrate that residual contaminant concentrations support future unrestricted land uses that can be represented by a rural-residential scenario. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  18. Microelectronic packaging

    CERN Document Server

    Datta, M; Schultze, J Walter

    2004-01-01

    Microelectronic Packaging analyzes the massive impact of electrochemical technologies on various levels of microelectronic packaging. Traditionally, interconnections within a chip were considered outside the realm of packaging technologies, but this book emphasizes the importance of chip wiring as a key aspect of microelectronic packaging, and focuses on electrochemical processing as an enabler of advanced chip metallization.Divided into five parts, the book begins by outlining the basics of electrochemical processing, defining the microelectronic packaging hierarchy, and emphasizing the impac

  19. Remaining Sites Verification Package for the 1607-F4 Sanitary Sewer System, Waste Site Reclassification Form 2004-131

    Energy Technology Data Exchange (ETDEWEB)

    L. M. Dittmer

    2007-12-03

    The 1607-F4 waste site is the former location of the sanitary sewer system that serviced the former 115-F Gas Recirculation Building. The system included a septic tank, drain field, and associated pipeline that were in use from 1944 to 1965. The 1607-F4 waste site received unknown amounts of sanitary sewage from the 115-F Gas Recirculation Building and may have potentially contained hazardous and radioactive contamination. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

  20. Data Package for Past and Current Groundwater Flow and Contamination beneath Single-Shell Tank Waste Management Areas

    Energy Technology Data Exchange (ETDEWEB)

    Horton, Duane G.

    2007-03-16

    This appendix summarizes historic and recent groundwater data collected from the uppermost aquifer beneath the 200 East and 200 West Areas. Although the area of interest is the Hanford Site Central Plateau, most of the information discussed in this appendix is at the scale of individual single-shell tank waste management areas. This is because the geologic, and thus the hydraulic, properties and the geochemical properties (i.e., groundwater composition) are different in different parts of the Central Plateau.

  1. Final evaluation & test report for the standard waste box (docket 01-53-7A) type A packaging

    Energy Technology Data Exchange (ETDEWEB)

    KELLY, D L

    2001-10-15

    This report documents the U.S. Department of Transportation Specification 7A Type A compliance test and evaluation results of the Standard Waste Box. Testing and evaluation activities documented herein are on behalf of the U.S. Department of Energy-Headquarters, Office of Safety, Health and Security (EM-5), Germantown, Maryland. Duratek Federal Services, Inc., Northwest Operations performed an evaluation of the changes as documented herein under Docket 01-53-7A.

  2. Investigation of the combustion characteristics of Zonguldak bituminous coal using DTA and DTG

    Energy Technology Data Exchange (ETDEWEB)

    Haykiri-Acma, H.; Yaman, S.; Kucukbayrak, S.; Okutan, H. [Istanbul Technical University, Istanbul (Turkey). Dept. of Chemical Engineering

    2006-06-21

    Combustion characteristics of coking, semicoking, and noncoking Turkish bituminous coal samples from Zonguldak basin were investigated applying differential thermal analysis (DTA) and differential thermogravimetry (DTG) techniques. Results were compared with that of the coke from Zonguldak bituminous coal, a Turkish lignite sample from Soma, and a Siberian bituminous coal sample. The thermal data from both techniques showed some differences depending on the proximate analyses of the samples. Noncombustible components of the volatile matter led to important changes in thermal behavior. The data front both methods were, evaluated jointly, and some thermal properties were interpreted considering these methods in a complementary combination.

  3. Remaining Sites Verification Package for the 100-B-23, 100-B/C Area Surface Debris. Attachment to Waste Site, Waste Site Reclassification Form 2008-027

    International Nuclear Information System (INIS)

    The 100-B-23, 100-B/C Surface Debris, waste consisted of multiple locations of surface debris and chemical stains that were identified during an Orphan Site Evaluation of the 100-B/C Area. Evaluation of the collected information for the surface debris features yielded four generic waste groupings: asbestos-containing material, lead debris, oil and oil filters, and treated wood. Focused verification sampling was performed concurrently with remediation. Site remediation was accomplished by selective removal of the suspect hazardous items and potentially impacted soils. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River

  4. Comparative study of thermal properties of bio-coal from aromatic spent with low rank sub-bituminous coals.

    Science.gov (United States)

    Yadav, Vineet; Baruah, B P; Khare, Puja

    2013-06-01

    In present investigation, biocoal samples were prepared from aromatic plant waste of two perennial grasses, i.e. Cymbopogon flexuosus (lemongrass) and Vetiveria zizanioides (khus) after oil extraction, root of Rosa damascene (rose), bark of Eucalyptus citriodora. These biocoals were characterized by proximate, ultimate, metal, thermogravimetric analysis (TGA), Fourier Transform Infra Red (FTIR) spectroscopy and ash analyses. Activation energies, initial temperature of devolatilization, maximum rate of weight loss (Rmax), fouling, slagging and alkali index were determined on the basis of TGA and ash analysis. These biocoals have good calorific values. There is possibility of slagging and fouling in combustion system but it is not severe. Owing to their similar fuel properties as high sulphur sub-bituminous coal, they can be good candidates for co-firing. Blending of these biocoals with high sulphur coals will serve dual purpose as (i) alternate fuel, and (ii) reduction in SO2 emission. PMID:23603187

  5. Effect of Cement on Properties of Over-Burnt Brick Bituminous Concrete Mixes

    Science.gov (United States)

    Sarkar, Dipankar; Pal, Manish

    2016-06-01

    The present investigation is carried out to propose the use of cement coated over burnt brick aggregate in the preparation of bituminous concrete mix. The effect of cement on various mechanical properties such as Marshall stability, flow, Marshall quotient (stability to flow ratio), indirect tensile strength, stripping, rutting and fatigue life of bituminous concrete overlay has been evaluated. In this study, different cement percentages such as 2, 3, 4 and 5 % by weight of aggregate have been mixed with Over Burnt Brick Aggregate (OBBA). The laboratory results indicate that bituminous concrete prepared by 4 % cement coated OBBA gives the highest Marshall stability. The bituminous concrete mix with 4 % cement shows considerable improvement in various mechanical properties of the mix compared to the plain OBBA concrete mix.

  6. Management of the radioactive waste treatment facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kil Jeong; An, Sum Jin; Lee, Kang Mu; Jeong, Kyeong Hwan; Lee, Young Hee; Sohn, Jong Sik; Bae, Sang Min; Kang, Kwon Ho; Yim, Kil Sung; Ui, Keum San; Kim, Tae Kuk; Sohn, Young Jun; You, Young Keol; Park, Young Yoong; Yoon, Bong Keun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-12-01

    The radioactive wastes generated in Korea Atomic Energy Research Institute (KAERI) in 1993 are about 107 m{sup 3} of liquid waste and 169 drums of solid waste. Liquid waste is treated by the evaporation process, the bituminization process, and the solar evaporation process. The solid waste is treated by the compaction process and the cementation process. The radioactive wastes treated in 1993 are about 194 m{sup 3} of liquid waste and 31 drums of solid waste, respectively. 28 tabs., 12 figs. (Author) .new.

  7. Specific transport and storage solutions: Waste management facing current and future stakes of the nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Deniau, Helene; Gagner, Laurent; Gendreau, Francoise; Presta, Anne [COGEMA LOGISTICS - AREVA Group, 1 rue des herons, Montigny le bretonneux 78180 (France)

    2006-07-01

    development process; - Current solutions proposed by COGEMA LOGISTICS; - Transport of Low Level and Very Low Level waste (LLW and VLLW); - Shipment of large volumes; - Non-reusable transport packaging; - Packaging transport permit the storage in final disposal area; - A steep increase of transport flows in the next 10 years; - Future challenges, waste from dismantling operations; - Transport packaging of IP 2 type and IP 2 package; - Some examples of standardized packages; - DV78 Package; - CC102 over pack; - A solution for liquid waste from NPPs, TNTMCIEL; - Transport of Intermediate Level Waste (ILW) and High Level Waste (HLW); - Alpha waste (or TRU waste); - TNTMGEMINI; - RD 26 package; - Bituminized waste (ILW): the TNTM833; - Technological waste in CBFC'2 (ILW): the TNTM837; - Universal Canisters for Vitrified waste (HLW) and Compacted waste (ILW); - Vitrified waste (HLW); - Transport solutions for CSD-V canisters: the TNTM28 VT; - Dual purpose solutions for CSD-V canisters, the TS 28; - CSD-V programme of return; - Compacted waste (ILW): the TNTM843; - Some future stakes and new developments. The following conclusions completes the paper. COGEMA LOGISTICS experience in designing packages and transporting nuclear materials is part of the COGEMA / AREVA waste management policy, which aims to waste volume reduction through waste sorting and packaging in universal canisters or standard waste containers. For the design and manufacture of special shipping casks, as for nuclear transportation and storage, COGEMA LOGISTICS tailors its solutions to its international customers' requirements. Personnel safety, transportation safety and environmental protection are put first. These reasons have lead COGEMA LOGISTICS to become the world's leading designer of nuclear fuel packages, with unique expertise in designing packages for back-end transportation and storage of nuclear materials. Beyond being a transport company and packaging designer, COGEMA LOGISTICS is a

  8. MEMS packaging

    CERN Document Server

    Hsu , Tai-Ran

    2004-01-01

    MEMS Packaging discusses the prevalent practices and enabling techniques in assembly, packaging and testing of microelectromechanical systems (MEMS). The entire spectrum of assembly, packaging and testing of MEMS and microsystems, from essential enabling technologies to applications in key industries of life sciences, telecommunications and aerospace engineering is covered. Other topics included are bonding and sealing of microcomponents, process flow of MEMS and microsystems packaging, automated microassembly, and testing and design for testing.The Institution of Engineering and Technology is

  9. Remaining Sites Verification Package for the 126-F-2, 183-F Clearwells, Waste Site Reclassification Form 2006-017

    Energy Technology Data Exchange (ETDEWEB)

    R. A. Carlson

    2006-05-04

    The 126-F-2 site is the clearwell facility formerly used as part of the reactor cooling water treatment at the 183-F facility. During demolition operations in the 1970s, potentially contaminated debris was disposed in the eastern clearwell structure. The site has been remediated by removing all debris in the clearwell structure to the Environmental Restoration Disposal Facility. The results of radiological surveys and visual inspection of the remediated clearwell structure show neither residual contamination nor the potential for contaminant migration beyond the clearwell boundaries. The results of verification sampling at the remediation waste staging area demonstrated that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also showed that residual contaminant concentrations are protective of groundwater and the Columbia River.

  10. Natural gas storage with activated carbon from a bituminous coal

    Science.gov (United States)

    Sun, Jielun; Rood, M.J.; Rostam-Abadi, M.; Lizzio, A.A.

    1996-01-01

    Granular activated carbons ( -20 + 100 mesh; 0.149-0.84 mm) were produced by physical activation and chemical activation with KOH from an Illinois bituminous coal (IBC-106) for natural gas storage. The products were characterized by BET surface area, micropore volume, bulk density, and methane adsorption capacities. Volumetric methane adsorption capacities (Vm/Vs) of some of the granular carbons produced by physical activation are about 70 cm3/cm3 which is comparable to that of BPL, a commercial activated carbon. Vm/Vs values above 100 cm3/cm3 are obtainable by grinding the granular products to - 325 mesh (adsorption capacity increases with increasing pore surface area and micropore volume when normalizing with respect to sample bulk volume. Compared with steam-activated carbons, granular carbons produced by KOH activation have higher micropore volume and higher methane adsorption capacities (g/g). Their volumetric methane adsorption capacities are lower due to their lower bulk densities. Copyright ?? 1996 Elsevier Science Ltd.

  11. Gasification of high ash, high ash fusion temperature bituminous coals

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Guohai; Vimalchand, Pannalal; Peng, WanWang

    2015-11-13

    This invention relates to gasification of high ash bituminous coals that have high ash fusion temperatures. The ash content can be in 15 to 45 weight percent range and ash fusion temperatures can be in 1150.degree. C. to 1500.degree. C. range as well as in excess of 1500.degree. C. In a preferred embodiment, such coals are dealt with a two stage gasification process--a relatively low temperature primary gasification step in a circulating fluidized bed transport gasifier followed by a high temperature partial oxidation step of residual char carbon and small quantities of tar. The system to process such coals further includes an internally circulating fluidized bed to effectively cool the high temperature syngas with the aid of an inert media and without the syngas contacting the heat transfer surfaces. A cyclone downstream of the syngas cooler, operating at relatively low temperatures, effectively reduces loading to a dust filtration unit. Nearly dust- and tar-free syngas for chemicals production or power generation and with over 90%, and preferably over about 98%, overall carbon conversion can be achieved with the preferred process, apparatus and methods outlined in this invention.

  12. The hydrogasification of lignite and sub-bituminous coals

    Science.gov (United States)

    Bhatt, B.; Fallon, P. T.; Steinberg, M.

    1981-02-01

    A North Dakota lignite and a New Mexico sub-bituminous coal have been hydrogenated at up to 900°C and 2500 psi hydrogen pressure. Yields of gaseous hydrocarbons and aromatic liquids have been studied as a function of temperature, pressure, residence time, feed rates and H2/coal ratio. Coal feed rates in excess of 10 lb/hr have been achieved in the 1 in. I. D.×8 ft reactor and methane concentration as high as 55% have been observed. A four-step reaction model was developed for the production and decomposition of the hydrocarbon products. A single object function formulated from the weighted errors for the four dependent process, variables, CH4, C2H6, BTX, and oil yields, was minimized using a program containing three independent iterative techniques. The results of the nonlinear regression analysis for lignite show that a first-order chemical reaction model with respect to C conversion satisfactorily describes the dilute phase hydrogenation. The activation energy for the initial products formation was estimated to be 42,700 cal/gmole and the power of hydrogen partial pressure was found to be +0.14. The overall correlation coefficient was 0.83. The mechanism, the rate expressions, and the design curves developed can be used for scale-up and reactor design.

  13. Corrosion studies and mechanical tests on metallic materials for the design of packagings for vitrified high level wastes (HLW)

    International Nuclear Information System (INIS)

    Selective corrosion studies and mechanical tests were performed on various metallic materials in order to find out appropriate materials for the design of HLW packagings serving as a barrier in the repository. Besides the Cr-Ni steels to be used as the canister material according to the previous concept, three nickel-base alloys were investigated as well as a mild steel and the materials Ti 99.8-Pd and the steel X 1 CrMoTi 182 (ELA-ferrite). Since within the framework of accident studies for a repository in a rock salt, salt solutions are being considered as a potential corrosion medium, the corrosion tests were made under selecting conditions in a quinary salt solution at 1700C and 1 bar. The mechanical tests have shown that the nickel-base alloys and most of the Cr-Ni steels undergo slight changes of their mechanical properties as a result of welding and annealing treatment, respectively. This provided the evidence for the suitability of these materials as canister materials. Owing to their high susceptibility to local corrosion attacks, above all to stress corrosion, Cr-Ni steels cannot take over the function of a barrier against brines. The materials Inconel 625, Incoloy 825, ELA ferrite and Ti-Pd resisted stress corrosion, but these materials are susceptible to pitting corrosion. High resistance to local corrosion attacks was exhibited by the materials Hastelloy C 4 and mild steel. After a period of testing of 432 days Hastelloy C 4 showed a uniform corrosion rate of less than 0.03 mm/a. The corrosion rate for mild steel 1.0566 was less than 0.25 mm/a in case of the initial material and less than 0.5 mm/a in case of the annealed material. (orig.)

  14. Remaining Sites Verification Package for the 128-F-2, 100-F Burning Pit Waste Site, Waste Site Reclassification Form 2008-031

    Energy Technology Data Exchange (ETDEWEB)

    J. M. Capron

    2008-12-01

    The 128-F-2 waste site consisted of multiple burn and debris filled pits located directly east of the 107-F Retention Basin and approximately 30.5 m east of the northeast corner of the 100-F Area perimeter road that runs along the riverbank. The burn pits were used for incinerating nonradioactive, combustible materials from 1945 to 1965. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The current site conditions achieve the remedial action objectives and the corresponding remedial action goals established in the Remaining Sites ROD. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  15. Small boiler uses waste coal

    Energy Technology Data Exchange (ETDEWEB)

    Virr, M.J. [Spinheat Ltd. (United States)

    2009-07-15

    Burning coal waste in small boilers at low emissions poses considerable problem. While larger boiler suppliers have successfully installed designs in the 40 to 80 MW range for some years, the author has been developing small automated fluid bed boiler plants for 25 years that can be applied in the range of 10,000 to 140,000 lbs/hr of steam. Development has centered on the use of an internally circulating fluid bed (CFB) boiler, which will burn waste fuels of most types. The boiler is based on the traditional D-shaped watertable boiler, with a new type of combustion chamber that enables a three-to-one turndown to be achieved. The boilers have all the advantages of low emissions of the large fluid boilers while offering a much lower height incorporated into the package boiler concept. Recent tests with a waste coal that had a high nitrogen content of 1.45% demonstrated a NOx emission below the federal limit of 0.6 lbs/mm Btu. Thus a NOx reduction on the order of 85% can be demonstrate by combustion modification alone. Further reductions can be made by using a selective non-catalytic reduction (SNCR) system and sulfur absorption of up to 90% retention is possible. The article describes the operation of a 30,000 lbs/hr boiler at the Fayette Thermal LLC plant. Spinheat has installed three ICFB boilers at a nursing home and a prison, which has been tested on poor-grade anthracite and bituminous coal. 2 figs.

  16. Numerical studies of fluid and heat flow near high-level nuclear waste packages emplaced in partially saturated fractured tuff

    International Nuclear Information System (INIS)

    We have performed modeling studies on the simultaneous transport of heat, liquid water, vapor, and air in partially saturated fractured porous rock. Formation parameters were chosen as representative of the potential repository horizon in the Topopah Spring Unit of the Yucca Mountain tuffs. The presence of fractures makes the transport problem very complex, both in terms of flow geometry and physics. The numerical simulator ''TOUGH'' used for our flow calculations takes into account most of the physical effects which are important in multi-phase fluid and heat flow. It has provisions for handling the extreme non-linearities which arise in phase transitions, component disappearances, and capillary discontinuities at fracture faces. We model a region around an infinite linear string of nuclear waste canisters, taking into account both the discrete fractures and the porous matrix. From an analysis of the results obtained with explicit fractures, we develop equivalent continuum models which can reproduce the temperature, saturation, and pressure variation, and gas and liquid flow rates of the discrete fracture-porous matrix calculations. The equivalent continuum approach makes use of a generalized relative permeability concept to take into account the fracture effects. This results in a substantial simplification of the flow problem which makes larger scale modeling of complicated unsaturated fractured porous systems feasible. Potential applications for regional scale simulations and limitations of the continuum approach are discussed. 35 refs., 14 figs., 4 tabs

  17. Remaining Sites Verification Package for the 100-F-50 Stormwater Runoff Culvert, Waste Site Reclassification Form 2007-001

    Energy Technology Data Exchange (ETDEWEB)

    J. M. Capron

    2008-04-15

    The 100-F-50 waste site, part of the 100-FR-2 Operable Unit, is a steel stormwater runoff culvert that runs between two railroad grades in the south-central portion of the 100-F Area. The culvert exiting the west side of the railroad grade is mostly encased in concrete and surrounded by a concrete stormwater collection depression partially filled with soil and vegetation. The drain pipe exiting the east side of the railroad grade embankment is partially filled with soil and rocks. The 100-F-50 stormwater diversion culvert confirmatory sampling results support a reclassification of this site to no action. The current site conditions achieve the remedial action objectives and corresponding remedial action goals established in the Remaining Sites ROD. The results of confirmatory sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  18. Material Efficiency in Dutch Packaging Policy

    NARCIS (Netherlands)

    Worrell, E.; van Sluisveld, M.A.E.

    2013-01-01

    Packaging materials are one of the largest contributors to municipal solid waste generation. In this paper, we evaluate the material impacts of packaging policy in The Netherlands, focusing on the role of material efficiency (or waste prevention). Since 1991, five different policies have been implem

  19. Conceptual design for HIC transport package

    International Nuclear Information System (INIS)

    Korea Hydro and Nuclear Power Co. Ltd.(KHNP) is developing a HIC(High Integrity Container) transport package to transport a high integrity container from the radioactive waste treatment facility of the nuclear power plant to the interim storage building of radioactive waste. The transport package is designed according to the requirements of the Korean and the IAEA transport regulations for Type B package. The transport package provides containment, radiation shielding, structural integrity and passive heat removal for normal transport and hypothetical accident conditions. This paper describes the conceptual design for the HIC transport package

  20. RH Packaging Program Guidance

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions LLC

    2006-11-07

    The purpose of this program guidance document is to provide the technical requirements for use, operation, inspection, and maintenance of the RH-TRU 72-B Waste Shipping Package and directly related components. This document complies with the requirements as specified in the RH-TRU 72-B Safety Analysis Report for Packaging (SARP), and Nuclear Regulatory Commission (NRC) Certificate of Compliance (C of C) 9212. If there is a conflict between this document and the SARP and/or C of C, the C of C shall govern. The C of C states: "...each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application." It further states: "...each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application." Chapter 9.0 of the SARP tasks the Waste Isolation Pilot Plant (WIPP) Management and Operating (M&O) Contractor with assuring the packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with 10 Code of Federal Regulations (CFR) §71.8, "Deliberate Misconduct." Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the U.S. Department of Energy (DOE) Carlsbad Field Office (CBFO) shall be notified immediately. CBFO will evaluate the issue and notify the NRC if required. In accordance with 10 CFR Part 71, "Packaging and Transportation of Radioactive Material," certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21, "Reporting of Defects and Noncompliance," regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to

  1. RH Packaging Program Guidance

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions LLC

    2008-01-12

    The purpose of this program guidance document is to provide the technical requirements for use, operation, inspection, and maintenance of the RH-TRU 72-B Waste Shipping Package (also known as the "RH-TRU 72-B cask") and directly related components. This document complies with the requirements as specified in the RH-TRU 72-B Safety Analysis Report for Packaging (SARP), and Nuclear Regulatory Commission (NRC) Certificate of Compliance (C of C) 9212. If there is a conflict between this document and the SARP and/or C of C, the C of C shall govern. The C of C states: "...each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application." It further states: "...each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application." Chapter 9.0 of the SARP tasks the Waste Isolation Pilot Plant (WIPP) Management and Operating (M&O) Contractor with assuring the packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) §71.8, "Deliberate Misconduct." Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the U.S. Department of Energy (DOE) Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required.In accordance with 10 CFR Part 71, "Packaging and Transportation of Radioactive Material," certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21, "Reporting of Defects and Noncompliance," regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a

  2. Treatment of concentrated waste for storage

    International Nuclear Information System (INIS)

    The french experience in bituminization of radioactive wastes is described through the successive items, an outline on the industrial realizations is followed by the inventories of the coatable wastes, the constraints to be respected for embedding and quality of bitumen to be used. The technological aspect is described with the example of brennilis, characteristics and properties of coated wastes are given in conclusion. (orig./RW)

  3. 废旧BGA器件二次利用的焊接强度研究%Soldering strength study of waste BGA package during reuse

    Institute of Scientific and Technical Information of China (English)

    张旻澍; 谢安; 卢智铨; 李世玮

    2014-01-01

    BGA器件首先经历2000 h的温循老化以接近废旧状态,再以自重跌落,然后用锡膏印刷的方法完成焊锡球的修复,最后利用Dage4000测试机对不同状态的器件进行焊锡球推拉力测试。结果发现,随着温循老化的时间增加,失效模式逐渐由韧性失效主导转向脆性失效主导。当焊锡球被修复后,失效模式变为混合模式。通过强度对比可以发现,器件的焊接强度随着温循老化时间的增加而降低。当焊锡球被修复后,焊接强度可以恢复到初始状态的80%左右。修复的焊锡球产生的再生金属间化合物可能是造成焊接强度变化的主要原因。%The BGA went through 2 000 h thermal aging to reach the waste package condition, then the waste BGA were broken up and repaired by falling weight and solder paste printing processes. Finally, the solder ball shear&pull tests were performed to measure the soldering strength by using Dage4000 tester. Results show that with the increasing aging time, the dominant failure mode turns from ductile to brittle. After solder joints repairing, the dominant failure mode is mixed modes. From the strength comparison, it is found that with the increasing aging time, the soldering strength decreases. After solder joints repairing, the soldering strength can recover to about 80% of original condition. The difference of soldering strength may be induced by the re-born inter metallic compound during the solder joints repairing.

  4. Transuranic contaminated waste form characterization and data base

    Energy Technology Data Exchange (ETDEWEB)

    Kniazewycz, B.G.; McArthur, W.C.

    1980-07-01

    This volume contains appendices A to F. The properties of transuranium (TRU) radionuclides are described. Immobilization of TRU wastes by bituminization, urea-formaldehyde polymers, and cements is discussed. Research programs at DOE facilities engaged in TRU waste characterization and management studies are described.

  5. Transuranic contaminated waste form characterization and data base

    International Nuclear Information System (INIS)

    This volume contains appendices A to F. The properties of transuranium (TRU) radionuclides are described. Immobilization of TRU wastes by bituminization, urea-formaldehyde polymers, and cements is discussed. Research programs at DOE facilities engaged in TRU waste characterization and management studies are described

  6. Effect of thickness variation of bituminous layer in the structural responses of flexible pavements

    Directory of Open Access Journals (Sweden)

    Juliana Pavan Vidotto

    2014-05-01

    Full Text Available In this work, a sensitivity analysis was performed to investigate the effect of thickness variation of bituminous layer in the structural responses of flexible pavements and, therefore in performance. For this, it used traffic data from a weigh station located at Km 28 of the Immigrants Highway, at state of Sao Paulo, whose data collection was realized in 2008. The loads deriving from traffic were applied to the computational program ELSYM5 to obtain structural responses related to damage such as fatigue cracks and rutting. With these values, the damage factors were calculated and then the sensitivity analysis related to the thickness variation of bituminous layer was performed. Through the obtained results was concluded that the variation of thickness of bituminous layer have an influence on structural responses, and that damage factors related to fatigue cracks were more sensitive than those factors related to rutting.

  7. Global market trade policy analysis for petroleum oils and oils obtained from bituminous minerals, crude

    Directory of Open Access Journals (Sweden)

    Bagheri, F.

    2012-01-01

    Full Text Available This article is based on surveying the custom tariffs imposed on the world export market of Petroleum Oils and Oils Obtained from Bituminous Minerals, Crude. We obtained the data according to the most updated available data provided online by UNCTAD and World Bank. The results indicate that none of the 142 countries in the world market of this product have imposed non-tariff trade barriers on the import of Petroleum Oils and Oils Obtained from Bituminous Minerals, Crude. The developed countries and the countries with transition economies are the main world import partners. European Union, United States, China, Japan, South Korea, Canada, Singapore, Taiwan, Thailand, South Africa, Australia, Turkey, Brazil, Sweden and Belarus are the examples and have imposed low custom tariffs on Petroleum Oils and Oils Obtained from Bituminous Minerals, Crude.

  8. Food Packaging Materials

    Science.gov (United States)

    1978-01-01

    The photos show a few of the food products packaged in Alure, a metallized plastic material developed and manufactured by St. Regis Paper Company's Flexible Packaging Division, Dallas, Texas. The material incorporates a metallized film originally developed for space applications. Among the suppliers of the film to St. Regis is King-Seeley Thermos Company, Winchester, Ma'ssachusetts. Initially used by NASA as a signal-bouncing reflective coating for the Echo 1 communications satellite, the film was developed by a company later absorbed by King-Seeley. The metallized film was also used as insulating material for components of a number of other spacecraft. St. Regis developed Alure to meet a multiple packaging material need: good eye appeal, product protection for long periods and the ability to be used successfully on a wide variety of food packaging equipment. When the cost of aluminum foil skyrocketed, packagers sought substitute metallized materials but experiments with a number of them uncovered problems; some were too expensive, some did not adequately protect the product, some were difficult for the machinery to handle. Alure offers a solution. St. Regis created Alure by sandwiching the metallized film between layers of plastics. The resulting laminated metallized material has the superior eye appeal of foil but is less expensive and more easily machined. Alure effectively blocks out light, moisture and oxygen and therefore gives the packaged food long shelf life. A major packaging firm conducted its own tests of the material and confirmed the advantages of machinability and shelf life, adding that it runs faster on machines than materials used in the past and it decreases product waste; the net effect is increased productivity.

  9. CH Packaging Program Guidance

    International Nuclear Information System (INIS)

    The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: ''each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.'' They further state: ''each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.'' Chapter 9.0 of the SARP charges the Waste Isolation Pilot Plant (WIPP) management and operating (M and O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) 71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required.

  10. Food Packaging for Sustainable Development

    OpenAIRE

    Williams, Helén

    2011-01-01

    Packaging has been on the environmental agenda for decades. It has been discussed and debated within the society mainly as an environmental problem. Production, distribution and consumption of food and drinks contribute significant to the environmental impact. However, consumers in the EU waste about 20% of the food they buy. The function of packaging in reducing the amount of food losses is an important but often neglected environmental issue. This thesis focuses on the functions of packagin...

  11. FAME. The container measuring equipment with extended performance characteristics for optimization of waste package documentation processes for the final storage facility Konrad; FAME. Die FAssMEssanlage mit erweiterten Leistungsmerkmalen zur Optimierung der Prozessablaeufe bei der Erstellung von Abfallgebindedokumentationen fuer das Endlager Konrad

    Energy Technology Data Exchange (ETDEWEB)

    Vallentin, R.; Risch, R. [WTI Wissenschaftlich-Technische Ingenieurberatung GmbH, Juelich (Germany); Buettner, M. [KGG Kernkraftwerk Gundremmingen GmbH (Germany)

    2008-07-01

    The FAME equipment allows significant time and resource savings in waste package documentation for the final storage facility Konrad. The facility is supposed to ensure qualified measurements using automated machine surveillance, quality assurance of waste products after conditioning by respective data acquisition in accordance with the radiation protection regulations, and appropriate documentation.

  12. CH Packaging Program Guidance

    International Nuclear Information System (INIS)

    The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: 'each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.' They further state: 'each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.' Chapter 9.0 of the SARP charges the U.S. Department of Energy (DOE) or the Waste Isolation Pilot Plant (WIPP) management and operating (M and O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) 71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required. In accordance with 10 CFR Part 71, certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21 regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations

  13. CH Packaging Program Guidance

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2008-09-11

    The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: "each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the pplication." They further state: "each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application." Chapter 9.0 of the SARP charges the U.S. Department of Energy (DOE) or the Waste Isolation Pilot Plant (WIPP) management and operating (M&O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) §71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required. In accordance with 10 CFR Part 71, certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21 regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations are

  14. CH Packaging Program Guidance

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2009-06-01

    The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: "each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application." They further state: "each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application." Chapter 9.0 of the SARP charges the U.S. Department of Energy (DOE) or the Waste Isolation Pilot Plant (WIPP) management and operating (M&O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) §71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required. In accordance with 10 CFR Part 71, certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21 regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations are

  15. 8th RILEM International Symposium on Testing and Characterization of Sustainable and Innovative Bituminous Materials

    CERN Document Server

    Partl, Manfred

    2016-01-01

    This work presents the results of RILEM TC 237-SIB (Testing and characterization of sustainable innovative bituminous materials and systems). The papers have been selected for publication after a rigorous peer review process and will be an invaluable source to outline and clarify the main directions of present and future research and standardization for bituminous materials and pavements. The following topics are covered: - Characterization of binder-aggregate interaction - Innovative testing of bituminous binders, additives and modifiers - Durability and aging of asphalt pavements - Mixture design and compaction analysis - Environmentally sustainable materials and technologies - Advances in laboratory characterization of bituminous materials - Modeling of road materials and pavement performance prediction - Field measurement and in-situ characterization - Innovative materials for reinforcement and interlayer systems - Cracking and damage characterization of asphalt pavements - Rec...

  16. Towards an Accurate Stress Dependant Time & Frequency Domain VE Response Model for Bituminous Binders

    NARCIS (Netherlands)

    Woldekidan, M.F.; Huurman, M.; Pronk, A.C.

    2010-01-01

    Linear viscoelastic properties of bituminous binders for short loading times are analyzed using dynamic mechanical analysis methods. Dynamic Shear Rheometer (DSR) test with parallel plate (PP) configuration is widely used for this purpose. Due to the complex stress distribution over the cross-sectio

  17. CH Packaging Program Guidance

    International Nuclear Information System (INIS)

    The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: 'each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.' They further state: 'each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.' Chapter 9.0 of the SARP charges the U.S. Department of Energy (DOE) or the Waste Isolation Pilot Plant| (WIPP) management and operating (M and O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations(CFR) 71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required.In accordance with 10 CFR Part 71, certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21 regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations

  18. CH Packaging Program Guidance

    International Nuclear Information System (INIS)

    The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: 'each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.' They further state: 'each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.' Chapter 9.0 of the SARP charges the U.S. Department of Energy (DOE) or the Waste Isolation Pilot Plant (WIPP) management and operating (M and O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) 71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required.In accordance with 10 CFR Part 71, certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21 regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations

  19. CH Packaging Program Guidance

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2006-04-25

    The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package TransporterModel II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: "each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application." They further state: "each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application." Chapter 9.0 of the SARP charges the U.S. Department of Energy (DOE) or the Waste Isolation Pilot Plant| (WIPP) management and operating (M&O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations(CFR) §71.8. Any time a user suspects or has indications that the conditions ofapproval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required.In accordance with 10 CFR Part 71, certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21 regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations are

  20. CH Packaging Program Guidance

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2007-12-13

    The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: "each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application." They further state: "each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application." Chapter 9.0 of the SARP charges the U.S. Department of Energy (DOE) or the Waste Isolation Pilot Plant (WIPP) management and operating (M&O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) §71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required.In accordance with 10 CFR Part 71, certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21 regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations are