WorldWideScience

Sample records for nonradiological waste resulting

  1. Idaho National Engineering Laboratory nonradiological waste management information for 1994 and record to date

    International Nuclear Information System (INIS)

    French, D.L.; Lisee, D.J.; Taylor, K.A.

    1995-08-01

    This document provides detailed data and graphics on airborne and liquid effluent releases, fuel oil and coal consumption, water usage, and hazardous and mixed waste generated for calendar year 1994. This report summarizes industrial waste data records compiled since 1971 for the Idaho National Engineering Laboratory (INEL). The data presented are from the INEL Nonradiological Waste Management Information System

  2. Idaho National Engineering Laboratory Nonradiological Waste Management Information for 1993 and record to date

    International Nuclear Information System (INIS)

    Sims, A.M.; Taylor, K.A.

    1994-08-01

    This document provides detailed data and graphics on airborne and liquid effluent releases, fuel oil and coal consumption, water usage, and hazardous and mixed waste generated for calendar year 1993. This report summarizes industrial waste data records compiled since 1971 for the Idaho National Engineering Laboratory (INEL). The data presented are from the INEL Nonradiological Waste Management Information System

  3. Process waste treatment system upgrades: Clarifier startup at the nonradiological wastewater treatment plant

    International Nuclear Information System (INIS)

    Lucero, A.J.; McTaggart, D.R.; Van Essen, D.C.; Kent, T.E.; West, G.D.; Taylor, P.A.

    1998-07-01

    The Waste Management Operations Division at Oak Ridge National Laboratory recently modified the design of a reactor/clarifier at the Nonradiological Wastewater Treatment Plant, which is now referred to as the Process Waste Treatment Complex--Building 3608, to replace the sludge-blanket softener/clarifier at the Process Waste Treatment Plant, now referred to as the Process Waste Treatment Complex-Building 3544 (PWTC-3544). This work was conducted because periodic hydraulic overloads caused poor water-softening performance in the PWTC-3544 softener, which was detrimental to the performance and operating costs of downstream ion-exchange operations. Over a 2-month time frame, the modified reactor/clarifier was tested with nonradiological wastewater and then with radioactive wastewater to optimize softening performance. Based on performance to date, the new system has operated more effectively than the former one, with reduced employee radiological exposure, less downtime, lower costs, and improved effluent quality

  4. Nonradiological groundwater quality at low-level radioactive waste disposal sites

    International Nuclear Information System (INIS)

    Goode, D.J.

    1986-04-01

    The NRC is investigating appropriate regulatory options for disposal of low-level radioactive waste containing nonradiological hazardous constituents, as defined by EPA regulations. Standard EPA/RCRA procedures to determine hazardous organics, metals, indicator parameters, and general water quality are applied to samples from groundwater monitoring wells at two commercial low-level radioactive waste disposal sites. At the Sheffield, IL site (nonoperating), several typical organic solvents are identified in elevated concentrations in onsite wells and in an offsite area exhibiting elevated tritium concentrations. At the Barnwell, SC site (operating), only very low concentrations of three organics are found in wells adjacent to disposal units. Hydrocarbons associated with petroleum products are detected at both sites. Hazardous constituents associated with previosuly identified major LLW mixed waste streams, toluene, xylene, chromium, and lead, are at or below detection limits or at background levels in all samples. Review of previously collected data also supports the conclusion that organic solvents are the primary nonradiological contaminants associated with LLW disposal

  5. Non-radiological air quality modeling for the high-level waste tank closure environmental impact statement

    International Nuclear Information System (INIS)

    Hunter, C.H.

    2000-01-01

    Dispersion modeling of potential non-radiological air emissions associated with the proposed closure of high-level waste (HLW) tanks at the Savannah River Site has been completed, as requested (TtNUS, 1999). Estimated maximum ground-level concentrations of applicable regulated air pollutants at the site boundary and at the distance to the co-located onsite worker (640 meters) are summarized. In all cases, the calculated concentrations were much less than regulatory standards

  6. Socio-economic and other non-radiological impacts of the near surface disposal of radioactive waste

    International Nuclear Information System (INIS)

    2002-09-01

    The objective of this report is to introduce, in a generic sense, the elements that could comprise a socio-economic and non-radiological environmental impact assessment. The various social, economic and environmental impacts that could be associated with surface and near surface disposal are discussed through factors that could apply at the local, regional or national level. Impact management is also discussed. The report also introduces concepts to help Member States develop their own approaches to undertaking impact assessment and management. The report is intended to complement IAEA documents on the technology and safety aspects of the near surface disposal of radioactive waste. The scope of this report includes a discussion of a range of social, economic and nonradiological environmental impacts relevant to surface and near surface disposal and illustrations of some impact management measures

  7. A comprehensive inventory of radiological and nonradiological contaminants in waste buried or projected to be buried in the subsurface disposal area of the INEL RWMC during the years 1984-2003, Volume 2

    International Nuclear Information System (INIS)

    1995-05-01

    This is the second volume of this comprehensive report of the inventory of radiological and nonradiological contaminants in waste buried or projected to be buried in the subsurface disposal area of the Idaho National Engineering Laboratory. Appendix B contains a complete printout of contaminant inventory and other information from the CIDRA Database and is presented in volumes 2 and 3 of the report

  8. Evaluation of the non-radiological environmental problems relating to the WIPP

    International Nuclear Information System (INIS)

    Baca, T.E.

    1983-02-01

    The major non-radiological environmental problems addressed are: air pollution, water pollution and sanitary waste, solid waste, domestic drinking water, occupational health and safety and toxic chemicals

  9. A comprehensive inventory of radiological and nonradiological contaminants in waste buried or projected to be buried in the subsurface disposal area of the INEL RWMC during the years 1984-2003, Volume 1

    International Nuclear Information System (INIS)

    1995-05-01

    This report presents a comprehensive inventory of the radiological and nonradiological contaminants in waste buried or projected to be buried from 1984 through 2003 in the Subsurface Disposal Area (SDA) at the Radioactive Waste Management Complex (RWMC) of the Idaho National Engineering Laboratory. The project to compile the inventory is referred to as the recent and projected data task. The inventory was compiled primarily for use in a baseline risk assessment under the Comprehensive Environmental Response, Compensation, and Liability Act. The compiled information may also be useful for environmental remediation activities that might be necessary at the RWMC. The information that was compiled has been entered into a database termed CIDRA-the Contaminant Inventory Database for Risk Assessment. The inventory information was organized according to waste generator and divided into waste streams for each generator. The inventory is based on waste information that was available in facility operating records, technical and programmatic reports, shipping records, and waste generator forecasts. Additional information was obtained by reviewing the plant operations that originally generated the waste, by interviewing personnel formerly employed as operators, and by performing nuclear physics and engineering calculations. In addition to contaminant inventories, information was compiled on the physical and chemical characteristics and the packaging of the 99 waste streams. The inventory information for waste projected to be buried at the SDA in the future was obtained from waste generator forecasts. The completeness of the contaminant inventories was confirmed by comparing them against inventories in previous reports and in other databases, and against the list of contaminants detected in environmental monitoring performed at the RWMC

  10. Non-radiological contaminants from uranium mining and milling at Ranger, Jabiru, Northern Territory, Australia.

    Science.gov (United States)

    Noller, B N

    1991-10-01

    Protection from the hazards from radioactivity is of prime importance in the management of uranium mine and mill wastes. Such wastes also contain non-radiological contaminants (heavy metals, acids and neutralising agents) which give rise to potential long-term health and environmental hazards and short-term hazards to the aquatic ecosystem, e.g. as a result of release of waste water. This study seeks to identify non-radiological contaminants (elements) transferred to waste water at the Ranger uranium mine/mill complex at Jabiru, which are likely to hazardous to the aquatic environment.The two principal sources of contaminants are: (i) ore and waste rock mobilised from mining; and (ii) process reagents used in the milling and mineral extraction process. These substances may or may not already be present in the natural environment but may lead to deleterious effects on the aquatic environment if increased above threshold levels.Rhenium, derived from the ore body, was found to be significantly enriched in waste water from Ranger, indicating its suitability as an indicator element for water originating from the mining and milling process, but only uranium, likewise derived from the ore, and magnesium, manganese and sulfur (as sulfate) from the milling process were found to be significant environmental contaminants.

  11. Treatability studies in support of the nonradiological wastewater treatment project

    Energy Technology Data Exchange (ETDEWEB)

    Begovich, J.M.; Brown, C.H. Jr.; Villiers-Fisher, J.F.; Fowler, V.L.

    1986-07-01

    The Nonradiological Wastewater Treatment Project (NRWTP) will treat nonradiological wastewaters generated at the Oak Ridge National Laboratory (ORNL) to pollutant levels acceptable under restrictions imposed by the effluent limits of best available technology (BAT) regulations of the US Environmental Protection Agency (EPA), according to the goals established by the Clean Water Act. A three-phase treatability study was conducted to resolve many of the uncertainties facing the NRWTP. The first phase consisted of batch simulation of the proposed NRWTP flowsheet in the laboratory. The Phase I results revealed no major problems with the proposed flowsheet. Phase II consisted of more-detailed parametric studies of the flowsheet processes at a bench-scale level in the laboratory. The Phase II results were used to guide the planning and design of the Phase III study, which consisted of flowsheet simulation on a continuous basis using a mini-pilot plant (MPP) facility. This facility is contained within two connected semitrailer vans and an analytical trailer.

  12. Assessment of radiological and non-radiological hazards in the nuclear fuel cycle - The Indian experience

    International Nuclear Information System (INIS)

    Krishnamony, S.; Gopinath, D.V.

    1996-01-01

    Design and operational aspects of nuclear fuel cycle facilities have several features that distinguish them from nuclear power plants. These are related to (i) the nature of operations which are chiefly mining, metallurgical and chemical; (ii) the nature and type of radio-active materials handled, their specific activities and inventories; and (iii) the physical and chemical processes involved and the associated containment provisions. Generally the radioactive materials are present in an already highly dispersible or mobile form, in the form of solutions, slurries and powders, often associated with a wide variety of reactive and corrosive chemicals. There are further marked differences between the front-end and back-end of the fuel cycle. Whereas the front-end is characterized by the presence of large quantities of low specific activity naturally occurring radioactive materials, the back-end is characterized by high specific activities and concentrations of fission products and actinides. Radioactive characteristics of waste arisings are also different in different phases of the nuclear fuel cycle. Potential for internal exposure in the occupational environment is another distinguishing feature as compared with the more common designs of nuclear power reactors. Potential for accidents, their phenomenology and the resulting consequences are also markedly different in fuel cycle operations. The non-radiological hazards in fuel cycle operations are also of significance, since the operations are mostly mining, metallurgical and chemical in nature. These aspects are examined and evaluated in this paper, based on the Indian experience. (author). 12 refs, 10 tabs

  13. Comparison of the distribution of non-radiological and radiological fatal risk in Ontario industries (addendum)

    International Nuclear Information System (INIS)

    Davis, C.K.; Forbes, W.F.; Hayward, L.M.

    1986-09-01

    Occupational limits for exposure to ionizing radiation, in force in Canada, are based on recommendations of international bodies, particularly the International Commission on Radiological Protection (ICRP). To determine whether the ICRP assertions concerning the similarity of the distributions of occupational risk at the higher risk levels (from non-radiation and from radiation work) to Canada a study of the high end of the distributions of non-radiological risk of occupational fatalities in the province of Ontario was performed. For the present study total doses from exposure to sources of ionizing radiation for Ontario workers were converted to relative risk rates to allow direct comparison with the non-radiological results. In addition, absolute values for the radiological risk rates (RRR) were derived. The radiological risk estimates are based on workers who work both from nuclear reactions and from X-rays. The conclusion is made that the radiological and non-radiological risk rate (NRRR) distributions are similar in shape, but the RRR are approximately 1 to 27 percent of the NRRR, depending on the industry concerned

  14. Identification and monitoring of non-radiological carcinogens

    Energy Technology Data Exchange (ETDEWEB)

    Chuaqui, C A; Petkau, A; Greenstock, C L; Brown, C P [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs.

    1995-09-01

    This study examines the feasibility of identifying and monitoring occupational exposures to non-radiological carcinogens in the workplace at Canadian nuclear establishments (Whiteshell Laboratories, Pickering Nuclear Generating Station, Cameco Limited and Canadian General Electric Company Limited). Recent epidemiological studies recommended that potential confounding factors of a non-radiological nature be identified and analyzed, particularly non-radiological carcinogens that may be present in the workplace at nuclear facilities. The feasibility of identifying and measuring occupational exposures to non-radiological carcinogens in Canadian nuclear facilities is examined. Also, the report describes the problem of chemical carcinogens and the mechanisms involved in chemical carcinogenesis; the epidemiology related to the problem, followed by a description of the analytical aspects of detection, monitoring and analysis of carcinogens, as well as a discussion on the regulatory aspects and the regulations in place; and the findings, recommendations and concluding remarks of this study. Several problem areas became apparent as the study proceeded. For example, the classification of a chemical as a human carcinogen is a difficult problem, as is its adequate monitoring and analysis. This situation reflects, in turn, the regulatory aspects in the workplace. A list of chemical carcinogens used industrially at the four Canadian nuclear facilities has been identified. The list includes arsenic, asbestos, benzene, cadmium, beryllium, nickel, polychlorinated biphenyls, lead and trichloroethylene. Several recommendations are made in relation to the need for practical and efficient monitoring methods for chemical carcinogens, the definition of radiation and chemical dose equivalencies, and the classification of human chemical carcinogens, as well as their disposal. (author). 122 refs., 8 tabs., 6 figs.

  15. Identification and monitoring of non-radiological carcinogens

    International Nuclear Information System (INIS)

    Chuaqui, C.A.; Petkau, A.; Greenstock, C.L.; Brown, C.P.

    1995-09-01

    This study examines the feasibility of identifying and monitoring occupational exposures to non-radiological carcinogens in the workplace at Canadian nuclear establishments (Whiteshell Laboratories, Pickering Nuclear Generating Station, Cameco Limited and Canadian General Electric Company Limited). Recent epidemiological studies recommended that potential confounding factors of a non-radiological nature be identified and analyzed, particularly non-radiological carcinogens that may be present in the workplace at nuclear facilities. The feasibility of identifying and measuring occupational exposures to non-radiological carcinogens in Canadian nuclear facilities is examined. Also, the report describes the problem of chemical carcinogens and the mechanisms involved in chemical carcinogenesis; the epidemiology related to the problem, followed by a description of the analytical aspects of detection, monitoring and analysis of carcinogens, as well as a discussion on the regulatory aspects and the regulations in place; and the findings, recommendations and concluding remarks of this study. Several problem areas became apparent as the study proceeded. For example, the classification of a chemical as a human carcinogen is a difficult problem, as is its adequate monitoring and analysis. This situation reflects, in turn, the regulatory aspects in the workplace. A list of chemical carcinogens used industrially at the four Canadian nuclear facilities has been identified. The list includes arsenic, asbestos, benzene, cadmium, beryllium, nickel, polychlorinated biphenyls, lead and trichloroethylene. Several recommendations are made in relation to the need for practical and efficient monitoring methods for chemical carcinogens, the definition of radiation and chemical dose equivalencies, and the classification of human chemical carcinogens, as well as their disposal. (author). 122 refs., 8 tabs., 6 figs

  16. 1994 Environmental monitoring drinking water and nonradiological effluent programs annual report

    International Nuclear Information System (INIS)

    Andersen, B.D.; Brock, T.A.; Meachum, T.R.

    1995-10-01

    EG ampersand G Idaho, Inc., initiated monitoring programs for drinking water in 1988 and for nonradiological parameters and pollutants in liquid effluents in 1985. These programs were initiated for the facilities operated by EG ampersand G Idaho for the US Department of Energy at the Idaho National Engineering Laboratory. On October 1, 1994, Lockheed Idaho Technologies Company (LITCO) replaced EG ampersand G Idaho as the prime contractor at the INEL and assumed responsibility for these programs. Section I discusses the general site characteristics, the analytical laboratories, and sampling methodology general to both programs. Section 2, the Drinking Water Program, tracks the bacteriological, chemical, and radiological parameters required by State and Federal regulations. This section describes the drinking water monitoring activities conducted at 17 LITCO-operated production wells and 11 distribution systems. It also contains all of the drinking water parameters detected and the regulatory limits exceeded during calendar year 1994. In addition, groundwater quality is discussed as it relates to contaminants identified at the wellhead for LITCO production wells. Section 3 discusses the nonradiological liquid effluent monitoring results for 27 liquid effluent streams. These streams are presented with emphasis on calendar year 1994 activities. All parameter measurements and concentrations were below the Resource Conservation and Recovery Act toxic characteristics limits

  17. HARVESTING EMSP RESEARCH RESULTS FOR WASTE CLEANUP

    International Nuclear Information System (INIS)

    Guillen, Donna Post; Nielson, R. Bruce; Phillips, Ann Marie; Lebow, Scott

    2003-01-01

    The extent of environmental contamination created by the nuclear weapons legacy combined with expensive, ineffective waste cleanup strategies at many U.S. Department of Energy (DOE) sites prompted Congress to pass the FY96 Energy and Water Development Appropriations Act, which directed the DOE to: ''provide sufficient attention and resources to longer-term basic science research, which needs to be done to ultimately reduce cleanup costs'', ''develop a program that takes advantage of laboratory and university expertise, and'' ''seek new and innovative cleanup methods to replace current conventional approaches which are often costly and ineffective.'' In response, the DOE initiated the Environmental Management Science Program (EMSP)-a targeted, long-term research program intended to produce solutions to DOE's most pressing environmental problems. EMSP funds basic research to lower cleanup cost and reduce risk to workers, the public, and the environment; direct the nation's scientific infrastructure towards cleanup of contaminated waste sites; and bridge the gap between fundamental research and technology development activities. EMSP research projects are competitively awarded based on the project's scientific, merit coupled with relevance to addressing DOE site needs. This paper describes selected EMSP research projects with long, mid, and short-term deployment potential and discusses the impacts, focus, and results of the research. Results of EMSP research are intended to accelerate cleanup schedules, reduce cost or risk for current baselines, provide alternatives for contingency planning, or provide solutions to problems where no solutions exist

  18. Dossier: management of nuclear wastes. Research, results

    International Nuclear Information System (INIS)

    Anon.

    2001-01-01

    The researches carried out since many years on nuclear wastes have led to two main ways of management: the long-term conditioning of radio-elements and their advanced separation. The French atomic energy commission (CEA) has chosen to take up also the transmutation challenge, a way to transform long-living radioactive wastes into short-living radioactive wastes or stable compounds. The transmutation programs are based both on simulation and experiments with a huge international collaboration. This dossier presents in a digest way the research activity carried out on nuclear wastes processing and management at the CEA. (J.S.)

  19. HARVESTING EMSP RESEARCH RESULTS FOR WASTE CLEANUP

    Energy Technology Data Exchange (ETDEWEB)

    Guillen, Donna Post; Nielson, R. Bruce; Phillips, Ann Marie; Lebow, Scott

    2003-02-27

    The extent of environmental contamination created by the nuclear weapons legacy combined with expensive, ineffective waste cleanup strategies at many U.S. Department of Energy (DOE) sites prompted Congress to pass the FY96 Energy and Water Development Appropriations Act, which directed the DOE to: ''provide sufficient attention and resources to longer-term basic science research, which needs to be done to ultimately reduce cleanup costs'', ''develop a program that takes advantage of laboratory and university expertise, and'' ''seek new and innovative cleanup methods to replace current conventional approaches which are often costly and ineffective.'' In response, the DOE initiated the Environmental Management Science Program (EMSP)-a targeted, long-term research program intended to produce solutions to DOE's most pressing environmental problems. EMSP funds basic research to lower cleanup cost and reduce risk to workers, the public, and the environment; direct the nation's scientific infrastructure towards cleanup of contaminated waste sites; and bridge the gap between fundamental research and technology development activities. EMSP research projects are competitively awarded based on the project's scientific, merit coupled with relevance to addressing DOE site needs. This paper describes selected EMSP research projects with long, mid, and short-term deployment potential and discusses the impacts, focus, and results of the research. Results of EMSP research are intended to accelerate cleanup schedules, reduce cost or risk for current baselines, provide alternatives for contingency planning, or provide solutions to problems where no solutions exist.

  20. Nuclear wastes future: results and forecasting

    International Nuclear Information System (INIS)

    2001-01-01

    Since many years, the CEA is greatly involved in the research programs on the long-dated management of radioactive wastes. This document presents the CEA development in the following domains: the spent fuel processing, the high separation process, the environmental behavior of wastes packages, the glass performance and the apatites behavior in the new matrix. (A.L.B.)

  1. ORNL necessary and sufficient standards for environment, safety, and health. Final report of the Identification Team for other industrial, radiological, and non-radiological hazard facilities

    International Nuclear Information System (INIS)

    1998-07-01

    This Necessary and Sufficient (N and S) set of standards is for Other Industrial, Radiological, and Non-Radiological Hazard Facilities at Oak Ridge National Laboratory (ORNL). These facility classifications are based on a laboratory-wide approach to classify facilities by hazard category. An analysis of the hazards associated with the facilities at ORNL was conducted in 1993. To identify standards appropriate for these Other Industrial, Radiological, and Non-Radiological Hazard Facilities, the activities conducted in these facilities were assessed, and the hazards associated with the activities were identified. A preliminary hazards list was distributed to all ORNL organizations. The hazards identified in prior hazard analyses are contained in the list, and a category of other was provided in each general hazard area. A workshop to assist organizations in properly completing the list was held. Completed hazard screening lists were compiled for each ORNL division, and a master list was compiled for all Other Industrial, Radiological Hazard, and Non-Radiological facilities and activities. The master list was compared against the results of prior hazard analyses by research and development and environment, safety, and health personnel to ensure completeness. This list, which served as a basis for identifying applicable environment, safety, and health standards, appears in Appendix A

  2. ORNL necessary and sufficient standards for environment, safety, and health. Final report of the Identification Team for other industrial, radiological, and non-radiological hazard facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-07-01

    This Necessary and Sufficient (N and S) set of standards is for Other Industrial, Radiological, and Non-Radiological Hazard Facilities at Oak Ridge National Laboratory (ORNL). These facility classifications are based on a laboratory-wide approach to classify facilities by hazard category. An analysis of the hazards associated with the facilities at ORNL was conducted in 1993. To identify standards appropriate for these Other Industrial, Radiological, and Non-Radiological Hazard Facilities, the activities conducted in these facilities were assessed, and the hazards associated with the activities were identified. A preliminary hazards list was distributed to all ORNL organizations. The hazards identified in prior hazard analyses are contained in the list, and a category of other was provided in each general hazard area. A workshop to assist organizations in properly completing the list was held. Completed hazard screening lists were compiled for each ORNL division, and a master list was compiled for all Other Industrial, Radiological Hazard, and Non-Radiological facilities and activities. The master list was compared against the results of prior hazard analyses by research and development and environment, safety, and health personnel to ensure completeness. This list, which served as a basis for identifying applicable environment, safety, and health standards, appears in Appendix A.

  3. Assessment of alternatives for management of ORNL retrievable transuranic waste. Nuclear Waste Program: transuranic waste (Activity No. AR 05 15 15 0; ONL-WT04)

    Energy Technology Data Exchange (ETDEWEB)

    1980-10-01

    Since 1970, solid waste with TRU or U-233 contamination in excess of 10 ..mu..Ci per kilogram of waste has been stored in a retrievable fashion at ORNL, such as in ss drums, concrete casks, and ss-lined wells. This report describes the results of a study performed to identify and evaluate alternatives for management of this waste and of the additional waste projected to be stored through 1995. The study was limited to consideration of the following basic strategies: Strategy 1: Leave waste in place as is; Strategy 2: Improve waste confinement; and Strategy 3: Retrieve waste and process for shipment to a Federal repository. Seven alternatives were identified and evaluated, one each for Strategies 1 and 2 and five for Strategy 3. Each alternative was evaluated from the standpoint of technical feasibility, cost, radiological risk and impact, regulatory factors and nonradiological environmental impact.

  4. Assessment of alternatives for management of ORNL retrievable transuranic waste. Nuclear Waste Program: transuranic waste (Activity No. AR 05 15 15 0; ONL-WT04)

    International Nuclear Information System (INIS)

    1980-10-01

    Since 1970, solid waste with TRU or U-233 contamination in excess of 10 μCi per kilogram of waste has been stored in a retrievable fashion at ORNL, such as in ss drums, concrete casks, and ss-lined wells. This report describes the results of a study performed to identify and evaluate alternatives for management of this waste and of the additional waste projected to be stored through 1995. The study was limited to consideration of the following basic strategies: Strategy 1: Leave waste in place as is; Strategy 2: Improve waste confinement; and Strategy 3: Retrieve waste and process for shipment to a Federal repository. Seven alternatives were identified and evaluated, one each for Strategies 1 and 2 and five for Strategy 3. Each alternative was evaluated from the standpoint of technical feasibility, cost, radiological risk and impact, regulatory factors and nonradiological environmental impact

  5. Pathway analysis and exposure assessment: MEPAS modeling for nonradiological chemical contaminants at the Hanford Site

    International Nuclear Information System (INIS)

    Blanton, M.L.; Dirkes, R.; Buck, J.; Cooper, A.; Castieton, K.; Glantz, C.

    1995-01-01

    A Chemical Pathway Analysis and Exposure Assessment was performed by the Surface Environmental Surveillance Project (SESP). The SESP monitors air, surface water, sediment, agricultural products, vegetation, soil, and wildlife in order to assess onsite of offsite environmental impacts and offsite human health risk at the Hanford Site. The objectives of this study are (1) determine if a nonradiological chemical monitoring program is warranted for the Hanford Site, (2) ensure that the selection of surveillance parameters such as media, sampling location, and analytes are chosen in a manner that is scientifically sound and cost-efficient, and (3) identify specific nonradiological chemicals of concern (COC) for the Hanford Site. The basis for identification of COC for the Hanford Site was an extensive literature review. The model was also used to predict COC concentrations required onsite to achieve an offsite cancer incidence of 1 E-6 and a hazard quotient of 1.0. This study indicated that nonradiological chemical contamination occurring onsite does not pose a significant offsite human health risk. The highest cancer incidence to the offsite maximally exposed individual from COC was from arsenic (1.76E-1 0); the highest hazard quotient was chromium VI (1.48E-04)

  6. Results of field testing of radioactive waste forms using lysimeters

    International Nuclear Information System (INIS)

    McConnell, J.W., Jr.; Rogers, R.D.; Jastrow, J.D.; Wickliff, D.S.

    1992-01-01

    The Field Lysimeter Investigation: Low-Level Waste Data Base Development Program is obtaining informaiton on the performance of radioactive waste in a disposal environment. Waste forms fabricated using ion-exchange resins from EPICOR-II prefilters employed in the cleanup of the Three Mile Island (TMI) Nuclear Power Station are being tested to develop a low-level waste data base and to obtain information on survivability of waste forms in a disposal environment. In this paper, radionuclide releases from waste forms in the first six years of sampling are presented and discussed. Application of lysimeter data to use in performance assessment models is presented. Initial results from use of data in a performance assessment model are discussed

  7. The Radiation Effect to Waste Glass that Resulting of Vitrification

    International Nuclear Information System (INIS)

    Herlan Martono; Aisyah

    2002-01-01

    The high level liquid waste (HLLW) is generated from the first step extraction of the nuclear fuel reprocessing. This waste was contain of few of actinide and many of fission product. The alpha radiation of actinide that contain on the HLLW cause the change the waste glass characteristic. The experiment was conducted by the doping, irradiation and heating of waste glass resulting from vitrification. The alpha radiation cause the change of composition that could be detected from change of waste glass density and mechanical strength. The increasing of alpha radiation dose cause the increasing change of density and mechanical strength, although the change of mechanical strength is not significant. Degree of change of waste glass density also depend on type of waste-glass and reach for saturated point at over of 5x10 24 alpha decay/m 3 . The gamma radiation of fission product that contain on the HLLW can increasing of waste glass temperature that cause the structure change, so devitrification was occur. The devitrification can the increasing of leaching rate. The cumulative of gamma dose rate was not cause the devitrification. (author)

  8. Non-radiological consequences to the aquatic biota and fisheries of the Susquehanna River from the 1979 accident at Three Mile Island Nuclear Station

    International Nuclear Information System (INIS)

    Hickey, C.R. Jr.; Samworth, R.B.

    1979-11-01

    The non-radiological consequences to the aquatic biota and fishes of the Susquehanna River from the March 28, 1979 accident at Three Mile Island Nuclear Station were assessed through the post-accident period of July 1979. Thermal and chemical discharges during the period did not exceed required effluent limitations. Several million gallons of treated industrial waste effluents were released into the river which were not of unusual volumes compared with normal operation and were a very small proportion of the seasonally high river flows. The extent and relative location of the effluent plume were defined and the fisheries known to have been under its immediate influence were identified, including rough, forage, and predator/sport fishery species

  9. A Transparent Framework for guiding Radiological and Non-Radiological Contaminated Land Risk Assessments

    International Nuclear Information System (INIS)

    Lee, Alex; Mathers, Dan

    2003-01-01

    A framework is presented that may be used as a transparent guidance to both radiological and non-radiological risk assessments. This framework has been developed by BNFL, with external consultation, to provide a systematic approach for identifying key system drivers and to guide associated research packages in light of data deficiencies and sources of model uncertainty. The process presented represents an advance on existing working practices yet combines regulator philosophy to produce a robust, comprehensive, cost-effective and transparent work package. It aims at lending added confidence to risk models thereby adding value to the decision process

  10. Principles and practices in managing the wastes resulting from decommissioning

    International Nuclear Information System (INIS)

    Vladescu, Gabriela; Oprescu, Theodor; Niculae, Ortenzia; Stan, Camelia

    2004-01-01

    . The chapter 2.4 presents a proposal for constituting a statistical basis for radioactive materials classification and the associated measuring procedures. The chapter 2.5 illustrates the principles by applying them to classification of solid. liquid and gaseous radioactive materials and their assignation to one of the categories: excluded, excepted, clean or radioactive. The results of this study can be applied in classifying the radioactive wastes produced in Romania in different nuclear activities such as Cernavoda NPP operation, nuclear research, industry, and medicine, decommissioning of different nuclear facilities, etc

  11. EU-CIS joint study project 2. Intervention criteria in CIS, risk assessments and non-radiological factors in decision-making

    Energy Technology Data Exchange (ETDEWEB)

    Hedemann Jensen, P. [Risoe National Lab., Roskilde (Denmark); Demin, V.F. [Russian Reserch Centre `Kurchatov Inst.`, Moscow (Russian Federation); Konstantinov, Y.O. [Research Inst. of Radiation Hygiene, St. Petersburg (Russian Federation); Likhtarev, I.A. [Ukrainian Scientific Centre for Radiation Medicine, Kiev (Ukraine); Rolevich, I.V. [Chernobyl State Commiettee, Minsk (Belarus); Schneider, T. [Centre d`etudes sur l`Evaluation de la Protection dans le domaine Nucleaire, CEPN, Paris (France)

    1996-05-01

    An extensive radiation risk estimation methodology has recently been developed in Russia and used for estimates of risk in exposed populations in the republics of Russia, Belarus and Ukraine. Results based on demographic data for the three republics are presented and compared with risk estimates from the EU risk model ASQRAD. The intervention criteria in the CIS republics have been evolving since the Chernobyl accident. The development of criteria in each of the three republics has been analysed and the CIS-Criteria have been compared to international guidance on intervention. After a nuclear or radiological emergency both radiological and non-radiological protection factors will influence the level of protective actions being introduced. The role of non-radiological protection factors in the overall optimization of health protection is addressed. It is argued that optimization of the overall health protection is not a question of developing radiation radiation protection philosophy to fully include socio-psychological factors. It is rather a question of including these factors - in parallel with the radiological protection factors - in cooperation between radiation protection experts and psychological specialists under the responsibility of the decision maker. (au) 19 tabs., 10 ills., 45 refs.

  12. EU-CIS joint study project 2. Intervention criteria in CIS, risk assessments and non-radiological factors in decision-making

    International Nuclear Information System (INIS)

    Hedemann Jensen, P.; Demin, V.F.; Konstantinov, Y.O.; Likhtarev, I.A.; Rolevich, I.V.; Schneider, T.

    1996-05-01

    An extensive radiation risk estimation methodology has recently been developed in Russia and used for estimates of risk in exposed populations in the republics of Russia, Belarus and Ukraine. Results based on demographic data for the three republics are presented and compared with risk estimates from the EU risk model ASQRAD. The intervention criteria in the CIS republics have been evolving since the Chernobyl accident. The development of criteria in each of the three republics has been analysed and the CIS-Criteria have been compared to international guidance on intervention. After a nuclear or radiological emergency both radiological and non-radiological protection factors will influence the level of protective actions being introduced. The role of non-radiological protection factors in the overall optimization of health protection is addressed. It is argued that optimization of the overall health protection is not a question of developing radiation radiation protection philosophy to fully include socio-psychological factors. It is rather a question of including these factors - in parallel with the radiological protection factors - in cooperation between radiation protection experts and psychological specialists under the responsibility of the decision maker. (au) 19 tabs., 10 ills., 45 refs

  13. Site study plan for utilities and solid waste, Deaf Smith County Site, Texas: Environmental Field Program: Preliminary draft

    International Nuclear Information System (INIS)

    1987-06-01

    This site plan describes utilities and solid waste studies to be conducted during the characterization of the Deaf Smith County, Texas, site for the US Department of Energy's Salt Repository Project. After utilities and solid waste information needs derived from Federal, State, and local statutes and regulations and the project specifications are briefly described, the site study plan describes the study design and rationale, the field data collection procedures and equipment, and data analysis methods and application of results, the data management strategy, the schedule of field activities, the management of the study, and the study's quality assurance program. The field data collection activities are organized into programs to characterize electrical power, natural gas, communication, water, wastewater sludge, nonradiological solid waste, nonradiological hazardous waste, and low-level radiological waste. These programs include details for the collection of project needs, identification of utilities and solid waste disposal contractor capabilities, and verification of the obtained data. Utilities and solid waste field activities will begin approximately at the time of site access. Utilities and solid waste characterization will be completed within the first year of activity. 29 refs., 6 figs., 2 tabs

  14. Management of radiological and non-radiological risks in a decommissioning project

    International Nuclear Information System (INIS)

    Deboodt, Pascal

    2002-01-01

    real commitment of each partner. We think that the ALARA approach is a very good way to provide an adapted language as well as such commitment. Without any doubt, this approach is mainly responsible for the good results we got. Thirdly, as far as the removal of asbestos at BR3 is concerned, it is obvious that the 'radiological approach' has brought some technical improvements to the 'non radiological' approach. Examples can be found in the use of the masks, of in the daily control for potential contamination. But, on the other end, the workers of the BR3 installation are now more aware of the potential existence of other sources of risks and of the rules, which have to be followed in such cases. Working into the nuclear field leads sometimes to a lack of awareness regarding 'industrial risks'. Some questions are still remaining as 'open questions'. Some of these have still been pointed out. How do we have to optimise such operations where more than one 'recognized' risk is involved? How did we cope up to now with such 'interactive' situations? How far do we have to optimise? What's the meaning of 'optimisation' in such cases? These are examples of questions we hope to deal with during the discussions with partners from the radiological and non-radiological fields

  15. Field test results for radioactive waste drum characterization with Waste Inspection Tomography (WIT)

    Energy Technology Data Exchange (ETDEWEB)

    Bernardi, R.T. [Bio-Imaging Research, Inc., Lincolnshire, IL (United States)

    1997-11-01

    This paper summarizes the design, fabrication, factory testing, evaluation and demonstration of waste inspection tomography (WIT). WIT consists of a self-sufficient, mobile semi-trailer for Non-Destructive Evaluation and Non-Destructive Assay (NDE/NDA) characterization of nuclear waste drums using X-ray and gamma-ray tomographic techniques. The 23-month WIT Phase I initial test results include 2 MeV Digital Radiography (DR), Computed Tomography (CT), Anger camera imaging, Single Photon Emission Computed Tomography (SPECT), Gamma-Ray Spectroscopy, Collimated Gamma Scanning (CGS), and Active and Passive Computed Tomography (A&PCT) using a 1.4 mCi source of {sup 166}Ho. These techniques were initially demonstrated on a 55-gallon phantom drum with three simulated waste matrices of combustibles, heterogeneous metals, and cement using check sources of gamma active isotopes. Waste matrix identification, isotopic identification, and attenuation-corrected gamma activity determination were all demonstrated nondestructively and noninvasively. Preliminary field tests results with nuclear waste drums are summarized. WIT has inspected drums with 0 to 20 grams plutonium 239. The minimum measured was 0.131 gram plutonium 239 in cement. 8 figs.

  16. Results from simulated contact-handled transuranic waste experiments at the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Molecke, M.A.; Sorensen, N.R.; Krumhansl, J.L.

    1993-01-01

    We conducted in situ experiments with nonradioactive, contact-handled transuranic (CH TRU) waste drums at the Waste Isolation Pilot Plant (WIPP) facility for about four years. We performed these tests in two rooms in rock salt, at WIPP, with drums surrounded by crushed salt or 70 wt % salt/30 wt % bentonite clay backfills, or partially submerged in a NaCl brine pool. Air and brine temperatures were maintained at ∼40C. These full-scale (210-L drum) experiments provided in situ data on: backfill material moisture-sorption and physical properties in the presence of brine; waste container corrosion adequacy; and, migration of chemical tracers (nonradioactive actinide and fission product simulants) in the near-field vicinity, all as a function of time. Individual drums, backfill, and brine samples were removed periodically for laboratory evaluations. Waste container testing in the presence of brine and brine-moistened backfill materials served as a severe overtest of long-term conditions that could be anticipated in an actual salt waste repository. We also obtained relevant operational-test emplacement and retrieval experience. All test results are intended to support both the acceptance of actual TRU wastes at the WIPP and performance assessment data needs. We provide an overview and technical data summary focusing on the WIPP CH TRU envirorunental overtests involving 174 waste drums in the presence of backfill materials and the brine pool, with posttest laboratory materials analyses of backfill sorbed-moisture content, CH TRU drum corrosion, tracer migration, and associated test observations

  17. Critical Protection Item classification for a waste processing facility at Savannah River Site

    International Nuclear Information System (INIS)

    Ades, M.J.; Garrett, R.J.

    1993-01-01

    This paper describes the methodology for Critical Protection Item (CPI) classification and its application to the Structures, Systems and Components (SSC) of a waste processing facility at the Savannah River Site (SRS). The WSRC methodology for CPI classification includes the evaluation of the radiological and non-radiological consequences resulting from postulated accidents at the waste processing facility and comparison of these consequences with allowable limits. The types of accidents considered include explosions and fire in the facility and postulated accidents due to natural phenomena, including earthquakes, tornadoes, and high velocity straight winds. The radiological analysis results indicate that CPIs are not required at the waste processing facility to mitigate the consequences of radiological release. The non-radiological analysis, however, shows that the Waste Storage Tank (WST) and the dike spill containment structures around the formic acid tanks in the cold chemical feed area and waste treatment area of the facility should be identified as CPIs. Accident mitigation options are provided and discussed

  18. Thermal plasma treatment of cell-phone waste : preliminary result

    Energy Technology Data Exchange (ETDEWEB)

    Ruj, B. [Central Mechanical Engineering Research Inst., Durgapur (India). Thermal Engineering Group; Chang, J.S.; Li, O.L. [McMaster Univ., Hamilton, ON (Canada). Dept. of Engineering Physics; Pietsch, G. [RWTH Aachen Univ., Aachen (Germany)

    2010-07-01

    The cell phone is an indispensable service facilitator, however, the disposal and recycling of cell phones is a major problem. While the potential life span of a mobile phone, excluding batteries, is over 10 years, most of the users upgrade their phones approximately four times during this period. Cell phone waste is significantly more hazardous than many other municipal wastes as it contains thousands of components made of toxic chemicals and metals like lead, cadmium, chromium, mercury, polyvinyl chlorides (PVC), brominated flame retardants, beryllium, antimony and phthalates. Cell phones also use many expensive rare metals. Since cell phones are made up of plastics, metals, ceramics, and trace other substances, primitive recycling or disposal of cell phone waste to landfills and incinerators creates irreversible environmental damage by polluting water and soil, and contaminating air. In order to minimize releases into the environment and threat to human health, the disposal of cell phones needs to be managed in an environmentally friendly way. This paper discussed a safer method of reducing the generation of syngas and hydrocarbons and metal recovery through the treatment of cell phone wastes by a thermal plasma. The presentation discussed the experiment, with particular reference to sample preparation; experimental set-up; and results four samples with different experimental conditions. It was concluded that the plasma treatment of cell phone waste in reduced condition generates gaseous components such as hydrogen, carbon monoxide, and hydrocarbons which are combustible. Therefore, this system is an energy recovery system that contributes to resource conservation and reduction of climate change gases. 5 refs., 2 tabs., 2 figs.

  19. Waste conditioning for tank heel transfer. Preliminary data and results

    International Nuclear Information System (INIS)

    Ebadian, M.A.

    1999-01-01

    This report summarizes the research carried out at Florida International University's Hemispheric Center for Environmental Technology (FIU-HCET) for the fiscal year 1998 (FY98) under the Tank Focus Area (TFA) project ''Waste Conditioning for Tank Slurry Transfer.'' The objective of this project is to determine the effect of chemical and physical properties on the waste conditioning process and transfer. The focus of this research consisted in building a waste conditioning experimental facility to test different slurry simulants under different conditions, and analyzing their chemical and physical properties. This investigation would provide experimental data and analysis results that can make the tank waste conditioning process more efficient, improve the transfer system, and influence future modifications to the waste conditioning and transfer system. A waste conditioning experimental facility was built in order to test slurry simulants. The facility consists of a slurry vessel with several accessories for parameter control and sampling. The vessel also has a lid system with a shaft-mounted propeller connected to an air motor. In addition, a circulation system is connected to the slurry vessel for simulant cooling and heating. Experimental data collection and analysis of the chemical and physical properties of the tank slurry simulants has been emphasized. For this, one waste slurry simulant (Fernald) was developed, and another two simulants (SRS and Hanford) obtained from DOE sites were used. These simulants, composed of water, soluble metal salts, and insoluble solid particles, were used to represent the actual radioactive waste slurries from different DOE sites. The simulants' chemical and physical properties analyzed include density, viscosity, pH, settling rate, and volubility. These analyses were done to samples obtained from different experiments performed at room temperature but different mixing time and strength. The experimental results indicate that the

  20. Site environmental report for Calendar Year 1994 on radiological and nonradiological parameters

    International Nuclear Information System (INIS)

    1995-01-01

    Battelle Memorial Institute's nuclear research facilities are currently being maintained in a surveillance and maintenance (S ampersand M) mode with continual decontamination and decommissioning (D ampersand D) activities being conducted under Department of Energy (DOE) Contract W-7405-ENG-92. These activities are referred to under the Contract as the Battelle Columbus Laboratories Decommissioning Project (BCLDP). Operations referenced in this report are performed in support of S ampersand M and D ampersand D activities. Battelle's King Avenue facility is not considered in this report to the extent that the West Jefferson facility is. The source term at the King Avenue site is a small fraction of the source term at the West Jefferson site. Off site levels of radionuclides that could be attributed to the west Jefferson and King Avenue nuclear operations wereindistinguishable from background levels at specific locations where air, water, and direct radiation measurements were performed. Environmental monitoring continued to demonstrate compliance by Battelle with federal, state and local regulations. Routine, nonradiological activities performed include monitoring liquid effluents and monitoring the ground water system for the West Jefferson North site. Samples of various environmental media including air, water, grass, fish, field and garden crops, sediment and soil were collected from the region surrounding the two sites and analyzed

  1. Non-radiological factors and decision making on the radiological protection of the environment

    International Nuclear Information System (INIS)

    Simcock, A.

    2002-01-01

    'Non-radiological factors' can cover both physical and non-physical issues. As far as physical issues are concerned, the appropriate course is not to forget that radioactive substances have to be considered in the same way as other substances in respect of their non-radioactive properties. 'Damage to amenities' and 'interference with legitimate uses of the sea' are long-standing descriptions of the non-physical aspects of marine pollution and degradation. A framework for a taxonomy of the interests involved in such aspects is suggested, using the three dimensions of the degree of linkage to the marine environment, the nature of the interaction with the marine environment, and the economic nature of the interest concerned. Questions of remoteness also arise. A multi-dimensional analysis of the risks to the interests concerned is suggested. The dimension of 'public response' is particularly significant for the non-physical aspects of marine pollution and degradation. This dimension is complex, being influenced by 'fright factors' and subject to media amplification. These influences can include special local economic circumstances and past experiences. Finally, the process for integrating physical and non-physical factors is examined. Early consideration is recommended of how to achieve a transparent presentation of the issues and the way in which decisions are to be taken. (author)

  2. Simulation for Teaching and Assessment of Nodule Perception on Chest Radiography in Nonradiology Health Care Trainees.

    Science.gov (United States)

    Auffermann, William F; Henry, Travis S; Little, Brent P; Tigges, Stefan; Tridandapani, Srini

    2015-11-01

    Simulation has been used as an educational and assessment tool in several fields, generally involving training of physical skills. To date, simulation has found limited application in teaching and assessment of skills related to image perception and interpretation. The goal of this pilot study was to evaluate the feasibility of simulation as a tool for teaching and assessment of skills related to perception of nodules on chest radiography. This study received an exemption from the institutional review board. Subjects consisted of nonradiology health care trainees. Subjects underwent training and assessment of pulmonary nodule identification skills on chest radiographs at simulated radiology workstations. Subject performance was quantified by changes in area under the localization receiver operating characteristic curve. At the conclusion of the study, all subjects were given a questionnaire with five questions comparing learning at a simulated workstation with training using conventional materials. Statistical significance for questionnaire responses was tested using the Wilcoxon signed rank test. Subjects demonstrated statistically significant improvement in nodule identification after training at a simulated radiology workstation (change in area under the curve, 0.1079; P = .015). Subjects indicated that training on simulated radiology workstations was preferable to conventional training methods for all questions; P values for all questions were less than .01. Simulation may be a useful tool for teaching and assessment of skills related to medical image perception and interpretation. Further study is needed to determine which skills and trainee populations may be most amenable to training and assessment using simulation. Copyright © 2015 American College of Radiology. Published by Elsevier Inc. All rights reserved.

  3. Ceramic waste form qualification using results from witness tubes

    International Nuclear Information System (INIS)

    O'Holleran, T.P.; Johnson, S.G.; Bateman, K.J.

    2002-01-01

    A ceramic waste form has been developed to immobilize the salt waste stream from electrometallurgical treatment of spent nuclear fuel. The ceramic waste form is prepared in a hot isostatic press (HIP). The use of small, easily fabricated HIP capsules called witness tubes has been proposed as a practical way to obtain representative samples of ceramic waste form material for process monitoring, waste form qualification, and archiving. Witness tubes are filled with the same material used to fill the corresponding HIP can, and are HIPed along with the HIP can. Relevant physical, chemical, and performance (leach test) data are analyzed and compared. Differences between witness tube and HIP can materials are shown to be statistically insignificant, demonstrating that witness tubes do provide ceramic waste form material representative of the material in the corresponding HIP can.

  4. PSA results for Hanford high level waste Tank 101-SY

    Energy Technology Data Exchange (ETDEWEB)

    MacFarlane, D.R.; Bott, T.F.; Brown, L.F.; Stack, D.W. [Los Alamos National Lab., NM (United States); Kindinger, J.; Deremer, R.K.; Medhekar, S.R.; Mikschl, T.J. [PLG, Inc., Newport Beach, CA (United States)

    1993-10-01

    Los Alamos National Laboratory has performed a comprehensive probabilistic safety assessment (PSA) that includes consideration of external events for the weapons-production wastes stored in tank number 241-SY-101, commonly known as Tank 101-SY, as configured in December 1992. This tank, which periodically releases (``burps``) a gaseous mixture of hydrogen, nitrous oxide, ammonia, and nitrogen, was analyzed because of public safety concerns associated with the potential for release of radioactive tank contents should this gas mixture be ignited during one of the burps. In an effort to mitigate the burping phenomenon, an experiment is underway in which a large pump has been inserted into the tank to determine if pump-induced circulation of the tank contents will promote a slow, controlled release of the gases. This PSA for Tank 101-SY, which did not consider the pump experiment or future tank-remediation activities, involved three distinct tasks. First, the accident sequence analysis identified and quantified those potential accidents whose consequences result in tank material release. Second, characteristics and release paths for the airborne and liquid radioactive source terms were determined. Finally, the consequences, primarily onsite and offsite potential health effects resulting from radionuclide release, were estimated, and overall risk curves were constructed. An overview of each of these tasks and a summary of the overall results of the analysis are presented in the following sections.

  5. PSA results for Hanford high level waste Tank 101-SY

    International Nuclear Information System (INIS)

    MacFarlane, D.R.; Bott, T.F.; Brown, L.F.; Stack, D.W.; Kindinger, J.; Deremer, R.K.; Medhekar, S.R.; Mikschl, T.J.

    1993-01-01

    Los Alamos National Laboratory has performed a comprehensive probabilistic safety assessment (PSA) that includes consideration of external events for the weapons-production wastes stored in tank number 241-SY-101, commonly known as Tank 101-SY, as configured in December 1992. This tank, which periodically releases (''burps'') a gaseous mixture of hydrogen, nitrous oxide, ammonia, and nitrogen, was analyzed because of public safety concerns associated with the potential for release of radioactive tank contents should this gas mixture be ignited during one of the burps. In an effort to mitigate the burping phenomenon, an experiment is underway in which a large pump has been inserted into the tank to determine if pump-induced circulation of the tank contents will promote a slow, controlled release of the gases. This PSA for Tank 101-SY, which did not consider the pump experiment or future tank-remediation activities, involved three distinct tasks. First, the accident sequence analysis identified and quantified those potential accidents whose consequences result in tank material release. Second, characteristics and release paths for the airborne and liquid radioactive source terms were determined. Finally, the consequences, primarily onsite and offsite potential health effects resulting from radionuclide release, were estimated, and overall risk curves were constructed. An overview of each of these tasks and a summary of the overall results of the analysis are presented in the following sections

  6. Radioactive Waste Management Information for 1992 and record-to-date

    International Nuclear Information System (INIS)

    Litteer, D.L.; Randall, V.C.; Sims, A.M.; Taylor, K.A.

    1993-07-01

    This document provides detailed data and graphics on air borne and liquid effluent releases, fuel oil and coal consumption, water usage, and hazardous and mixed waste generated for calendar year 1992. This report summarizes industrial waste data records compiled since 1971 for the Idaho National Engineering Laboratory (INEL). The data presented are from the INEL Nonradiological Waste Management Information System

  7. The Research Results of Radioactive Waste Management Technology Center Year 1997/1998

    International Nuclear Information System (INIS)

    1998-12-01

    The research results of Radioactive Waste Management Technology Center, National Atomic Energy Agency of Indonesia year 1997/1998 contain paper as form of research results on radioactive waste management related fields. There were included many aspects such as radioactive waste processing, storage, decontamination, decommissioning, safety and environmental aspects. There are 26 papers indexed individually (ID)

  8. The Research Results of Radioactive Waste Management Technology Center Year 1996/1997

    International Nuclear Information System (INIS)

    Budiman, P.; Martono, H.; Las, T.; Lubis, E.; Mulyanto; Wisnubroto, D. S.; Sucipta

    1997-12-01

    The research results of Radioactive Waste Management Technology Center, National Atomic Energy Agency of Indonesia year 1996/1997 contain paper as form of research results on radioactive waste management related fields. There were included many aspects such as radioactive waste processing, storage, decontamination, decommissioning, safety and environmental aspects. There are 24 papers and 12 short communications indexed individually(ID)

  9. Impact of radioactive waste management operations

    International Nuclear Information System (INIS)

    Paine, D.; Rogers, L.E.; Uresk, D.W.

    1977-01-01

    Impact assessment of radioactive waste management operations is considered separately for nonradiological impact on biota, impact on ecosystem structure and function and radiological impact on biota. Localized effects related to facility construction and maintenance activities probably occur but the large expanse of relatively undisturbed surrounding landscape minimizes any overall effects

  10. Chemical risks from nuclear waste repositories

    International Nuclear Information System (INIS)

    Persson, L.

    1988-01-01

    Studies concerning the chemical risks of nuclear waste are reviewed. The radiological toxicity of the material is of primary concern but the potential nonradiological toxicity should not be overlooked as the chemotoxic substances may reach the biosphere from a nuclear waste repository. In the report is concluded that the possible chemotoxic effects of a repository for nuclear waste should be studied as a part of the formal risk assessment of the disposal concept. (author)

  11. Waste Receiving and Processing (WRAP) Weight Scale Analysis Results

    International Nuclear Information System (INIS)

    JOHNSON, M.D.

    2000-01-01

    Fairbanks Weight Scales are used at the Waste Receiving and Processing (WRAP) facility to determine the weight of waste drums as they are received, processed, and shipped. Due to recent problems, discovered during calibration, the WRAP Engineering Department has completed this document which outlines both the investigation of the infeed conveyor scale failure in September of 1999 and recommendations for calibration procedure modifications designed to correct deficiencies in the current procedures

  12. Results of field testing of waste forms using lysimeters

    International Nuclear Information System (INIS)

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

    1988-01-01

    The purpose of the field testing task, using lysimeter arrays, is to expose samples of solidified resin waste to the actual physical, chemical, and microbiological conditions of disposal enviroment. Wastes used in the experiment include a mixture of synthetic organic ion exchange resins and a mixture of organic exchange resins and an inorganic zeolite. Solidification agents used to produce the 4.8-by 7.6-cm cylindrical waste forms used in the study were Portland Type I-II cement and Dow vinyl ester-styrene. Seven of these waste forms were stacked end-to-end and inserted into each lysimeter to provide a 1-L volume. There are 10 lysimeters, 5 at ORNL and 5 at ANL-E. Lysimeters used in this study were designed to be self-contained units which will be disposed at the termination of the 20-year study. Each is a 0.91-by 3.12-m right-circular cylinder divided into an upper compartment, which contains fill material, waste forms, and instrumentation, and an empty lower compartment, which collects leachate. Four lysimeters at each site are filled with soil, while a fifth (used as a control) is filled with inert silica oxide sand. Instrumentation within each lysimeter includes porous cup soil-water samplers and soil moisture/temperature probes. The probes are connected to an on-site data acquisition and storage system (DAS) which also collects data from a field meteorological station located at each site. 9 refs

  13. Preliminary analysis of the cost and risk of transporting nuclear waste to potential candidate commercial repository sites

    International Nuclear Information System (INIS)

    Wilmot, E.L.; Madsen, M.M.; Cashwell, J.W.; Joy, D.S.

    1983-06-01

    This report documents preliminary cost and risk analyses that were performed in support of the Nuclear Waste Terminal Storage (NWTS) program. The analyses compare the costs and hazards of transporting wastes to each of five regions that contain potential candidate nuclear waste repository sites being considered by the NWTS program. These regions are: the Gulf Interior Region, the Permian Basin, the Paradox Basin, Yucca Mountain, and Hanford. Two fuel-cycle scenarios were analyzed: once-through and reprocessing. Transportation was assumed to be either entirely by truck or entirely by rail for each of the scenarios. The results from the risk analyses include those attributable to nonradiological causes and those attributable to the radioactive character of the wastes being transported. 17 references

  14. FY-87 packing fabrication techniques (commercial waste form) results

    International Nuclear Information System (INIS)

    Werry, E.V.; Gates, T.E.; Cabbage, K.S.; Eklund, J.D.

    1988-04-01

    This report covers the investigation of fabrication techniques associated with the development of suitable materials and methods to provide a prefabricated packing for waste packages for the Basalt Waste Isolation Project (BWIP). The principal functions of the packing are to minimize container corrosion during the 300 to 1000 years following repository closure and provide long-term control of the release of radionuclides from the waste package. The investigative work, discussed in this report, was specifically conceived to develop the design criteria for production of full-scale prototypical packing rings. The investigative work included the preparation of procedures, the preparation of fabrication materials, physical properties, and the determination of the engineering properties. The principal activities were the preparation of the materials and the determination of the physical properties. 21 refs., 20 figs., 14 tabs

  15. Waste Management's LNG Truck Fleet: Final Results

    Energy Technology Data Exchange (ETDEWEB)

    Chandler, K. [Battelle (US); Norton, P. [National Renewable Energy Laboratory (US); Clark, N. [West Virginia University (US)

    2001-01-25

    Waste Management, Inc., began operating a fleet of heavy-duty LNG refuse trucks at its Washington, Pennsylvania, facility. The objective of the project was to provide transportation professionals with quantitative, unbiased information on the cost, maintenance, operational, and emissions characteristics of LNG as one alternative to conventional diesel for heavy-duty trucking applications.

  16. Metals partitioning resulting from rotary kiln incineration of hazardous waste

    International Nuclear Information System (INIS)

    Richards, M.K.; Fournier, D.J. Jr.

    1992-01-01

    In response to the need for date on the partitioning of trace metals from hazardous waste incinerators, an extensive series of test was conducted in the summer of 1991 at the USEPA Incineration Research Facility (IRF) in Jefferson, Arkansas. These tests were conducted in the IRF's rotary kiln incinerator system (RKS) equipped with a pilot-scale Calvert Flux-Force/Condensation scrubber as the primary air pollution control system (APCS). The purpose of this test series was to extend the data base on trace metal partitioning and to investigate the effects of variations in incinerator operation on metal partitioning. Another objective was to evaluate the effectiveness of the scrubber for collecting flue gas metals. This series is a continuation of an ongoing IRF research program investigating trace metal partitioning and APCS collection efficiencies. Two previous test series were conducted using the RKS equipped with a venturi/packed-column scrubber and a single-state ionizing wet scrubber. The primary objective of this test series was to determine the fate of six hazardous and four nonhazardous trace metals fed to the RKS in a synthetic, organic-contaminated solid waste matrix. The six hazardous trace metals used were arsenic, barium, cadmium, chromium, mercury, and lead. The four nonhazardous trace metals--bismuth, copper, magnesium, and strontium--were included primarily to supply data to evaluate their potential for use as surrogates. The temperature, waste feed chlorine content, and scrubber pressure drop. The test program objectives were to identify. The partitioning of metals among kiln ash, scrubber liquor, and flue gas. Changes in metal partitioning related to variations in kiln exit gas temperature and waste feed chlorine content. The efficiency of the Calvert scrubber for collecting flue gas metals. The effects of scrubber pressure drop on metal collection efficiencies. 2 figs., 2 tabs

  17. Results after nine years of field testing low-level radioactive waste forms using lysimeters

    International Nuclear Information System (INIS)

    McConnell, J.W. Jr.; Rogers, R.D.; Jastrow, J.D.; Sanford, W.E.; Sullivan, T.M.

    1995-01-01

    The Field Lysimeter Investigations: Low-Level Waste Data Base Development Program is obtaining information on the performance of radioactive waste forms. Ion-exchange resins from a nuclear power station were solidified into waste forms using Portland cement and vinyl ester-styrene. These waste forms are being tested to develop a low-level waste data base and to obtain information on survivability of waste forms in a disposal environment. This paper reviews radionuclide releases from those waste forms in the first 9 years of sampling. Included is a discussion of the recently discovered upward migration of radionuclides. Also, lysimeter data are applied to a performance assessment source term model, and initial results are presented

  18. Waste Minimization Improvements Achieved Through Six Sigma Analysis Result In Significant Cost Savings

    International Nuclear Information System (INIS)

    Mousseau, Jeffrey D.; Jansen, John R.; Janke, David H.; Plowman, Catherine M.

    2003-01-01

    Improved waste minimization practices at the Department of Energy's (DOE) Idaho National Engineering and Environmental Laboratory (INEEL) are leading to a 15% reduction in the generation of hazardous and radioactive waste. Bechtel, BWXT Idaho, LLC (BBWI), the prime management and operations contractor at the INEEL, applied the Six Sigma improvement process to the INEEL Waste Minimization Program to review existing processes and define opportunities for improvement. Our Six Sigma analysis team: composed of an executive champion, process owner, a black belt and yellow belt, and technical and business team members used this statistical based process approach to analyze work processes and produced ten recommendations for improvement. Recommendations ranged from waste generator financial accountability for newly generated waste to enhanced employee recognition programs for waste minimization efforts. These improvements have now been implemented to reduce waste generation rates and are producing positive results

  19. The influence of slaughterhouse waste on fermentative H2 production from food waste: Preliminary results

    International Nuclear Information System (INIS)

    Boni, Maria Rosaria; Sbaffoni, Silvia; Tuccinardi, Letizia

    2013-01-01

    Highlights: • Co-digestion process finalized to bio-H 2 production was tested in batch tests. • Slaughterhouse waste (SHW) and food waste (FW) were co-digested in different proportions. • The presence of SHW affected the H 2 production from FW. • When SHW ranging between 50% and 70% the H 2 production is improved. • SHW percentages above 70%, led to a depletion in H 2 production. - Abstract: The aim of this study was to evaluate the influence of slaughterhouse waste (SHW; essentially the skin, fats, and meat waste of pork, poultry, and beef) in a fermentative co-digestion process for H 2 production from pre-selected organic waste taken from a refectory (food waste [FW]). Batch tests under mesophilic conditions were conducted in stirred reactors filled with different proportions of FW and SHW. The addition of 60% and 70% SHW to a mixture of SHW and FW improved H 2 production compared to that in FW only, reaching H 2 -production yields of 145 and 109 ml gVS 0 -1 , respectively, which are 1.5–2 times higher than that obtained with FW alone. Although the SHW ensured a more stable fermentative process due to its high buffering capacity, a depletion of H 2 production occurred when SHW fraction was higher than 70%. Above this percentage, the formation of foam and aggregated material created non-homogenous conditions of digestion. Additionally, the increasing amount of SHW in the reactors may lead to an accumulation of long chain fatty acids (LCFAs), which are potentially toxic for anaerobic microorganisms and may inhibit the normal evolution of the fermentative process

  20. Soil washing results for mixed waste pond soils at Hanford

    International Nuclear Information System (INIS)

    Gerber, M.A.; Freeman, H.D.; Baker, E.G.; Riemath, W.F.

    1991-01-01

    Soil washing technology was assessed as a means for remediating soil contaminated with mixed wastes primarily composed of heavy metals and radionuclides. The soils at the US Department of Energy's Hanford Site are considered suitable for soil washing because of their relatively low quantities of silt and clay. However, in a limited number of soil washing experiments using soils from different locations in the north pond of the 300 Area, the degree of decontamination achieved for the coarse fraction of the soil varied considerably. Part of this variation appears to be due to the presence of a discrete layer of contaminated sediment found in some of the samples

  1. Alternative Electrochemical Salt Waste Forms, Summary of FY2010 Results

    International Nuclear Information System (INIS)

    Riley, Brian J.; Rieck, Bennett T.; Crum, Jarrod V.; Matyas, Josef; McCloy, John S.; Sundaram, S.K.; Vienna, John D.

    2010-01-01

    In FY2009, PNNL performed scoping studies to qualify two waste form candidates, tellurite (TeO2-based) glasses and halide minerals, for the electrochemical waste stream for further investigation. Both candidates showed promise with acceptable PCT release rates and effective incorporation of the 10% fission product waste stream. Both candidates received reprisal for FY2010 and were further investigated. At the beginning of FY2010, an in-depth literature review kicked off the tellurite glasses study. The review was aimed at ascertaining the state-of-the-art for chemical durability testing and mixed chloride incorporation for tellurite glasses. The literature review led the authors to 4 unique binary and 1 unique ternary systems for further investigation which include TeO2 plus the following: PbO, Al2O3-B2O3, WO3, P2O5, and ZnO. Each system was studied with and without a mixed chloride simulated electrochemical waste stream and the literature review provided the starting points for the baseline compositions as well as starting points for melting temperature, compatible crucible types, etc. The most promising glasses in each system were scaled up in production and were analyzed with the Product Consistency Test, a chemical durability test. Baseline and PCT glasses were analyzed to determine their state, i.e., amorphous, crystalline, phase separated, had undissolved material within the bulk, etc. Conclusions were made as well as the proposed direction for FY2011 plans. Sodalite was successfully synthesized by the sol-gel method. The vast majority of the dried sol-gel consisted of sodalite with small amounts of alumino-silicates and unreacted salt. Upon firing the powders made by sol-gel, the primary phase observed was sodalite with the addition of varying amounts of nepheline, carnegieite, lithium silicate, and lanthanide oxide. The amount of sodalite, nepheline, and carnegieite as well as the bulk density of the fired pellets varied with firing temperature, sol

  2. Alternative Electrochemical Salt Waste Forms, Summary of FY2010 Results

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J.; Rieck, Bennett T.; Crum, Jarrod V.; Matyas, Josef; McCloy, John S.; Sundaram, S. K.; Vienna, John D.

    2010-08-01

    In FY2009, PNNL performed scoping studies to qualify two waste form candidates, tellurite (TeO2-based) glasses and halide minerals, for the electrochemical waste stream for further investigation. Both candidates showed promise with acceptable PCT release rates and effective incorporation of the 10% fission product waste stream. Both candidates received reprisal for FY2010 and were further investigated. At the beginning of FY2010, an in-depth literature review kicked off the tellurite glasses study. The review was aimed at ascertaining the state-of-the-art for chemical durability testing and mixed chloride incorporation for tellurite glasses. The literature review led the authors to 4 unique binary and 1 unique ternary systems for further investigation which include TeO2 plus the following: PbO, Al2O3-B2O3, WO3, P2O5, and ZnO. Each system was studied with and without a mixed chloride simulated electrochemical waste stream and the literature review provided the starting points for the baseline compositions as well as starting points for melting temperature, compatible crucible types, etc. The most promising glasses in each system were scaled up in production and were analyzed with the Product Consistency Test, a chemical durability test. Baseline and PCT glasses were analyzed to determine their state, i.e., amorphous, crystalline, phase separated, had undissolved material within the bulk, etc. Conclusions were made as well as the proposed direction for FY2011 plans. Sodalite was successfully synthesized by the sol-gel method. The vast majority of the dried sol-gel consisted of sodalite with small amounts of alumino-silicates and unreacted salt. Upon firing the powders made by sol-gel, the primary phase observed was sodalite with the addition of varying amounts of nepheline, carnegieite, lithium silicate, and lanthanide oxide. The amount of sodalite, nepheline, and carnegieite as well as the bulk density of the fired pellets varied with firing temperature, sol

  3. Waste glass corrosion modeling: Comparison with experimental results

    International Nuclear Information System (INIS)

    Bourcier, W.L.

    1993-11-01

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

  4. Knowledge, attitudes and behaviour regarding waste management options in Romania: results from a school questionnaire

    Directory of Open Access Journals (Sweden)

    Karin KOLBE

    2014-12-01

    The analysis revealed that knowledge is highly developed in Romania regarding the potential of recycling, while the concepts of waste management technologies are far less known about and understood. Landfill is seen as a problem for human health and the environment. However, recycling behaviour is low - partly as a result of limited possibilities. In general, the treatment hierarchy that is recommended in the "European waste hierarchy" is only partly reflected in students’ attitudes towards waste management options.

  5. The influence of slaughterhouse waste on fermentative H2 production from food waste: preliminary results.

    Science.gov (United States)

    Boni, Maria Rosaria; Sbaffoni, Silvia; Tuccinardi, Letizia

    2013-06-01

    The aim of this study was to evaluate the influence of slaughterhouse waste (SHW; essentially the skin, fats, and meat waste of pork, poultry, and beef) in a fermentative co-digestion process for H2 production from pre-selected organic waste taken from a refectory (food waste [FW]). Batch tests under mesophilic conditions were conducted in stirred reactors filled with different proportions of FW and SHW. The addition of 60% and 70% SHW to a mixture of SHW and FW improved H2 production compared to that in FW only, reaching H2-production yields of 145 and 109 ml g VS 0(-1), respectively, which are 1.5-2 times higher than that obtained with FW alone. Although the SHW ensured a more stable fermentative process due to its high buffering capacity, a depletion of H2 production occurred when SHW fraction was higher than 70%. Above this percentage, the formation of foam and aggregated material created non-homogenous conditions of digestion. Additionally, the increasing amount of SHW in the reactors may lead to an accumulation of long chain fatty acids (LCFAs), which are potentially toxic for anaerobic microorganisms and may inhibit the normal evolution of the fermentative process. Copyright © 2013 Elsevier Ltd. All rights reserved.

  6. Oconee Nuclear Station, Units 1, 2, and 3. Annual operating report for 1976, volume 1: nonradiological environmental surveillance report; II: summary of operations

    International Nuclear Information System (INIS)

    1977-01-01

    The non-radiological environmental surveillance program including thermal and chemical effluents, water quality, fish populations, benthos, fish impingement, gas-bubble disease, and plankton, fish larvae and fish egg entrainment is described. Information is also presented concerning operations, personnel radiation exposures, and fuel examinations

  7. Results Of The Extraction-Scrub-Strip Testing Using An Improved Solvent Formulation And Salt Waste Processing Facility Simulated Waste

    International Nuclear Information System (INIS)

    Peters, T.; Washington, A.; Fink, S.

    2012-01-01

    The Office of Waste Processing, within the Office of Technology Innovation and Development, is funding the development of an enhanced solvent - also known as the next generation solvent (NGS) - for deployment at the Savannah River Site to remove cesium from High Level Waste. The technical effort is a collaborative effort between Oak Ridge National Laboratory (ORNL) and Savannah River National Laboratory (SRNL). As part of the program, the Savannah River National Laboratory (SRNL) has performed a number of Extraction-Scrub-Strip (ESS) tests. These batch contact tests serve as first indicators of the cesium mass transfer solvent performance with actual or simulated waste. The test detailed in this report used simulated Tank 49H material, with the addition of extra potassium. The potassium was added at 1677 mg/L, the maximum projected (i.e., a worst case feed scenario) value for the Salt Waste Processing Facility (SWPF). The results of the test gave favorable results given that the potassium concentration was elevated (1677 mg/L compared to the current 513 mg/L). The cesium distribution value, DCs, for extraction was 57.1. As a comparison, a typical D Cs in an ESS test, using the baseline solvent formulation and the typical waste feed, is ∼15. The Modular Caustic Side Solvent Extraction Unit (MCU) uses the Caustic-Side Solvent Extraction (CSSX) process to remove cesium (Cs) from alkaline waste. This process involves the use of an organic extractant, BoBCalixC6, in an organic matrix to selectively remove cesium from the caustic waste. The organic solvent mixture flows counter-current to the caustic aqueous waste stream within centrifugal contactors. After extracting the cesium, the loaded solvent is stripped of cesium by contact with dilute nitric acid and the cesium concentrate is transferred to the Defense Waste Processing Facility (DWPF), while the organic solvent is cleaned and recycled for further use. The Salt Waste Processing Facility (SWPF), under

  8. Soil washing results for mixed waste pond soils at Hanford

    International Nuclear Information System (INIS)

    Gerber, M.A.

    1991-09-01

    Soil washing technology was assessed as a means for remediating soil contaminated with mixed wastes primarily composed of heavy metals and radionuclides. The soils at the US Department of Energy's Hanford Site are considered suitable for soil washing because of their relatively low quantities of silt and clay. However, in a limited number of soil washing experiments using soils from different locations in the north pond of the 300 Area, the degree of decontamination achieved for the coarse fraction of the soil varied considerably. Part of this variation appears to be due to the presence of a discrete layer of contaminated sediment found in some of the samples. 7 refs., 2 figs., 4 tabs

  9. Long-term management of wastes resulting from dismantling operations. Storing the very low-level activity wastes at Morvilliers

    International Nuclear Information System (INIS)

    Duret, F.; Dutzer, M.; Beranger, V.; Lecoq, P.

    2003-01-01

    Extension of dismantling operations in France in the years to come poses the question of availability of long-term waste facility. Large amount of such wastes will be produced after progressive shutdown of the 58 pressurized water reactors now in operation, not before 2010. However, France is already confronted with dismantling of 9 power reactors (6 of which of gas cooled graphite type), the first reprocessing plant at Marcoule, as well as, dismantling of other installations, for instance the CEA reactors or laboratories. The systems of processing the dismantling waste are not different from those used for wastes resulting from nuclear operations. For the high-level or long-term intermediate level activity disposal the debates must start by 2006, as based on the results of the research conducted according to different provisions of the December 30, 1991 law. These wastes represent however small amounts from the dismantling (around 2000 t for the 9 reactors at shutdown) and they will be stored until a decision will be made. A specific storing system should be implemented by 2008-2010 for the graphite wastes (around 23,000 t) which contain significant amount of long-lived radioelements, although their gross activity is low. But the most significant amount will come from low-level or intermediate-level of short lifetime or from wastes of very low activity. The first category is stored at Storage Center at Aube (CSA), its capacity being of 1,000,000 m 3 of drums. The total volume stored by the end of 2002 amounted 136,500 m 3 with an annual delivering of 12-15,000 m 3 at design rate of 30,000 m 3 /y. This center will be able to absorb the flux increase resulting from dismantling of the decommissioned nuclear installations (around 50,000 t from the dismantling of the 9 power reactor). The Center at Aube can be also adapted for storing wastes of large sizes as for instance the lid of the reactor vessel. According to the French regulation, the wastes produced within a

  10. Radiological protection from radioactive waste management in existing exposure situations resulting from a nuclear accident.

    Science.gov (United States)

    Sugiyama, Daisuke; Hattori, Takatoshi

    2013-01-01

    In environmental remediation after nuclear accidents, radioactive wastes have to be appropriately managed in existing exposure situations with contamination resulting from the emission of radionuclides by such accidents. In this paper, a framework of radiation protection from radioactive waste management in existing exposure situations for application to the practical and reasonable waste management in contaminated areas, referring to related ICRP recommendations was proposed. In the proposed concept, intermediate reference levels for waste management are adopted gradually according to the progress of the reduction in the existing ambient dose in the environment on the basis of the principles of justification and optimisation by taking into account the practicability of the management of radioactive waste and environmental remediation. It is essential to include the participation of relevant stakeholders living in existing exposure situations in the selection of reference levels for the existing ambient dose and waste management.

  11. Radioactive waste disposal by UKAEA establishments during 1980 and associated environmental monitoring results

    International Nuclear Information System (INIS)

    Flew, E.M.

    1981-09-01

    This report gives details of the amounts of solid and liquid radioactive waste disposed of by the principal establishments of the UKAEA during 1980. Waste arising at the UKAEA Nuclear Power Development Laboratories at Windscale and Springfields, which are both situated on British Nuclear Fuels Ltd. (BNFL)-sites, is disposed of by BNFL and included in their authorisations. Discharges to atmosphere of airborne radioactive waste are also included in the report. A summary of the results of the environmental monitoring programmes carried out in connection with the radioactive waste discharges is given. (author)

  12. Radioactive waste disposal by UKAEA establishments during 1978 and associated environmental monitoring results

    International Nuclear Information System (INIS)

    Flew, E.M.

    1979-05-01

    This report gives details of the amounts of solid and liquid radioactive waste disposed of by the principal establishments of the UKAEA during 1978. Waste arising at the UKAEA Nuclear Power Development Laboratories at Windscale and Springfields, which are both situated on British Nuclear Fuels Ltd. (BNFL) sites, is disposed of by BNFL and included in their authorisations. Discharges to atmosphere of airborne radioactive waste are also included in the report. A summary of the results of the environmental monitoring programmes carried out in connection with the radioactive waste discharges is given. (author)

  13. MCNP Modeling Results for Location of Buried TRU Waste Drums

    International Nuclear Information System (INIS)

    Steinman, D K; Schweitzer, J S

    2006-01-01

    In the 1960's, fifty-five gallon drums of TRU waste were buried in shallow pits on remote U.S. Government facilities such as the Idaho National Engineering Laboratory (now split into the Idaho National Laboratory and the Idaho Completion Project [ICP]). Subsequently, it was decided to remove the drums and the material that was in them from the burial pits and send the material to the Waste Isolation Pilot Plant in New Mexico. Several technologies have been tried to locate the drums non-intrusively with enough precision to minimize the chance for material to be spread into the environment. One of these technologies is the placement of steel probe holes in the pits into which wireline logging probes can be lowered to measure properties and concentrations of material surrounding the probe holes for evidence of TRU material. There is also a concern that large quantities of volatile organic compounds (VOC) are also present that would contaminate the environment during removal. In 2001, the Idaho National Engineering and Environmental Laboratory (INEEL) built two pulsed neutron wireline logging tools to measure TRU and VOC around the probe holes. The tools are the Prompt Fission Neutron (PFN) and the Pulsed Neutron Gamma (PNG), respectively. They were tested experimentally in surrogate test holes in 2003. The work reported here estimates the performance of the tools using Monte-Carlo modelling prior to field deployment. A MCNP model was constructed by INEEL personnel. It was modified by the authors to assess the ability of the tools to predict quantitatively the position and concentration of TRU and VOC materials disposed around the probe holes. The model was used to simulate the tools scanning the probe holes vertically in five centimetre increments. A drum was included in the model that could be placed near the probe hole and at other locations out to forty-five centimetres from the probe-hole in five centimetre increments. Scans were performed with no chlorine in the

  14. Creep of ocean sediments resulting from the isolation of radioactive wastes

    International Nuclear Information System (INIS)

    Dawson, P.R.; Chavez, P.F.; Lipkin, J.; Silva, A.J.

    1980-01-01

    Predictive models for the creep of deep ocean sediments resulting from the disposal of radioactive wastes are presented and preliminary observations of a program for evaluation of creep constitutive equation parameters are discussed. The models are used to provide calculated response of sediments under waste disposal conditions

  15. Results of a hospital waste survey in private hospitals in Fars province, Iran

    International Nuclear Information System (INIS)

    Askarian, Mehrdad; Vakili, Mahmood; Kabir, Gholamhosein

    2004-01-01

    Hospital waste is considered dangerous because it may possess pathogenic agents and can cause undesirable effects on human health and the environment. In Iran, neither rules have been compiled nor does exact information exist regarding hospital waste management. The survey presented in this article was carried out in all 15 private hospitals of Fars province (Iran) from the total numbers of 50 governmental and private hospitals located in this province, in order to determine the amount of different kinds of waste produced and the present situation of waste management. The results indicated that the waste generation rate is 4.45 kg/bed/day, which includes 1830 kg (71.44%) of domestic waste, 712 kg (27.8%) of infectious waste, and 19.6 kg (0.76%) of sharps. Segregation of the different types of waste is not carried out perfectly. Two (13.3%) of the hospitals use containers without lids for on-site transport of wastes. Nine (60%) of the hospitals are equipped with an incinerator and six of them (40%) have operational problems with the incinerators. In all hospitals municipal workers transport waste outside the hospital premises daily or at the most on alternative days. In the hospitals under study, there aren't any training courses about hospital waste management and the hazards associated with them. The training courses that are provided are either ineffective or unsuitable. Performing extensive studies all over the country, compiling and enacting rules, establishing standards and providing effective personnel training are the main challenges for the concerned authorities and specialists in this field

  16. Perceived risks of radioactive waste transport through Oregon: Results of a statewide survey

    International Nuclear Information System (INIS)

    MacGregor, D.; Slovic, P.; Mason, R.G.; Detweiler, J.; Binney, S.E.; Dodd, B.

    1994-01-01

    Transportation of hazardous materials, and particularly radioactive wastes, on public highways has become an important risk management issue. The unfavorability of public attitudes regarding hazardous and nuclear waste signals the potential for strong public opposition to programs for transporting these materials. This paper presents the results of a survey conducted to assess public reactions to a long-term nuclear waste transport program planned to follow a route through a portion of rural Oregon. The survey assessed a number of key risk perception issues, including perceived health and safety risks of nuclear waste transport, relative risks of transport vs. storage at an existing site, trust in state officials, and satisfaction with life in communities along the transport route. The survey identified a number of attitudes and concerns that need to be understood and considered by those in charge of designing and implementing the waste-transportation program. 22 refs., 1 fig., 5 tabs

  17. Results after ten years of field testing low-level radioactive waste forms using lysimeters

    International Nuclear Information System (INIS)

    McConnell, J.W. Jr.; Rogers, R.D.; Jastrow, J.D.; Sanford, W.E.; Larsen, I.L.; Sullivan, T.M.

    1995-01-01

    The Field Lysimeter Investigations: Low-Level Waste Data Base Development Program is obtaining information on the performance of radioactive waste forms. Ion-exchange resins from a commercial nuclear power station were solidified into waste forms using portland cement and vinyl esterstyrene. These waste forms are being tested to: (a) obtain information on performance of waste forms in typical disposal environments, (b) compare field results with bench leach studies, (c) develop a low-level waste data base for use in performance assessment source term calculations, and (d) apply the DUST computer code to compare predicted cumulative release to actual field data. The program, funded by the Nuclear Regulatory Commission (NRC), includes observed radionuclide releases from waste forms in field lysimeters. The purpose of this paper is to present the experimental results of two lysimeter arrays over 10 years of operation, and to compare those results to bench test results and to DUST code predicted releases. Further analysis of soil cores taken to define the observed upward migration of radionuclides in one lysimeter is also presented

  18. Characterization of secondary solid waste anticipated from the treatment of trench water from Waste Area Grouping 6 at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    Kent, T.E.; Taylor, P.A.

    1992-09-01

    This project was undertaken to demonstrate that new liquid waste streams, generated as a consequence of closure activities at Waste Area Grouping (WAG) 6, can be treated adequately by existing wastewater treatment facilities at Oak Ridge National Laboratory (ORNL) without producing hazardous secondary solid wastes. Previous bench-scale treatable studies indicated that ORNL treatment operations will adequately remove the contaminants although additional study was required in order to characterize the secondary waste materials produced as a result of the treatment A 0.5-L/min pilot plant was designed and constructed to accurately simulate the treatment capabilities of ORNL fill-scale (490 L/min) treatment facilities-the Process Waste Treatment Plant (PWTP) and Nonradiological Wastewater Treatment Plant (NRWTP). This new test system was able to produce secondary wastes in the quantities necessary for US Environmental Protection Agency toxicity characteristic leaching procedure (TCLP) testing. The test system was operated for a 45-d test period with a minimum of problems and downtime. The pilot plant operating data verified that the WAG 6 trench waters can be treated at the PWTP and NRWTP to meet the discharge limits. The results of TCLP testing indicate that none of the secondary solid wastes will be considered hazardous as defined by the Resource Conservation and Recovery Act

  19. Development of threshold guidance: National Low-Level Radioactive Waste Management Program

    International Nuclear Information System (INIS)

    1986-09-01

    The current study has been conducted to provide DOE with a technical basis for the development of threshold guidance. The objective of the study was to develop the necessary background information and recommendations to assist the DOE in implementing the threshold limit concept for the disposal of DOE wastes at DOE facilities. The nature of low-level radioactive waste (LLW) varies greatly in both form and radionuclide content. While some low-level waste streams can contain substantial quantities of radioactive constituents, a potentially significant fraction of low-level waste is contaminated either very slightly or not at all. There is a strong likelihood that managing wastes with extremely low levels of radioactivity as nonradioactive waste would pose no significant safety problems and could result in substantial cost savings relative to its handling as LLW. Since all materials, including waste products, contain some radioactivity, it is necessary to distinguish between those wastes that would require disposal as LLW and those that have sufficiently low levels of radiological content to be managed according to their nonradiological properties. 131 refs., 9 figs., 24 tabs

  20. Processing results of 1,800 gallons of mercury and radioactively contaminated mixed waste rinse solution

    International Nuclear Information System (INIS)

    Thiesen, B.P.

    1993-01-01

    The mercury-contaminated rinse solution (INEL waste ID number-sign 123; File 8 waste) was successfully treated at the Idaho National Engineering Laboratory (INEL). This waste was generated during the decontamination of the Heat Transfer Reactor Experiment 3 (HTRE-3) reactor shield tank. Approximately 1,800 gal of waste was generated and was placed into 33 drums. Each drum contained precipitated sludge material ranging from 1--10 in. in depth, with the average depth of about 2.5 in. The pH of each drum varied from 3--11. The bulk liquid waste had a mercury level of 7.0 mg/l, which exceeded the Resource Conservation and Recovery Act (RCRA) limit of 0.2 mg/l. The average liquid bulk radioactivity was about 2.1 pCi/ml, while the average sludge contamination was about 13,800 pci/g. Treatment of the waste required separation of the liquid from the sludge, filtration, pH adjustment, and ion exchange. Because of difficulties in processing, three trials were required to reduce the mercury levels to below the RCRA limit. In the first trial, insufficient filtration of the waste allowed solid particulate produced during pH adjustment to enter into the ion exchange columns and ultimately the waste storage tank. In the second trial, the waste was filtered down to 0.1 μ to remove all solid mercury compounds. However, before filtration could take place, a solid mercury complex dissolved and mercury levels exceeded the RCRA limit after filtration. In the third trial, the waste was filtered through 0.3-A filters and then passed through the S-920 resin to remove the dissolved mercury. The resulting solut

  1. Results of Sludge Mobilization Testing at Hanford High Level Waste (HLW) Tank

    International Nuclear Information System (INIS)

    STAEHR, T.W.

    2001-01-01

    Waste stored in the Tank 241-AZ-101 at the US DOE Hanford is scheduled as the initial feed for high-level waste vitrification. Tank 241-AZ-101 currently holds over 3,000,000 liters of waste made up of a settled sludge layer covered by a layer of liquid supernant. To retrieve the waste from the tank, it is necessary to mobilize and suspend the settled sludge so that the resulting slurry can be pumped from the tank for treatment and vitrification. Two 223.8-kilowatt mixer pumps have been installed in Tank 241-AZ-101 to mobilize the settled sludge layer of waste for retrieval. In May of 2000, the mixer pumps were subjected to a series of tests to determine (1) the extent to which the mixer pumps could mobilize the settle sludge layer of waste, (2) if the mixer pumps could function within operating parameters, and (3) if state-of-the-art monitoring equipment could effectively monitor and quantify the degree of sludge mobilization and suspension. This paper presents the major findings and results of the Tank 241-AZ-101 mixer pump tests, based on analysis of data and waste samples that were collected during the testing. Discussion of the results focuses on the effective cleaning radius achieved and the volume and concentration of sludge mobilized, with both one and two pumps operating in various configurations and speeds. The Tank 241-AZ-101 mixer pump tests were unique in that sludge mobilization parameters were measured using actual waste in an underground storage tank at the hanford Site. The methods and instruments that were used to measure waste mobilization parameters in Tank 241-AZ-101 can be used in other tanks. It can be concluded from the testing that the use of mixer pumps is an effective retrieval method for the mobilization of settled solids in Tank 241-AZ-101

  2. Performance test results of noninvasive characterization of RCRA surrogate waste by prompt gamma neutron activation analysis

    International Nuclear Information System (INIS)

    Gehrke, R.J.; Propp, W.A.

    1997-11-01

    A performance evaluation to determine the feasibility of using prompt gamma neutron activation analysis (PGNAA) for noninvasive, quantitative assay of mixed waste containers was sponsored by DOE's Office of Technology Development (OTD), the Mixed Waste Focus Area (MWFA), and the Idaho National Engineering and Environmental Laboratory (INEEL). The evaluation was conducted using a surrogate waste, based on Portland cement, that was spiked with three RCRA metals, mercury, cadmium, and lead. The results indicate that PGNAA has potential as a process monitor. However, further development is required to improve its sensitivity to meet regulatory requirements for determination of these RCRA metals

  3. Waste disposal in granite: preliminary results from Stripa, Sweden

    International Nuclear Information System (INIS)

    Cook, N.G.W.; Gale, J.E.; Witherspoon, P.A.

    1979-03-01

    The results of this experiment to date indicate that the temperature fields in a rock mass contaning geologic discontinuities can be predicted with accuracy using the simple linear theory of heat conduction. Geologic discontinuities appear to introduce significant non-linear thermomechanical behavior into the rock mass as a result of which the thermally induced displacements are much less than those predicted by the simple theory of thermoelasticity, using laboratory values for Poisson's ratio and the coefficient of thermal expansion. The additional compliance introduced into the rock mass by geologic discontinuities affects the thermally induced stresses but to a lesser degree than the displacements. Further analytical, laboratory and field studies are expected to resolve many of the current uncertainties, especially field data gathered during a planned cooling down period following the switching off of the heaters scheduled in the near future

  4. Improving radiation awareness and feeling of personal security of non-radiological medical staff by implementing a traffic light system in computed tomography

    Energy Technology Data Exchange (ETDEWEB)

    Heilmaier, C.; Mayor, A.; Zuber, N.; Weishaupt, D. [Stadtspital Triemli, Zurich (Switzerland). Dept. of Radiology; Fodor, P. [Stadtspital Triemli, Zurich (Switzerland). Dept. of Anesthesiology and Intensive Care Medicine

    2016-03-15

    Non-radiological medical professionals often need to remain in the scanning room during computed tomography (CT) examinations to supervise patients in critical condition. Independent of protective devices, their position significantly influences the radiation dose they receive. The purpose of this study was to assess if a traffic light system indicating areas of different radiation exposure improves non-radiological medical staff's radiation awareness and feeling of personal security. Phantom measurements were performed to define areas of different dose rates and colored stickers were applied on the floor according to a traffic light system: green = lowest, orange = intermediate, and red = highest possible radiation exposure. Non-radiological medical professionals with different years of working experience evaluated the system using a structured questionnaire. Kruskal-Wallis and Spearman's correlation test were applied for statistical analysis. Fifty-six subjects (30 physicians, 26 nursing staff) took part in this prospective study. Overall rating of the system was very good, and almost all professionals tried to stand in the green stickers during the scan. The system significantly increased radiation awareness and feeling of personal protection particularly in staff with ? 5 years of working experience (p < 0.05). The majority of non-radiological medical professionals stated that staying in the green stickers and patient care would be compatible. Knowledge of radiation protection was poor in all groups, especially among entry-level employees (p < 0.05). A traffic light system in the CT scanning room indicating areas with lowest, in-termediate, and highest possible radiation exposure is much appreciated. It increases radiation awareness, improves the sense of personal radiation protection, and may support endeavors to lower occupational radiation exposure, although the best radiation protection always is to re-main outside the CT room during the scan.

  5. Improving radiation awareness and feeling of personal security of non-radiological medical staff by implementing a traffic light system in computed tomography

    International Nuclear Information System (INIS)

    Heilmaier, C.; Mayor, A.; Zuber, N.; Weishaupt, D.; Fodor, P.

    2016-01-01

    Non-radiological medical professionals often need to remain in the scanning room during computed tomography (CT) examinations to supervise patients in critical condition. Independent of protective devices, their position significantly influences the radiation dose they receive. The purpose of this study was to assess if a traffic light system indicating areas of different radiation exposure improves non-radiological medical staff's radiation awareness and feeling of personal security. Phantom measurements were performed to define areas of different dose rates and colored stickers were applied on the floor according to a traffic light system: green = lowest, orange = intermediate, and red = highest possible radiation exposure. Non-radiological medical professionals with different years of working experience evaluated the system using a structured questionnaire. Kruskal-Wallis and Spearman's correlation test were applied for statistical analysis. Fifty-six subjects (30 physicians, 26 nursing staff) took part in this prospective study. Overall rating of the system was very good, and almost all professionals tried to stand in the green stickers during the scan. The system significantly increased radiation awareness and feeling of personal protection particularly in staff with ? 5 years of working experience (p < 0.05). The majority of non-radiological medical professionals stated that staying in the green stickers and patient care would be compatible. Knowledge of radiation protection was poor in all groups, especially among entry-level employees (p < 0.05). A traffic light system in the CT scanning room indicating areas with lowest, in-termediate, and highest possible radiation exposure is much appreciated. It increases radiation awareness, improves the sense of personal radiation protection, and may support endeavors to lower occupational radiation exposure, although the best radiation protection always is to re-main outside the CT room during the scan.

  6. Alternative Electrochemical Salt Waste Forms, Summary of FY11-FY12 Results

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Mccloy, John S. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Crum, Jarrod V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lepry, William C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rodriguez, Carmen P. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Windisch, Charles F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Matyas, Josef [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Westman, Matthew P. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rieck, Bennett T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lang, Jesse B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Olszta, Matthew J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pierce, David A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2014-01-17

    The Fuel Cycle Research and Development Program, sponsored by the U.S. Department of Energy Office of Nuclear Energy, is currently investigating alternative waste forms for wastes generated from nuclear fuel processing. One such waste results from an electrochemical separations process, called the “Echem” process. The Echem process utilizes a molten KCl-LiCl salt to dissolve the fuel. This process results in a spent salt containing alkali, alkaline earth, lanthanide halides and small quantities of actinide halides, where the primary halide is chloride with a minor iodide fraction. Pacific Northwest National Laboratory (PNNL) is concurrently investigating two candidate waste forms for the Echem spent-salt: high-halide minerals (i.e., sodalite and cancrinite) and tellurite (TeO2)-based glasses. Both of these candidates showed promise in fiscal year (FY) 2009 and FY2010 with a simplified nonradioactive simulant of the Echem waste. Further testing was performed on these waste forms in FY2011 and FY2012 to assess the possibility of their use in a sustainable fuel cycle. This report summarizes the combined results from FY2011 and FY2012 efforts.

  7. Generation and management of solid waste resulting from tourist activities of the Porto de Galinhas - P

    Directory of Open Access Journals (Sweden)

    Jaqueline Guimarães Santos

    2015-04-01

    Full Text Available The significant solid waste generation, coupled with the lack of proper management of the waste generated, has been one of the issues of concern and conducting research on the part of scholars in the field. Tourism as an activity that positively impacts and negativity a given location, has emerged as an activity that can generate a lot of waste, especially in periods of high season, considering the increase of people moving to the tourist destinations. Accordingly, this study aims to analyze the generation and management of solid waste resulting from tourism in Porto de Galinhas, PE. We performed an exploratory, descriptive, qualitative study, conducted in the form of a case study in Porto de Galinhas, PE. The data collection was done interviews together social actors, as well as non-participant observation during data collection. The results showed that tourism activities in Porto de Galinhas result in a high amount of solid waste, and these are directed to inappropriate places. Although fate presents a combination of recyclable materials, RECYCLE, reuses this not a significant amount, given the proportion of waste generated.

  8. National Waste Terminal Storage repository in a bedded salt formation for spent unreprocessed fuel. Occupational exposure and health physics studies. KE report No. 78-21-R

    International Nuclear Information System (INIS)

    1978-09-01

    This report includes the Occupational Exposure and Health Physics studies of the National Waste Terminal Storage Repository Number 2 (NWTSR2). Section 1 deals with occupational radiation exposures. The results of dose equivalent and dose commitment calculations are summarized. The man-rem total is a summation of all doses to all personnel throughout a year. It would have to be divided by the number of total personnel involved, to obtain an average annual dose per person. Section 2 presents the occupational exposures due to nonradiological pollutants. The activities of workers and the equipment used during the construction, storage/retrieval and decommissioning of the facility are outlined. Tabulations are presented of the substances (dusts, fumes, gases and vapors) and physical agents (heat, vibration, and nonionizing radiation) to which personnel will be exposed in various surface and underground work areas during construction and decommissioning, and during storage and retrieval operations. Some significant nonradiological occupational exposures are summarized. Section 3 outlines the health physics program for the NWTSR2 facility. It is important to initiate the health physics program one or two years before the facility is placed in operation to establish the radiation background levels at the site and its surrounding area, and to collect environmental samples, both on-site and off-site, prior to waste storage and retrieval operations. The health physics organization consists of 15 persons, including five health physicists

  9. The public visits a nuclear waste site: Survey results from the West Valley Demonstration Project

    International Nuclear Information System (INIS)

    Hoffman, W.D.

    1987-01-01

    This paper discusses the results of the 1986 survey taken at the West Valley Demonstration Project Open House where a major nuclear waste cleanup is in progress. Over 1400 people were polled on what they think is most effective in educating the public on nuclear waste. A demographic analysis describes the population attending the event and their major interests in the project. Responses to attitudinal questions are examined to evaluate the importance of radioactive waste cleanup as an environmental issue and a fiscal responsibility. Additionally, nuclear power is evaluated on its public perception as an energy resource. The purpose of the study is to find out who visits a nuclear waste site and why, and to measure their attitudes on nuclear issues

  10. Idaho Nuclear Technology and Engineering Center (INTEC) Sodium Bearing Waste - Waste Incidental to Reprocessing Determination

    International Nuclear Information System (INIS)

    Jacobson, Victor Levon

    2002-01-01

    U.S. Department of Energy Manual 435.1-1, Radioactive Waste Management, Section I.1.C, requires that all radioactive waste subject to Department of Energy Order 435.1 be managed as high-level radioactive waste, transuranic waste, or low-level radioactive waste. Determining the radiological classification of the sodium-bearing waste currently in the Idaho Nuclear Technology and Engineering Center Tank Farm Facility inventory is important to its proper treatment and disposition. This report presents the technical basis for making the determination that the sodium-bearing waste is waste incidental to spent fuel reprocessing and should be managed as mixed transuranic waste. This report focuses on the radiological characteristics of the sodium-bearing waste. The report does not address characterization of the nonradiological, hazardous constituents of the waste in accordance with Resource Conservation and Recovery Act requirements

  11. Risks associated with nuclear material recovery and waste preparation

    Energy Technology Data Exchange (ETDEWEB)

    Fullwood, R R; Erdmann, R C

    1983-01-01

    An analysis of the risk associated with nuclear material recovery and waste preparation is presented. The steps involve: reprocessing of spent fuel to recycle fissionable material, refabrication of the recovered material for use as reactor fuel, and the transportation links connecting these plants with the power plants and waste repositories. The risks considered are radiological and non-radiological, accident and routine effects on the public and workers during plant construction, operation and decommissioning.

  12. Nonradiological chemical pathway analysis and identification of chemicals of concern for environmental monitoring at the Hanford Site

    International Nuclear Information System (INIS)

    Blanton, M.L.; Cooper, A.T.; Castleton, K.J.

    1995-11-01

    Pacific Northwest's Surface Environmental Surveillance Project (SESP) is an ongoing effort tot design, review, and conducted monitoring on and off the Hanford site. Chemicals of concern that were selected are listed. Using modeled exposure pathways, the offsite cancer incidence and hazard quotient were calculated and a retrospective pathway analysis performed to estimate what onsite concentrations would be required in the soil for each chemical of concern and other detected chemicals that would be required to obtain an estimated offsite human-health risk of 1.0E-06 cancer incidence or 1.0 hazard quotient. This analysis indicates that current nonradiological chemical contamination occurring on the site does not pose a significant offsite human-health risk; the highest cancer incidence to the offsite maximally exposed individual was from arsenic (1.76E-10); the highest hazard quotient was chromium(VI) (1.48E-04). The most sensitive pathways of exposure were surfacewater and aquatic food consumption. Combined total offsite excess cancer incidence was 2.09E-10 and estimated hazard quotient was 2.40E-04. Of the 17 identified chemicals of concern, the SESP does not currently (routinely) monitor arsenic, benzo(a)pyrene, bis(2- ethylhexyl)phthalate (BEHP), and chrysene. Only 3 of the chemicals of concern (arsenic, BEHP, chloroform) could actually occur in onsite soil at concern high enough to cause a 1.0E-06 excess cancer incidence or a 1.0 hazard index for a given offsite exposure pathway. During the retrospective analysis, 20 other chemicals were also evaluated; only vinyl chloride and thallium could reach targeted offsite risk values

  13. Cost avoidance techniques through the Fernald controlled area trash segregation program and the RIMIA solid waste reduction program

    International Nuclear Information System (INIS)

    Menche, C.E.

    1997-01-01

    The Fernald Environmental Management Project is a Department of Energy owned facility that produced high quality uranium metals for military defense. The Fernald mission has changed from one of production to remediation. Remediation is intended to clean up legacy (primary) waste from past practices. Little opportunity is available to reduce the amount of primary waste. However, there is an opportunity to reduce secondary waste generation, primarily through segregation. Two programs which accomplish this are the Controlled Area Trash Segregation Program and the RIMIA Solid Waste Reduction Program. With these two programs now in place at the FEMP, it has been estimated that a 60% reduction has been achieved in unnecessary clean waste being disposed as Low Level Waste at the Nevada Test Site. The cost savings associated with these programs (currently 79,000 cubic feet, $428,000) could easily run into the millions of dollars based on the upcoming restoration activities to be undertaken. The segregation of non-radiological waste in the radiologically Controlled Area not only establishes a firm commitment to send only low-level radioactive waste to the Nevada Test Site, but also results in substantial cost avoidance

  14. Critical evaluation of the nonradiological environmental technical specifications. Program description, summary, and recommendations. Vol. 1

    International Nuclear Information System (INIS)

    Adams, S.M.; Cunningham, P.A.; Gray, D.D.; Kumar, K.D.; Witten, A.J.

    1976-01-01

    A comprehensive study of the data collected as part of the environmental Technical Specifications program for eight nuclear power plants was conducted for the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory commission. This report includes a summary of the screening phase in which the adequacy of the hydrothermal and ecological monitoring data for each plant were evaluated, and the summary and recommendations resulting from a detailed examination of the three nuclear power plants selected in the initial screening

  15. Radioactive waste disposal by UKAEA establishments during 1979 and associated environmental monitoring results

    International Nuclear Information System (INIS)

    Flew, E.M.

    1980-07-01

    This report gives details of the amounts of solid and liquid radioactive waste disposed of by the principal establishments of the UKAEA during 1979. Waste arising at the UKAEA Nuclear Power Development Laboratories at Windscale and Springfields, which are both situated on British Nuclear Fuels Ltd. (BNFL) sites, is disposed of by BNFL and included in their authorisations. Discharges to atmosphere of airborne radioactive waste are also included in the report. A summary of the results of the environmental monitoring programmes carried out in connection with the radioactive waste discharges is given. To facilitate an appreciation of the standard of safety achieved, the discharges are, where appropriate, shown as a percentage of those authorised. In the case of atmospheric discharges no quantitative limits are yet specified in the authorisations, but the results and estimates of discharges from stacks are compared with Derived Working Limits (DWL's) (i.e. a limit derived from the dose limits recommended by The International Commission on Radiological Protection in such a way that compliance with it implies virtual certainty of compliance with the relevant dose limits). Environmental monitoring results are also compared with appropriate DWL's. The principles underlying the control of the discharge of radioactive waste to the environment are summarised in an Appendix to the report. (author)

  16. Results of the CRCPD survey of 1984 low-level radioactive waste: progress to mid-September, 1986

    International Nuclear Information System (INIS)

    Devine, T.L.

    1987-01-01

    The survey of 1984 low-level radioactive waste by the Conference of Radiation Control Program Directors, Inc., is the second such survey. The previous survey was for waste generated during 1982. The CRCPD survey of 1984 LLRW requested information concerning the license, the effluents and other on-site managed wastes, details of exported waste type, the capacity for storing waste prior to shipment and its average utilization during 1984. Details of the exported waste included waste type, processing and packaging, NRC class, burial site or broker to which the waste was sent, and anticipated waste generation by year and by class through 1989. Shortcomings of the questionnaire and preliminary results are discussed. Based on the results of the two surveys of low-level radioactive waste conducted by the CRCPD, and the serious discrepancies which exist between data on waste shipped by generators and that on waste received by disposal sites, the following recommendation is made. That a single, national repository be established for all data on the generation and ultimate disposition of low-level radioactive waste. 1 figure, 1 table

  17. BCLDP site environmental report for calendar year 1997 on radiological and nonradiological parameters

    International Nuclear Information System (INIS)

    Fry, J.

    1998-01-01

    Battelle Memorial Institute currently maintains its retired nuclear research facilities in a surveillance and maintenance (S and M) mode and continues decontamination and decommissioning (D and D) activities. The activities are referred to as the Battelle Columbus Laboratories Decommissioning Project (BCLDP). Operations reference in this report are performed in support of S and M and D and D activities. The majority of this report is devoted to discussion of the West Jefferson facility, because the source term at this facility is larger than the source term at Battelle's King Avenue site. The contamination found at the King Avenue site consists of small amounts of residual radioactive material in solid form, which has become embedded or captured in nearby surfaces such as walls, floors, ceilings, drains, laboratory equipment, and soils. By the end of calendar year (CY) 1997, most remediation activities were completed at the King Avenue site. The contamination found at the West Jefferson site is the result of research and development activities with irradiated materials. During CY 1997, multiple tests at the West Jefferson Nuclear Sciences Area found no isotopes present above the minimum detectable activity (MDA) for air releases or for liquid discharges to Big Darby Creek. Data obtained from downstream sampling locations were statistically indistinguishable from background levels

  18. Using slaughterhouse waste in a biochemical-based biorefinery - results from pilot scale tests.

    Science.gov (United States)

    Schwede, Sebastian; Thorin, Eva; Lindmark, Johan; Klintenberg, Patrik; Jääskeläinen, Ari; Suhonen, Anssi; Laatikainen, Reino; Hakalehto, Elias

    2017-05-01

    A novel biorefinery concept was piloted using protein-rich slaughterhouse waste, chicken manure and straw as feedstocks. The basic idea was to provide a proof of concept for the production of platform chemicals and biofuels from organic waste materials at non-septic conditions. The desired biochemical routes were 2,3-butanediol and acetone-butanol fermentation. The results showed that hydrolysis resulted only in low amounts of easily degradable carbohydrates. However, amino acids released from the protein-rich slaughterhouse waste were utilized and fermented by the bacteria in the process. Product formation was directed towards acidogenic compounds rather than solventogenic products due to increasing pH-value affected by ammonia release during amino acid fermentation. Hence, the process was not effective for 2,3-butanediol production, whereas butyrate, propionate, γ-aminobutyrate and valerate were predominantly produced. This offered fast means for converting tedious protein-rich waste mixtures into utilizable chemical goods. Furthermore, the residual liquid from the bioreactor showed significantly higher biogas production potential than the corresponding substrates. The combination of the biorefinery approach to produce chemicals and biofuels with anaerobic digestion of the residues to recover energy in form of methane and nutrients that can be utilized for animal feed production could be a feasible concept for organic waste utilization.

  19. Research results of sewage sludge and waste oil disposal by entrained bed gasification

    Energy Technology Data Exchange (ETDEWEB)

    Schingnitz, M.; Goehler, P.; Wenzel, W.; Seidel, W. (Noell-DBI Energie- und Entsorgungstechnik GmbH, Freiberg (Germany))

    1992-01-01

    Presents results of gasifying sewage sludge and waste oil with the GSP technology, developed by the Freiberg Fuel Institute (FRG). The GSP reactor was developed in 1976 for gasification of pulverized brown coal. An industrial reactor of this design operated for over 5 years with a total coal throughput of more than 300,000 t. The design of the gasification generator and the flowsheet of a 3 MW experimental pilot plant for waste gasification are presented. The PCB content in the gasification sludge is 6.14 mg/kg, in waste oil - 160 mg/kg. Gasification takes place at high temperatures of more than 1,400 C for complete destruction of toxic pollutants. Gasification results compare composition of raw gas produced by gasification of brown coal, sewage sludge and waste oil. A detailed list of content of pollutants (PCDD, PCDF, PAH, dioxin and furan) in the gasification gas, in process waters and in solid residue of the process water is provided. It is concluded that the GSP gasification process is suitable for safe disposal of waste with toxic content. 3 refs.

  20. New developments and improvements in processing of 'problematic' radioactive waste. Results of a coordinated research project 2003-2007

    International Nuclear Information System (INIS)

    2007-12-01

    This report addresses a category of wastes termed 'problematic wastes', wastes for which safe, efficient and cost effective methods for processing are not readily available. Processing options for many of these are identified and addressed. Results presented, illustrate the strategy for breaking 'problematic' waste streams down into a sequence of 'standard' issues which are amenable to solution. Decision makers and facility managers faced with problematic waste streams should be able to use this information to identify and pursue solutions to meet their needs. In this report, processing options for a total of 27 problematic waste streams that were identified and addressed by the individual laboratories participating in the Coordinated Research Project are discussed. These waste streams covered an extremely broad spectrum, ranging from simple, one component aqueous solutions originating from a research laboratory to very complex aqueous concentrates of waste resulting from reprocessing activities or reactor operation. These challenging wastes included: waste contaminated by tritium, wastes containing transuranic elements, and solid health care waste. The range of aqueous wastes included those contaminated by organic complexing agents and surfactants to pure organic waste such as contaminated oil. Correspondingly, the scale of approaches and technologies used to address these wastes is very broad. Use of this report is likely to be most effective as an initial screening tool to identify technologies best able to meet specific waste management objectives in terms of the waste generated, the technical complexity, the available economic resources, the environmental impact considerations, and the desired end product (output) of the technology. The report should assist the user to compare technologies and to reach an informed decision based on safety, technological maturity, economics, and other local needs

  1. Characterization of a low-level radioactive waste grout: Sampling and test results

    International Nuclear Information System (INIS)

    Martin, P.F.C.; Lokken, R.O.

    1992-12-01

    WHC manages and operates the grout treatment facility at Hanford as part of a DOE program to clean up wastes stored at federal nuclear production sites. PNL provides support to the grout disposal program through pilot-scale tests, performance assessments, and formulation verification activities. in 1988 and 1989, over one million gallons of a low-level radioactive liquid waste was processed through the facility to produce a grout waste that was then deposited in an underground vault. The liquid waste was phosphate/sulfate waste (PSW) generated in decontamination of the N Reactor. PNL sampled and tested the grout produced during the second half of the PSW campaign to support quality verification activities prior to grout vault closure. Samples of grout were obtained by inserting nested-tube samplers into the grout slurry in the vault. After the grout had cured, the inner tube of the sampler was removed and the grout samples extracted. Tests for compressive strength, sonic velocity, and leach testing were used to assess grout quality; results were compared to those from pilot-scale test grouts made with a simulated PSW. The grout produced during the second half of the PSW campaign exceeded compressive strength and leachability formulation criteria. The nested tube samplers were effective in collecting samples of grout although their use introduced greater variability into the compressive strength data

  2. Status of test results of electrochemical organic oxidation of a tank 241-SY-101 simulated waste

    International Nuclear Information System (INIS)

    Colby, S.A.

    1994-06-01

    This report presents scoping test results of an electrochemical waste pretreatment process to oxidize organic compounds contained in the Hanford Site's radioactive waste storage tanks. Electrochemical oxidation was tested on laboratory scale to destroy organics that are thought to pose safety concerns, using a nonradioactive, simulated tank waste. Minimal development work has been applied to alkaline electrochemical organic destruction. Most electrochemical work has been directed towards acidic electrolysis, as in the metal purification industry, and silver catalyzed oxidation. Alkaline electrochemistry has traditionally been associated with the following: (1) inefficient power use, (2) electrode fouling, and (3) solids handling problems. Tests using a laboratory scale electrochemical cell oxidized surrogate organics by applying a DC electrical current to the simulated tank waste via anode and cathode electrodes. The analytical data suggest that alkaline electrolysis oxidizes the organics into inorganic carbonate and smaller carbon chain refractory organics. Electrolysis treats the waste without adding chemical reagents and at ambient conditions of temperature and pressure. Cell performance was not affected by varying operating conditions and supplemental electrolyte additions

  3. 60-Day waste compatibility safety issues and final results for AY-102 grab samples

    Energy Technology Data Exchange (ETDEWEB)

    Nuzum, J.L.

    1997-01-31

    Four grab samples (2AY-96-15, 2AY-96-16, 2AY-96-17, and 2AY-96-18) were taken from Riser 15D of Tank 241-AY-102 on October 8, 1996, and received by 222-S Laboratory on October 8, 1996. These samples were analyzed in accordance with Compatibility Grab Sampling and Analysis Plan (TSAP) and Data Quality Objectives for Tank Farms Waste Compatibility Program (DQO) in support of the Waste Compatibility Program. No notifications were required based on sample results.

  4. Liquid and Gaseous Waste Operations Department annual operating report CY 1996

    International Nuclear Information System (INIS)

    Maddox, J.J.; Scott, C.B.

    1997-03-01

    This annual report summarizes operating activities dealing with the process waste system, the liquid low-level waste system, and the gaseous waste system. It also describes upgrade activities dealing with the process and liquid low-level waste systems, the cathodic protection system, a stack ventilation system, and configuration control. Maintenance activities are described dealing with nonradiological wastewater treatment plant, process waste treatment plant and collection system, liquid low-level waste system, and gaseous waste system. Miscellaneous activities include training, audits/reviews/tours, and environmental restoration support

  5. Demonstration Results on the Effects of Mercury Speciation on the Stabilization of Wastes

    International Nuclear Information System (INIS)

    Conley, T.B.; Hulet, G.A.; Morris, M.I.; Osborne-Lee, I.W.

    1999-01-01

    Mercury-contaminated wastes are currently being stored at approximately 19 Department of Energy sites, the volume of which is estimated to be about 16m(sup)3. These wastes exist in various forms including soil, sludges, and debris, which present a particular challenge regarding possible mercury stabilization methods. This reports provides the test results of three vendors, Allied Technology Group, IT Corporation, and Nuclear Fuel Services, Inc., that demonstrate the effects of mercury speciation on the stabilization of the mercury wastes. Mercury present in concentrations that exceed 260 parts per million must be removed by extraction methods and requires stabilization to ensure that the final wasteforms leach less than 0.2mg/L of mercury by the Toxicity Characteristic Leaching Procedure or 0.025 mg/L using the Universal Treatment Standard

  6. Study of the use waste resulting from the mining of emerald for the production refractory ceramic

    International Nuclear Information System (INIS)

    Esteves, P.J.C.; Coelho, R.E.; Cruz, R.M.S.; Cavalcanti, R.F.

    2009-01-01

    Full text: The great impact caused by excess mineral waste in ambient of the emerald exploration, in determined locals of Brazil, where are deposited, it has caused inconvenience to their various people residents. The jungles, rivers and lakes are directly harmed by the aggressions imposed by neglect in the destination of such waste. Considering the importance of the issue outlined to the goal of this work, this paper can back report a study for utilizing emerald waste, focused the possibility manufacture for obtained refractory ceramic. The results show that the specimens prepared by the ball milling, cold pressing and sintering method had better high temperature properties, due to a higher mica volume percent and finer crystallite size. Specimens it was characterized by X-ray diffractometer and fluorescence. Test was realized in the materials, submitted in high temperature was observed good thermal stability, the processed ceramics could be recommended for the adequate applications. (author)

  7. A literature-based preliminary characterization of risks in the nuclear waste management system

    International Nuclear Information System (INIS)

    Daling, P.M.; Rhoads, R.E.; Van Luik, A.E.

    1990-04-01

    The objectives of this study were to (1) review the literature containing information on risks in the nuclear waste management system and (2) use this information to develop preliminary estimates of the potential magnitudes of these risks. Information was collected on a broad range of risk categories to assist the US Department of Energy (DOE) in communicating information about the risks in the waste management system. The study, which was completed prior to passage of the Nuclear Waste Policy Amendments Act of 1987, examined all of the portions of the nuclear waste management system envisioned by the DOE in the 1985 ''Mission Plant for the Civilian Radioactive Waste Management Program.'' As such, there may be statements in this paper that are not consistent with current DOE positions. The scope of this paper includes the repository, the integral Monitored Retrievable Storage (MRS) facility, and the transportation system that supports the repository and the MRS facility. Based on the results of this analysis, it is concluded that the radiological risks in the waste management system are small relative to nonradiological risks and relative to the risks of exposure to natural background radiation. 6 refs., 2 figs., 2 tabs

  8. Converting Simulated Sodium-bearing Waste into a Single Solid Waste Form by Evaporation: Laboratory- and Pilot-Scale Test Results on Recycling Evaporator Overheads

    Energy Technology Data Exchange (ETDEWEB)

    Griffith, D.; D. L. Griffith; R. J. Kirkham; L. G. Olson; S. J. Losinski

    2004-01-01

    Conversion of Idaho National Engineering and Environmental Laboratory radioactive sodium-bearing waste into a single solid waste form by evaporation was demonstrated in both flask-scale and pilot-scale agitated thin film evaporator tests. A sodium-bearing waste simulant was adjusted to represent an evaporator feed in which the acid from the distillate is concentrated, neutralized, and recycled back through the evaporator. The advantage to this flowsheet is that a single remote-handled transuranic waste form is produced in the evaporator bottoms without the generation of any low-level mixed secondary waste. However, use of a recycle flowsheet in sodium-bearing waste evaporation results in a 50% increase in remote-handled transuranic volume in comparison to a non-recycle flowsheet.

  9. Initial Laboratory-Scale Melter Test Results for Combined Fission Product Waste

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J.; Crum, Jarrod V.; Buchmiller, William C.; Rieck, Bennett T.; Schweiger, Michael J.; Vienna, John D.

    2009-10-01

    This report describes the methods and results used to vitrify a baseline glass, CSLNTM-C-2.5 in support of the AFCI (Advanced Fuel Cycle Initiative) using a Quartz Crucible Scale Melter at the Pacific Northwest National Laboratory. Document number AFCI-WAST-PMO-MI-DV-2009-000184.

  10. Hazards assessment for the Waste Experimental Reduction Facility

    Energy Technology Data Exchange (ETDEWEB)

    Calley, M.B.; Jones, J.L. Jr.

    1994-09-19

    This report documents the hazards assessment for the Waste Experimental Reduction Facility (WERF) located at the Idaho National Engineering Laboratory, which is operated by EG&G Idaho, Inc., for the US Department of Energy (DOE). The hazards assessment was performed to ensure that this facility complies with DOE and company requirements pertaining to emergency planning and preparedness for operational emergencies. DOE Order 5500.3A requires that a facility-specific hazards assessment be performed to provide the technical basis for facility emergency planning efforts. This hazards assessment was conducted in accordance with DOE Headquarters and DOE Idaho Operations Office (DOE-ID) guidance to comply with DOE Order 5500.3A. The hazards assessment identifies and analyzes hazards that are significant enough to warrant consideration in a facility`s operational emergency management program. This hazards assessment describes the WERF, the area surrounding WERF, associated buildings and structures at WERF, and the processes performed at WERF. All radiological and nonradiological hazardous materials stored, used, or produced at WERF were identified and screened. Even though the screening process indicated that the hazardous materials could be screened from further analysis because the inventory of radiological and nonradiological hazardous materials were below the screening thresholds specified by DOE and DOE-ID guidance for DOE Order 5500.3A, the nonradiological hazardous materials were analyzed further because it was felt that the nonradiological hazardous material screening thresholds were too high.

  11. Hazards assessment for the Waste Experimental Reduction Facility

    International Nuclear Information System (INIS)

    Calley, M.B.; Jones, J.L. Jr.

    1994-01-01

    This report documents the hazards assessment for the Waste Experimental Reduction Facility (WERF) located at the Idaho National Engineering Laboratory, which is operated by EG ampersand G Idaho, Inc., for the US Department of Energy (DOE). The hazards assessment was performed to ensure that this facility complies with DOE and company requirements pertaining to emergency planning and preparedness for operational emergencies. DOE Order 5500.3A requires that a facility-specific hazards assessment be performed to provide the technical basis for facility emergency planning efforts. This hazards assessment was conducted in accordance with DOE Headquarters and DOE Idaho Operations Office (DOE-ID) guidance to comply with DOE Order 5500.3A. The hazards assessment identifies and analyzes hazards that are significant enough to warrant consideration in a facility's operational emergency management program. This hazards assessment describes the WERF, the area surrounding WERF, associated buildings and structures at WERF, and the processes performed at WERF. All radiological and nonradiological hazardous materials stored, used, or produced at WERF were identified and screened. Even though the screening process indicated that the hazardous materials could be screened from further analysis because the inventory of radiological and nonradiological hazardous materials were below the screening thresholds specified by DOE and DOE-ID guidance for DOE Order 5500.3A, the nonradiological hazardous materials were analyzed further because it was felt that the nonradiological hazardous material screening thresholds were too high

  12. E-waste collection in Italy: Results from an exploratory analysis.

    Science.gov (United States)

    Favot, Marinella; Grassetti, Luca

    2017-09-01

    This study looks at the performance of household electrical and electronic waste (WEEE) collection in 20 Italian regions from 2008 to 2015. The impact of several explicative variables on the results of e-waste collection is evaluated. The independent variables are socio-economic and demographic ones (age, gender, household size, education level, migration and income) along with technical-organisational variables (population density, presence of metropoles, macro regions, characteristics of the territory, percentage of household waste collected separately and number of e-waste collection points). The results show that the presence of collection points, the percentage of household waste collected separately and the percentage of females are positively correlated with the kg collected per inhabitant per year. For example, a variation of 1% of input (presence of collection points) corresponds to a 0.25% variation in the output (collection results) while 1% difference in the percentage of females in the population corresponds to a 7.549% difference in the collection rate. Population density, instead, is negatively correlated. It is interesting to note that there is a discrepancy between the Southern regions and the Centre regions (the former have an outcome 0.66 times lower than the latter) while the Northern regions perform similarly to the Centre ones. Moreover, the first year (2008) had a very low performance compared to the following years when the scheme constantly improved, mainly due to the additional collection points available. The Stochastic Frontier Model allows for the identification of the optimal production function among the 20 Italian regions. The best performing region is Tuscany (in the Centre), followed by Sardinia and Sicily (in the South). Copyright © 2017. Published by Elsevier Ltd.

  13. Environmental impact assessment and socio political issues of nuclear waste management

    International Nuclear Information System (INIS)

    Harmaajaervi, I.; Tolsa, H.

    1997-09-01

    The study is a part of the Publicly Administrated Nuclear Waste Management Research Programme (JYT2) which was carried out in 1994-1996. The principal goal of the research programme has been to provide the authorities with information and research results relevant for the safety of nuclear waste management in order to support the various activities of the authorities. The main emphasis of the research programme focuses on the disposal of spent fuel. In addition to nuclear waste research in the field of natural sciences and technology, the research program- me has focused mostly on societal issues associated with nuclear waste disposal facilities and on the non-radiological environmental effects in the environs of the disposal site. Some of the local effects are already revealed in the research phase, before any final decisions are made as to the selection of the disposal site. The study has focused primarily on local and regional issues. The statutory requirement to conduct environ- mental impact assessment (EIA) chiefly concerns those who are responsible for waste management, but the authorities also need to acquire systematic information in the field to support developing requirements for the content and scope of EIA procedure and preparedness to check the assessments made. This is a report of the first parts of the study in 1994-1995. The report deals with the subject matter generally based on earlier studies in Finland and other countries. The results of the study will be reported later

  14. Effects of container material on PCT leach test results for high-level nuclear waste glasses

    International Nuclear Information System (INIS)

    Xing, S.B.; Pegg, I.L.

    1994-01-01

    A glass-based waste form used for the immobilization of high-level nuclear wastes should exhibit good resistance to aqueous corrosion since typically this is the primary process by which radionucleides could be released into the environment upon failure of other barriers. In the USA, the Waste Acceptance Product Specifications (WAPS) provides a set of requirements to ensure the consistency of the waste forms produced and specifies the Product Consistency Test (PCT) as a measure of relative chemical durability. While the PCT procedure permits usage of both Teflon and stainless steel vessels for testing of simulated development glasses, Teflon is not permitted for testing of production glasses due to radiative degradation. The results presented in this paper indicate that there are very significant differences between tests conducted in the two types of vessels due to the well-known permeability of Teflon to atmospheric carbon dioxide which results in lowering of the solution pH and a consequent reduction in the leach rate of silicate glasses. A wide range of nuclear waste glass compositions was subjected to the PCT procedure using both Teflon and stainless steel vessels. The magnitude of the effect (up to a factor of four for B, Na, Li concentrations) depends strongly on glass composition, therefore the isolated checks performed previously were inconclusive. The permeability to CO, of two types of Teflon vessels specified in the PCT procedure was directly measured using buffer solutions: ingress of CO, is linear in time, strongly pH-dependent, and was as high as 100 ppm after 7 days. In actual PCT tests in Teflon vessels, the total CO, content was 560 ppm after 87 days and 1930 ppm after one year

  15. Plasma/arc melter review for vitrification of mixed wastes: Results

    Energy Technology Data Exchange (ETDEWEB)

    Eddy, T.L.; Soelberg, N.R.; Raivo, B.D. [MeltTran, Inc., Idaho Falls, ID (United States)

    1995-12-31

    In October of 1994, the Idaho Waste Treatment Program (IWTP) sponsored a workshop to review the results of a plasma/arc melter system preliminary design for treating mixed waste. Attention focused on (1) the melter design, (2) the offgas system design, and (3) the overall system design. The inclusion of feed preparation and handling systems, as well as monitoring and control systems, were considered premature until decisions regarding the melter and offgas treatment were resolved. The evaluation was based on the constraints of the transuranic-contaminated mixed waste in the Radioactive Waste Management Complex (RWMC) at the Idaho National Engineering Laboratory (INEL). Major factors are the retention of the transuranics in the basaltic slag, maintenance in a radioactive environment, reliability of components to prevent any major problems, upsets, or safety concerns, and the collection, elimination, or reduction of hazardous materials for appropriate stabilization. Several modifications were recommended by the group at large, discussed by the subcommittees, and accepted as the preferred options by the design team. Though all questions were not answered, the preferred systems for mixed waste treatment were the arc melters with graphite electrode systems with appropriate cooling which reduced maintenance and the possibility of eruptions that have occurred with plasma torches. Arc melters can also result in the minimum footprint and shielding. The preferred offgas systems were the wet/dry systems, that essentially eliminate the formation of carcinogenic compounds so they do not have to be destroyed down stream. This system also puts all of the particulate matter into one stream, instead of two.

  16. Processing results of 1800 gallons of mercury and radioactively contaminated mixed waste rinse solution

    International Nuclear Information System (INIS)

    Thiesen, B.P.

    1993-01-01

    Mercury-contaminated rinse solution was successfully treated at the Idaho National Engineering Laboratory. This waste was generated during the decontamination of the Heat Transfer Reactor Experiment 3 reactor shield tank. Approximately 6.8 m 3 (1,800 pi) of waste was generated and placed into 33 drums. Each drum contained precipitated sludge material ranging from 2--5 cm in depth, with the average depth of about 6 cm. The pH of each drum varied from 3--11. The bulk liquid waste had a mercury level of 7.0 mg/l, which exceeded the Resource Conservation and Recovery Act limit of 0.2 mg/l. The average liquid bulk radioactivity was about 2.1 pCi/mL while the average sludge contamination was about 13,800 pCi/g. Treatment of the waste required separation of the liquid from the sludge, filtration, pH adjustment, and ion exchange. The resulting solution after treatment had mercury levels at 0.0186 mg/l and radioactivity of 0.282 pCi/ml

  17. Transportation risk assessment of radioactive wastes generated by the N-Reactor stabilization program at the Hanford Site, Washington

    International Nuclear Information System (INIS)

    Wheeler, T.

    1994-12-01

    The potential radiological and nonradiological risks associated with specific radioactive waste shipping campaigns at the Hanford Site are estimated. The shipping campaigns analyzed are associated with the transportation of wastes from the N-Reactor site at the 200-W Area, both within the Hanford Reservation, for disposal. The analysis is based on waste that would be generated from the N-Reactor stabilization program

  18. Environmental and other evaluations of alternatives for long-term management of stored INEL transuranic waste

    International Nuclear Information System (INIS)

    1979-02-01

    This study identifies, develops, and evaluates, in a preliminary manner, alternatives for long-term management of TRU waste stored at the Radioactive Waste Management Complex (RWMC) at the INEL. The evaluations concern waste currently at the RWMC and waste expected to be received by the beginning of the year 1985. The effects of waste that might be received after that date are addressed in an appendix. The technology required for managing the waste, the environmental effects, the risks to the public, the radiological and nonradiological hazards to workers, and the estimated costs are discussed

  19. Environmental and other evaluations of alternatives for long-term management of stored INEL transuranic waste

    International Nuclear Information System (INIS)

    1979-12-01

    This study identifies, develops, and evaluates, in a preliminary manner, alternatives for long-term management of TRU waste stored at the Radioactive Waste Management Complex (RWMC) at the INEL. The evaluations concern waste currently at the RWMC and waste expected to be received by the beginning of the year 1985. The effects of waste that might be received after that data are addressed in an appendix. The technology required for managing the waste, the environmental effects, the risks to the public, the radiological and nonradiological hazards to workers, and the estimated costs are discussed

  20. Environmental and other evaluations of alternatives for long-term management of stored INEL transuranic waste

    Energy Technology Data Exchange (ETDEWEB)

    1979-02-01

    This study identifies, develops, and evaluates, in a preliminary manner, alternatives for long-term management of TRU waste stored at the Radioactive Waste Management Complex (RWMC) at the INEL. The evaluations concern waste currently at the RWMC and waste expected to be received by the beginning of the year 1985. The effects of waste that might be received after that date are addressed in an appendix. The technology required for managing the waste, the environmental effects, the risks to the public, the radiological and nonradiological hazards to workers, and the estimated costs are discussed.

  1. Safety assessment of Novi Han radioactive waste repository - features, problems, results and perspectives

    International Nuclear Information System (INIS)

    Mateeva, M.

    2000-01-01

    This paper summarizes the work done and the achievements reached in the Novi Han radioactive waste repository safety assessment within the IAEA Model Project 'Increasing the safety of Novi Han radioactive waste repository BUL 4/005'. The overall safety assessment has a wide context, but the work reported here relates only to some details and results concerning the development and implementation of the appropriate methodology approach, model and computer code used for the calculations. Different steps and procedures are included for a better practical understanding of the obtained results during the safety assessment performance. The methodology approach is widely based on an international experience in safety analysis and implemented for evaluation computer code AMBER, which is one of the recommended from the safety assessments experts. (author)

  2. DEWATERING TREATMENT SCALE-UP TESTING RESULTS OF HANFORD TANK WASTES

    International Nuclear Information System (INIS)

    TEDESCHI AR

    2008-01-01

    This report documents CH2M HILL Hanford Group Inc. (CH2M HILL) 2007 dryer testing results in Richland, WA at the AMEC Nuclear Ltd., GeoMelt Division (AMEC) Horn Rapids Test Site. It provides a discussion of scope and results to qualify the dryer system as a viable unit-operation in the continuing evaluation of the bulk vitrification process. A 10,000 liter (L) dryer/mixer was tested for supplemental treatment of Hanford tank low-activity wastes, drying and mixing a simulated non-radioactive salt solution with glass forming minerals. Testing validated the full scale equipment for producing dried product similar to smaller scale tests, and qualified the dryer system for a subsequent integrated dryer/vitrification test using the same simulant and glass formers. The dryer system is planned for installation at the Hanford tank farms to dry/mix radioactive waste for final treatment evaluation of the supplemental bulk vitrification process

  3. Waste Preparation and Transport Chemistry: Results of the FY 2001 Studies

    Energy Technology Data Exchange (ETDEWEB)

    Hunt, R.D.

    2002-03-25

    of researchers from AEA Technology, Florida International University (FIU), Fluor Hanford, Mississippi State University (MSU), Oak Ridge National Laboratory (ORNL), and Savannah River Technology Center (SRTC) to evaluate various aspects of the waste preparation and transport chemistry. The majority of this effort was focused on saltcake dissolution and saltwell pumping. The results of the AEA Technology, FIU, and MSU studies of saltcake dissolution and slurry transfers for Hanford are discussed in detail in a companion report prepared by T. D. Welch in 2001 (ORNIJTM-2001097). Staff members at Fluor Hanford have continued to conduct saltcake dissolution tests on actual tank waste (documented in reports prepared by D. L. Herting in 2000 and 2001). It should be noted that full-scale saltcake dissolution at Hanford is scheduled to begin in FY 2002. While the Hanford effort is focused on the transfer of waste from one tank to another, the objective of the SRTC study is the formation of aluminosilicates at elevated temperatures, which are present in the waste evaporator.

  4. Preliminary characterization of risks in the nuclear waste management system based on information in the literature

    International Nuclear Information System (INIS)

    Daling, P.M.; Rhoads, R.E.; Van Luick, A.E.; Fecht, B.A.; Nilson, S.A.; Sevigny, N.L.; Armstrong, G.R.; Hill, D.H.; Rowe, M.; Stern, E.

    1992-01-01

    This document presents preliminary information on the radiological and nonradiological risks in the nuclear waste management system. The objective of the study was to (1) review the literature containing information on risks in the nuclear waste management system and (2) use this information to develop preliminary estimates of the potential magnitude of these risks. Information was collected on a broad range of risk categories to assist the US Department of Energy (DOE) in communicating information about the risks in the waste management systems. The study examined all of the portions of the nuclear waste management system currently expected to be developed by the DOE. The scope of this document includes the potential repository, the integral MRS facility, and the transportation system that supports the potential repository and the MRS facility. Relevant literature was reviewed for several potential repository sites and geologic media. A wide range of ''risk categories'' are addressed in this report: (1) public and occupational risks from accidents that could release radiological materials, (2) public and occupational radiation exposure resulting from routine operations, (3) public and occupational risks from accidents involving hazards other than radioactive materials, and (4) public and occupational risks from exposure to nonradioactive hazardous materials during routine operations. The report is intended to provide a broad spectrum of risk-related information about the waste management system. This information is intended to be helpful for planning future studies

  5. Cost effectiveness of below-threshold waste disposal at DOE sites

    International Nuclear Information System (INIS)

    Smith, C.F.; Cohen, J.J.

    1987-01-01

    A minimal health and environmental risk, limitations on disposal capacity, and the relatively high costs of low level waste (LLW) disposal are basic driving forces that lead to consideration of less restrictive disposal of wastes with very low levels of radiological contamination. The term threshold limit describes radioactive wastes that have sufficiently low-levels of radiological content to be managed according to their nonradiological properties. Given the efforts described elsewhere to provide guidance on the definition of below threshold (BT) doses and concentration levels, the purpose of this study was to quantify the resultant quantities, costs and cost effectiveness of BT disposal. For purposes of consistency with the previous demonstrations of the application of the threshold concept, available data for waste streams at the Idaho National Engineering Laboratory (INEL) and the Savannah River Plant (SRP) sites were collected and analyzed with regard to volumes, radionuclide concentrations, and disposal costs. From this information, quantities of BT waste, potential cost savings and cost effectiveness values were estimated. 1 reference, 5 tables

  6. Methods and results of a probabilistic risk assessment for radioactive waste transports

    International Nuclear Information System (INIS)

    Lange, F.; Gruendler, D.; Schwarz, G.

    1993-01-01

    The radiological risk from accidents has been analyzed for the expected annual transport volume (3400 shipping units) of low and partially intermediate level radioactive wastes to be shipped to a final repository. In order to take account of these variable quantities and conditions a computer code was developed to simulate a wide spectrum of waste transport and accident configurations using Monte Carlo sampling techniques. Typically some 10.000 source terms were generated to represent possible releases of radionuclides from transport accidents. Accident events in which the integrity of waste packagings is retained and consequently no releases occur are included. Potential radiological consequences are then calculated for each of the release categories by using an accident consequence code which takes into account atmospheric dispersion statistics. Finally cumulative complementary frequency distributions of radiological consequences are generated by superposing the results for all release categories. Radiological consequences are primarily expressed as potential effective individual doses resulting from airborne and deposited radionuclides. The results of the risk analysis show that expected frequencies of effective doses comparable to the natural radiation exposure of one year are quite low and very low for potential radiation exposures in the range of 50 mSv. (J.P.N.)

  7. Predisposal management of low and intermediate level radioactive waste. Safety guide

    International Nuclear Information System (INIS)

    2003-01-01

    Radioactive waste is generated in the generation of electricity in nuclear power reactors and in the use of radioactive material in industry, research and medicine. The importance of the safe management of radioactive waste for the protection of human health and the environment has long been recognized. The principles and requirements that govern the safety of the management of radioactive waste are presented in 'The Principles of Radioactive Waste Management', 'Legal and Governmental Infrastructure for Nuclear, Radiation, Radioactive Waste and Transport Safety' and 'Predisposal Management of Radioactive Waste, Including Decommissioning'. The objective of this Safety Guide is to provide regulatory bodies and the operators that generate and manage radioactive waste with recommendations on how to meet the principles and requirements established in Refs for the predisposal management of LLW. This Safety Guide deals with the safety issues associated with the predisposal management of LLW from nuclear fuel cycle facilities, large research and development installations and radioisotope production facilities. This includes all steps and activities in the management of waste, from its initial generation to its final acceptance at a waste disposal facility or the removal of regulatory control. The predisposal management of radioactive waste includes decommissioning. The term 'decommissioning' encompasses both the process of decommissioning a facility and the management of the waste that results (prior to its disposal). Recommendations on the process of decommissioning are provided in Refs. Recommendations on the management of the waste resulting from decommissioning are included in this Safety Guide. Although the mining and milling of uranium and thorium ores is part of the nuclear fuel cycle, the management of the operational waste (e.g. waste rock, tailings and effluent treatment waste) from these activities is not within the scope of this Safety Guide. The LLW that is

  8. The role and the results of the European Community's R and D work on radioactive waste management

    International Nuclear Information System (INIS)

    Orlowski, S.; Girardi, F.

    1986-01-01

    The role and results of the European Community's research and development (R and D) work on radioactive waste management are described. The R and D work includes: radioactive waste conditioning, characterization and storage, materials science studies for the storage, geological media confinement studies, and radionuclide migration investigations. Financial management and the long term, and the socio-political aspects of waste management, are also discussed. (U.K.)

  9. Solid Waste Management Requirements Definition for Advanced Life Support Missions: Results

    Science.gov (United States)

    Alazraki, Michael P.; Hogan, John; Levri, Julie; Fisher, John; Drysdale, Alan

    2002-01-01

    Prior to determining what Solid Waste Management (SWM) technologies should be researched and developed by the Advanced Life Support (ALS) Project for future missions, there is a need to define SWM requirements. Because future waste streams will be highly mission-dependent, missions need to be defined prior to developing SWM requirements. The SWM Working Group has used the mission architecture outlined in the System Integration, Modeling and Analysis (SIMA) Element Reference Missions Document (RMD) as a starting point in the requirement development process. The missions examined include the International Space Station (ISS), a Mars Dual Lander mission, and a Mars Base. The SWM Element has also identified common SWM functionalities needed for future missions. These functionalities include: acceptance, transport, processing, storage, monitoring and control, and disposal. Requirements in each of these six areas are currently being developed for the selected missions. This paper reviews the results of this ongoing effort and identifies mission-dependent resource recovery requirements.

  10. Uncertainty and sensitivity results for pre-waste-emplacement groundwater travel time

    International Nuclear Information System (INIS)

    Kaplan, P.G.

    1992-01-01

    In this paper uncertainty and sensitivity analyses for pre-waste-emplacement groundwater travel time conducted. Although preliminary, a numbed of interesting results were obtained. Uncertainty in the ground water travel time statistics, as measured by the coefficient of variation, increases and then decrease as the modeled system transitions from matrix-dominated to fracture-dominated flow. The uncertainty analysis also suggests that the median, as opposed to the mean, may be a better indicator of performance with respect to the regulatory criterion. The sensitivity analysis shows a strong correlation between an effective fracture property, fracture porosity, and failure to meet the regulatory pre-waste-emplacement groundwater travel time criterion of 1,000 years

  11. Radioactive waste storage in mined caverns in crystalline rock: results of field investigations at Stripa, Sweden

    International Nuclear Information System (INIS)

    Witherspoon, P.A.

    1980-10-01

    It is generally agreed that the most practicable method of isolating nuclear wastes from the biosphere is by deep burial in suitable geologic formations. Such burial achieves a high degree of physical isolation but raises questions concerning the rate at which some of these wastes may return to the biosphere through transport by groundwater. Any suitable repository site will be disturbed first by excavation and second by the thermal pulse caused by the radioactive decay of the wastes. To assess the effectiveness of geologic isolation it is necessary to develop the capability of predicting the response of a rock mass to such a thermal pulse. Ultimately, this requires field measurements below the surface in media representative of those likely to be encountered at an actual repository. Access to a granitic rock mass adjacent to a defunct iron ore mine at Stripa, Sweden, at a depth of about 350 m below surface has provided a unique opportunity to conduct a comprehensive suite of hydrological and thermo-mechanical experiments under such conditions. The results of these field tests have shown the importance of geologic structure and the functional dependence of the thermo-mechanical properties on temperature in developing a valid predictive model. The results have also demonstrated the vital importance of carrying out large-scale investigations in a field test facility

  12. Improved Process Used to Treat Aqueous Mixed Waste Results in Cost Savings and Improved Worker Safety

    International Nuclear Information System (INIS)

    Hodge, D.S.; Preuss, D.E.; Belcher, K.J.; Rock, C.M.; Bray, W.S.; Herman, J.P.

    2006-01-01

    This paper describes an improved process implemented at Argonne National Laboratory (ANL) to treat aqueous mixed waste. This waste is comprised of radioactively-contaminated corrosive liquids with heavy metals. The Aqueous Mixed Waste Treatment System (AMWTS) system components include a reaction tank and a post-treatment holding tank with ancillary piping and pumps; and a control panel with pumping/mixing controls; tank level, temperature and pH/Oxidation Reduction Potential (ORP) indicators. The process includes a neutralization step to remove the corrosive characteristic, a chromium reduction step to reduce hexavalent chromium to trivalent chromium, and a precipitation step to convert the toxic metals into an insoluble form. Once the toxic metals have precipitated, the resultant sludge is amenable to stabilization and can be reclassified as a low-level waste if the quantity of leachable toxic metals, as determined by the TCLP, is below Universal Treatment Standards (UTS). To date, six batches in eight have passed the UTS. The AMWTS is RCRA permitted and allows for the compliant treatment of mixed waste prior to final disposal at a Department of Energy (DOE) or commercial radioactive waste disposal facility. Mixed wastes eligible for treatment include corrosive liquids (pH 12.5) containing EPA-regulated toxic metals (As, Ba, Pb, Cd, Cr, Ag, Se, Hg) at concentrations greater than the RCRA Toxicity Characteristic Leaching Procedure (TCLP) limit. The system has also been used to treat corrosive wastes with small quantities of fissionable materials. The AMWTS is a significant engineered solution with many improvements over the more labor intensive on-site treatment method being performed within a ventilation hood used previously. The previously used treatment system allowed for batch sizes of only 15-20 gallons whereas the new AMWTS allows for the treatment of batches up to 75 gallons; thereby reducing batch labor and supply costs by 40-60% and reducing analytical

  13. Results from five years of treatability studies using hydraulic binders to stabilize low-level mixed waste at the INEL

    International Nuclear Information System (INIS)

    Gering, K.L.; Schwendiman, G.L.

    1997-01-01

    This paper summarizes work involving bench-scale solidification of nonincinerable, land disposal restricted low-level mixed waste. Waste forms included liquids, sludges, and solids; treatment techniques included hydraulic systems (Portland cement with and without additives), proprietary commercial formulations, and sulphur polymer cement. Solidification was performed to immobilize hazardous heavy metals (including mercury, lead, chromium, and cadmium), and volatile and semivolatile organic compounds. Pretreatment options for mixed wastes are discussed, using a decision tree based on the form of mixed waste and the type of hazardous constituents. Hundreds of small concrete monoliths were formed for a variety of waste types. The experimental parameters used for the hydraulic concrete systems include the ratio of waste to dry binder (Portland cement, proprietary materials, etc.), the total percentage of water in concrete, and the amount of concrete additives. The only parameter that was used for the sulfur polymer-based monoliths is ratio of waste to binder. Optimum concrete formulations or open-quotes recipesclose quotes for a given type of waste were derived through this study, as based on results from the Toxicity Characteristic Leaching Procedure analyses and a free liquids test. Overall results indicate that high waste loadings in the concrete can be achieved while the monolithic mass maintains excellent resistance to leaching of heavy metals. In our study the waste loadings in the concrete generally fell within the range of 0.5 to 2.0 kg mixed waste per kg dry binder. Likewise, the most favorable amount of water in concrete, which is highly dependent upon the concrete constituents, was determined to be generally within the range of 300 to 330 g/kg (30-33% by weight). The results of this bench-scale study will find applicability at facilities where mixed or hazardous waste solidification is a planned or ongoing activity. 19 refs., 1 fig., 5 tabs

  14. Transport of radioactive wastes to the planned final waste repository Konrad: Radiation exposure resulting from normal transport and radiological risks from transport accidents

    International Nuclear Information System (INIS)

    Lange, F.; Fett, H.J.; Gruendler, D.; Schwarz, G.

    1993-01-01

    Radiation exposures of members of critical groups of the general population and of transport personnel resulting from normal transport of radioactive wastes to the planned final waste repository Konrad have been evaluated in detail. By applying probabilistic safety assessment techniques radiological risks from transport accidents have been analysed by quantifying potential radiation exposures and contaminations of the biosphere in connection with their expected frequencies of occurrence. The Konrad transport study concentrates on the local region of the waste repository, where all transports converge. (orig.) [de

  15. Wastes

    International Nuclear Information System (INIS)

    Bovard, Pierre

    The origin of the wastes (power stations, reprocessing, fission products) is determined and the control ensuring the innocuity with respect to man, public acceptance, availability, economics and cost are examined [fr

  16. The role and results of the European Community's R and D work on radioactive waste management

    International Nuclear Information System (INIS)

    Orlowski, S.; Girardi, F.

    1985-01-01

    The titles of R and D programmes generally relate to a scientific discipline, a technology or a project: biotechnology, nuclear fission, etc. This is not so in the case of radioactive wastes, where R and D is focused on the management aspect. The role of R and D in general, and the contribution made by the Community programme in particular, are described and discussed with this in mind. Community R and D in the field of radioactive waste emerges as a powerful tool for establishing a broad consensus on delicate scientific questions such as the feasibility and long-term safety of the final storage of high activity wastes. Such a consensus is based on the many results obtained jointly by Community research teams over the last ten years. The implementation of three projects concerning experimental underground facilities in the context of the Community's new five-year (1985-1989) programme will provide the additional information that is needed before the large industrial disposal facilities of the future can be built

  17. Fertilization effects of organic waste resources and bottom wood ash: results from a pot experiment

    Directory of Open Access Journals (Sweden)

    Eva Brod

    2012-12-01

    Full Text Available We conducted a pot experiment to study the fertilization effects of four N- and P-rich organic waste resources alone and in combination with K-rich bottom wood ash at two application rates (150 kg N ha–1 + 120 kg K ha–1, 300 kg N ha-1 + 240 kg K ha–1. Plant-available N was the growth-limiting factor. 48–73% of N applied with meat and bone meal (MBM and composted fish sludge (CFS was taken up in aboveground biomass, resulting in mineral fertilizer equivalents (MFE% of 53–81% for N uptake and 61–104% for yield. MFE% of MBM and CFS decreased for increasing application rates. Two industrial composts had weak N fertilization effects and are to be considered soil conditioners rather than fertilizers. Possible P and K fertilization effects of waste resources were masked by the soil’s ability to supply plant-available P and K, but effects on plant-available P and K contents in soil suggest that the waste resources may have positive effects under more nutrient-deficient conditions.

  18. Management of the radioactive waste resulting from the Romanian VVR-S research reactor decommissioning

    International Nuclear Information System (INIS)

    Ene, D.; Cepraga, D.G.

    2002-01-01

    The paper consists in a waste study of the Romanian VVR-S reactor which will be prepared for decommissioning operations after the permanent shutdown (23.12.1997). Calculations were carried out to determine the activity arising from neutron activation of structural materials inside the reactor, considering the design of the facility and its operating rules. To this end, the following method was used: i) Neutron flux distribution within the reactor was calculated using the DORT transport code, based on DLC23 shielding library relating to three cylindrical reference systems of the reactor structure: reactor core, horizontal tube and thermal column; ii) Calculation of the activity of each reactor component at different cooling times was performed by the ANITA2000 code, using the neutron flux, compositional data for each material and the power history of the reactor; iii) Unconditional clearance indexes for all material at various cooling times were calculated using the clearance levels defined in IAEA-TECDOC-855; iv) Total activities and masses by material type, within the waste category and for each decay time were calculated by summation of the data previously classified for each reactor component. The resulting activation inventory and waste masses, falling in IAEA defined waste categories are presented in the paper at periods of 100 days, and 6, 10, 25, and 50 years after reactor the shutdown. For some components of the reactor as: aluminum central vessel, the central iron shielding ring, the time behaviour of both the fin spatial activity distribution and the radionuclide contributions to the total activity are plotted in the paper. (author)

  19. Development of radiological performance objectives interim results: trade-offs in attitudes toward radioactive waste

    International Nuclear Information System (INIS)

    Lathrop, J.W.

    1978-07-01

    In order to measure the risk associated with radioactive waste it is necessary to ascertain public opinion concerning the relative significance of the different possible health effects of radiation, and public attitudes towards uncertainty. LLL has directed Decisions and Designs, Incorporated (DDI), to elicit such views from various members of the public. Purpose of this note is to give a brief account of some of the views so far obtained, provide some interpretation of these results, and briefly demonstrate how these results can be used to guide the drafting of regulations

  20. The Behaviours of Cementitious Materials in Long Term Storage and Disposal of Radioactive Waste. Results of a Coordinated Research Project

    International Nuclear Information System (INIS)

    2013-09-01

    Radioactive waste with widely varying characteristics is generated from the operation and maintenance of nuclear power plants, nuclear fuel cycle facilities, research laboratories and medical facilities. This waste must be treated and conditioned, as necessary, to provide waste forms acceptable for safe storage and disposal. Many countries use cementitious materials (concrete, mortar, etc.) as a containment matrix for immobilization, as well as for engineered structures of disposal facilities. Radionuclide release is dependent on the physicochemical properties of the waste forms and packages, and on environmental conditions. In the use of cement, the diffusion process and metallic corrosion can induce radionuclide release. The advantage of cementitious materials is the added stability and mechanical support during storage and disposal of waste. Long interim storage is becoming an important issue in countries where it is difficult to implement low level waste and intermediate level waste disposal facilities, and in countries where cement is used in the packaging of waste that is not suitable for shallow land disposal. This coordinated research project (CRP), involving 24 research organizations from 21 Member States, investigated the behaviour and performance of cementitious materials used in an overall waste conditioning system based on the use of cement - including waste packaging (containers), waste immobilization (waste form) and waste backfilling - during long term storage and disposal. It also considered the interactions and interdependencies of these individual elements (containers, waste, form, backfill) to understand the processes that may result in degradation of their physical and chemical properties. The main research outcomes of the CRP are summarized in this report under four topical sections: (i) conventional cementitious systems; (ii) novel cementitious materials and technologies; (iii) testing and waste acceptance criteria; and (iv) modelling long

  1. The Behaviours of Cementitious Materials in Long Term Storage and Disposal of Radioactive Waste. Results of a Coordinated Research Project

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-09-15

    Radioactive waste with widely varying characteristics is generated from the operation and maintenance of nuclear power plants, nuclear fuel cycle facilities, research laboratories and medical facilities. This waste must be treated and conditioned, as necessary, to provide waste forms acceptable for safe storage and disposal. Many countries use cementitious materials (concrete, mortar, etc.) as a containment matrix for immobilization, as well as for engineered structures of disposal facilities. Radionuclide release is dependent on the physicochemical properties of the waste forms and packages, and on environmental conditions. In the use of cement, the diffusion process and metallic corrosion can induce radionuclide release. The advantage of cementitious materials is the added stability and mechanical support during storage and disposal of waste. Long interim storage is becoming an important issue in countries where it is difficult to implement low level waste and intermediate level waste disposal facilities, and in countries where cement is used in the packaging of waste that is not suitable for shallow land disposal. This coordinated research project (CRP), involving 24 research organizations from 21 Member States, investigated the behaviour and performance of cementitious materials used in an overall waste conditioning system based on the use of cement - including waste packaging (containers), waste immobilization (waste form) and waste backfilling - during long term storage and disposal. It also considered the interactions and interdependencies of these individual elements (containers, waste, form, backfill) to understand the processes that may result in degradation of their physical and chemical properties. The main research outcomes of the CRP are summarized in this report under four topical sections: (i) conventional cementitious systems; (ii) novel cementitious materials and technologies; (iii) testing and waste acceptance criteria; and (iv) modelling long

  2. Development of a comprehensive radioactive waste classification system

    International Nuclear Information System (INIS)

    Smith, C.F.; Cohen, J.J.

    1989-01-01

    Several previous studies have been conducted with the intent of developing a rational system for classification of radioactive wastes. Although none of the proposed systems has gained general acceptance, certain waste classes, specifically high-level waste and low-level waste suitable for shallow land burial have been essentially defined by regulation. Wastes which remain undefined include: those intermediate level wastes which require more restrictive controls than that provided by shallow land burial but not the high degree of isolation needed for high level wastes, and wastes below regulatory concern (BRC) which entail so low a radiological risk that they can be managed according to their nonradiological properties. This study has developed a framework within which the complete spectrum of radioactive wastes can be defined

  3. Environmental impact assessment and socio political issues of nuclear waste management; Ydinjaetehuollon ympaeristoevaikutusten arviointi ja sosiopoliittiset kysymykset

    Energy Technology Data Exchange (ETDEWEB)

    Harmaajaervi, I; Tolsa, H [VTT Communities and Infrastructure, Espoo (Finland). Urban Planning; Vuori, S [VTT Energy, Espoo (Finland). Nuclear Energy; Litmanen, T [Jyvaeskylae Univ. (Finland)

    1997-09-01

    The study is a part of the Publicly Administrated Nuclear Waste Management Research Programme (JYT2) which was carried out in 1994-1996. The principal goal of the research programme has been to provide the authorities with information and research results relevant for the safety of nuclear waste management in order to support the various activities of the authorities. The main emphasis of the research programme focuses on the disposal of spent fuel. In addition to nuclear waste research in the field of natural sciences and technology, the research program- me has focused mostly on societal issues associated with nuclear waste disposal facilities and on the non-radiological environmental effects in the environs of the disposal site. Some of the local effects are already revealed in the research phase, before any final decisions are made as to the selection of the disposal site. The study has focused primarily on local and regional issues. The statutory requirement to conduct environ- mental impact assessment (EIA) chiefly concerns those who are responsible for waste management, but the authorities also need to acquire systematic information in the field to support developing requirements for the content and scope of EIA procedure and preparedness to check the assessments made. This is a report of the first parts of the study in 1994-1995. The report deals with the subject matter generally based on earlier studies in Finland and other countries. The results of the study will be reported later. 101 refs.

  4. Quarry waste management and recovery: first results connected to Carrara marble ravaneti (Italy)

    Science.gov (United States)

    Antonella Dino, Giovanna; Chiappino, Claudia; Rossetti, Piergiorgio

    2017-04-01

    Quarry waste (QW) represents a huge economic and environmental issue, due to loss of resources and to economic and environmental costs connected to waste management and landfilling activities. In many cases, valuable Raw Materials (RM) and Secondary Raw Materials (SRM) can be supplied by enhancing the QW recovery. In Italy large amounts of QW have been and still are dumped: such materials, if their quality (chemical, mineralogical, physical characteristics) and quantity are adequate, and if the impacts connected to their management are positive, can represent a valuable resource for SRM exploitation. Several dimension stone quarries have been and are interested by researches as for QW exploitation. Some researches show positive results, which are the basis for QW recovery (both from waste streams and from quarry dumps exploitation): a noticeable example is represented by Carrara marble waste. The Carrara quarry basin is characterized by ca. one hundred quarries for colored and white marble exploitation. The waste production can be summarized in: 80 Mm3 waste present in old quarry dumps (Ravaneti) and 3 Mm3/y of waste stream from quarrying activities. At present only 0.5 Mm3/y of QW is exploited for SRM production, causing a huge loss of resource. This has been the background for a preliminary research, on Carrara marble Ravaneti characterization, which was carried out thanks to the close cooperation between University of Torino, Società Apuana Marmi srl, and SET srl. In 2015, two QW dumping areas, Calocara and Lorano, were selected as representative for sampling activities. Three main sample categories were individuated based on granulometry (0.5-4 mm, 0-25 mm, 0-150 mm) to be characterized (size distribution, density, Atterberg limits, Los Angeles test, freezing and heat tests, flat and shape indexes, geochemistry, mineralogy). The results obtained are promising: the physical characterization shows an attitude for Carrara QW to be recovered as crushed materials

  5. The influence of slaughterhouse waste on fermentative H{sub 2} production from food waste: Preliminary results

    Energy Technology Data Exchange (ETDEWEB)

    Boni, Maria Rosaria; Sbaffoni, Silvia; Tuccinardi, Letizia, E-mail: letizia.tuccinardi@uniroma1.it

    2013-06-15

    Highlights: • Co-digestion process finalized to bio-H{sub 2} production was tested in batch tests. • Slaughterhouse waste (SHW) and food waste (FW) were co-digested in different proportions. • The presence of SHW affected the H{sub 2} production from FW. • When SHW ranging between 50% and 70% the H{sub 2} production is improved. • SHW percentages above 70%, led to a depletion in H{sub 2} production. - Abstract: The aim of this study was to evaluate the influence of slaughterhouse waste (SHW; essentially the skin, fats, and meat waste of pork, poultry, and beef) in a fermentative co-digestion process for H{sub 2} production from pre-selected organic waste taken from a refectory (food waste [FW]). Batch tests under mesophilic conditions were conducted in stirred reactors filled with different proportions of FW and SHW. The addition of 60% and 70% SHW to a mixture of SHW and FW improved H{sub 2} production compared to that in FW only, reaching H{sub 2}-production yields of 145 and 109 ml gVS{sub 0}{sup -1}, respectively, which are 1.5–2 times higher than that obtained with FW alone. Although the SHW ensured a more stable fermentative process due to its high buffering capacity, a depletion of H{sub 2} production occurred when SHW fraction was higher than 70%. Above this percentage, the formation of foam and aggregated material created non-homogenous conditions of digestion. Additionally, the increasing amount of SHW in the reactors may lead to an accumulation of long chain fatty acids (LCFAs), which are potentially toxic for anaerobic microorganisms and may inhibit the normal evolution of the fermentative process.

  6. Thermally induced motion of marine sediments resulting from disposal of radioactive wastes

    International Nuclear Information System (INIS)

    Chavez, P.F.; Dawson, P.R.

    1981-01-01

    Coupled creep and heat transfer calculations have been performed to assess the sensitivity of heat load, viscosity, and canister density on the motion of waste canisters buried in marine sediments. Results indicate that no upward movement is predicted for heat loads remaining within the metallurgical and geochemical constraints placed on the temperature of sediments near the canister for the times analyzed. Upward movement of the canister is again not observed in calculations involving reasonable variations of the sediment viscosity and canister density. Maximum effective deviatoric stress levels due to thermally induced differential body forces are significantly less than the sediment's short term peak strength

  7. Characterization of Secondary Solid Wastes in Trench Water in Waste Area Grouping 6 at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    Taylor, P.A.; Kent, T.E.

    1994-02-01

    This project was undertaken to demonstrate that new liquid waste streams, generated as a consequence of closure activities at Waste Area Grouping (WAG) 6 and other sites, can be treated at the existing wastewater treatment facilities at Oak Ridge National Laboratory (ORNL) to meet discharge requirements without producing hazardous secondary solid wastes. Previous bench and pilot-scale treatability studies have shown that ORNL treatment operations will adequately remove the contaminants and that the secondary solid wastes produced were not hazardous when treating water from two trenches in WAG 6. This study used WAG 6 trench water spiked with the minimum concentration of Toxicity Characteristics Leaching Procedure (TCLP) constituents (chemicals that can make a waste hazardous) found in any groundwater samples at ORNL. The Wastewater Treatment Test Facility (WTTF), a 0.5 L/min pilot plant that simulates the treatment capabilities of the Process Waste Treatment Plant (PWPT) and Nonradiological Wastewater Treatment Plant (NRWTP), was used for this test. This test system, which is able to produce secondary wastes in the quantities necessary for TCLP testing, was operated for a 59-d test period with a minimum of problems and downtime. The pilot plant operating data verified that WAG 6 trench waters, spiked with the minimum concentration of TCLP contaminants measured to date, can be treated at the PWTP and NRWTP to meet current discharge limits. The results of the TCLP analysis indicated that none of the secondary solid wastes produced during the treatment of these wastewaters will be considered hazardous as defined by the Resource Conservation and Recovery Act

  8. Results of two years' operation of the waste processing cell PROLIXE

    International Nuclear Information System (INIS)

    Lecomte, M.; Madic, C.; Broudic, J.C.

    1990-01-01

    Solid wastes, contaminated by alpha, beta, gamma radioisotopes, are produced by spent fuel reprocessing and isotope production. The PROLIXE plant, prototype for leaching and encapsulation was put into operation in March 1988 for waste management with the following aims: development of decontamination by oxidative leaching of alpha wastes, to obtain less than 0.1 Ci/t for surface storage; recycling radioactive isotope recovered especially transuranium elements; define a versatile process for various solid radioactive waste for an industrial plant [fr

  9. Treatment and conditioning of low-level radioactive waste in Belgium: initial operating results of the Cilva facility

    International Nuclear Information System (INIS)

    Monsch, O.; Renard, C.; Deckers, J.; Luycx, P.

    1995-01-01

    The Belgian National Radioactive Waste and Enriched Fissile Material Agency (ONDRAF), which is responsible for the management of all radioactive waste in Belgium, recently decided to commission the CILVA facility. Operation of this facility, which comprises a number of units for the treatment of low-level radwaste, has been contracted to ONDRAF's Belgoprocess subsidiary based at the Dessel site. A consortium comprising SGN and Fabricom was in charge of building the CILVA facility's waste preparation and conditioning (concrete solidification) units. The concrete solidification processes, which were devised and developed by SGN, have been qualified to secure ONDRAF certification of the process and the facility. This enabled active commissioning of the waste conditioning unit in mid-August 1994. Active commissioning of the waste preparation unit was carried out in several stages up to the beginning of 1995 in accordance with operating requirements. Initial operating results of the two units are presented. (author)

  10. Effect of COSMOS technologies in detoxifying municipal solid waste incineration fly ash, preliminary results

    Science.gov (United States)

    Piccinelli, Elsa; Lasagni, Marina; Collina, Elena; Bonaiti, Stefania; Bontempi, Elza

    2017-05-01

    This study investigates the effect of technologies for heavy metal stabilization on the concentration of PolyChlorinatedDibenzo-p-Dioxins (PCDD) and PolyChlorinatedDibenzoFurans (PCDF), abbreviated PCDD/F, in Municipal Solid Waste Incineration (MSWI) fly ash. We determined the variation of the Total Organic Carbon (TOC) and PCDD/F concentration between raw and stabilized material. The technologies, that already proved to be very promising for heavy metal entrapment, showed encouraging results also for PCDD/F detoxification. This result could be very impacting on the management of MSWI fly ash: at the best of our knowledge, there are no methods, in literature, that can provide good results in stabilization of heavy metals, and abatement of chlorinated organic pollutants contained in the same matrix.

  11. Removing Phosphate from Hanford High-Phosphate Tank Wastes: FY 2010 Results

    Energy Technology Data Exchange (ETDEWEB)

    Lumetta, Gregg J.; Braley, Jenifer C.; Edwards, Matthew K.; Qafoku, Odeta; Felmy, Andrew R.; Carter, Jennifer C.; MacFarlan, Paul J.

    2010-09-22

    The U.S. Department of Energy (DOE) is responsible for environmental remediation at the Hanford Site in Washington State, a former nuclear weapons production site. Retrieving, processing, immobilizing, and disposing of the 2.2 × 105 m3 of radioactive wastes stored in the Hanford underground storage tanks dominates the overall environmental remediation effort at Hanford. The cornerstone of the tank waste remediation effort is the Hanford Tank Waste Treatment and Immobilization Plant (WTP). As currently designed, the capability of the WTP to treat and immobilize the Hanford tank wastes in the expected lifetime of the plant is questionable. For this reason, DOE has been pursuing supplemental treatment options for selected wastes. If implemented, these supplemental treatments will route certain waste components to processing and disposition pathways outside of WTP and thus will accelerate the overall Hanford tank waste remediation mission.

  12. Results of Toxicity Studies Conducted on Outfall X-08 and Its Contributing Waste Streams, November 1999 - June 2000

    International Nuclear Information System (INIS)

    Specht, W.L.

    2000-01-01

    This interim report summarizes the results of toxicity tests, Toxicity Identification Evaluations, and chemical analyses that have been conducted on SRS's NPDES Outfall X-08 and its contributing waste streams between November 1999 and June 2000

  13. Aspects on the acceptance of waste for disposal in SFR

    International Nuclear Information System (INIS)

    Torstenfelt, Boerje

    2006-01-01

    When licensing a final repository for radioactive waste certain assumptions have to be made concerning the waste. These assumptions cover radionuclide inventory and nonradiological materials and its physical and chemical impact on the waste, the repository and on the environment. Development of new waste treatment systems and waste packages at the waste producer site aim at finding solutions and products that can be stored, transported and disposed of safely and are economically sound. This paper discusses some aspects concerning development of new or modified waste products. It highlights the importance of analysing the whole sequence in treatment, handling and disposing the waste. The process should be to find an optimal solution for the whole system, considering the fact that what is best in one step it not necessary best for the whole system, including the post closure issues. (author)

  14. Radioactive waste discharges from UKAEA establishments during 1996 and associated monitoring results

    International Nuclear Information System (INIS)

    Morton, A.K.M.; Forbes, S.A.; Hughes, B.; Richardson, E.

    1997-08-01

    This annual report is published by the Safety Directorate of the United Kingdom Atomic Energy Authority (UKAEA) and provides information on radioactive discharges from its sites. The Culcheth site was closed and then redeveloped during the end of 1993 and the Springfields site became part of BNFL in October 1994. No operations involving the need to discharge radioactivity are undertaken at the Risley site. After discussions with the Authorising Departments at that time, the discharge authorisations were revoked on 1 July 1994. These sites are therefore no longer included in this report. UKAEA has published annual radioactive waste discharges and associated monitoring results since 1963. This report is intended to give a relatively short factual overview of UKAEA waste discharge and disposal, and its impact on the environment. Additional information may be found in annual discharge reports published by the individual UKAEA establishments and the UKAEA Report on Safety and the Environment 1996-97 due to be issued at the end of September 1997. (UK)

  15. Accelerator driven systems. ADS benchmark calculations. Results of stage 2. Radiotoxic waste transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Freudenreich, W.E.; Gruppelaar, H

    1998-12-01

    This report contains the results of calculations made at ECN-Petten of a benchmark to study the neutronic potential of a modular fast spectrum ADS (Accelerator-Driven System) for radiotoxic waste transmutation. The study is focused on the incineration of TRans-Uranium elements (TRU), Minor Actinides (MA) and Long-Lived Fission Products (LLFP), in this case {sup 99}Tc. The benchmark exercise is made in the framework of an IAEA Co-ordinated Research Programme. A simplified description of an ADS, restricted to the reactor part, with TRU or MA fuel (k{sub eff}=0.96) has been analysed. All spectrum calculations have been performed with the Monte Carlo code MCNP-4A. The burnup calculations have been performed with the code FISPACT coupled to MCNP-4A by means of our OCTOPUS system. The cross sections are based upon JEF-2.2 for transport calculations and supplemented with EAF-4 data for inventory calculations. The determined quantities are: core dimensions, fuel inventories, system power, sensitivity on external source spectrum and waste transmutation rates. The main conclusions are: The MA-burner requires only a small accelerator current increase during burnup, in contrast to the TRU-burner. The {sup 99} Tc-burner has a large initial loading; a more effective design may be possible. 5 refs.

  16. Accelerator driven systems. ADS benchmark calculations. Results of stage 2. Radiotoxic waste transmutation

    International Nuclear Information System (INIS)

    Freudenreich, W.E.; Gruppelaar, H.

    1998-12-01

    This report contains the results of calculations made at ECN-Petten of a benchmark to study the neutronic potential of a modular fast spectrum ADS (Accelerator-Driven System) for radiotoxic waste transmutation. The study is focused on the incineration of TRans-Uranium elements (TRU), Minor Actinides (MA) and Long-Lived Fission Products (LLFP), in this case 99 Tc. The benchmark exercise is made in the framework of an IAEA Co-ordinated Research Programme. A simplified description of an ADS, restricted to the reactor part, with TRU or MA fuel (k eff =0.96) has been analysed. All spectrum calculations have been performed with the Monte Carlo code MCNP-4A. The burnup calculations have been performed with the code FISPACT coupled to MCNP-4A by means of our OCTOPUS system. The cross sections are based upon JEF-2.2 for transport calculations and supplemented with EAF-4 data for inventory calculations. The determined quantities are: core dimensions, fuel inventories, system power, sensitivity on external source spectrum and waste transmutation rates. The main conclusions are: The MA-burner requires only a small accelerator current increase during burnup, in contrast to the TRU-burner. The 99 Tc-burner has a large initial loading; a more effective design may be possible. 5 refs

  17. Field Lysimeter Investigations - test results: Low-Level Waste Data Base Development Program: Test results for fiscal years 1994-1995

    International Nuclear Information System (INIS)

    McConnell, J.W. Jr.; Rodgers, R.D.; Hilton, L.D.; Neilson, R.M. Jr.

    1996-06-01

    The Field Lysimeter Investigations: Low-Level Waste Data Base Development Program, funded by the U.S. Nuclear Regulatory Commission (NRC), is (1) studying the degradation effects in EPICOR-II organic ion-exchange resins caused by radiation, (2) examining the adequacy of test procedures recommended in the Branch Technical Position on Waste Form to meet the requirements of 10 CFR 61 using solidified EPICOR-II resins, (3) obtaining performance information on solidified EPICOR-II ion-exchange resins in a disposal environment, and (4) determining the condition of EPICOR-II liners. Results of the final 2 (10 total) years of data acquisition from operation of the field testing are presented and discussed. During the continuing field testing, both portland type I-II cement and Dow vinyl ester-styrene waste forms are being tested in lysimeter arrays located at Argonne National Laboratory-East in Illinois and at Oak Ridge National Laboratory. The experimental equipment is described and results of waste form characterization using tests recommended by the NRC's open-quotes Technical Position on Waste Formclose quotes are presented. The study is designed to provide continuous data on nuclide release and movement, as well as environmental conditions, over a 20-year period. At the end of the tenth year, the experiment was closed down. Examination of soil and waste forms is planned to be conducted next and will be reported later

  18. Sealing a nuclear waste repository in Columbia river basalt: preliminary results

    International Nuclear Information System (INIS)

    Hodges, F.N.

    1980-01-01

    The long containment time required of repositories for nuclear waste (10 4 to 10 6 years) requires that materials used for repository seals be stable in the geologic environment of the repository and of proven longevity. A list of candidate materials for sealing a repository in Columbia River Basalts has been prepared and refined through laboratory testing. The most feasible techniques for emplacing preferred plug materials have been identified and the resultant plugs have been evaluated on the basis of design functions. Preconceptual designs for tunnel, shaft, and borehole seals consist of multiple zone plugs with each zone fulfilling one or more design functions. Zones of disturbed rock around tunnels and shafts, resulting from excavation and subsequent stress release, are zones of higher permeability and of possible fluid migration. In preliminary designs the disturbed zones are blocked by cut-off collars filled with low permeability materials

  19. Method and equipment for treating waste water resulting from the technological testing processes of NPP equipment

    International Nuclear Information System (INIS)

    Radulescu, M. C.; Valeca, S.; Iorga, C.

    2016-01-01

    Modern methods and technologies coupled together with advanced equipment for treating residual substances resulted from technological processes are mandatory measures for all industrial facilities. The correct management of the used working agents and of the all wastes resulted from the different technological process (preparation, use, collection, neutralization, discharge) is intended to reduce up to removal of their potential negative impact on the environment. The high pressure and temperature testing stands from INR intended for functional testing of nuclear components (fuel bundles, fuelling machines, etc.) were included in these measures since the use of oils, demineralized water chemically treated, greases, etc. This paper is focused on the method and equipment used at INR Pitesti in the chemical treatment of demineralized waters, as well as the equipment for collecting, neutralizing and discharging them after use. (authors)

  20. Results of technical and economical examinations for substantiation of special plant design for reprocessing and radioactive wastes disposal

    International Nuclear Information System (INIS)

    Galkin, A.V.; Baldov, A.N.

    2001-01-01

    In the paper the results of technical and economical examinations for substantiation of special plant design for reprocessing and radioactive wastes disposal are presented. Ground for the examination conducting was Health of Nation Programme ratified by the President and a number of Governmental decisions. The special plant is planned in the Mangystau Region. In the framework of feasibility study the data base by the worldwide known technologies was implemented, on reprocessing and experience of radioactive waste disposal. The technical requirements for the special plant construction are determined. The alternative options by structure content and site location of the special plant and radioactive waste disposal are cited

  1. Overview of french P and T programme and results for waste management

    International Nuclear Information System (INIS)

    Warin, D.; Courtois, C.

    2005-01-01

    We will present here the French program and an update on the progress made by the research conducted on partitioning and transmutation. Studies on partitioning and transmutation aim at isolating the most radio toxic long-lived elements present in the waste then at transmuting them through recycling in nuclear reactors, in order to change them into non-radioactive or shorter-lived elements. The partitioning of minor actinides (americium, curium and neptunium), followed by their transmutation, would reduce to a few hundred years the time necessary for the radiotoxicity of the vitrified waste to become similar to that contained in the natural uranium ore originally used. The feasibility of partitioning, which did not appear easily accessible at the time the research began since lanthanides and actinides have rather similar chemical properties, was nevertheless demonstrated in 2001 thanks to a series of tests conducted on solutions of dissolved spent fuel, in the CEA Atalante facility at Marcoule. The 2002-2005 program encompasses technological demonstration of the selected liquid-liquid process, with representative equipment, and economic evaluation of industrial implementation of partitioning. Studies on transmutation, which were initiated before the 1991 Law, rapidly led to concluding that transmutation of minor actinides (Americium, Curium, and Neptunium) was feasible in particular in fast neutron spectra. Results obtained confirm that the feasibility of transmutation is demonstrated, both in pressurized-water reactors (recycling and transmutation of plutonium, optionally but with more difficulty of americium and neptunium) and in advanced systems of nuclear-energy production (GEN IV fast-spectrum reactors, with recycling and transmutation of all heavy nuclides, uranium, plutonium, the minor actinides) or in dedicated incinerator reactors, either critical or sub critical. Work on transmutation is now focusing on technical elements necessary for the demonstration of

  2. Tank Inspection NDE Results for Fiscal Year 2014, Waste Tanks 26, 27, 28 and 33

    Energy Technology Data Exchange (ETDEWEB)

    Elder, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Vandekamp, R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2014-09-29

    Ultrasonic nondestructive examinations (NDE) were performed on waste storage tanks 26, 27, 28 and 33 at the Savannah River Site as a part of the “In-Service Inspection (ISI) Program for High Level Waste Tanks.” No reportable conditions were identified during these inspections. The results indicate that the implemented corrosion control program continues to effectively mitigate corrosion in the SRS waste tanks. Ultrasonic inspection (UT) is used to detect general wall thinning, pitting and interface attack, as well as vertically oriented cracks through inspection of an 8.5 inch wide strip extending over the accessible height of the primary tank wall and accessible knuckle regions. Welds were also inspected in tanks 27, 28 and 33 with no reportable indications. In a Type III/IIIA primary tank, a complete vertical strip includes scans of five plates (including knuckles) so five “plate/strips” would be completed at each vertical strip location. In FY 2014, a combined total of 79 plate/strips were examined for thickness mapping and crack detection, equating to over 45,000 square inches of area inspected on the primary tank wall. Of the 79 plate/strips examined in FY 2014 all but three have average thicknesses that remain at or above the construction minimum thickness which is nominal thickness minus 0.010 inches. There were no service induced reportable thicknesses or cracking encountered. A total of 2 pits were documented in 2014 with the deepest being 0.032 inches deep. One pit was detected in Tank 27 and one in Tank 33. No pitting was identified in Tanks 26 or 28. The maximum depth of any pit encountered in FY 2014 is 5% of nominal thickness, which is less than the minimum reportable criteria of 25% through-wall for pitting. In Tank 26 two vertical strips were inspected, as required by the ISI Program, due to tank conditions being outside normal chemistry controls for more than 3 months. Tank 28 had an area of localized thinning on the exterior wall of the

  3. Extraction, scrub, and strip test results for the solvent transfer to salt waste processing facility

    Energy Technology Data Exchange (ETDEWEB)

    Peters, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-09-07

    The Savannah River National Laboratory (SRNL) prepared approximately 240 gallons of Caustic-Side Solvent Extraction (CSSX) solvent for use at the Salt Waste Processing Facility (SWPF). An Extraction, Scrub, and Strip (ESS) test was performed on a sample of the prepared solvent using a salt solution prepared by Parsons to determine cesium distribution ratios (D(Cs)), and cesium concentration in the strip effluent (SE) and decontaminated salt solution (DSS) streams. This data will be used by Parsons to help qualify the solvent for use at the SWPF. The ESS test showed acceptable performance of the solvent for extraction, scrub, and strip operations. The extraction D(Cs) measured 15.5, exceeding the required value of 8. This value is consistent with results from previous ESS tests using similar solvent formulations. Similarly, scrub and strip cesium distribution ratios fell within acceptable ranges.

  4. Results of Washington's phase two study on closure requirements for the Hanford commercial low-level waste facility

    International Nuclear Information System (INIS)

    Anderson, D.C.; Hana, S.L.

    1989-01-01

    This paper reports on the closure design objectives and cover alternatives resulting from the state of Washington's phase two study on closure and long-term care for the Hanford commercial low-level radioactive waste disposal facility. Four approaches to dealing with subsidence and two cover design alternatives are discussed in this paper, along with information on each layer of each cover. Objectives for closure of the Hanford low-level waste facility are also discussed

  5. Mechanisms of gas retention and release: Experimental results for Hanford waste tanks 241-AW-101 and 241-AN-103

    Energy Technology Data Exchange (ETDEWEB)

    Rassat, S.D.; Gauglitz, P.A.; Bredt, P.R.; Mahoney, L.A.; Forbes, S.V.; Tingey, S.M.

    1997-09-01

    The 177 storage tanks at Hanford contain a vast array of radioactive waste forms resulting, primarily, from nuclear materials processing. Through radiolytic, thermal, and other decomposition reactions of waste components, gaseous species including hydrogen, ammonia, and the oxidizer nitrous oxide are generated within the waste tanks. Many of these tanks are known to retain and periodically release quantities of these flammable gas mixtures. The primary focus of the Flammable Gas Project is the safe storage of Hanford tank wastes. To this end, we strive to develop an understanding of the mechanisms of flammable gas retention and release in Hanford tanks through laboratory investigations on actual tank wastes. These results support the closure of the Flammable Gas Unreviewed Safety Question (USQ) on the safe storage of waste tanks known to retain flammable gases and support resolution of the broader Flammable Gas Safety Issue. The overall purpose of this ongoing study is to develop a comprehensive and thorough understanding of the mechanisms of flammable gas retention and release. The first objective of the current study was to classify bubble retention and release mechanisms in two previously untested waste materials from Tanks 241-AN-103 (AN-103) and 241-AW-101 (AW-101). Results were obtained for retention mechanisms, release characteristics, and the maximum gas retention. In addition, unique behavior was also documented and compared with previously studied waste samples. The second objective was to lengthen the duration of the experiments to evaluate the role of slowing bubble growth on the retention and release behavior. Results were obtained for experiments lasting from a few hours to a few days.

  6. Mechanisms of gas retention and release: Experimental results for Hanford waste tanks 241-AW-101 and 241-AN-103

    International Nuclear Information System (INIS)

    Rassat, S.D.; Gauglitz, P.A.; Bredt, P.R.; Mahoney, L.A.; Forbes, S.V.; Tingey, S.M.

    1997-09-01

    The 177 storage tanks at Hanford contain a vast array of radioactive waste forms resulting, primarily, from nuclear materials processing. Through radiolytic, thermal, and other decomposition reactions of waste components, gaseous species including hydrogen, ammonia, and the oxidizer nitrous oxide are generated within the waste tanks. Many of these tanks are known to retain and periodically release quantities of these flammable gas mixtures. The primary focus of the Flammable Gas Project is the safe storage of Hanford tank wastes. To this end, we strive to develop an understanding of the mechanisms of flammable gas retention and release in Hanford tanks through laboratory investigations on actual tank wastes. These results support the closure of the Flammable Gas Unreviewed Safety Question (USQ) on the safe storage of waste tanks known to retain flammable gases and support resolution of the broader Flammable Gas Safety Issue. The overall purpose of this ongoing study is to develop a comprehensive and thorough understanding of the mechanisms of flammable gas retention and release. The first objective of the current study was to classify bubble retention and release mechanisms in two previously untested waste materials from Tanks 241-AN-103 (AN-103) and 241-AW-101 (AW-101). Results were obtained for retention mechanisms, release characteristics, and the maximum gas retention. In addition, unique behavior was also documented and compared with previously studied waste samples. The second objective was to lengthen the duration of the experiments to evaluate the role of slowing bubble growth on the retention and release behavior. Results were obtained for experiments lasting from a few hours to a few days

  7. A study of inter-particle bonds in dry bauxite waste resulting in atmospheric aerosols

    Science.gov (United States)

    Wagh, Arun S.; Thompson, Bentley

    1988-02-01

    Bauxite and Alumina production are one of the main activities of several third world countries such as Jamaica, Brazil, India, Guinea, eastern European countries such as Hungary and Rumania and advanced countries such as Australia, West Germany, Japan and the United States. The mining operations lead to dust pollution, but the refining of bauxite to alumina yield large amounts of highly caustic sludge waste, called "Red Mud". Millions of tons of the waste produced in every country are stored in containment dams or natural valleys. This leads to ground water pollution, destruction of plant and bird life and is hazardous to human settlement in earthquake prone regions like Jamaica. As a result several companies have been looking into dry mud stacking which involves thickening the mud in the refining plants and sprying it on the slopes to sun dry it. Typically it involves a drying field of about two hundred acres, which could act as a potential source of caustic dust. In Jamaica one company has started disposing of the mud in this way. The aerosol formation from such areas depends mainly on the integrity of the top dry layers. Presently this is done by studying the approximate parameters such as the friability of the mud. However, following the recent advances in powder technology it has been possible for us to develop an instrument to study the average interparticle forces between the red mud particles. The instrument is based on the principle of a tensometer and a split cell is used to load specimens. A load cell is used to measure the force and a chart recorder is used for plotting separation and the force. The present study reports elemental composition of the dust and its health hazards. It also reports the physical measurement of the average interparticle force as a function of their separation in the Jamaican mud. The effect of ultraviolet radiation on the strength of the material is studied to see the effect of sun-drying of the waste. The five-fold increase

  8. Waste Tank Vapor Program: Vapor space characterization of waste tank 241-T-111. Results from samples collected on January 20, 1995

    International Nuclear Information System (INIS)

    Klinger, G.S.; Clauss, T.W.; Ligotke, M.W.; Pool, K.H.; McVeety, B.D.; Olsen, K.B.; Bredt, O.P.; Fruchter, J.S.; Goheen, S.C.

    1995-10-01

    This document presents the details of the inorganic and organic analysis that was performed on samples from the headspace of Hanford waste tank 241-T-111. The results described were obtained to support the safety and toxicological evaluations. A summary of the results for the inorganic and organic analytes is included, as well as, a detailed description of the results which appears in the text

  9. Waste Tank Vapor Program: Vapor space characterization of waste tank 241; C-102: Results from samples collected on August 23, 1994

    International Nuclear Information System (INIS)

    Klinger, G.S.; Clauss, T.W.; Ligotke, M.W.

    1995-10-01

    This document presents the details of the inorganic and organic analysis that was performed on samples from the headspace of Hanford waste tank 241-C-102. The results described were obtained to support the safety and toxicological evaluations. A summary of the results for the inorganic and organic analytes is included, as well as, a detailed description of the results which appears in the text

  10. Vapor space characterization of waste tank 241-U-111: Results from samples collected on February 28, 1995. Waste Tank Vapor Program

    International Nuclear Information System (INIS)

    Clauss, T.W.; Pool, K.H.; McVeety, B.D.; Bredt, O.P.; Goheen, S.C.; Ligotke, M.W.; Lucke, R.B.; Klinger, G.S.; Fruchter, J.S.

    1995-07-01

    This document presents the details of the inorganic and organic analysis that was performed on samples from the headspace of Hanford waste tank 241-U-111. The results described were obtained to support the safety and toxicological evaluations. A summary of the results for the inorganic and organic analytes is included, as well as, a detailed description of the results which appears in the text

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-12-01

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

    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 ∼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 ∼1 g/m 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

  14. New results on long term aging tests for rad-waste container alloy selection

    International Nuclear Information System (INIS)

    Alves, H.; Wahl, V.; Ibas, O.; Stenner, F.

    2004-01-01

    The current design of containers for high level nuclear waste proceeds on using an outer barrier of corrosion resistant Ni-based super alloy. The current alloy of choice is alloy 22 (UNS N06022). It is a quaternary Ni-Cr- Mo-W alloy system. The new but well established alloy 59 (UNS N06059) is an excellent equal or even a superior alternative to alloy 22 for the 10,000 years reliability being sought. Alloy 59 is a pure ternary alloy in the Ni-Cr-Mo alloy system. Objective of this paper is to present data comparing these two alloys. Therefore the behaviour of alloy 59 and alloy 22 was characterised after aging in air for 10,000 h and 20,000 h at different temperatures (200, 300 and 427 deg. C). Since the performance of weldments is of great concern, both welded and unwelded specimens were studied. Mechanical properties of the air aged alloys were measured at room temperature by tensile and notch impact-bending test. Thermal stability and aqueous corrosion are considered to be the key issues in the long-term performance of container materials proposed for the geological disposal of high level nuclear waste. The long-term thermal stability and corrosion resistance of the alloy 59 compared to alloy 22 is discussed. Corrosion resistance was evaluated in ASTM G28 A and 'green death' solution laboratory tests; hereby corrosion rates and depth of attack were determined. Metallo-graphical studies were performed in mill annealed and air aged conditions. The results of the aging tests at 10,000 h and 20,000 h show that alloy 59 is an equal or better candidate material due to its superior localised corrosion resistance behaviour (pitting and crevice corrosion resistance) and better thermal stability needed especially in multi-pass welding of thick sections. Therefore alloy 59 seems to be the most promising alternative to alloy 22. (authors)

  15. New results on long term aging tests for rad-waste container alloy selection

    Energy Technology Data Exchange (ETDEWEB)

    Alves, H.; Wahl, V.; Ibas, O.; Stenner, F. [ThyssenKrupp VDM GmbH, Altena (Germany)

    2004-07-01

    The current design of containers for high level nuclear waste proceeds on using an outer barrier of corrosion resistant Ni-based super alloy. The current alloy of choice is alloy 22 (UNS N06022). It is a quaternary Ni-Cr- Mo-W alloy system. The new but well established alloy 59 (UNS N06059) is an excellent equal or even a superior alternative to alloy 22 for the 10,000 years reliability being sought. Alloy 59 is a pure ternary alloy in the Ni-Cr-Mo alloy system. Objective of this paper is to present data comparing these two alloys. Therefore the behaviour of alloy 59 and alloy 22 was characterised after aging in air for 10,000 h and 20,000 h at different temperatures (200, 300 and 427 deg. C). Since the performance of weldments is of great concern, both welded and unwelded specimens were studied. Mechanical properties of the air aged alloys were measured at room temperature by tensile and notch impact-bending test. Thermal stability and aqueous corrosion are considered to be the key issues in the long-term performance of container materials proposed for the geological disposal of high level nuclear waste. The long-term thermal stability and corrosion resistance of the alloy 59 compared to alloy 22 is discussed. Corrosion resistance was evaluated in ASTM G28 A and 'green death' solution laboratory tests; hereby corrosion rates and depth of attack were determined. Metallo-graphical studies were performed in mill annealed and air aged conditions. The results of the aging tests at 10,000 h and 20,000 h show that alloy 59 is an equal or better candidate material due to its superior localised corrosion resistance behaviour (pitting and crevice corrosion resistance) and better thermal stability needed especially in multi-pass welding of thick sections. Therefore alloy 59 seems to be the most promising alternative to alloy 22. (authors)

  16. Savannah River Plant low-level waste incinerator: Operational results and technical development

    International Nuclear Information System (INIS)

    Irujo, M.J.; Bucci, J.R.

    1987-04-01

    Volume reduction of solid and liquid low-level waste has been demonstrated at the Savannah River Plant (SRP) in the Waste Management Beta-Gamma Incinerator facility (BGI). The BGI uses a two-stage, controlled-air incinerator capable of processing 180 kg/hr (400 lbs/hr) of solid waste or 150 liters/hr (40 gal/hr) of liquid waste. These wastes are pyrolyzed in a substoichiometric air environment at 900 to 1100 degrees Celsius in the primary chamber. Products of partial combustion from the primary chamber are oxidized at 950 to 1150 degrees Celsius in the secondary chamber. A spray dryer, baghouse,and HEPA filter unit cool and filter the incinerator offgases. 2 refs., 9 tabs

  17. Techniques for improving shuffler assay results for 55-gallon waste drums

    International Nuclear Information System (INIS)

    Rinard, P.M.; Prettyman, T.H.; Stuenkel, D.

    1994-01-01

    Accurate assays of the fissile contents in waste drums are needed to ensure the most proper and economical handling and disposal of the waste. An improvement of accuracy will mean fewer drums disposed as transuranic waste when they really contain low-level waste, saving both money and burial sites. Shufflers are used for assaying waste drums and are very accurate with nonmoderating matrices (such as iron). In the active mode they count delayed neutrons released after fissions are induced by irradiation neutrons from a 252 Cf source. However, as the hydrogen density from matrices such as paper or gloves increases, the accuracy can suffer without proper attention. The neutron transport and fission probabilities change with the hydrogen density, causing the neutron count rate to vary with the position of the fissile material within the drum. The magnitude of this variation grows with the hydrogen density

  18. The extent of food waste generation across EU-27: different calculation methods and the reliability of their results.

    Science.gov (United States)

    Bräutigam, Klaus-Rainer; Jörissen, Juliane; Priefer, Carmen

    2014-08-01

    The reduction of food waste is seen as an important societal issue with considerable ethical, ecological and economic implications. The European Commission aims at cutting down food waste to one-half by 2020. However, implementing effective prevention measures requires knowledge of the reasons and the scale of food waste generation along the food supply chain. The available data basis for Europe is very heterogeneous and doubts about its reliability are legitimate. This mini-review gives an overview of available data on food waste generation in EU-27 and discusses their reliability against the results of own model calculations. These calculations are based on a methodology developed on behalf of the Food and Agriculture Organization of the United Nations and provide data on food waste generation for each of the EU-27 member states, broken down to the individual stages of the food chain and differentiated by product groups. The analysis shows that the results differ significantly, depending on the data sources chosen and the assumptions made. Further research is much needed in order to improve the data stock, which builds the basis for the monitoring and management of food waste. © The Author(s) 2014.

  19. Summary Of Cold Crucible Vitrification Tests Results With Savannah River Site High Level Waste Surrogates

    Energy Technology Data Exchange (ETDEWEB)

    Stefanovsky, Sergey; Marra, James; Lebedev, Vladimir

    2014-01-13

    The cold crucible inductive melting (CCIM) technology successfully applied for vitrification of low- and intermediate-level waste (LILW) at SIA Radon, Russia, was tested to be implemented for vitrification of high-level waste (HLW) stored at Savannah River Site, USA. Mixtures of Sludge Batch 2 (SB2) and 4 (SB4) waste surrogates and borosilicate frits as slurries were vitrified in bench- (236 mm inner diameter) and full-scale (418 mm inner diameter) cold crucibles. Various process conditions were tested and major process variables were determined. Melts were poured into 10L canisters and cooled to room temperature in air or in heat-insulated boxes by a regime similar to Canister Centerline Cooling (CCC) used at DWPF. The products with waste loading from ~40 to ~65 wt.% were investigated in details. The products contained 40 to 55 wt.% waste oxides were predominantly amorphous; at higher waste loadings (WL) spinel structure phases and nepheline were present. Normalized release values for Li, B, Na, and Si determined by PCT procedure remain lower than those from EA glass at waste loadings of up to 60 wt.%.

  20. Genotoxicity studies in semiconductor industry. 1. In vitro mutagenicity and genotoxicity studies of waste samples resulting from plasma etching

    Energy Technology Data Exchange (ETDEWEB)

    Braun, R.; Huettner, E.M.; Merten, H.; Raabe, F. (Institute of Plant Genetics and Crop Plant Research, Gatersleben (Germany))

    1993-07-01

    Solid waste samples taken from the etching reactor, the turbo pump, and the waste air system of a plasma etching technology line in semiconductor production were studied as to their genotoxic properties in a bacterial repair test, in the Ames/Salmonella microsome assay, in the SOS chromotest, in primary mouse hepatocytes, and in Chinese hamster V79 cell cultures. All three waste samples were found to be active by inducing of unscheduled DNA-synthesis in mouse hepatocytes in vitro. In the bacterial rec-type repair test with Proteus mirabilis, waste samples taken from the turbo pump and the vacuum pipe system were not genotoxic. The waste sample taken from the chlorine-mediated plasma reactor was clearly positive in the bacterial repair assay and in the SOS chromotest with Escherichia coli. Mutagenic activity was demonstrated for all samples in the presence and absence of S9 mix made from mouse liver homogenate. Again, highest mutagenic activity was recorded for the waste sample taken from the plasma reactor, while samples collected from the turbo pump and from the waste air system before dilution and liberation of the air were less mutagenic. For all samples chromosomal damage in V79 cells was not detected, indicating absence of clastogenic activity in vitro. Altogether, these results indicate generation of genotoxic and mutagenic products as a consequence of chlorine-mediated plasma etching in the microelectronics industry and the presence of genotoxins even in places distant from the plasma reactor. Occupational exposure can be expected both from the precipitated wastes and from chemicals reaching the environment with the air stream.

  1. Removal of strontium and transuranics from Hanford waste via hydrothermal processing -- FY 1994/95 test results

    International Nuclear Information System (INIS)

    Orth, R.J.; Schmidt, A.J.; Elmore, M.R.; Hart, T.R.; Neuenschwander, G.G.; Gano, S.R.; Lehmann, R.W.; Momont, J.A.

    1995-09-01

    Under the Tank Waste Remediation System (TWRS) Pretreatment Technology Development Project, Pacific Northwest Laboratory (PNL) is evaluating and developing organic destruction technologies that may be incorporated into the Initial Pretreatment Module (IPM) to treat Hanford tank waste. Organic (and ferrocyanide) destruction removes the compounds responsible for waste safety issues, and conditions the supernatant for low-level waste disposal by removing compounds that may be responsible for promoting strontium and transuranic (TRU) components solubility. Destruction or defunctionalization of complexing organics in tank wastes eliminates organic species that can reduce the efficiency of radionuclide (E.g., 90 Sr) separation processes, such as ion exchange, solvent extraction, and precipitation. The technologies being evaluated and tested for organic destruction are low-temperature hydrothermal processing (HTP) and wet air oxidation (WAO). Four activities are described: Batch HTP/WAO testing with Actual Tank Waste (Section 3.0), Batch HTP Testing with Simulant (Section 4.0), Batch WAO testing with Simulant (Section 5.0), and Continuous Bench-scale WAO Testing with Simulant (Section 6.0). For each of these activities, the objectives, test approach, results, status, and direction of future investigations are discussed. The background and history of the HTP/WAO technology is summarized below. Conclusions and Recommendations are provided in Section 2.0. A continuous HTP off-gas safety evaluation conducted in FY 1994 is included as Appendix A

  2. Results of interagency effort to determine carbon-14 source term in low-level radioactive waste

    International Nuclear Information System (INIS)

    Gruhlke, J.M.; Meyer, G.L.; Neiheisel, J.

    1987-01-01

    A preliminary estimate of the risks from the shallow land disposal of low-level radioactive wastes by EPA in 1984-1985 indicated that Carbon-14 caused virtually all of the risk and that these risks were relatively high. Therefore, an informal interagency group, which included the US Department of Energy, US Geological Survey, US Nuclear Regulatory Commission, and US Environmental Protection Agency, formed in 1985 to obtain up-to-date information on the activity and chemical form of Carbon-14 in the different types of LLW and how Carbon-14 behaves after disposal. The EPA acted as a focal point for collating the information collected by all of the Agencies and will publish a report in Fall 1986 on the results of the Carbon-14 data collection effort. Of particular importance, the study showed that Carbon-14 activity in LLW was overestimated approximately 2000%. This paper summarizes results of the Carbon-14 data collection effort. 40 references, 1 figure, 3 tables

  3. The use of bitumen for storing radioactive waste resulting from oil industry containing Ra-226

    International Nuclear Information System (INIS)

    Takriti, S.; Shweikani, R.; Abdulhafiz, M.; Salman, M.

    2009-12-01

    The releases of radon gas from NORM waste contained in two different forms of bitumen samples have been investigated. The artificial NORM source samples were made by mixing NORM with bitumen. The sources surrounded by different thickness of bitumen layers to prepare the first form of samples. While, the NORM powder was put inside bitumen samples prepared as a cylindrical shape with different thickness. The results showed that the release of radon from the bitumen samples was different in case of sources and powder. The results illustrated that the release of radon from the bitumen samples was decrees linearly with the samples thicknesses (in both cases source and powder). On the other hand, the release from the cement samples was proportional inversely with the different thickness (for comparison). In addition, many other supporting experiments were performed, as γ-ray spectroscopy measurements showed that the cement is better than the bitumen in shielding process, while the bitumen is better than cement to prevent the releases of radon. (authors)

  4. Test Results and Comparison of Triaxial Strength Testing of Waste Isolation Pilot Plant Clean Salt

    Energy Technology Data Exchange (ETDEWEB)

    Buchholz, Stuart A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-12-01

    This memorandum documents laboratory thermomechanical triaxial strength testing of Waste Isolation Pilot Plant (WIPP) clean salt. The limited study completed independent, adjunct laboratory tests in the United States to assist in validating similar testing results being provided by the German facilities. The testing protocol consisted of completing confined triaxial, constant strain rate strength tests of intact WIPP clean salt at temperatures of 25°C and 100°C and at multiple confining pressures. The stratigraphy at WIPP also includes salt that has been labeled “argillaceous.” The much larger test matrix conducted in Germany included both the so-called clean and argillaceous salts. When combined, the total database of laboratory results will be used to develop input parameters for models, assess adequacy of existing models, and predict material behavior. These laboratory studies are also consistent with the goals of the international salt repository research program. The goal of this study was to complete a subset of a test matrix on clean salt from the WIPP undertaken by German research groups. The work was performed at RESPEC in Rapid City, South Dakota. A rigorous Quality Assurance protocol was applied, such that corroboration provides the potential of qualifying all of the test data gathered by German research groups.

  5. Disposal of radwastes and recycling of wastes and structural materials -fundamental principles, concepts, results

    International Nuclear Information System (INIS)

    Schaller, G.; Arens, G.; Brennecke, P.; Goertz, R.; Poschner, J.; Thieme, M.

    1997-01-01

    This report describes the German concept for the disposal of radioactive waste, and the re-use or recycling of contaminated materials. All radioactive waste can be disposed of in deep geological formations (practised at ERAM disposal site, planned for Konrad disposal site). Radioactively contaminated material below clearance levels can proceed for disposal at waste disposal sites and incineration plants, or for re-use and recycling, especially where the material consists of contaminated steel and of buildings. The basic principles (dose limits and model structures for deriving recommendations), reference values, or limits are described. The latest concepts are described in greater detail. Waste management in Germany is compared with international concepts. (orig.) [de

  6. Knowledge, attitudes and behaviour regarding waste management options in Romania: results from a school questionnaire

    OpenAIRE

    Karin KOLBE

    2014-01-01

    This study analyses knowledge, attitudes and behaviour in the area of different waste management approaches of pupils in Romania. Examining school students' knowledge about waste management options and finding out the reasons that prevent them from participating in environmentally sound disposal options is essential for teachers and legislators. For this purpose, questionnaires were designed and distributed in two schools in Romania. The analysis revealed that knowledge is highly developed in...

  7. Waste Receiving and Processing (WRAP) Facility Weight Scale Analysis Fairbanks Weight Scale Evaluation Results

    International Nuclear Information System (INIS)

    JOHNSON, M.D.

    1999-01-01

    Fairbanks Weight Scales are used at the Waste Receiving and Processing (WRAP) facility to determine the weight of waste drums as they are received, processed, and shipped. Due to recent problems, discovered during calibration, the WRAP Engineering Department has completed this document which outlines both the investigation of the infeed conveyor scale failure in September of 1999 and recommendations for calibration procedure modifications designed to correct deficiencies in the current procedures

  8. Results of research and development works 1981 of the Institute for Nuclear Waste Management Techniques

    International Nuclear Information System (INIS)

    1982-02-01

    The emphasis of the work is on the development and characterization of waste products suitable for final storage, such as actinides and fission products from MAW/LAW, as well as on the development of methods for the treatment and solidification of radioactive wastes, e.g. wet chemical oxidation, vitrification and cementation. Investigations on the HAW-storage in salt are carried out in cooperation with the GSF. (HP) [de

  9. Results of the freeze resistance test, swelling index and coefficient of permeability of finegrained mining waste reinforced with cements

    Science.gov (United States)

    Morman, Justyna

    2018-04-01

    The article presents the result of laboratory tests for mining waste with grain size of 0 to 2 mm stabilized with cement. Used for stabilization of cement CEM I 42.5 R and blast furnace cement CEM III / A 42.5N - LH / HSR / NA and a plasticizer sealant. Cement was added to the mining waste test in the proportions of 5 - 8% in relation to the skeleton's weight. For the cemented samples, the freeze resistance test, swelling index, coefficient of permeability and pH of water leachate were tested. The addition of a cement binder resulted in diminishing the water permeability of mining waste and limiting the leaching of fine particles from the material.

  10. A convenient method for estimating the contaminated zone of a subsurface aquifer resulting from radioactive waste disposal into ground

    International Nuclear Information System (INIS)

    Fukui, Masami; Katsurayama, Kousuke; Uchida, Shigeo.

    1981-01-01

    Studies were conducted to estimate the contamination spread resulting from the radioactive waste disposal into a subsurface aquifer. A general equation, expressing the contaminated zone as a function of radioactive decay, the physical and chemical parameters of soil is presented. A distribution coefficient was also formulated which can be used to judge the suitability of a site for waste disposal. Moreover, a method for predicting contaminant concentration in groundwater at a site boundary is suggested for a heterogeneous media where the subsurface aquifer has different values of porosity, density, flow velocity, distribution coefficient and so on. A general equation was also developed to predict the distribution of radionuclides resulting from the disposal of a solid waste material. The distributions of contamination was evaluated for 90 Sr and 239 Pu which obey a linear adsorption model and a first order kinetics respectively. These equations appear to have practical utility for easily estimating groundwater contamination. (author)

  11. Waste Preparation and Transport Chemistry: Results of the FY 2002 Studies

    International Nuclear Information System (INIS)

    Hunt, R.D.

    2003-01-01

    The initial step in the remediation of nuclear waste stored at Hanford and the Savannah River Site (SRS) involves the retrieval and transfer of the waste to another tank or to a treatment facility. The retrieved waste can range from a filtered supernatant to a slurry. Nearly all of the recent solid formation problems encountered during waste transfers and subsequent treatment steps have involved decanted or filtered supernatants. Problems with slurry transfers have not yet surfaced, because tank farm operations at Hanford and the SRS have focused primarily on supernatant transfers and treatment. For example, the interim stabilization program at Hanford continues to reduce the level of supernatants and interstitial liquids in its single-shell tanks through saltwell pumping of filtered liquid. In addition, at present, the cross-site transfer lines at Hanford can be used only to transfer liquids. Another reason for fewer problems with slurry transfers involves the additions of large quantities of dilution water prior to the transfer. When the waste is transferred, a drop in temperature is expected because most transfer lines are not heated. However, the dilution water reduces or eliminates solid formation caused by this temperature drop. In sharp contrast, decanted or filtered supernatants are near or at saturation for certain compounds. In such cases, tank farm operators must continue to evaporate their liquid waste since available tank space is quite limited. Solid formation can occur when the temperature of saturated solutions drops even slightly. The evaporation step can also lead to the formation of problematic solids. At the SRS, the evaporation of a relatively dilute waste stream was suspended due to the formation of deposits in the evaporator system. Therefore, small drops in temperature or evaporation can lead to problematic solid formations

  12. Waste Preparation and Transport Chemistry: Results of the FY 2002 Studies

    Energy Technology Data Exchange (ETDEWEB)

    Hunt, R.D.

    2003-07-10

    The initial step in the remediation of nuclear waste stored at Hanford and the Savannah River Site (SRS) involves the retrieval and transfer of the waste to another tank or to a treatment facility. The retrieved waste can range from a filtered supernatant to a slurry. Nearly all of the recent solid formation problems encountered during waste transfers and subsequent treatment steps have involved decanted or filtered supernatants. Problems with slurry transfers have not yet surfaced, because tank farm operations at Hanford and the SRS have focused primarily on supernatant transfers and treatment. For example, the interim stabilization program at Hanford continues to reduce the level of supernatants and interstitial liquids in its single-shell tanks through saltwell pumping of filtered liquid. In addition, at present, the cross-site transfer lines at Hanford can be used only to transfer liquids. Another reason for fewer problems with slurry transfers involves the additions of large quantities of dilution water prior to the transfer. When the waste is transferred, a drop in temperature is expected because most transfer lines are not heated. However, the dilution water reduces or eliminates solid formation caused by this temperature drop. In sharp contrast, decanted or filtered supernatants are near or at saturation for certain compounds. In such cases, tank farm operators must continue to evaporate their liquid waste since available tank space is quite limited. Solid formation can occur when the temperature of saturated solutions drops even slightly. The evaporation step can also lead to the formation of problematic solids. At the SRS, the evaporation of a relatively dilute waste stream was suspended due to the formation of deposits in the evaporator system. Therefore, small drops in temperature or evaporation can lead to problematic solid formations.

  13. Surrogate formulations for thermal treatment of low-level mixed waste, Part II: Selected mixed waste treatment project waste streams

    Energy Technology Data Exchange (ETDEWEB)

    Bostick, W.D.; Hoffmann, D.P.; Chiang, J.M.; Hermes, W.H.; Gibson, L.V. Jr.; Richmond, A.A. [Martin Marietta Energy Systems, Inc., Oak Ridge, TN (United States); Mayberry, J. [Science Applications International Corp., Idaho Falls, ID (United States); Frazier, G. [Univ. of Tennessee, Knoxville, TN (United States)

    1994-01-01

    This report summarizes the formulation of surrogate waste packages, representing the major bulk constituent compositions for 12 waste stream classifications selected by the US DOE Mixed Waste Treatment Program. These waste groupings include: neutral aqueous wastes; aqueous halogenated organic liquids; ash; high organic content sludges; adsorbed aqueous and organic liquids; cement sludges, ashes, and solids; chloride; sulfate, and nitrate salts; organic matrix solids; heterogeneous debris; bulk combustibles; lab packs; and lead shapes. Insofar as possible, formulation of surrogate waste packages are referenced to authentic wastes in inventory within the DOE; however, the surrogate waste packages are intended to represent generic treatability group compositions. The intent is to specify a nonradiological synthetic mixture, with a minimal number of readily available components, that can be used to represent the significant challenges anticipated for treatment of the specified waste class. Performance testing and evaluation with use of a consistent series of surrogate wastes will provide a means for the initial assessment (and intercomparability) of candidate treatment technology applicability and performance. Originally the surrogate wastes were intended for use with emerging thermal treatment systems, but use may be extended to select nonthermal systems as well.

  14. Surrogate formulations for thermal treatment of low-level mixed waste, Part II: Selected mixed waste treatment project waste streams

    International Nuclear Information System (INIS)

    Bostick, W.D.; Hoffmann, D.P.; Chiang, J.M.; Hermes, W.H.; Gibson, L.V. Jr.; Richmond, A.A.; Mayberry, J.; Frazier, G.

    1994-01-01

    This report summarizes the formulation of surrogate waste packages, representing the major bulk constituent compositions for 12 waste stream classifications selected by the US DOE Mixed Waste Treatment Program. These waste groupings include: neutral aqueous wastes; aqueous halogenated organic liquids; ash; high organic content sludges; adsorbed aqueous and organic liquids; cement sludges, ashes, and solids; chloride; sulfate, and nitrate salts; organic matrix solids; heterogeneous debris; bulk combustibles; lab packs; and lead shapes. Insofar as possible, formulation of surrogate waste packages are referenced to authentic wastes in inventory within the DOE; however, the surrogate waste packages are intended to represent generic treatability group compositions. The intent is to specify a nonradiological synthetic mixture, with a minimal number of readily available components, that can be used to represent the significant challenges anticipated for treatment of the specified waste class. Performance testing and evaluation with use of a consistent series of surrogate wastes will provide a means for the initial assessment (and intercomparability) of candidate treatment technology applicability and performance. Originally the surrogate wastes were intended for use with emerging thermal treatment systems, but use may be extended to select nonthermal systems as well

  15. Conditioning matrices from high level waste resulting from pyrochemical processing in fluorine salt

    International Nuclear Information System (INIS)

    Grandjean, Agnes; Advocat, Thierry; Bousquet, Nicolas; Jegou, Christophe

    2007-01-01

    Separating the actinides from the fission products through reductive extraction by aluminium in a LiF/AlF 3 medium is a process investigated for pyrometallurgical reprocessing of spent fuel. The process involves separation by reductive salt-metal extraction. After dissolving the fuel or the transmutation target in a salt bath, the noble metal fission products are first extracted by contacting them with a slightly reducing metal. After extracting the metal fission products, then the actinides are selectively separated from the remaining fission products. In this hypothesis, all the unrecoverable fission products would be conditioned as fluorides. Therefore, this process will generate first a metallic waste containing the 'reducible' fission products (Pd, Mo, Ru, Rh, Tc, etc.) and a fluorine waste containing alkali-metal, alkaline-earth and rare earth fission products. Immobilization of these wastes in classical borosilicate glasses is not feasible due to the very low solubility of noble metals, and of fluoride in these hosts. Alternative candidates have therefore been developed including silicate glass/ceramic system for fluoride fission products and metallic ones for noble metal fission products. These waste-forms were evaluated for their confinement properties like homogeneity, waste loading, volatility during the elaboration process, chemical durability, etc. using appropriate techniques. (authors)

  16. Results from NNWSI [Nevada Nuclear Waste Storage Investigations] Series 2 bare fuel dissolution tests

    International Nuclear Information System (INIS)

    Wilson, C.N.

    1990-09-01

    The dissolution and radionuclide release behavior of spent fuel in groundwater is being studied by the Nevada Nuclear Waste Storage Investigations (NNWSI) Project. Two bare spent fuel specimens plus the empty cladding hulls were tested in NNWSI J-13 well water in unsealed fused silica vessels under ambient hot cell air conditions (25 degree C) in the currently reported tests. One of the specimens was prepared from a rod irradiated in the H. B. Robinson Unit 2 reactor and the other from a rod irradiated in the Turkey Point Unit 3 reactor. Results indicate that most radionuclides of interest fall into three groups for release modeling. The first group principally includes the actinides (U, Np, Pu, Am, and Cm), all of which reached solubility-limited concentrations that were orders of magnitude below those necessary to meet the NRC 10 CFR 60.113 release limits for any realistic water flux predicted for the Yucca Mountain repository site. The second group is nuclides of soluble elements such as Cs, Tc, and I, for which release rates do not appear to be solubility-limited and may depend on the dissolution rate of fuel. In later test cycles, 137 Cs, 90 Sr, 99 Tc, and 129 I were continuously released at rates between about 5 x 10 -5 and 1 x 10 -4 of inventory per year. The third group is radionuclides that may be transported in the vapor phase, of which 14 C is of primary concern. Detailed test results are presented and discussed. 17 refs., 15 figs., 21 tabs

  17. Laboratory corrosion tests on candidate high-level waste container materials: Results from the Belgian programme

    International Nuclear Information System (INIS)

    Druyts, F.; Kursten, B.; Iseghem, P. Van

    2004-01-01

    The Belgian SAFIR-2 concept foresees the geological disposal of conditioned high-level radioactive waste in stainless steel containers and overpacks placed in a concrete gallery backfilled with Boom clay or a bentonite-type backfill. In addition to earlier in situ experiments, we used a laboratory approach to investigate the corrosion properties of selected stainless steels in Boom clay and bentonite environments. In the SAFIR-2 concept, AISI 316L hMo is the main candidate overpack material. As an alternative, we also investigated the higher alloyed stainless steel UHB 904L. Our study focused on localised corrosion and in particular pitting. We used cyclic potentiodynamic polarisation measurements to determine the pit nucleation potential E NP and the protection potential E PP . The evolution of the corrosion potential with time was determined by monitoring the open circuit potential in synthetic clay-water over extended periods. In this paper we present and discuss some results from our laboratory programme, focusing on long-term interactions between the stainless steel overpack and the backfill materials. We describe in particular the influence of chloride and thio-sulphate ions on the pitting corrosion behaviour. The results show that, under geochemical conditions typical for geological disposal, i.e. [Cl-] ∼ 30 mg/L for a Boom clay backfill and [Cl-] ∼ 90 mg/L for a bentonite backfill, neither AISI 316L hMo nor UHB 904L is expected to present pitting problems. An important factor in the long-term prediction of the corrosion behaviour however, is the robustness of the model for the evolution of the geochemistry of the backfill. Indeed, at chloride levels higher than 1000 mg/L, we predict pitting corrosion for AISI 316L hMo. (authors)

  18. Preliminary results from uranium/americium affinity studies under experimental conditions for cesium removal from NPP ''Kozloduy'' simulated wastes solutions

    International Nuclear Information System (INIS)

    Nikiforova, A.; Kinova, L.; Peneva, C.; Taskaeva, I.; Petrova, P.

    2005-01-01

    We use the approach described by Westinghouse Savannah River Company using ammonium molybdophosphate (AMP) to remove elevated concentrations of radioactive cesium to facilitate handling waste samples from NPP K ozloduy . Preliminary series of tests were carried out to determine the exact conditions for sufficient cesium removal from five simulated waste solutions with concentrations of compounds, whose complexing power complicates any subsequent processing. Simulated wastes solutions contain high concentrations of nitrates, borates, H 2 C 2 O 4 , ethylenediaminetetraacetate (EDTA) and Citric acid, according to the composition of the real waste from the NPP. On this basis a laboratory treatment protocol was created. This experiment is a preparation for the analysis of real waste samples. In this sense the results are preliminary. Unwanted removal of non-cesium radioactive species from simulated waste solutions was studied with gamma spectrometry with the aim to find a compromise between on the one hand the AMP effectiveness and on the other hand unwanted affinity to AMP of Uranium and Americium. Success for the treatment protocol is defined by proving minimal uptake of U and Am, while at the same time demonstrating good removal effectiveness through the use of AMP. Uptake of U and Am were determined as influenced by oxidizing agents at nitric acid concentrations, proposed by Savannah River National laboratory. It was found that AMP does not significantly remove U and Am when concentration of oxidizing agents is more than 0.1M for simulated waste solutions and for contact times inherent in laboratory treatment protocol. Uranium and Americium affinity under experimental conditions for cesium removal were evaluated from gamma spectrometric data. Results are given for the model experiment and an approach for the real waste analysis is chosen. Under our experimental conditions simulated wastes solutions showed minimal affinity to AMP when U and Am are most probably in

  19. Colloid Genesis/Transport and Flow Pathway Alterations Resulting From Interactions of Reactive Waste Solutions and Hanford Vadose Zone Sediments

    International Nuclear Information System (INIS)

    Wan, Jiamin; Tokunaga, Tetsu K.

    2001-01-01

    Leakage of underground tanks containing high-level nuclear waste solutions has been identified at various DOE facilities. The Hanford Site is one the main facilities of concern, with about 2,300 to 3,400 m3 of leaked waste liquids. Radionuclides and other contaminants have been found in elevated concentrations in the vadose zone and groundwater underneath single shell tank farms. We do not currently know the mechanisms responsible for the unexpected deep migration of some contaminants through the vadose zone, and such understanding is urgently needed for planning remediation. Due to the extreme chemical conditions of the tank waste solutions (very high pH, aluminum concentration, and ionic strength), interactions between the highly reactive waste solutions and sediments underneath the tanks can result in dissolution of primary minerals of the sediments and precipitation of secondary phases including colloidal particles. Contaminants can sorb onto and/or co-precipitate with the secondary phases. Therefore transport of strongly associated contaminants on mobile colloids can be substantially greater than without colloids. The overall objective of this research is to improve our understanding on the effects of interactions between the tank waste solution and sediments on deep contaminant migration under Hanford Site conditions. This objective will be achieved through the following four tasks: (1) colloid generation and transport studies, (2) studies on sediment permeability and chemical composition alterations, (3) quantifying associations of contaminants with secondary colloids, and (4) studies on the combined effects of the aforementioned processes on deep contaminant migration

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

    Directory of Open Access Journals (Sweden)

    Mansutti Mauro

    2008-04-01

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

  1. Technical solution for radioactive waste management resulting from the I-131 therapy

    International Nuclear Information System (INIS)

    Jerez V, P.; Lopez F, Y.; Quevedo G, J.; Betancourt H, L.

    1996-01-01

    The paper discusses a system designed for the collection and storage of biological wastes arising from the therapy with l-131. This system is based on the use of either retention or septic tanks, in which the waste is stored or delayed until the activity decays to acceptable levels, in order to comply with authorized limits established by the Regulatory Authority for discharge to environment. A method for estimating waste activity concentration as a function of the number of patients, the activity delivered to each one of them, as well as other parameter related to the system design are discussed. The general requirements to be met by the system are also included. (authors). 4 refs., 4 figs

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

    International Nuclear Information System (INIS)

    Chesnut, D.A.

    1991-09-01

    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 10 7 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 10 5 years to provide the same reliability. Other examples illustrate how multiple barriers can provide a strategy for not only achieving but demonstrating regulatory compliance

  3. Management of toxic waste resulting from decommissioning and environmental remediation of nuclear facilities in Northwest Russia

    International Nuclear Information System (INIS)

    Vysotskij, V.L.; Nikitin, V.S.; Kulikov, K.N.; Ivanov, S.A.; Bogdanova, G.S.; Zakharov, A.A.

    2008-01-01

    Integrated information on toxic wastes formed during utilization and rehabilitation of shutdown naval nuclear object at Northwest Russia is performed. Dynamics of their accumulation to 2025 is estimated. Necessity of present waste management review and search of new methods with the view of decrease of environmental risks by means of systematic reprocessing or economic favorable destruction. Several strategies are treated. Advantages and imperfections of each of them are estimated by safety factors and economic costs, and the most acceptable strategy is selected. Functional model is found. Lists of projects, technical means are given, periods, costs for its realization are evaluated. Guidelines are provided [ru

  4. BIOREFINE-2G — Result In Brief: Novel biopolymers from biorefinery waste-streams

    DEFF Research Database (Denmark)

    Stovicek, Vratislav; Chen, Xiao; Borodina, Irina

    Second generation biorefineries are all about creating value from waste, so it seems only right that the ideal plant should leave nothing behind. With this in mind, the BIOREFINE-2G project has developed novel processes to convert pentose-rich side-streams into biopolymers.......Second generation biorefineries are all about creating value from waste, so it seems only right that the ideal plant should leave nothing behind. With this in mind, the BIOREFINE-2G project has developed novel processes to convert pentose-rich side-streams into biopolymers....

  5. Pilot plant SERSE: Description and results of the experimental tests under treatment of simulated chemical liquid waste

    International Nuclear Information System (INIS)

    Calle, C.; Gili, M.; Luce, A.; Marrocchelli, A.; Pietrelli, L.; Troiani, F.

    1989-11-01

    The chemical processes for the selective separation of the actinides and long lived fission products from aged liquid wastes is described. The SERSE pilot plant is a cold facility which has been designed, by ENEA, for the engineering scale demonstration of the chemical separation processes. The experimental tests carried out in the plant are described and the results confirm the laboratory data. (author)

  6. Initial performance assessment of the disposal of spent nuclear fuel and high-level waste stored at Idaho National Engineering Laboratory. Volume 1, Methodology and results

    Energy Technology Data Exchange (ETDEWEB)

    Rechard, R.P. [ed.

    1993-12-01

    This performance assessment characterized plausible treatment options conceived by the Idaho National Engineering Laboratory (INEL) for its spent fuel and high-level radioactive waste and then modeled the performance of the resulting waste forms in two hypothetical, deep, geologic repositories: one in bedded salt and the other in granite. The results of the performance assessment are intended to help guide INEL in its study of how to prepare wastes and spent fuel for eventual permanent disposal. This assessment was part of the Waste Management Technology Development Program designed to help the US Department of Energy develop and demonstrate the capability to dispose of its nuclear waste. Although numerous caveats must be placed on the results, the general findings were as follows: Though the waste form behavior depended upon the repository type, all current and proposed waste forms provided acceptable behavior in the salt and granite repositories.

  7. Response of soil microorganisms to radioactive oil waste: results from a leaching experiment

    Science.gov (United States)

    Galitskaya, P.; Biktasheva, L.; Saveliev, A.; Ratering, S.; Schnell, S.; Selivanovskaya, S.

    2015-06-01

    Oil wastes produced in large amounts in the processes of oil extraction, refining, and transportation are of great environmental concern because of their mutagenicity, toxicity, high fire hazardousness, and hydrophobicity. About 40% of these wastes contain radionuclides; however, the effects of oil products and radionuclides on soil microorganisms are frequently studied separately. The effects on various microbial parameters of raw waste containing 575 g of total petroleum hydrocarbons (TPH) kg-1 waste, 4.4 of 226Ra, 2.8 of 232Th, and 1.3 kBq kg-1 of 40K and its treated variant (1.6 g kg-1 of TPH, 7.9 of 226Ra, 3.9 of 232Th, and 183 kBq kg-1 of 40K) were examined in a leaching column experiment to separate the effects of hydrocarbons from those of radioactive elements. The raw waste sample (H) was collected from tanks during cleaning and maintenance, and a treated waste sample (R) was obtained from equipment for oil waste treatment. Thermal steam treatment is used in the production yard to reduce the oil content. The disposal of H waste samples on the soil surface led to an increase in the TPH content in soil: it became 3.5, 2.8, and 2.2 times higher in the upper (0-20 cm), middle (20-40 cm), and lower (40-60cm) layers, respectively. Activity concentrations of 226Ra and 232Th increased in soil sampled from both H- and R- columns in comparison to their concentrations in control soil. The activity concentrations of these two elements in samples taken from the upper and middle layers were much higher for the R-column compared to the H-column, despite the fact that the amount of waste added to the columns was equalized with respect to the activity concentrations of radionuclides. The H waste containing both TPH and radionuclides affected the functioning of the soil microbial community, and the effect was more pronounced in the upper layer of the column. Metabolic quotient and cellulase activity were the most sensitive microbial parameters as their levels were changed 5

  8. LETTUCE AND BROCCOLI RESPONSE AND SOIL PROPERTIES RESULTING FROM TANNERY WASTE APPLICATIONS

    Science.gov (United States)

    Broccoli (Brassica oleracea L. var. italica) and lettuce (Lactuca sativa L.) were grown on Willamette sil (Pachic Ultic Argixerolls) amended 1 and 2 yr earlier with chrome tannery wastes at rates up to 192 Mg ha to determine nutrient and trace element availability. Soils were sam...

  9. Wet Oxidation Pretreatment of Tobacco Stalks and Orange Waste for Bioethanol Production. Preliminary results

    DEFF Research Database (Denmark)

    Martin, Carlos; Fernandez, Teresa; Garcia, Ariel

    2009-01-01

    Wet oxidation (WO) was used as a pretreatment method prior to enzymatic hydrolysis of tobacco stalks and orange waste. The pretreatment, performed at 195 degrees C and an oxygen pressure of 1.2 MPa, for 15 min, in the presence of Na2CO3, increased the cellulose content of the materials and gave c...

  10. Independent monitoring of a release from the waste isolation pilot plant in New Mexico, USA. Results and purpose

    Energy Technology Data Exchange (ETDEWEB)

    Thakur, Punam; Ballard, Sally [Carlsbad Environmental Monitoring and Research Center, Carlsbad, NM (United States)

    2015-07-01

    The Waste Isolation Pilot Plant (WIPP) is a transuranic (TRU) waste repository operated by the U.S. Department of Energy (DOE). The repository is emplacing defense-related transuranic (TRU) wastes into a bedded salt formation approximately 655 m (2150 ft.) below the surface of the Earth. Located near Carlsbad, New Mexico, an area with less than 30,000 people, the WIPP facility is licensed to accept TRU waste with activity concentrations of alpha-emitting isotopes >3700 Bq/m{sup 3} (> 100 nCi/g) and half-life >20 years. The upper waste acceptance limit is 0.85 TBq/liter (<23 Ci/liter) of total activity and 10 Sv/hr dose rate on contact. The repository, which opened in March 1999 will eventually contain the equivalent of ∝176,000 m{sup 3} of TRU waste. The vast majority of the waste disposed in the WIPP repository is ''contact-handled'' waste, meaning it has a surface dose rate less than 2 mSv per hour. Local acceptance of WIPP is in part due to an independent environmental monitoring program that began before and continues after WIPP began receiving nuclear waste. This independent monitoring is being conducted by the Carlsbad Environmental Monitoring and Research Center (CEMRC), which is associated with New Mexico State University. CEMRC is funded by DOE through a grant process that respects its independence in carrying out and reporting the results of environmental monitoring at and near the WIPP site. The primary focus of CEMRC monitoring is on airborne radioactive particulate; however other pathways are also monitored. Pre-disposal baseline data of various anthropogenic radionuclides present in the WIPP environment is essential for the proper evaluation of the WIPP integrity. These data are compared against disposal phase data to assess whether or not there is any radiological impact from the presence of WIPP on workers and on the regional public. The program has capabilities to detect radionuclides rapidly in case of accidental releases

  11. Vapor space characterization of waste tank 241-C-106: Results from samples collected on February 15, 1994

    International Nuclear Information System (INIS)

    McVeety, B.D.; Clauss, T.W.; Young, J.S.; Ligotke, M.W.; Goheen, S.C.; Lucke, R.B.; Pool, K.H.; McCulloch, M.; Fruchter, J.S.

    1995-06-01

    This document presents the details of the inorganic and organic analysis that was performed on samples from the headspace of Hanford waste tank 241-C-106. The results described were obtained to support the safety and toxicological evaluations. A summary of the results for the inorganic and organic analytes is included, as well as, a detailed description of the results which appears in the text

  12. Headspace vapor characterization of Hanford Waste Tank 241-U-112: Results from samples collected on 7/09/96

    International Nuclear Information System (INIS)

    Evans, J.C.; Pool, K.H.; Thomas, B.L.; Olsen, K.B.; Fruchter, J.S.; Silvers, K.L.

    1997-01-01

    This report describes the analytical results of vapor samples taken from the headspace of the waste storage tank 241-U-112 at the Hanford Site in Washington State. The results described in this report were obtained to characterize the vapors present in the tank headspace and to support safety evaluations and tank farm operations. The results include air concentrations of selected inorganic and organic analytes and grouped compounds from samples obtained by Westinghouse Hanford Company

  13. Results of questionaire survey for the measurement of radioactivity in waste water

    International Nuclear Information System (INIS)

    1992-01-01

    A questionaire for radioactivity in waste water was sent to 388 facilities, including 158 medical facilities, and all (100%) answered. Information requested included: (1) kinds and annual usage of unsealed RI, (2) measuring method of radioactivity in waste water, (3) kinds of measuring instruments and the detection limits, (4) prior treatment of measurement materials, (5) level of radioactive waste exhausted during 3 months, (6) personnel and time per month required for radioactivity measurement, (7) problems and comments in waste water management, and (8) kinds of facilities. A total of 36 unsealed RI were used. The most commonly used RI was I-125 (n=240), followed by H-3 (n=189) and P-32 (n=179). Annual level of RI was 4 GBq or less in 90% of the facilities. The most common method for measuring radioactivity was sampling method (n=241). The most common instrument for measuring radioactivity was a gamma counter for I-125 (45% of the facilities), and a liquid scintillation counter for P-32 (80%) and for C-14 and H-3 (90%). The detection limits for I-125 exceeded the radioactivity limits in 24% of the facilities. The amount of sampler was 5 cc or less in 80% of the facilities. Prio treatment was not carried out in 62.7%. Prior treatment methods reported were enrichment, evaporation, pH adjustment, and sedimentation. Half of the facilities exhausted 10 cm 3 or less of waste water during 3 months. The number of persons engaging in radioactivity measurement per month was reported to be one in 282 facilities (87%). (N.K.)

  14. Groundwater monitoring at three Oak Ridge National Laboratory inactive waste impoundments: results after one year

    Energy Technology Data Exchange (ETDEWEB)

    Francis, C. W.; Stansfield, R. G.

    1986-10-01

    To determine if the migration of potential contaminants from three inactive waste impoundments at Oak Ridge National Laboratory poses a threat to groundwater quality, at least one upgradient groundwater monitoring well and threee downgradient monitoring wells were installed at each impoundment in early 1985. These three unlined impoundments, formerly used to collect and, in some instances, treat wastewater are: the 3513 impoundment; the Old Hydrofracture Facility (OHF) impoundment; and the Homogeneous Reactor Experimnt No. 2 impoundment. Groundwater samples were collected quarterly for one year. Analyses were conducted for the groundwater protection parameters promulgated by the Resource Conservation and Recovery Act. The groundwater samples were also analyzed for polychlorinated biphenyls, copper, nickel, zinc, /sup 90/Sr, /sup 137/Cs, and tritium. The contaminants found most often to affect groundwater quality at all three waste impoundments were radionuclides. For example, mean concentrations of gross beta and gross alpha activity exceeded drinking water limits at all three sites. The gross beta limit was exceeded at the 3513 and OHF impoundments by either /sup 90/Sr or tritium levels. At the 3513 impoundment, there was substantial evidence that the downgradient groundwater has been contaminated by chromium and lead and possibly by halogenated organic compounds. At the OHF impoundment, the mean level of tritium measured in the upgradient well (about 91,000 Bq/L as compared with 80,000 Bq/L in the downgradient wells) indicated that the groundwater quality has been affected by the radioactive wastes buried in the low-level radioactive waste burial ground solid waste storage area-5 upgradient of the impoundment. Testing for groundwater contamination, disclosed statistically significant contamination at all three sites.

  15. Groundwater monitoring at three Oak Ridge National Laboratory inactive waste impoundments: results after one year

    International Nuclear Information System (INIS)

    Francis, C.W.; Stansfield, R.G.

    1986-10-01

    To determine if the migration of potential contaminants from three inactive waste impoundments at Oak Ridge National Laboratory poses a threat to groundwater quality, at least one upgradient groundwater monitoring well and threee downgradient monitoring wells were installed at each impoundment in early 1985. These three unlined impoundments, formerly used to collect and, in some instances, treat wastewater are: the 3513 impoundment; the Old Hydrofracture Facility (OHF) impoundment; and the Homogeneous Reactor Experimnt No. 2 impoundment. Groundwater samples were collected quarterly for one year. Analyses were conducted for the groundwater protection parameters promulgated by the Resource Conservation and Recovery Act. The groundwater samples were also analyzed for polychlorinated biphenyls, copper, nickel, zinc, 90 Sr, 137 Cs, and tritium. The contaminants found most often to affect groundwater quality at all three waste impoundments were radionuclides. For example, mean concentrations of gross beta and gross alpha activity exceeded drinking water limits at all three sites. The gross beta limit was exceeded at the 3513 and OHF impoundments by either 90 Sr or tritium levels. At the 3513 impoundment, there was substantial evidence that the downgradient groundwater has been contaminated by chromium and lead and possibly by halogenated organic compounds. At the OHF impoundment, the mean level of tritium measured in the upgradient well (about 91,000 Bq/L as compared with 80,000 Bq/L in the downgradient wells) indicated that the groundwater quality has been affected by the radioactive wastes buried in the low-level radioactive waste burial ground solid waste storage area-5 upgradient of the impoundment. Testing for groundwater contamination, disclosed statistically significant contamination at all three sites

  16. Performance assessment of the direct disposal in unsaturated tuff or spent nuclear fuel and high-level waste owned by USDOE: Volume 2, Methodology and results

    Energy Technology Data Exchange (ETDEWEB)

    Rechard, R.P. [ed.

    1995-03-01

    This assessment studied the performance of high-level radioactive waste and spent nuclear fuel in a hypothetical repository in unsaturated tuff. The results of this 10-month study are intended to help guide the Office of Environment Management of the US Department of Energy (DOE) on how to prepare its wastes for eventual permanent disposal. The waste forms comprised spent fuel and high-level waste currently stored at the Idaho National Engineering Laboratory (INEL) and the Hanford reservations. About 700 metric tons heavy metal (MTHM) of the waste under study is stored at INEL, including graphite spent nuclear fuel, highly enriched uranium spent fuel, low enriched uranium spent fuel, and calcined high-level waste. About 2100 MTHM of weapons production fuel, currently stored on the Hanford reservation, was also included. The behavior of the waste was analyzed by waste form and also as a group of waste forms in the hypothetical tuff repository. When the waste forms were studied together, the repository was assumed also to contain about 9200 MTHM high-level waste in borosilicate glass from three DOE sites. The addition of the borosilicate glass, which has already been proposed as a final waste form, brought the total to about 12,000 MTHM.

  17. Performance assessment of the direct disposal in unsaturated tuff or spent nuclear fuel and high-level waste owned by USDOE: Volume 2, Methodology and results

    International Nuclear Information System (INIS)

    Rechard, R.P.

    1995-03-01

    This assessment studied the performance of high-level radioactive waste and spent nuclear fuel in a hypothetical repository in unsaturated tuff. The results of this 10-month study are intended to help guide the Office of Environment Management of the US Department of Energy (DOE) on how to prepare its wastes for eventual permanent disposal. The waste forms comprised spent fuel and high-level waste currently stored at the Idaho National Engineering Laboratory (INEL) and the Hanford reservations. About 700 metric tons heavy metal (MTHM) of the waste under study is stored at INEL, including graphite spent nuclear fuel, highly enriched uranium spent fuel, low enriched uranium spent fuel, and calcined high-level waste. About 2100 MTHM of weapons production fuel, currently stored on the Hanford reservation, was also included. The behavior of the waste was analyzed by waste form and also as a group of waste forms in the hypothetical tuff repository. When the waste forms were studied together, the repository was assumed also to contain about 9200 MTHM high-level waste in borosilicate glass from three DOE sites. The addition of the borosilicate glass, which has already been proposed as a final waste form, brought the total to about 12,000 MTHM

  18. Disposal of flow-level radioactive waste in Belgium: A safety analysis for inorganic chemotoxic elements

    International Nuclear Information System (INIS)

    Mallants, D.; Volckaert, G.; Marivoet, J.; Neerdael, B.

    2000-01-01

    Low-level radioactive waste often contains large quantities of inorganic chemical substances. Due attention should therefore be given to the safety implications of both the radiological and chemical substances in the waste. Our study develops the safety assessment methodology for surface disposal with emphasis on the potential effects of inorganic nonradiological elements on human health. Contamination of groundwater was considered as the major exposure pathway. The applied methodology first screens all elements on the basis of five criteria. Conservative screening calculations were used to screen out the elements that do not pose danger to humans, and to select those that could have a negative impact and thus require further analysis. The latter was done by first calculating the elemental mass fluxes out of the repository and into the aquifer followed by the calculation of groundwater concentrations. The results showed that on the basis of the screening calculations, 75% of all elements could be classified as non-hazardous. The detailed calculations showed that the majority of the remaining elements had groundwater concentrations below the drinking water or groundwater standards. The results further showed that for a few elements the maximum groundwater concentration was above the standard, but below the background concentrations. (author)

  19. Annual Report, Fall 2016: Alternative Chemical Cleaning of Radioactive High Level Waste Tanks - Corrosion Test Results

    International Nuclear Information System (INIS)

    Wyrwas, R. B.

    2016-01-01

    The testing presented in this report is in support of the investigation of the Alternative Chemical Cleaning program to aid in developing strategies and technologies to chemically clean radioactive High Level Waste tanks prior to tank closure. The data and conclusions presented here were the examination of the corrosion rates of A285 carbon steel and 304L stainless steel exposed to two proposed chemical cleaning solutions: acidic permanganate (0.18 M nitric acid and 0.05M sodium permanganate) and caustic permanganate. (10 M sodium hydroxide and 0.05M sodium permanganate). These solutions have been proposed as a chemical cleaning solution for the retrieval of actinides in the sludge in the waste tanks, and were tested with both HM and PUREX sludge simulants at a 20:1 ratio.

  20. Initial formulation results for in situ grouting of a waste trench at ORNL Site No. 6

    International Nuclear Information System (INIS)

    Tallent, O.K.; McDaniel, E.W.; Spence, R.D.; Godsey, T.T.

    1987-01-01

    An investigation is being conducted by the Chemical Technology Division to assist the Environmental Sciences Division in developing a grout formulation for use in testing in situ grouting in a waste trench at ORNL Site 6. This final report satisfies the milestone of Subtack 12 entitled, ''Low Level Waste (LLW) Trench Grouting Assessment,'' which was initially issued as RAP-86-7, December 31, 1985. Grouts prepared from dry-solid blends containing Type I Portland cement, ASTM Class C or Class F fly ash, and bentonite, mixed water at ratios of 10 to 15 lb/gal, were evaluated. The grouts prepared with ASTM Class C fly ash exhibited significantly better properties than those prepared with ASTM Class F fly ash. The grouts containing ASTM Class C fly ash satisfy tentative performance criteria for the project. 8 refs., 7 tabs

  1. Annual Report, Fall 2016: Alternative Chemical Cleaning of Radioactive High Level Waste Tanks - Corrosion Test Results

    Energy Technology Data Exchange (ETDEWEB)

    Wyrwas, R. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-09-01

    The testing presented in this report is in support of the investigation of the Alternative Chemical Cleaning program to aid in developing strategies and technologies to chemically clean radioactive High Level Waste tanks prior to tank closure. The data and conclusions presented here were the examination of the corrosion rates of A285 carbon steel and 304L stainless steel exposed to two proposed chemical cleaning solutions: acidic permanganate (0.18 M nitric acid and 0.05M sodium permanganate) and caustic permanganate. (10 M sodium hydroxide and 0.05M sodium permanganate). These solutions have been proposed as a chemical cleaning solution for the retrieval of actinides in the sludge in the waste tanks, and were tested with both HM and PUREX sludge simulants at a 20:1 ratio.

  2. Composition, preparation, and gas generation results from simulated wastes of Tank 241-SY-101

    International Nuclear Information System (INIS)

    Bryan, S.A.; Pederson, L.R.

    1994-08-01

    This document reviews the preparation and composition of simulants that have been developed to mimic the wastes temporarily stored in Tank 241-SY-101 at Hanford. The kinetics and stoichiometry of gases that are generated using these simulants are also compared, considering the roles of hydroxide, chloride, and transition metal ions; the identities of organic constituents; and the effects of dilution, radiation, and temperature. Work described in this report was conducted for the Flammable Gas Safety Program at Pacific Northwest Laboratory, (a) whose purpose is to develop information that is necessary to mitigate potential safety hazards associated with waste tanks at the Hanford Site. The goal of this research and of related efforts at the Georgia Institute of Technology (GIT), Argonne National Laboratory (ANL), and Westinghouse Hanford Company (WHC) is to determine the thermal and thermal/radiolytic mechanisms by which flammable and other gases are produced in Hanford wastes, emphasizing those stored in Tank 241-SY-101. A variety of Tank 241-SY-101 simulants have been developed to date. The use of simulants in laboratory testing activities provides a number of advantages, including elimination of radiological risks to researchers, lower costs associated with experimentation, and the ability to systematically alter simulant compositions to study the chemical mechanisms of reactions responsible for gas generation. The earliest simulants contained the principal inorganic components of the actual waste and generally a single complexant such as N-(2-hydroxyethyl) ethylenediaminetriacetic acid (HEDTA) or ethylenediaminetriacetic acid (EDTA). Both homogeneous and heterogeneous compositional forms were developed. Aggressive core sampling and analysis activities conducted during Windows C and E provided information that was used to design new simulants that more accurately reflected major and minor inorganic components

  3. Pathogens\\' Reduction in Vermicompost Process Resulted from the Mixed Sludge Treatments-Household Wastes

    OpenAIRE

    Hossien Karimi; Mohammad Rezvani; Morteza Mohammadzadeh; Yaser Eshaghi; Mehdi Mokhtari

    2016-01-01

    Introduction: The presence of pathogenic microbial agents and pathogens in organic fertilizers causes health problems and disease transmission. The aim of this study was to evaluate the efficiency of vermicomposting process in improve the microbial quality of the compost produced. Materials and Methods: This experimental study was conducted as a pilot-scale one, in the laboratory of school of Health. In order to produce vermicompost, some perishable domestic waste were mixed whit sludge o...

  4. Bedfordshire County Structure Plan. Proposed alterations. Results of public consultation. Policy 97: Nuclear waste

    International Nuclear Information System (INIS)

    1984-01-01

    The document refers to Alterations to the County Structure Plan, proposed by Bedfordshire County Council and submitted to the Secretary of State for the Environment. An additional Alteration initiated at the County Council's meeting, dealing with nuclear waste, had not been the subject of prior public consultation. Consultation had since been arranged, and the present document summarises the responses that have been received, and describes the next action to be taken. (U.K.)

  5. Incineration of a typical LWR combustible waste and analysis of the resulting ash

    International Nuclear Information System (INIS)

    Treat, R.L.; Lokken, R.O.; Schliebe, M.J.

    1983-05-01

    In this study 4540 kg (10,000 lb) of simulated nuclear power plant combustion wastes were burned in a controlled-air incinerators. The purpose of this work was to generate ashes suitable for solidification, the products of which will be analyzed to determine if they are suitable for disposal. Two different types of waste were burned: resin and simulated crud, and general trash (paper, plastics, wood, rubber, and cloth). Volume-reduction ratios (unburned waste: ash) were 13:1 and 22:1, respectively. Approximately 20% of the ash was lost due to adherence to incinerator walls and entrainment in the off-gas stream. Losses of the volatile species cesium and iodine were 79% and 100%, respectively. The ashes were not hygroscopic, but they exhibited a pH of 4.6 to 5.0 when water was added. Corrosion of mild steel drums would occur within this pH range. The ashes contained a significant quantity of clinkers haveing lengths as great as 20 cm (8 in.). Most of the clinkers were fully incinerated and easy to crush, suggesting that standard comminuting equipment should be effective in reducing the size of clinkers to allow their solidification with the fine ashes

  6. Performance test results of noninvasive characterization of Resource Conservation and Recovery Act surrogate waste by prompt gamma neutron activation analysis

    Energy Technology Data Exchange (ETDEWEB)

    Gehrke, R.J.; Streier, G.G.

    1997-03-01

    During FY-96, a performance test was carried out with funding from the Mixed Waste Focus Area (MWFA) of the Department of Energy (DOE) to determine the noninvasive elemental assay capabilities of commercial companies for Resource Conservation and Recovery Act (RCRA) metals present in 8-gal drums containing surrogate waste. Commercial companies were required to be experienced in the use of prompt gamma neutron activation analysis (PGNAA) techniques and to have a prototype assay system with which to conduct the test assays. Potential participants were identified through responses to a call for proposals advertised in the Commerce Business Daily and through personal contacts. Six companies were originally identified. Two of these six were willing and able to participate in the performance test, as described in the test plan, with some subsidizing from the DOE MWFA. The tests were conducted with surrogate sludge waste because (1) a large volume of this type of waste awaits final disposition and (2) sludge tends to be somewhat homogeneous. The surrogate concentrations of the above RCRA metals ranged from {approximately} 300 ppm to {approximately} 20,000 ppm. The lower limit was chosen as an estimate of the expected sensitivity of detection required by noninvasive, pretreatment elemental assay systems to be of value for operational and compliance purposes and to still be achievable with state-of-the-art methods of analysis. The upper limit of {approximately} 20,000 ppm was chosen because it is the opinion of the author that assay above this concentration level is within current state-of-the-art methods for most RCRA constituents. This report is organized into three parts: Part 1, Test Plan to Evaluate the Technical Status of Noninvasive Elemental Assay Techniques for Hazardous Waste; Part 2, Participants` Results; and Part 3, Evaluation of and Comments on Participants` Results.

  7. Performance test results of noninvasive characterization of Resource Conservation and Recovery Act surrogate waste by prompt gamma neutron activation analysis

    International Nuclear Information System (INIS)

    Gehrke, R.J.; Streier, G.G.

    1997-03-01

    During FY-96, a performance test was carried out with funding from the Mixed Waste Focus Area (MWFA) of the Department of Energy (DOE) to determine the noninvasive elemental assay capabilities of commercial companies for Resource Conservation and Recovery Act (RCRA) metals present in 8-gal drums containing surrogate waste. Commercial companies were required to be experienced in the use of prompt gamma neutron activation analysis (PGNAA) techniques and to have a prototype assay system with which to conduct the test assays. Potential participants were identified through responses to a call for proposals advertised in the Commerce Business Daily and through personal contacts. Six companies were originally identified. Two of these six were willing and able to participate in the performance test, as described in the test plan, with some subsidizing from the DOE MWFA. The tests were conducted with surrogate sludge waste because (1) a large volume of this type of waste awaits final disposition and (2) sludge tends to be somewhat homogeneous. The surrogate concentrations of the above RCRA metals ranged from ∼ 300 ppm to ∼ 20,000 ppm. The lower limit was chosen as an estimate of the expected sensitivity of detection required by noninvasive, pretreatment elemental assay systems to be of value for operational and compliance purposes and to still be achievable with state-of-the-art methods of analysis. The upper limit of ∼ 20,000 ppm was chosen because it is the opinion of the author that assay above this concentration level is within current state-of-the-art methods for most RCRA constituents. This report is organized into three parts: Part 1, Test Plan to Evaluate the Technical Status of Noninvasive Elemental Assay Techniques for Hazardous Waste; Part 2, Participants' Results; and Part 3, Evaluation of and Comments on Participants' Results

  8. Results of sampling the contents of the liquid low-level waste evaporator feed tank W-22 at ORNL

    International Nuclear Information System (INIS)

    Sears, M.B.

    1996-09-01

    This report summarizes the results of the fall 1994 sampling of the contents of the liquid low- level waste (LLLW) tank W-22 at the Oak Ridge National Laboratory (ORNL). Tank W-22 is the central collection and holding tank for LLLW at ORNL before the waste is transferred to the evaporators. Samples of the tank liquid and sludge were analyzed to determine (1) the major chemical constituents, (2) the principal radionuclides, (3) the metals listed on the U.S. Environmental Protection Agency (EPA) Contract Laboratory Program Inorganic Target Analyte List, (4) organic compounds, and (5) some physical properties. The organic chemical characterization consisted of the determinations of the EPA Contract Laboratory Program Target Compound List semivolatile compounds, pesticides, and polychlorinated biphenyls (PCBs). Water-soluble volatile organic compounds were also determined. Information provided in this report forms part of the technical basis in support of (1) waste management for the active LLLW system and (2) planning for the treatment and disposal of the waste

  9. Safety assessment for the transportation of NECSA's LILW to the Vaalputs waste disposal facility

    International Nuclear Information System (INIS)

    Maphoto, K.P.; Raubenheimer, E.; Swart, H.

    2008-01-01

    The transport safety assessment was carried out with a view to assess the impact on the environment and the people living in it, from exposure to radioactivity during transportation of the radioactive materials. It provides estimates of radiological risks associated with the envisaged transport scenarios for the road transport mode. This is done by calculating the human health impact and radiological risk from transportation of LILW along the R563 route, N14 and eventually to the Vaalputs National Waste Disposal Facility. Various parameters are needed by the RADTRAN code in calculating the human health impact and risk. These include: numbers of population densities following the routes undertaken, number of stops made, and the speed at which the transport will be traversing at towards the final destination. The human health impact with regard to the dose to the public, LCF and risk associated with transportation of Necsa's LILW to the Vaalputs Waste Disposal Facility by road have been calculated using RADTRAN 5 code. The results for both accident and incident free scenarios have shown that the overall risks are insignificant and can be associated with any non-radiological transportation. (authors)

  10. Chemical composition analysis and product consistency tests to support enhanced Hanford waste glass models: Results for the January, March, and April 2015 LAW glasses

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Riley, W. T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Best, D. R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-09-03

    In this report, the Savannah River National Laboratory provides chemical analyses and Product Consistency Test (PCT) results for several simulated low activity waste (LAW) glasses (designated as the January, March, and April 2015 LAW glasses) fabricated by the Pacific Northwest National Laboratory. The results of these analyses will be used as part of efforts to revise or extend the validation regions of the current Hanford Waste Treatment and Immobilization Plant glass property models to cover a broader span of waste compositions.

  11. Chemical composition analysis and product consistency tests to support Enhanced Hanford Waste Glass Models. Results for the Augusta and October 2014 LAW Glasses

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Best, D. R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-07-07

    In this report, the Savannah River National Laboratory provides chemical analyses and Product Consistency Test (PCT) results for several simulated low activity waste (LAW) glasses (designated as the August and October 2014 LAW glasses) fabricated by the Pacific Northwest National Laboratory. The results of these analyses will be used as part of efforts to revise or extend the validation regions of the current Hanford Waste Treatment and Immobilization Plant glass property models to cover a broader span of waste compositions.

  12. Mechanisms of gas retention and release: Experimental results for Hanford single-shell waste tanks 241-A-101, 241-S-106, and 241-U-103

    International Nuclear Information System (INIS)

    Rassat, S.D.; Caley, S.M.; Bredt, P.R.; Gauglitz, P.A.; Rinehart, D.E.; Forbes, S.V.

    1998-09-01

    The 177 underground waste storage tanks at the Hanford Site contain millions of gallons of radioactive waste resulting from the purification of nuclear materials and related processes. Through various mechanisms, flammable gas mixtures of hydrogen, ammonia, methane, and nitrous oxide are generated and retained in significant quantities within the waste in many (∼25) of these tanks. The potential for large releases of retained gas from these wastes creates a flammability hazard. It is a critical component of the effort to understand the flammability hazard and a primary goal of this laboratory investigation to establish an understanding of the mechanisms of gas retention and release in these wastes. The results of bubble retention experimental studies using waste samples from several waste tanks and a variety of waste types support resolution of the Flammable Gas Safety Issue. Gas bubble retention information gained in the pursuit of safe storage will, in turn, benefit future waste operations including salt-well pumping, waste transfers, and sluicing/retrieval

  13. Results of geo-radio-monitoring for radioactive waste storage in large diameter boreholes in clayey ground

    International Nuclear Information System (INIS)

    Dmitriev, S.; Litinsky, Y.; Tkachenko, A.

    2010-01-01

    Document available in extended abstract form only. Full text of publication follows: The main purpose of the work carried out at the site of SUE MosSIA 'Radon' is to develop the system of geo-radio-monitoring for new type of storage facility (large diameter borehole) integrated into existing monitoring system of the whole site, check its effectiveness and improve the system, obtain initial results on safety aspects for using large diameter boreholes for RAW storage. Technology of large diameter boreholes (LDB) construction for low- and intermediate-level waste (LILW) isolation in moraine loams is being under development at SUE MosSIA 'Radon' site since the end of the last century. A project for construction of a demonstration unit for LILW storage in large diameter boreholes at the SUE MosSIA 'Radon' site in Sergiev Posad region has been developed taking into account specific site conditions. The main aim of the project is to develop the technology of LDB repository construction, operational procedures such as loading and retrieval, to develop and improve monitoring system for the new repository type, to get practical data on safety of radioactive wastes storage in new repositories, hermeticity of construction, and behavior of waste, waste packages, construction materials and near-field. In the case of LDB applications for LILW storage, the waste are removed from the scope of human activity into a stable geological medium. Waste are placed below the frost zone where damage of engineered barriers due to climatic factors is practically impossible. Two boreholes with 1.5 m internal diameter and 38 m depth have been drilled in 1997, equipped with engineering barriers including bentonite-concrete stone, licensed as storage facilities in 2003 and are in use now for solid and solidified RAW storage. Specific automated system of geo-radio-monitoring has been developed especially for the LDB-type repository, covering both the interior and the

  14. Headspace vapor characterization of Hanford waste tank 241-U-108: Results from samples collected on 8/29/95

    International Nuclear Information System (INIS)

    Thomas, B.L.; Clauss, T.W.; Evans, J.C.; McVeety, B.D.; Pool, K.H.; Olsten, K.B.; Fruchter, J.S.; Ligotke, M.W.

    1996-05-01

    This report describes the analytical results of vapor samples taken from the headspace of the waste storage tank 241-U-108 (Tank U-108) at the Hanford Site in Washington State. The results described in the report were obtained to characterize the vapors present in the tank headspace and to support safety evaluations and tank farm operations. The results include air concentrations of selected inorganic and organic analytes and grouped compounds from samples obtained by Westinghouse Hanford Company (WHC) and provided for analysis to Pacific Northwest National Laboratory (PNNL). Analyte concentrations were based on analytical results and, where appropriate, sample volumes provided by WHC

  15. Process Testing Results and Scaling for the Hanford Waste Treatment and Immobilization Plant (WTP) Pretreatment Engineering Platform - 10173

    International Nuclear Information System (INIS)

    Kurath, Dean E.; Daniel, Richard C.; Baldwin, David L.; Rapko, Brian M.; Barnes, Steven M.; Gilbert, Robert A.; Mahoney, Lenna A.; Huckaby, James L.

    2010-01-01

    The U.S. Department of Energy-Office of River Protections Hanford Tank Waste Treatment and Immobilization Plant (WTP) is being designed and built to pretreat and then vitrify a large portion of the wastes in Hanfords 177 underground waste storage tanks at Richland, Washington. In support of this effort, engineering-scale tests at the Pretreatment Engineering Platform (PEP) have been completed to confirm the process design and provide improved projections of system capacity. The PEP is a 1/4.5-scale facility designed, constructed, and operated to test the integrated leaching and ultrafiltration processes being deployed at the WTP. The PEP replicates the WTP leaching processes with prototypic equipment and control strategies and non-prototypic ancillary equipment to support the core processing. The testing approach used a nonradioactive aqueous slurry simulant to demonstrate the unit operations of caustic and oxidative leaching, cross-flow ultrafiltration solids concentration, and solids washing. Parallel tests conducted at the laboratory scale with identical simulants provided results that allow scale-up factors to be developed between the laboratory and PEP performance. This paper presents the scale-up factors determined between the laboratory and engineering-scale results and presents arguments that extend these results to the full-scale process.

  16. Radionuclide Incorporation in Secondary Crystalline Minerals Resulting from Chemical Weathering of Selected Waste Glasses: Progress Report: Task kd.5b

    International Nuclear Information System (INIS)

    Mattigod, Shas V.; Serne, R. Jeffrey; Legore, Virginia L.; Parker, Kent E.; Orr, Robert D.; McCready, David E.; )

    2003-01-01

    Experiments were conducted by Pacific Northwest National Laboratory to evaluate potential incorporation of radionuclides in secondary mineral phases that form from weathering vitrified nuclear waste glasses. These experiments were conducted as part of the Immobilized Low-Activity Waste-Performance Assessment (ILAW-PA) to generate data on radionuclide mobilization and transport in a near-field environment of disposed vitrified wastes. The results of these experiments demonstrated that radionuclide sequestration can be significantly enhanced by promoting the formation of cage structured minerals such as sodalite from weathering glasses. These results have important implications regarding radionuclide sequestration/mobilization aspects that are not currently accounted for in the ILAW PA. Additional studies are required to confirm the results and to develop an improved understanding of the mechanisms of sequestration of radionuclides into the secondary and tertiary weathering products o f the ILAW glass to help refine how contaminants are released from the near-field disposal region out into the accessible environment. Of particular interest is to determine whether the contaminants remain sequestered in the glass weathering products for hundreds to thousands of years. If the sequestration can be shown to continue for long periods, another immobilization process can be added to the PA analysis and predicted risks should be lower than past predictions

  17. Estimation of waste water treatment plant methane emissions: methodology and results from a short campaign

    Science.gov (United States)

    Yver-Kwok, C. E.; Müller, D.; Caldow, C.; Lebegue, B.; Mønster, J. G.; Rella, C. W.; Scheutz, C.; Schmidt, M.; Ramonet, M.; Warneke, T.; Broquet, G.; Ciais, P.

    2013-10-01

    This paper describes different methods to estimate methane emissions at different scales. These methods are applied to a waste water treatment plant (WWTP) located in Valence, France. We show that Fourier Transform Infrared (FTIR) measurements as well as Cavity Ring Down Spectroscopy (CRDS) can be used to measure emissions from the process to the regional scale. To estimate the total emissions, we investigate a tracer release method (using C2H2) and the Radon tracer method (using 222Rn). For process-scale emissions, both tracer release and chamber techniques were used. We show that the tracer release method is suitable to quantify facility- and some process-scale emissions, while the Radon tracer method encompasses not only the treatment station but also a large area around. Thus the Radon tracer method is more representative of the regional emissions around the city. Uncertainties for each method are described. Applying the methods to CH4 emissions, we find that the main source of emissions of the plant was not identified with certainty during this short campaign, although the primary source of emissions is likely to be from solid sludge. Overall, the waste water treatment plant represents a small part (3%) of the methane emissions of the city of Valence and its surroundings,which is in agreement with the national inventories.

  18. Analytical Chemistry and Materials Characterization Results for Debris Recovered from Nitrate Salt Waste Drum S855793

    Energy Technology Data Exchange (ETDEWEB)

    Martinez, Patrick Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Chamberlin, Rebecca M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Schwartz, Daniel S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Worley, Christopher Gordon [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Garduno, Katherine [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Lujan, Elmer J. W. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Borrego, Andres Patricio [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Castro, Alonso [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Colletti, Lisa Michelle [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Fulwyler, James Brent [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Holland, Charlotte S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Keller, Russell C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Klundt, Dylan James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Martinez, Alexander [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Martin, Frances Louise [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Montoya, Dennis Patrick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Myers, Steven Charles [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Porterfield, Donivan R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Schake, Ann Rene [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Schappert, Michael Francis [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Soderberg, Constance B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Spencer, Khalil J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stanley, Floyd E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Thomas, Mariam R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Townsend, Lisa Ellen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Xu, Ning [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-09-16

    Solid debris was recovered from the previously-emptied nitrate salt waste drum S855793. The bulk sample was nondestructively assayed for radionuclides in its as-received condition. Three monoliths were selected for further characterization. Two of the monoliths, designated Specimen 1 and 3, consisted primarily of sodium nitrate and lead nitrate, with smaller amounts of lead nitrate oxalate and lead oxide by powder x-ray diffraction. The third monolith, Specimen 2, had a complex composition; lead carbonate was identified as the predominant component, and smaller amounts of nitrate, nitrite and carbonate salts of lead, magnesium and sodium were also identified. Microfocused x-ray fluorescence (MXRF) mapping showed that lead was ubiquitous throughout the cross-sections of Specimens 1 and 2, while heteroelements such as potassium, calcium, chromium, iron, and nickel were found in localized deposits. MXRF examination and destructive analysis of fragments of Specimen 3 showed elevated concentrations of iron, which were broadly distributed through the sample. With the exception of its high iron content and low carbon content, the chemical composition of Specimen 3 was within the ranges of values previously observed in four other nitrate salt samples recovered from emptied waste drums.

  19. Low-level radioactive waste disposal. Study of a conceptual nuclear energy center at Green River, Utah

    International Nuclear Information System (INIS)

    Card, D.H.; Hunter, P.H.; Barg, D.; de Souza, F.; Felthauser, K.; Winkler, V.; White, R.

    1982-02-01

    This document constitutes a segment of a feasibility study investigating the ramifications of constructing a nuclear energy center in an arid western region. In this phase of the study, the alternatives for disposing of the low-level waste on the site are compared with the alternative of transporting the waste to the nearest commercial waste disposal site for permanent disposal. Both radiological and nonradiological impacts on the local socioeconomic infrastructure and the environment are considered. Disposal on the site was found to cost considerably less than off-site disposal with only negligible impacts associated with the disposal option on either mankind or the environment

  20. The results of an ecological risk assessment screening at the Idaho National Engineering`s waste area group 2

    Energy Technology Data Exchange (ETDEWEB)

    VanHorn, R.

    1995-11-01

    The Idaho National Engineering Laboratory (INEL) is a Department of Energy (DOE) facility located in southeastern Idaho and occupies approximately 890 square miles on the northwestern portion of the eastern Snake River Plain. INEL has been devoted to nuclear energy research and related activities since its establishment in 1949. In the process of fulfilling this mission, wastes were generated, including radioactive and hazardous materials. Most materials were effectively stored or disposed of, however, some release of contaminants to the environment has occurred. For this reason, the INEL was listed by the US environmental Protection Agency on the National Priorities List (NPL), in November, 1989. This report describes the results of an ecological risk assessment performed for the Waste Area Groups 2 (WAG 2) at the INEL. It also summarizes the performance of screening level ecological risk assessments (SLERA).

  1. The results of an ecological risk assessment screening at the Idaho National Engineering's waste area group 2

    International Nuclear Information System (INIS)

    VanHorn, R.

    1995-01-01

    The Idaho National Engineering Laboratory (INEL) is a Department of Energy (DOE) facility located in southeastern Idaho and occupies approximately 890 square miles on the northwestern portion of the eastern Snake River Plain. INEL has been devoted to nuclear energy research and related activities since its establishment in 1949. In the process of fulfilling this mission, wastes were generated, including radioactive and hazardous materials. Most materials were effectively stored or disposed of, however, some release of contaminants to the environment has occurred. For this reason, the INEL was listed by the US environmental Protection Agency on the National Priorities List (NPL), in November, 1989. This report describes the results of an ecological risk assessment performed for the Waste Area Groups 2 (WAG 2) at the INEL. It also summarizes the performance of screening level ecological risk assessments (SLERA)

  2. Vitrification of Hanford wastes in a joule-heated ceramic melter and evaluation of resultant canisterized product

    International Nuclear Information System (INIS)

    Chapman, C.C.; Buelt, J.L.; Slate, S.C.; Katayama, Y.B.; Bunnell, L.R.

    1979-08-01

    Experience gained in the week-long vitrification test and characterization of the glass produced in the run support the following conclusions: The Hanford waste simulated in this test can be readily vitrified in a joule-heated ceramic melter. Physical properties of the molten glass were entirely compatible with melter operation. The average feed rate of 106 kg/h is high enough to make the ceramic melter a feasible piece of equipment for vitrifying Hanford wastes. The glass produced in this trial had good chemical durability, 6(10) -5 g/cm 2 -d. When one of the canisters was purposely dropped onto a steel pad, the damage was limited to deformation of the steel can in the impact area, cracking of a weld, and fracturing of glass in the immediate vicinity of the impact area. No glass was released from the canister as a result of the drop test. The results of this vitrification test support the technical feasibility of vitrifying Hanford wastes by means of a joule-heated ceramic melter. Surface area for large glass castings is equivalent to the mass median particle diameters between 4.27 cm (1.75 in.) and 8.91 cm (3.51 in.) even when allowed to cool rapidly by standing in ambient air. Large canisters (up to 0.91 m in dia) can be cast without large voids while standing in air if the fill rate is over 100 kg/h. 34 figures, 10 tables

  3. Introduction of microbial nutrients in a nuclear fuel waste disposal vault as a result of excavation and operation activities

    International Nuclear Information System (INIS)

    Stroes-Gascoyne, S.; Gascoyne, M.; Onagi, D.; Thomas, D.A.; Hamon, C.J.; Watson, R.; Porth, R.J.

    1996-08-01

    A nuclear fuel waste disposal vault would not likely be a sterile environment. Bacterial activity would be expected in those areas of the vault conducive to bacterial life, i.e., where effects of heat, moisture content, radiation and compaction would not prevent or severely restrict bacterial life and where suitable and sufficient nutrients would be present. An inventory of bacterial nutrients that would be emplaced 'intentionally' with vault materials (fuel waste, waste containers, buffer and backfill materials) has been made previously. This report assesses bacterial nutrients that would be added 'inadvertently' to a vault in the form of residues of materials used to excavate and operate a vault. Measurements of blasting material residues in the various water supplies, excavated broken rock (muck) and in cores drilled in old and new tunnel walls were made at AECL's Underground Research Laboratory. Results show that the largest potential nutrient addition (both carbon and nitrogen) to a vault would result from using untreated excavated broken rock as part of the backfill. (author). 16 refs., 4 tabs., 10 figs

  4. Introduction of microbial nutrients in a nuclear fuel waste disposal vault as a result of excavation and operation activities

    Energy Technology Data Exchange (ETDEWEB)

    Stroes-Gascoyne, S; Gascoyne, M; Onagi, D; Thomas, D A; Hamon, C J; Watson, R; Porth, R J

    1996-08-01

    A nuclear fuel waste disposal vault would not likely be a sterile environment. Bacterial activity would be expected in those areas of the vault conducive to bacterial life, i.e., where effects of heat, moisture content, radiation and compaction would not prevent or severely restrict bacterial life and where suitable and sufficient nutrients would be present. An inventory of bacterial nutrients that would be emplaced `intentionally` with vault materials (fuel waste, waste containers, buffer and backfill materials) has been made previously. This report assesses bacterial nutrients that would be added `inadvertently` to a vault in the form of residues of materials used to excavate and operate a vault. Measurements of blasting material residues in the various water supplies, excavated broken rock (muck) and in cores drilled in old and new tunnel walls were made at AECL`s Underground Research Laboratory. Results show that the largest potential nutrient addition (both carbon and nitrogen) to a vault would result from using untreated excavated broken rock as part of the backfill. (author). 16 refs., 4 tabs., 10 figs.

  5. Thermal, chemical, and mass transport processes induced in abyssal sediments by the emplacement of nuclear wastes: Experimental and modelling results

    International Nuclear Information System (INIS)

    McVey, D.F.; Erickson, K.L.; Seyfried, W.E. Jr.

    1983-01-01

    In this chapter the authors discuss the current status of heat and mass transport studies in the marine red clay sediments that are being considered as a nuclear waste isolation medium and review analytical and experimental studies. Calculations based on numerical models indicate that for a maximum allowable sediment-canister interface temperatures of 200 0 to 250 0 C, the sediment can absorb about 1.5kW initial power from waste buried 30 m in the sediment in a canister that is 3 m long and 0.3 m in diameter. The resulting fluid displacement due to convections is found to be small, less than 1 m. Laboratory studies of the geochemical effects induced by heating sediment-seawater mixtures indicate that the canister and waste form should be designed to resist a hot, relatively acidic oxidizing environment. Since the thermally altered sediment volume of about 5.5 m/sup 3/ is small relative to the sediment volume overlying the canister, the acid and oxidizing conditions should significantly affect the properties of the far field only if thermodiffusional process (Soret effect) prove to be significant. If thermodiffusional effects are important, however, near-field chemistry will differ considerably from that predicted from results of constant temperature sediment-seawater interaction experiments

  6. Critique of rationale for transmutation of nuclear waste

    International Nuclear Information System (INIS)

    Smith, C.F.; Cohen, J.J.

    1980-07-01

    It has been suggested that nuclear transmutation could be used in the elimination or reduction of hazards from radioactive wastes. The rationale for this suggestion is the subject of this paper. The objectives of partitioning-transmutation are described. The benefits are evaluated. The author concludes that transmutation would appear at best to offer the opportunity of reducing an already low risk. This would not seem to be justifiable considering the cost. If non-radiological risks are considered, there is a negative total benefit

  7. Results from Nevada Nuclear Waste Storage Investigations (NNWSI) Series 3 spent fuel dissolution tests

    International Nuclear Information System (INIS)

    Wilson, C.N.

    1990-06-01

    The dissolution and radionuclide release behavior of spent fuel in groundwater is being studied by the Yucca Mountain Project (YMP), formerly the Nevada Nuclear Waste Storage Investigations (NNWSI) Project. Specimens prepared from pressurized water reactor fuel rod segments were tested in sealed stainless steel vessels in Nevada Test Site J-13 well water at 85 degree C and 25 degree C. The test matrix included three specimens of bare-fuel particles plus cladding hulls, two fuel rod segments with artificially defected cladding and water-tight end fittings, and an undefected fuel rod section with watertight end fittings. Periodic solution samples were taken during test cycles with the sample volumes replenished with fresh J-13 water. Test cycles were periodically terminated and the specimens restarted in fresh J-13 water. The specimens were run for three cycles for a total test duration of 15 months. 22 refs., 32 figs., 26 tabs

  8. Evaluating long-term performance of in situ vitrified waste forms: Methodology and results

    International Nuclear Information System (INIS)

    McGrail, B.P.; Olson, K.M.

    1992-11-01

    In situ vitrification (ISV) is an emerging technology for the remediation of hazardous and radioactive waste sites. The concept relies on the principle of Joule heating to raise the temperature of a soil between an array of electrodes above the melting temperature. After cooling, the melt solidifies into a massive glass and crystalline block similar to naturally occurring obsidian. Determining the long-term performance of ISV products in a changing regulatory environment requires a fundamental understanding of the mechanisms controlling the dissolution behavior of the material. A series of experiments was performed to determine the dissolution behavior of samples produced from the ISV processing of typical soils from the Idaho National Engineering Laboratory subsurface disposal area. Dissolution rate constant measurements were completed at 90 degrees C over the pH range 2 to 11 for one sample obtained from a field test of the ISV process

  9. Integrated water management system - Description and test results. [for Space Station waste water processing

    Science.gov (United States)

    Elden, N. C.; Winkler, H. E.; Price, D. F.; Reysa, R. P.

    1983-01-01

    Water recovery subsystems are being tested at the NASA Lyndon B. Johnson Space Center for Space Station use to process waste water generated from urine and wash water collection facilities. These subsystems are being integrated into a water management system that will incorporate wash water and urine processing through the use of hyperfiltration and vapor compression distillation subsystems. Other hardware in the water management system includes a whole body shower, a clothes washing facility, a urine collection and pretreatment unit, a recovered water post-treatment system, and a water quality monitor. This paper describes the integrated test configuration, pertinent performance data, and feasibility and design compatibility conclusions of the integrated water management system.

  10. The study of the container types used for transport and final disposal of the radioactive wastes resulting from decommissioning of nuclear facilities

    International Nuclear Information System (INIS)

    Postelnicu, C.

    1998-01-01

    The purpose of the present paper is to select from a variety of package forms and capacities some containers which will be used for transport and disposal of the radioactive wastes resulting from decommissioning of nuclear facilities into the National Repository for Radioactive Waste - Baita, Bihor county. Taken into account the possibilities of railway and / or road transport and waste disposal in our country, detailed container classification was given in order to use them for radioactive waste transport and final disposal from decommissioning of IFIN-HH Research Reactor. (author)

  11. Long-term leaching behavior of simulated Savannah River Plant waste glass: Part 1, MCC-1 leachability results, four-year leaching data

    International Nuclear Information System (INIS)

    Wicks, G.G.; Stone, J.A.; Chandler, G.T.; Williams, S.

    1986-08-01

    Long-term leaching data were obtained on SRP 131/TDS waste glass using MCC-1 or slightly modified MCC-1 standard leaching tests. Experiments were conducted out to four years at 40 0 C and 3-1/2 years at 90 0 C. These experiments have produced the longest standardized leaching data currently available in the waste management community. Long-term leaching data provide important input to modeling of waste glass behavior and ultimate prediction of waste glass performance. In this study, the leaching behavior of SRP waste glass was found to be excellent; leachates based on a variety of elements were not only very low, but also improved with increasing time. In addition to these data, results are also reported from another independent Savannah River study. Leaching behavior at 40 0 C and 90 0 C was assessed not only for a similar SRP 131 waste glass composition, but also for extreme waste glass compositions involving high-iron and high-aluminum waste. In addition, these experiments were performed using not only a standard deionized water leachant, but also simplified brine and silicate groundwater simulations. These two large data bases will be summarized and correlated along with some of the more interesting results recently reported in another study, a two-year leaching program performed on a similar SRP waste glass composition at Battelle Pacific Northwest Laboratories

  12. Ferrocyanide Safety Program: Analysis of postulated energetic reactions and resultant aerosol generation in Hanford Site Waste Tanks

    International Nuclear Information System (INIS)

    Postma, A.K.; Dickinson, D.R.

    1995-09-01

    This report reviews work done to estimate the possible consequences of postulated energetic reactions in ferrocyanide waste stored in underground tanks at the Hanford Site. The issue of explosive reactions was raised in the 1987 Environmental Impact Statement (EIS), where a detonation-like explosion was postulated for the purpose of defining an upper bound on dose consequences for various disposal options. A review of the explosion scenario by the General Accounting Office (GAO) indicated that the aerosol generation and consequent radioactive doses projected for the explosion postulated in the EIS were understated by one to two orders of magnitude. The US DOE has sponsored an extensive study of the hazard posed by uncontrolled exothermic reactions in ferrocyanide waste, and results obtained during the past three years have allowed this hazard to be more realistically assessed. The objective of this report is to summarize the improved knowledge base that now indicates that explosive or vigorous chemical reactions are not credible in the ferrocyanide waste stored in underground tanks. This improved understanding supports the decision not to proceed with further analyses or predictions of the consequences of such an event or with aerosol tests in support of such predictions. 53 refs., 2 tabs

  13. National Waste Repository Novi Han operational safety analysis report. Safety assessment methodology

    International Nuclear Information System (INIS)

    2003-01-01

    The scope of the safety assessment (SA), presented includes: waste management functions (acceptance, conditioning, storage, disposal), inventory (current and expected in the future), hazards (radiological and non-radiological) and normal and accidental modes. The stages in the development of the SA are: criteria selection, information collection, safety analysis and safety assessment documentation. After the review the facilities functions and the national and international requirements, the criteria for safety level assessment are set. As a result from the 2nd stage actual parameters of the facility, necessary for safety analysis are obtained.The methodology is selected on the base of the comparability of the results with the results of previous safety assessments and existing standards and requirements. The procedure and requirements for scenarios selection are described. A radiological hazard categorisation of the facilities is presented. Qualitative hazards and operability analysis is applied. The resulting list of events are subjected to procedure for prioritization by method of 'criticality analysis', so the estimation of the risk is given for each event. The events that fall into category of risk on the boundary of acceptability or are unacceptable are subjected to the next steps of the analysis. As a result the lists with scenarios for PSA and possible design scenarios are established. PSA logical modeling and quantitative calculations of accident sequences are presented

  14. Waste tank vapor project: Vapor space characterization of waste tank 241-BY-104: Results from samples collected on June 24, 1994

    International Nuclear Information System (INIS)

    Clauss, T.W.; Ligotke, M.W.; McVeety, B.D.; Pool, K.H.; Lucke, R.B.; Fruchter, J.S.; Goheen, S.C.

    1994-11-01

    This report describes results of the analyses of tank-headspace samples taken from Hanford waste Tank 241-BY-104 (referred to as Tank BY-104) on June 24, 1994. The Pacific Northwest Laboratory (PNL) contracted with Westinghouse Hanford Company (WHC) to provide sampling devices and analyze inorganic and organic samples collected from the tank headspace. The sample job was designated S4019 and was performed by WHC on June 24, 1994 using the vapor sampling system (VSS). The results of the analyses are expected to be used in the determination of safety and toxicological issues related to the tank-headspace gas as described in the WHC report entitled Data Quality Objectives for Generic In-Tank Health and Safety Vapor Issue Resolution, WHC-SD-WM-DQO-002, Rev. 0. Sampling devices, including 16 sorbent trains (for inorganic analyses), and 5 SUMMA trademark canisters (for organic analyses), were supplied to the WHC sampling staff on June 20, 1994. Samples were taken (by WHC) on June 24. The samples were returned from the field on June 27. The inorganic samples delivered to PNL on chain-of-custody (COC) 006893 included 16 sorbent trains as described in Tables 2.2, 2.3, and 2.4. Additional inorganic blank spikes were obtained from related sample jobs. SUMMA trademark samples delivered to PNL on COC 006896 included one ambient air sample, one ambient-air sample through the sampling system, and three tank-headspace SUMMA trademark canister samples. The samples were inspected upon delivery to the 326/23B laboratory and logged into PNL laboratory record book 55408. Custody of the sorbent trains was transferred to PNL personnel performing the inorganic analysis and stored at refrigerated (≤10 degrees C) temperature until the time of analysis. Access to the 326/23B laboratory is limited to PNL personnel working on the waste-tank safety program

  15. A preliminary analysis of the risk of transporting nuclear waste to potential candidate commercial repository sites

    International Nuclear Information System (INIS)

    Madsen, M.M.

    1984-01-01

    In accordance with the provisions of the Nuclear Waste Policy Act of 1982, environmental assessments for potential candidate sites are required to provide a basis for selection of the first site for disposal of commercial radioactive waste in deep geologic repositories. A preliminary analysis of the impacts of transportation for each of the five potential sites will be described. Transportation was assumed to be entirely by truck or entirely by rail in order to obtain bounding impacts. This paper presents both radiological and nonradiological risks for the once-through fuel cycle

  16. Environmental surveillance for the INEL radioactive waste management complex. Annual report, 1979

    International Nuclear Information System (INIS)

    Wickham, L.E.; Janke, D.H.

    1980-12-01

    This document is the 1979 annual environmental surveillance report for the Radioactive Waste Management Complex (RWMC) of the Idaho National Engineering Laboratory. Included are tabulated data from and discussions about routine radiological monitoring of atmospheric, hydrologic, geologic, and biotic environments of the RWMC. Also included are discussions of selected nonradiological pollutants (e.g., sodium, etc.). It is concluded that (a) RWMC operations have not adversely affected local, existing environments; (b) environmental conditions within the Transuranic Storage Area are not corrosive enough to adversely affect transuranic waste storage containers, and (c) the addition of lakebed soil to pit, trench, and soil test plot areas has altered the moisture cycle characteristic of RWMC soil

  17. Waste Tank Vapor Program: Vapor space characterization of Waste Tank 241-T-107. Results from samples collected on January 18, 1995

    International Nuclear Information System (INIS)

    Pool, K.H.; Lucke, R.B.; McVeety, B.D.

    1995-06-01

    This report describes inorganic and organic analyses results from samples obtained from the headspace of the Hanford waste storage Tank 241-T-107 (referred to as Tank T-107). The results described here were obtained to support safety and toxicological evaluations. A summary of the results for inorganic and organic analytes is listed in Table 1. Detailed descriptions of the results appear in the text. Quantitative results were obtained for the inorganic compounds ammonia (NH 3 ), nitrogen dioxide (NO 2 ), nitric oxide (NO), and water (H 2 O). Sampling for hydrogen cyanide (HCN) and sulfur oxides (SO x ) was not requested. In addition, quantitative results were obtained for the 39 TO-14 compounds plus an additional 14 analytes. Of these, I was observed above the 5-ppbv reporting cutoff. Six organic tentatively identified compounds (TICs) were observed above the reporting cutoff of (ca.) 10 ppbv and are reported with concentrations that are semiquantitative estimates based on internal-standard response factors. The estimated concentration of all 7 organic analytes observed in the tank headspace are listed in Table I and account for approximately 100% of the total organic components in Tank T-107. Two permanent gases, carbon dioxide (CO 2 ) and nitrous oxide (N 2 O), were also detected in the tank-headspace samples

  18. Research and demonstration results for a new "Double-Solution" technology for municipal solid waste treatment.

    Science.gov (United States)

    Erping, Li; Haoyun, Chen; Yanyang, Shang; Jun, Pan; Qing, Hu

    2017-11-01

    In this paper, the pyrolysis characteristics of six typical components in municipal solid waste (MSW) were investigated through a TG-FTIR combined technique and it was concluded that the main pyrolysis process of the biomass components (including food residues, sawdust and paper) occurred at 150-600°C. The main volatiles were multi-component gas including H 2 O, CO 2 , and CO. The main pyrolysis temperatures of three artificial products (PP, PVC and leather) was ranged from 200to 500°C. The wavelength of small molecule gases (CH 4 , CO 2 and CO) and the the chemical bonds (CO and CC) were observed in the infrared spectrum Based on the pyrolysis temperature interval and volatile constituent, a new "double-solution" process of pyrolysis and oxygen-enrichment decomposition MSW was designed. To achieve this process, a double-solution project was built for the direct treatment of MSW (10t/d). The complete setup of equipment and analysis of the byproducts has been reported in this paper to indicate the performance of this process. Energy balance and economic benefits were analysed for the process supporting. It was successfully demonstrated that the double-solution process was the environmentally friendly alternative method for MSW treatment in Chinese rural areas. Copyright © 2017 Elsevier Ltd. All rights reserved.

  19. Creep of ocean sediments resulting from the isolation of radioactive wastes

    International Nuclear Information System (INIS)

    Dawson, P.R.; Chavez, P.F.; Lipkin, J.; Silva, A.J.

    1983-01-01

    Long-term disposal of high-level radioactive wastes in subseabed sediments requires that the sediments constitute the principal barrier to the release of radionuclides over very long times. In this chapter the development of the components for mathematical modelling of creep deformations of marine sediments is presented. This development includes formulation of the conservation equations and constitutive equations that describe coupled movement and heating of the fully saturated porous sediments. Numerical methods for solving the system of governing equations for complicated two-dimensional geometrics are discussed, and the program of laboratory tests for understanding the mechanical behavior of the ocean sediments is presented. Using properties taken from published literature on the creep of clays, two problems were analyzed to obtain preliminary estimates of the behavior. Analysis of cavity closure following emplacement showed that the sediment would flow around the canister before heating would significantly alter the temperature field. Large-scale motion caused by density gradients in the sediment was predicted to be small

  20. Actinide chemistry research supporting the Waste Isolation Pilot Plant (WIPP): FY94 results

    Energy Technology Data Exchange (ETDEWEB)

    Novak, C.F. [ed.

    1995-08-01

    This document contains six reports on actinide chemistry research supporting the Waste Isolation Pilot Plant (WIPP). These reports, completed in FY94, are relevant to the estimation of the potential dissolved actinide concentrations in WIPP brines under repository breach scenarios. Estimates of potential dissolved actinide concentrations are necessary for WIPP performance assessment calculations. The specific topics covered within this document are: the complexation of oxalate with Th(IV) and U(VI); the stability of Pu(VI) in one WIPP-specific brine environment both with and without carbonate present; the solubility of Nd(III) in a WIPP Salado brine surrogate as a function of hydrogen ion concentration; the steady-state dissolved plutonium concentrations in a synthetic WIPP Culebra brine surrogate; the development of a model for Nd(III) solubility and speciation in dilute to concentrated sodium carbonate and sodium bicarbonate solutions; and the development of a model for Np(V) solubility and speciation in dilute to concentrated sodium Perchlorate, sodium carbonate, and sodium chloride media.

  1. Actinide chemistry research supporting the Waste Isolation Pilot Plant (WIPP): FY94 results

    International Nuclear Information System (INIS)

    Novak, C.F.

    1995-08-01

    This document contains six reports on actinide chemistry research supporting the Waste Isolation Pilot Plant (WIPP). These reports, completed in FY94, are relevant to the estimation of the potential dissolved actinide concentrations in WIPP brines under repository breach scenarios. Estimates of potential dissolved actinide concentrations are necessary for WIPP performance assessment calculations. The specific topics covered within this document are: the complexation of oxalate with Th(IV) and U(VI); the stability of Pu(VI) in one WIPP-specific brine environment both with and without carbonate present; the solubility of Nd(III) in a WIPP Salado brine surrogate as a function of hydrogen ion concentration; the steady-state dissolved plutonium concentrations in a synthetic WIPP Culebra brine surrogate; the development of a model for Nd(III) solubility and speciation in dilute to concentrated sodium carbonate and sodium bicarbonate solutions; and the development of a model for Np(V) solubility and speciation in dilute to concentrated sodium Perchlorate, sodium carbonate, and sodium chloride media

  2. Headspace vapor characterization of Hanford waste tank 241-U-109: Results from samples collected on 8/10/95

    International Nuclear Information System (INIS)

    Evans, J.C.; Thomas, B.L.; Pool, K.H.; Olsen, K.B.; Fruchter, J.S.; Silvers, K.L.

    1996-05-01

    This report describes the analytical results of vapor samples taken from the headspace of the waste storage tank 241-U-109 (Tank U-109) At the Hanford Site in Washington State. The results described in this report were obtained to characterize the vapors present in the tank headspace and to support safety evaluations and tank farm operations. This tank is on the Hydrogen Waste List. The results include air concentrations of selected inorganic and organic analytes and grouped compounds from samples obtained by Westinghouse Hanford Company (WHC) and provided for analysis to Pacific Northwest National Laboratory (PNNL). Analyses were performed by the Vapor Analytical Laboratory (VAL) at PNNL. Analyte concentrations were based on analytical results and, where appropriate, sample volumes provided by WHC. A summary of the inorganic analytes, permanent gases and total non-methane hydrocarbons is listed in a table. The three highest concentration analytes detected in SUMMA trademark canister and triple sorbent trap samples is also listed in the table. Detailed descriptions of the analytical results appear in the text

  3. Tank Vapor Characterization Project: Vapor space characterization of waste Tank A-101, Results from samples collected on June 8, 1995

    International Nuclear Information System (INIS)

    Pool, K.H.; Clauss, T.W.; McVeety, B.D.; Evans, J.C.; Thomas, B.L.; Olsen, K.B.; Fruchter, J.S.; Ligotke, M.W.

    1995-11-01

    This report describes the analytical results of vapor samples taken from the headspace of the waste storage tank 241-A-101 (Tank A-101) at the Hanford Site in Washington State. The results described in this report were obtained to characterize the vapors present in the tank headspace and to support safety evaluations and tank-farm operations. The results include air concentrations of selected inorganic and organic analytes and grouped compounds from samples obtained by Westinghouse Hanford Company (WHC) and provided for analysis to Pacific Northwest National Laboratory (PNL). Analyses were performed by the Vapor Analytical Laboratory (VAL) at PNL. Analyte concentrations were based on analytical results and, where appropriate, sample volumes provided by WHC. A summary of the results is listed in Table 1. Detailed descriptions of the analytical results appear in the text

  4. Safety assessment for the transportation of NECSA's LILW to the Vaalputs waste disposal facility

    Energy Technology Data Exchange (ETDEWEB)

    Maphoto, K.P.; Raubenheimer, E.; Swart, H. [Nuclear Liabilities Management, NECSA, P O Box 582, Pretoria, 0001 (South Africa)

    2008-07-01

    The transport safety assessment was carried out with a view to assess the impact on the environment and the people living in it, from exposure to radioactivity during transportation of the radioactive materials. It provides estimates of radiological risks associated with the envisaged transport scenarios for the road transport mode. This is done by calculating the human health impact and radiological risk from transportation of LILW along the R563 route, N14 and eventually to the Vaalputs National Waste Disposal Facility. Various parameters are needed by the RADTRAN code in calculating the human health impact and risk. These include: numbers of population densities following the routes undertaken, number of stops made, and the speed at which the transport will be traversing at towards the final destination. The human health impact with regard to the dose to the public, LCF and risk associated with transportation of Necsa's LILW to the Vaalputs Waste Disposal Facility by road have been calculated using RADTRAN 5 code. The results for both accident and incident free scenarios have shown that the overall risks are insignificant and can be associated with any non-radiological transportation. (authors)

  5. Vapor space characterization of waste Tank 241-U-106: Results from samples collected on March 7, 1995. Waste Tank Vapor Program

    International Nuclear Information System (INIS)

    Klinger, G.S.; Lucke, R.B.; McVeety, B.D.

    1995-07-01

    This report describes inorganic and organic analyses results from samples obtained from the headspace of the Hanford waste storage Tank 241-U-106 (referred to as Tank U-106). The results described here were obtained to support safety and toxicological evaluations. Quantitative results were obtained for the inorganic compounds ammonia (NH 3 ), nitrogen dioxide (NO 2 ), nitric oxide (NO), and water (H 2 O) Sampling for hydrogen cyanide (HCN) and sulfur oxides (SO x ) was not requested. The NH 3 concentration was 16% greater than that determined from an ISS sample obtained in August 1994; the H 2 O concentration was about 10% less. In addition, quantitative results were obtained for the 39 TO-14 compounds plus an additional 14 analytes. Of these, 5 were observed in two or more canisters above the 5-ppbv reporting cutoff. Eleven organic tentatively identified compounds (TICS) were observed in two or more canisters above the reporting cutoff of (ca.) 10 ppbv and are reported with concentrations that are semiquantitative estimates based on internal-standard response factors. The 10 organic analytes with the highest estimated concentrations account for approximately 90% of the total organic components in Tank U-106. Three permanent gases, nitrous oxide (N 2 O), hydrogen (H 2 ) and carbon dioxide (COD were also detected

  6. Spent fuel and high level waste: Chemical durability and performance under simulated repository conditions. Results of a coordinated research project 1998-2004

    International Nuclear Information System (INIS)

    2007-10-01

    This publication contains the results of an IAEA Coordinated Research Project (CRP). It provides a basis for understanding the potential interactions of waste form and repository environment, which is necessary for the development of the design and safety case for deep disposal. Types of high level waste matrices investigated include spent fuel, glasses and ceramics. Of particular interest are the experimental results pertaining to ceramic forms such as SYNROC. This publication also outlines important areas for future work, namely, standardized, collaborative experimental protocols for package-release studies, structured development and calibration of predictive models linking the performance of packaged waste and the repository environment, and studies of the long term behaviour of the wastes, including active waste samples

  7. Summary of EPA's risk assessment results from the analysis of alternative methods of low-level waste disposal

    International Nuclear Information System (INIS)

    Bandrowski, M.S.; Hung, C.Y.; Meyer, G.L.; Rogers, V.C.

    1987-01-01

    Evaluation of the potential health risk and individual exposure from a broad number of disposal alternatives is an important part of EPA's program to develop generally applicable environmental standards for the land disposal of low-level radioactive wastes (LLW). The Agency has completed an analysis of the potential population health risks and maximum individual exposures from ten disposal methods under three different hydrogeological and climatic settings. This paper briefly describes the general input and analysis procedures used in the risk assessment for LLW disposal and presents their preliminary results. Some important lessons learned from simulating LLW disposal under a large variety of methods and conditions are identified

  8. Near-field performance assessment for a low-activity waste glass disposal system: laboratory testing to modeling results

    International Nuclear Information System (INIS)

    McGrail, B.P.; Bacon, D.H.; Icenhower, J.P.; Mann, F.M.; Puigh, R.J.; Schaef, H.T.; Mattigod, S.V.

    2001-01-01

    Reactive chemical transport simulations of glass corrosion and radionuclide release from a low-activity waste (LAW) disposal system were conducted out to times in excess of 20 000 yr with the subsurface transport over reactive multiphases (STORM) code. Time and spatial dependence of glass corrosion rate, secondary phase formation, pH, and radionuclide concentration were evaluated. The results show low release rates overall for the LAW glasses such that performance objectives for the site will be met by a factor of 20 or more. Parameterization of the computer model was accomplished by combining direct laboratory measurements, literature data (principally thermodynamic data), and parameter estimation methods

  9. Derivation of activity limits for the disposal of radioactive waste in near surface disposal facilities

    International Nuclear Information System (INIS)

    2003-12-01

    Radioactive waste must be managed safely, consistent with internationally agreed safety standards. The disposal method chosen for the waste should be commensurate with the hazard and longevity of the waste. Near surface disposal is an option used by many countries for the disposal of radioactive waste containing mainly short lived radionuclides and low concentrations of long lived radionuclides. The term 'near surface disposal' encompasses a wide range of design options, including disposal in engineered structures at or just below ground level, disposal in simple earthen trenches a few metres deep, disposal in engineered concrete vaults, and disposal in rock caverns several tens of metres below the surface. The use of a near surface disposal option requires design and operational measures to provide for the protection of human health and the environment, both during operation of the disposal facility and following its closure. To ensure the safety of both workers and the public (both in the short term and the long term), the operator is required to design a comprehensive waste management system for the safe operation and closure of a near surface disposal facility. Part of such a system is to establish criteria for accepting waste for disposal at the facility. The purpose of the criteria is to limit the consequences of events which could lead to radiation exposures and in addition, to prevent or limit hazards, which could arise from non-radiological causes. Waste acceptance criteria include limits on radionuclide content concentration in waste materials, and radionuclide amounts in packages and in the repository as a whole. They also include limits on quantity of free liquids, requirements for exclusion of chelating agents and pyrophoric materials, and specifications of the characteristics of the waste containers. Largely as a result of problems encountered at some disposal facilities operated in the past, in 1985 the IAEA published guidance on generic acceptance

  10. Headspace vapor characterization of Hanford waste Tank 241-C-201: Results from samples collected on 06/19/96

    International Nuclear Information System (INIS)

    Thomas, B.L.; Evans, J.C.; Pool, K.H.; Olsen, K.B.; Fruchter, J.S.; Silvers, K.L.

    1997-01-01

    This report describes the analytical results of vapor samples taken from the headspace of the waste storage tank 241-C-201 (Tank C-201) at the Hanford Site in Washington State. The results described in this report were obtained to characterize the vapors present in the tank headspace and to support safety evaluations and tank farm operations. The results include air concentrations of selected inorganic and organic analytes and grouped compounds from samples obtained by Westinghouse Hanford Company (WHC) and provided for analysis to Pacific Northwest National Laboratory (PNNL). Analyses were performed by the Vapor Analytical Laboratory (VAL) at PNNL. Analyte concentrations were based on analytical results and, where appropriate, on sample volumes provided by WHC. A summary, of the inorganic analytes, permanent gases, and total non-methane organic compounds is listed in a table. Detailed descriptions of the analytical results appear in the appendices

  11. The new revision of NPP Krsko decommissioning, radioactive waste and spent fuel management program: analyses and results

    International Nuclear Information System (INIS)

    Zeleznik, Nadja; Kralj, Metka; Lokner, Vladimir; Levanat, Ivica; Rapic, Andrea; Mele, Irena

    2010-01-01

    The preparation of the new revision of the Decommissioning and Spent Fuel (SF) and Low and Intermediate level Waste (LILW) Disposal Program for the NPP Krsko (Program) started in September 2008 after the acceptance of the Term of Reference for the work by Intergovernmental Committee responsible for implementation of the Agreement between the governments of Slovenia and Croatia on the status and other legal issues related to investment, exploitation, and decommissioning of the Nuclear power plant Krsko. The responsible organizations, APO and ARAO together with NEK prepared all new technical and financial data and relevant inputs for the new revision in which several scenarios based on the accepted boundary conditions were investigated. The strategy of immediate dismantling was analyzed for planned and extended NPP life time together with linked radioactive waste and spent fuel management to calculate yearly annuity to be paid by the owners into the decommissioning funds in Slovenia and Croatia. The new Program incorporated among others new data on the LILW repository including the costs for siting, construction and operation of silos at the location Vrbina in Krsko municipality, the site specific Preliminary Decommissioning Plan for NPP Krsko which included besides dismantling and decontamination approaches also site specific activated and contaminated radioactive waste, and results from the referenced scenario for spent fuel disposal but at very early stage. Important inputs for calculations presented also new amounts of compensations to the local communities for different nuclear facilities which were taken from the supplemented Slovenian regulation and updated fiscal parameters (inflation, interest, discount factors) used in the financial model based on the current development in economical environment. From the obtained data the nominal and discounted costs for the whole nuclear program related to NPP Krsko which is jointly owned by Slovenia and Croatia have

  12. Results of detailed ground geophysical surveys for locating and differentiating waste structures in waste management area 'A' at Chalk River Laboratories, Ontario

    International Nuclear Information System (INIS)

    Tomsons, D.K.; Street, P.J.; Lodha, G.S.

    1999-01-01

    Waste Management Area 'A' (WMA 'A'), located in the outer area of the Chalk River Laboratories (CRL) was in use as a waste burial site from 1946 to 1955. Waste management structures include debris-filled trenches, concrete bunkers and miscellaneous contaminated solid materials, and ditches and pits used for liquid dispersal. In order to update historical records, it was proposed to conduct detailed ground geophysical surveys to define the locations of waste management structures in WMA 'A', assist in planning of the drilling and sampling program to provide ground truth for the geophysics investigation and to predict the nature and locations of unknown/undefined shallow structures. A detailed ground geophysical survey grid was established with a total of 127 grid lines, oriented NNE and spaced one metre apart. The geophysical surveys were carried out during August and September, 1996. The combination of geophysical tools used included the Geonics EM61 metal detector, the GSM-19 magnetometer/gradiometer and a RAMAC high frequency ground penetrating radar system. The geophysical surveys were successful in identifying waste management structures and in characterizing to some extent, the composition of the waste. The geophysical surveys are able to determine the presence of most of the known waste management structures, especially in the western and central portions of the grid which contain the majority of the metallic waste. The eastern portion of the grid has a completely different geophysical character. While historical records show that trenches were dug, they are far less evident in the geophysical record. There is clear evidence for a trench running between lines 30E and 63E at 70 m. There are indications from the radar survey of other trench-like structures in the eastern portion. EM61 data clearly show that there is far less metallic debris in the eastern portion. The geophysical surveys were also successful in identifying previously unknown locations of waste

  13. Long-term management of the existing radioactive wastes and residues at the Niagara Falls Storage Site. Draft Environmental Impact Statement

    International Nuclear Information System (INIS)

    1984-08-01

    The statement assesses and compares several alternatives for long-term management of the existing radioactive wastes and residues at the Niagara Falls Storage Site (NFSS), Lewiston, New York. The alternatives include: (1) no action (continued interim storage at NFSS within a diked and capped containment area), (2) long-term management at NFSS (improved containment, with or without modified form of the residues), (3) long-term management at other DOE sites (Hanford, Washington, or Oak Ridge, Tennessee), and (4) offsite management of the residues at Hanford or Oak Ridge and either leaving the wastes at NFSS or removing them for disposal in the ocean. In addition to alternatives analyzed in depth, several options are also considered, including: other modifications of residue form, modification of the basic conceptual designs, other containment design options, transportation routes, and transportation modes. The radiological health effects (primarily increased risk of cancer) associated with long-term management of the wastes and residues are expected to be smaller than the nonradiological risks of occupational and transportation-related injuries and deaths. During the action period, the risk is highest for workers if all wastes and residues are moved to Hanford. The risk is highest for the general public if the residues are moved to Hanford and the wastes are moved to the ocean. Dispersal of the slightly contaminated wastes in the ocean is not expected to result in any significant impacts on the ocean environment or pose any significant radiological risk to humans. For all alternatives, if controls ceased, there would be eventual dispersion of the radioactive materials to the environment. If it is assumed that all controls cease, predicted time for loss of covers over the buried materials ranges from several hundred years to more than two million years, depending on the use of the land surface

  14. Headspace vapor characterization of Hanford waste tank 241-B-107: Results from samples collected on 7/23/96

    International Nuclear Information System (INIS)

    Evans, J.C.; Pool, K.H.; Thomas, B.L.; Olsen, K.B.; Fruchter, J.S.; Silvers, K.L.

    1997-01-01

    This report describes the analytical results of vapor samples taken from the headspace of the waste storage tank 241-B-107 (Tank B-107) at the Hanford Site in Washington State. The results described in this report were obtained to characterize the vapors present in the tank headspace and to support safety evaluations and tank farm operations. The results include air concentrations of selected inorganic and organic analytes and grouped compounds from samples obtained by Westinghouse Hanford Company (WHC) and provided for analysis to Pacific Northwestern National Laboratory (PNNL). A summary of the inorganic analytes, permanent gases, and total non-methane organic compounds is listed in a table. The three highest concentration analytes detected in SUMMA trademark canister and triple sorbent trap samples are also listed in the same table. Detailed descriptions of the analytical results appear in the appendices

  15. Headspace vapor characterization of Hanford waste tank 241-S-106: Results from samples collected on 06/13/96

    International Nuclear Information System (INIS)

    Evans, J.C.; Pool, K.H.; Thomas, B.L.; Olsen, K.B.; Fruchter, J.S.; Silvers, K.L.

    1997-01-01

    This report describes the analytical results of vapor samples taken from the headspace of the waste storage tank 241-S-106 (Tank S-106) at the Hanford Site in Washington State. The results described in this report were obtained to characterize the vapors present in the tank headspace and to support safety evaluations and tank farm operations. The results include air concentrations of selected inorganic and organic analytes and grouped compounds from samples obtained by Westinghouse Hanford Company (WHC) and provided for analysis to Pacific Northwest National Laboratory (PNNL). A summary of the inorganic analytes, permanent gases, and total non-methane organic compounds is listed in a table. The three highest concentration analytes detected in SUMMA trademark canister and triple sorbent trap samples are also listed in the same table. Detailed descriptions of the analytical results appear in the appendices

  16. Organic tank safety project: Preliminary results of energetics and thermal behavior studies of model organic nitrate and/or nitrite mixtures and a simulated organic waste

    International Nuclear Information System (INIS)

    Scheele, R.D.; Sell, R.L.; Sobolik, J.L.; Burger, L.L.

    1995-08-01

    As a result of years of production and recovery of nuclear defense materials and subsequent waste management at the Hanford Site, organic-bearing radioactive high-level wastes (HLW) are currently stored in large (up to 3. ML) single-shell storage tanks (SSTs). Because these wastes contain both fuels (organics) and the oxidants nitrate and nitrite, rapid energetic reactions at certain conditions could occur. In support of Westinghouse Hanford Company's (WHC) efforts to ensure continued safe storage of these organic- and oxidant-bearing wastes and to define the conditions necessary for reactions to occur, we measured the thermal sensitivities and thermochemical and thermokinetic properties of mixtures of selected organics and sodium nitrate and/or nitrite and a simulated Hanford organic-bearing waste using thermoanalytical technologies. These thermoanalytical technologies are used by chemical reactivity hazards evaluation organizations within the chemical industry to assess chemical reaction hazards

  17. Organic tank safety project: Preliminary results of energetics and thermal behavior studies of model organic nitrate and/or nitrite mixtures and a simulated organic waste

    Energy Technology Data Exchange (ETDEWEB)

    Scheele, R.D.; Sell, R.L.; Sobolik, J.L.; Burger, L.L.

    1995-08-01

    As a result of years of production and recovery of nuclear defense materials and subsequent waste management at the Hanford Site, organic-bearing radioactive high-level wastes (HLW) are currently stored in large (up to 3. ML) single-shell storage tanks (SSTs). Because these wastes contain both fuels (organics) and the oxidants nitrate and nitrite, rapid energetic reactions at certain conditions could occur. In support of Westinghouse Hanford Company`s (WHC) efforts to ensure continued safe storage of these organic- and oxidant-bearing wastes and to define the conditions necessary for reactions to occur, we measured the thermal sensitivities and thermochemical and thermokinetic properties of mixtures of selected organics and sodium nitrate and/or nitrite and a simulated Hanford organic-bearing waste using thermoanalytical technologies. These thermoanalytical technologies are used by chemical reactivity hazards evaluation organizations within the chemical industry to assess chemical reaction hazards.

  18. Product consistency test and toxicity characteristic leaching procedure results of the ceramic waste form from the electrometallurgical treatment process for spent fuel

    International Nuclear Information System (INIS)

    Johnson, S. G.; Adamic, M. L.: DiSanto, T.; Warren, A. R.; Cummings, D. G.; Foulkrod, L.; Goff, K. M.

    1999-01-01

    The ceramic waste form produced from the electrometallurgical treatment of sodium bonded spent fuel from the Experimental Breeder Reactor-II was tested using two immersion tests with separate and distinct purposes. The product consistency test is used to assess the consistency of the waste forms produced and thus is an indicator of a well-controlled process. The toxicity characteristic leaching procedure is used to determine whether a substance is to be considered hazardous by the Environmental Protection Agency. The proposed high level waste repository will not be licensed to receive hazardous waste, thus any waste forms destined to be placed there cannot be of a hazardous nature as defined by the Resource Conservation and Recovery Act. Results are presented from the first four fully radioactive ceramic waste forms produced and from seven ceramic waste forms produced from cold surrogate materials. The fully radioactive waste forms are approximately 2 kg in weight and were produced with salt used to treat 100 driver subassemblies of spent fuel

  19. Report of results and progress research (1982-1984) total research on long life radioactive waste management

    International Nuclear Information System (INIS)

    1985-03-01

    The specific research ''Synthetic research on long life radioactive waste management'' has been advanced in the Research Center for Nuclear Energy, University of Tokyo, for three years since 1982. This research was roughly divided into material science, biology and process engineering, and the research has been advanced according to 14 subthemes by the cooperation of the researchers in wide fields in the university. In this report, the report of the progress of research and the data on the results of researche from fiscal year 1982 to 1984 are summarized. The title of research, organization, the persons in charge, the period of research, the title of report, the objective, contents, state of progress, results obtained in 1984 and results obtained during three years of 5 material group papers, 7 process group papers and 4 biology group papers are given. (Kako, I.)

  20. Extraction, Scrub, and Strip Test Results for the Salt Waste Processing Facility Caustic Side Solvent Extraction Solvent Sample

    Energy Technology Data Exchange (ETDEWEB)

    Peters, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-10-06

    An Extraction, Scrub, and Strip (ESS) test was performed on a sample of Salt Waste Processing Facility (SWPF) Caustic-Side Solvent Extraction (CSSX) solvent and salt simulant to determine cesium distribution ratios (D(Cs)), and cesium concentration in the strip effluent (SE) and decontaminated salt solution (DSS) streams; this data will be used by Parsons to help determine if the solvent is qualified for use at the SWPF. The ESS test showed acceptable performance of the solvent for extraction, scrub, and strip operations. The extraction D(Cs) measured 12.5, exceeding the required value of 8. This value is consistent with results from previous ESS tests using similar solvent formulations. Similarly, scrub and strip cesium distribution ratios fell within acceptable ranges. This revision was created to correct an error. The previous revision used an incorrect set of temperature correction coefficients which resulted in slight deviations from the correct D(Cs) results.

  1. Gamma-ray spectrometry combined with acceptable knowledge (GSAK). A technique for characterization of certain remote-handled transuranic (RH-TRU) wastes. Part 2. Testing and results

    International Nuclear Information System (INIS)

    Hartwell, J.K.; McIlwain, M.E.

    2005-01-01

    Gamma-ray spectrometry combined with acceptable knowledge (GSAK) is a technique for the characterization of certain remote-handled transuranic (RH-TRU) wastes. GSAK uses gamma-ray spectrometry to quantify a portion of the fission product inventory of RH-TRU wastes. These fission product results are then coupled with calculated inventories derived from acceptable process knowledge to characterize the radionuclide content of the assayed wastes. GSAK has been evaluated and tested through several test exercises. These tests and their results are described; while the former paper in this issue presents the methodology, equipment and techniques. (author)

  2. Oxidation of hazardous waste in supercritical water: A comparison of modeling and experimental results for methanol destruction

    International Nuclear Information System (INIS)

    Butler, P.B.; Bergan, N.E.; Bramlette, T.T.; Pitz, W.J.; Westbrook, C.K.

    1991-01-01

    Recent experiments at Sandia National Laboratories conducted in conjunction with MODEC Corporation have demonstrated successful clean- up of contaminated water in a supercritical water reactor. These experiments targeted wastes of interest to Department of Energy production facilities. In this paper we present modeling and experimental results for a surrogate waste containing 98% water, 2% methanol, and parts per million of chlorinated hydrocarbons and laser dyes. Our initial modeling results consider only methanol and water. Experimental data are available for inlet and outlet conditions and axial temperature profiles along the outside reactor wall. The purpose of our model is to study the chemical and physical processes inside the reactor. We are particularly interested in the parameters that control the location of the reaction zone. The laboratory-scale reactor operates at 25 MPa., between 300 K and 900 K; it is modeled as a plug-flow reactor with a specified temperature profile. We use Chemkin Real-Gas to calculate mixture density, with the Peng-Robinson equation of state. The elementary reaction set for methanol oxidation and reactions of other C 1 and C 2 hydrocarbons is based on previous models for gas-phase kinetics. Results from our calculations show that the methanol is 99.9% destroyed at 1/3 the total reactor length. Although we were not able to measure composition of the fluid inside the experimental reactor, this prediction occurs near the location of the highest reactor temperature. This indicates that the chemical reaction is triggered by thermal effects, not kinetic rates. Results from ideal-gas calculations show nearly identical chemical profiles inside the reactor in dimensionless distance. However, reactor residence times are overpredicted by nearly 150% using an ideal-gas assumption. Our results indicate that this oxidation process can be successfully modeled using gas-phase chemical mechanisms. 23 refs., 8 figs

  3. Electrokinetic remediation of plutonium-contaminated nuclear site wastes: Results from a pilot-scale on-site trial

    International Nuclear Information System (INIS)

    Agnew, Kieran; Cundy, Andrew B.; Hopkinson, Laurence; Croudace, Ian W.; Warwick, Phillip E.; Purdie, Philip

    2011-01-01

    This paper examines the field-scale application of a novel low-energy electrokinetic technique for the remediation of plutonium-contaminated nuclear site soils, using soil wastes from the Atomic Weapons Establishment (AWE) Aldermaston site, Berkshire, UK as a test medium. Soils and sediments with varying composition, contaminated with Pu through historical site operations, were electrokinetically treated at laboratory-scale with and without various soil pre-conditioning agents. Results from these bench-scale trials were used to inform a larger on-site remediation trial, using an adapted containment pack with battery power supply. 2.4 m 3 (ca. 4 tonnes) of Pu-contaminated soil was treated for 60 days at a power consumption of 33 kW h/m 3 , and then destructively sampled. Radiochemical data indicate mobilisation of Pu in the treated soil, and migration (probably as a negatively charged Pu-citrate complex) towards the anodic compartment of the treatment cell. Soil in the cathodic zone of the treatment unit was remediated to a level below free-release disposal thresholds (1.7 Bq/g, or <0.4 Bq/g above background activities). The data show the potential of this method as a low-cost, on-site tool for remediation of radioactively contaminated soils and wastes which can be operated remotely on working sites, with minimal disruption to site infrastructure or operations.

  4. The Radioactive Waste Management Advisory Committee's advice to ministers on the establishment of scientific consensus on the interpretation and significance of the results of science programmes into radioactive waste disposal

    International Nuclear Information System (INIS)

    1999-04-01

    This document presents conclusions and recommendations on establishment of scientific consensus on the interpretation and significance of the results of science programmes into radioactive waste disposal. The topics discussed include: the nature of science and its limitations; societal views of science and the radioactive waste problem; issues upon which consensus will be needed; evidence of past attempts at greater involvement of the public; the linking of scientific and social consensus; communicating the nature of consensus to the public

  5. Waste treatment

    International Nuclear Information System (INIS)

    Hutson, G.V.

    1996-01-01

    Numerous types of waste are produced by the nuclear industry ranging from high-level radioactive and heat-generating, HLW, to very low-level, LLW and usually very bulky wastes. These may be in solid, liquid or gaseous phases and require different treatments. Waste management practices have evolved within commercial and environmental constraints resulting in considerable reduction in discharges. (UK)

  6. A recovery installation for sodium sulfates, thiosulfates and sulfides from waste water resulting from hydrogen sulfide fabrication

    International Nuclear Information System (INIS)

    Mazilu, Mihai; Costescu, Sanda

    2002-01-01

    An installation for recovery of sodium sulfate and sulfur suspensions from waste water was conceived. It consists from a preheater, vacuum evaporator and a refrigerating system with drum and scraper. This equipment concentration the solution by eliminating in the first stage the water in the vacuum evaporator. The water resulting at this stage is chemically pure and can be discharged in the sewage sludge system. The concentrated solution is then directed to the refrigerating system with drum and scrapper. Here the sodium sulfates, thiosulfates and sulfides get crystallized onto the drum surface. The resulting aqueous solution to be discharged in the sewage sludge system is previously analyzed as in case of the absent of the recovery installation, but the amount of pollutants will be much lower because sulfates, thiosulfates and sulfides were already recovered as scales from the drum. These solid scales can be used in detergent industry

  7. Comparison of elastic and inelastic analysis and test results for the defense high level waste shipping cask

    International Nuclear Information System (INIS)

    Zimmer, A.; Koploy, M.A.; Madsen, M.M.

    1991-01-01

    In the early 1980s, the US DOE/Defense Programs (DOE/DP) initiated a project to develop a safe and efficient transportation system for defense high level waste (DHLW). A long-standing objective of the DHLW transportation project is to develop a truck cask that represents the leading edge of cask technology as well as fully complies with all applicable DOE, Nuclear Regulatory Commission, and DOT regulations. General Atomics designed the DHLW Truck Shipping Cask using state-of-the-art analytical techniques verified by model testing performed by Sandia National Labs. (SNL). The analytical techniques include two approaches, inelastic analysis and elastic analysis. This paper will compare the results of the two analytical approaches and with model testing results. The purpose of this work is to provide data to support licensing of the DHLW cask and to support the acceptance by the NRC of inelastic analysis as a tool in packaging design and licensing

  8. The use of scientific and technical results from underground research laboratory investigations for the geological disposal of radioactive waste

    International Nuclear Information System (INIS)

    2001-09-01

    The objective of the report is to provide information on the use of results obtained from underground research laboratory investigations for the development of a deep geological repository system for long lived and/or high level radioactive waste including spent fuel. Specifically, it should provide Member States that intend to start development of a geological disposal system with an overview of existing facilities and of the sorts and quality of results that have already been acquired. The report is structured into six main themes: rock characterization methodologies and testing; assessment of the geological barrier; assessment of the engineered barrier system; respository construction techniques; demonstration of repository operations; confidence building and international co-operation

  9. Results of detailed ground geophysical surveys for locating and differentiating waste structures in waste management area 'A' at Chalk River Laboratories, Ontario

    Energy Technology Data Exchange (ETDEWEB)

    Tomsons, D.K.; Street, P.J.; Lodha, G.S

    1999-07-01

    Waste Management Area 'A' (WMA 'A'), located in the outer area of the Chalk River Laboratories (CRL) was in use as a waste burial site from 1946 to 1955. Waste management structures include debris-filled trenches, concrete bunkers and miscellaneous contaminated solid materials, and ditches and pits used for liquid dispersal. In order to update historical records, it was proposed to conduct detailed ground geophysical surveys to define the locations of waste management structures in WMA 'A', assist in planning of the drilling and sampling program to provide ground truth for the geophysics investigation and to predict the nature and locations of unknown/undefined shallow structures. A detailed ground geophysical survey grid was established with a total of 127 grid lines, oriented NNE and spaced one metre apart. The geophysical surveys were carried out during August and September, 1996. The combination of geophysical tools used included the Geonics EM61 metal detector, the GSM-19 magnetometer/gradiometer and a RAMAC high frequency ground penetrating radar system. The geophysical surveys were successful in identifying waste management structures and in characterizing to some extent, the composition of the waste. The geophysical surveys are able to determine the presence of most of the known waste management structures, especially in the western and central portions of the grid which contain the majority of the metallic waste. The eastern portion of the grid has a completely different geophysical character. While historical records show that trenches were dug, they are far less evident in the geophysical record. There is clear evidence for a trench running between lines 30E and 63E at 70 m. There are indications from the radar survey of other trench-like structures in the eastern portion. EM61 data clearly show that there is far less metallic debris in the eastern portion. The geophysical surveys were also successful in identifying

  10. Approach to defining de minimis, intermediate, and other classes of radioactive waste

    International Nuclear Information System (INIS)

    Cohen, J.J.; Smith, C.F.

    1986-01-01

    This study has developed a framework within which the complete spectrum of radioactive wastes can be defined. An approach has been developed that reflects both concerns in the framework of a radioactive waste classification system. In this approach, the class of any radioactive waste stream is dependent on its degree of radioactivity and its persistence. To be consistent with conventional systems, four waste classes are defined. In increasing order of concern due to radioactivity and/or duration, these are: 1. De Minimis Wastes: This waste has such a low content of radioactive material that it can be considered essentially nonradioactive and managed according to its nonradiological characteristics. 2. Low-Level Waste (LLW): Maximum concentrations for wastes considered to be in this class are prescribed in 10CFR61 as wastes that can be disposed of by shallow land burial methods. 3. Intermediate Level Waste (ILW): This category defines a class of waste whose content exceeds class C (10CFR61) levels, yet does not pose a sufficient hazard to justify management as a high-level waste (i.e., permanent isolation by deep geologic disposal). 4. High-Level Waste: HLW poses the most serious management problem and requires the most restrictive disposal methods. It is defined in NWPA as waste derived from the reprocessing of nuclear fuel and/or as highly radioactive wastes that require permanent isolation

  11. Assessment of management alternatives for LWR wastes. Volume 5. Assessment of the radiological impact to the public resulting from discharges of radioactive effluents

    International Nuclear Information System (INIS)

    Centner, B.

    1993-01-01

    This report deals with the assessment of the radiological impact to the public resulting from discharges of radioactive effluents (liquid and gaseous) in connection with the implementation of the Belgian scenario for the management of PWR waste. Both individual and collective doses have been estimated for a critical group of the population living around the nuclear power plants concerned. This study is part of an overall theoretical exercise aimed at evaluating a selection of management wastes for LWR waste based on economical and radiological criteria

  12. Finite-element model evaluation of barrier configurations to reduce infiltration into waste-disposal structures: preliminary results and design considerations

    International Nuclear Information System (INIS)

    Lu, A.H.; Phillips, S.J.; Adams, M.R.

    1982-09-01

    Barriers to reduce infiltration into waste burial disposal structures (trenches, pits, etc.) may be required to provide adequate waste confinement. The preliminary engineering design of these barriers should consider interrelated barrier performance factors. This paper summarizes preliminary computer simulation activities to further engineering barrier design efforts. Several barrier configurations were conceived and evaluated. Models were simulated for each barrier configuration using a finite element computer code. Results of this preliminary evaluation indicate that barrier configurations, depending on their morphology and materials, may significantly influence infiltration, flux, drainage, and storage of water through and within waste disposal structures. 9 figures

  13. H12: Examination of safety assessment aims, procedures and results from a wider perspective

    International Nuclear Information System (INIS)

    Neall, F.B; Smith, P.A.

    2004-04-01

    Safety assessment (SA) are a familiar tool for the evaluation of disposal concepts for radioactive waste. There is, however, often confusion in the wider community about the aims, methods and results used in SA. This report aims to present the H12 SA in a way that makes the assessment process clearer and the implications of the results more meaningful both to workers within the SA field and to a wider technical audience. The reasonableness of the assessment results, the quality of the models and databases and redundancy within the natural and engineered barrier system have been considered. A number of recent and somewhat older SAs that address a range of different waste types, host rocks and disposal concepts have been considered, and comparisons made to H12. A further aim is to put both doses and timescales in a more meaningful context. It has been necessary to: consider ways of demonstrating the meaningfulness of calculations that give results for many thousands of years in the future; provide a framework timescale as a context for SA results over long times; demonstrate the smallness of the risk associated with the doses by comparison with other radiological and non-radiological risks. The perception of risk, which is a critical issue for public acceptance of radioactive waste disposal and must be considered when seeking to present safety assessment results 'in perspective' to a wider audience, is also discussed. It is concluded that H12 is comparable in many ways to assessments carried out internationally. Some assumptions are somewhat arbitrary reflecting the generic stage of the Japanese programme, and are likely to become better founded in future exercises. Nevertheless, H12 provides a clear and well-founded message that it is feasible to site and construct a safe repository from HLW in Japan. (author)

  14. Tank 48H Waste Composition and Results of Investigation of Analytical Methods

    Energy Technology Data Exchange (ETDEWEB)

    Walker , D.D. [Westinghouse Savannah River Company, AIKEN, SC (United States)

    1997-04-02

    This report serves two purposes. First, it documents the analytical results of Tank 48H samples taken between April and August 1996. Second, it describes investigations of the precision of the sampling and analytical methods used on the Tank 48H samples.

  15. Comparison of the results of several heat transfer computer codes when applied to a hypothetical nuclear waste repository

    International Nuclear Information System (INIS)

    Claiborne, H.C.; Wagner, R.S.; Just, R.A.

    1979-12-01

    A direct comparison of transient thermal calculations was made with the heat transfer codes HEATING5, THAC-SIP-3D, ADINAT, SINDA, TRUMP, and TRANCO for a hypothetical nuclear waste repository. With the exception of TRUMP and SINDA (actually closer to the earlier CINDA3G version), the other codes agreed to within +-5% for the temperature rises as a function of time. The TRUMP results agreed within +-5% up to about 50 years, where the maximum temperature occurs, and then began an oscillary behavior with up to 25% deviations at longer times. This could have resulted from time steps that were too large or from some unknown system problems. The available version of the SINDA code was not compatible with the IBM compiler without using an alternative method for handling a variable thermal conductivity. The results were about 40% low, but a reasonable agreement was obtained by assuming a uniform thermal conductivity; however, a programming error was later discovered in the alternative method. Some work is required on the IBM version to make it compatible with the system and still use the recommended method of handling variable thermal conductivity. TRANCO can only be run as a 2-D model, and TRUMP and CINDA apparently required longer running times and did not agree in the 2-D case; therefore, only HEATING5, THAC-SIP-3D, and ADINAT were used for the 3-D model calculations. The codes agreed within +-5%; at distances of about 1 ft from the waste canister edge, temperature rises were also close to that predicted by the 3-D model

  16. Results of complex studies in radiation state of temporary areas for radioactive waste localization in the Chernobyl estrangement zone.; Rezul`taty kompleksnykh issledovanij radiatsionnogo sostoyaniya punktov vremennoj lokalizatsii radioaktivnykh otkhodov v Zone otchuzhdeniya ChAEhS.

    Energy Technology Data Exchange (ETDEWEB)

    Ledenev, A I; Ovcharov, P A; Mishunina, I B; Antropov, V M [Naukovo-Tekhnyichnij Tsentr z dezaktivatsyiyi ta kompleksnogo povodzhennya z radyioaktivnimi vyidkhodami, Zhovtyi Vodi (Ukraine)

    1994-12-31

    Describing complex studies in radiation state of temporary areas for radioactive waste localization in the nearest Chernobyl NPP zone, the paper provides results of these studies as well as results of inspection of radioactive waste hidden in 1990 - 1994.

  17. Potential for criticality in Hanford tanks resulting from retrieval of tank waste

    International Nuclear Information System (INIS)

    Whyatt, G.A.; Sterne, R.J.; Mattigod, S.V.

    1996-09-01

    This report assesses the potential during retrieval operations for segregation and concentration of fissile material to result in a criticality. The sluicing retrieval of C-106 sludge to AY-102 and the operation of mixer pumps in SY-102 are examined in some detail. These two tanks (C-106, SY-102) were selected because of the near term plans for retrieval of these tanks and their high plutonium inventories relative to other tanks. Although all underground storage tanks are subcritical by a wide margin if assumed to be uniform in composition, the possibility retrieval operations could preferentially segregate the plutonium and locally concentrate it sufficiently to result in criticality was a concern. This report examines the potential for this segregation to occur

  18. First results and future trends for the transmutation of long-lived radioactive wastes

    International Nuclear Information System (INIS)

    Prunier, C.; Salvatores, M.; Guerin, Y.; Zaetta, A.

    1993-01-01

    In the frame of the CEA SPIN program, a project has been set-up at the Direction of Nuclear Reactors of CEA, to study the transmutation of long-lived radioactive products (both minor actinides and fission products) resulting from the operation of current nuclear power plants. The program is focused on: transmutation in minor actinides (Np, Am) in fission reactors of known technology (both of the PWR or the fast reactor type), using the so-called ''homogeneous'' (mixed with Uranium or Uranium-Plutonium), and ''heterogeneous'' (mixed with inert matrices) recycling modes for both type of reactors. Transmutation studies in dedicated devices (both fission reactors with actinide/plutonium fuel or with high thermal flux, and particle accelerator-based systems). Fuel studies related to both homogeneous and heterogeneous recycling modes in fission reactors. For the homogeneous recycling mode, some experimental irradiations results are available from past PHENIX programs. For the heterogeneous mode, very limited experimental results are available, and new theoretical and experimental work is underway on the use of appropriate inert matrices. Basic data studies to assess the quality of existing nuclear data for fission reactor transmutation studies, future data needs of relevance, and model/data developments needed for accelerator-based systems. Strategy studies, to evaluate the consequences of the different transmutation options on the fuel cycle, according to different scenarios of nuclear power development. 7 refs., 3 figs., 5 tabs

  19. Quaternary-geological results and problems of the Gorleben project for final storage of radioactive waste

    International Nuclear Information System (INIS)

    Duphorn, K.

    1984-01-01

    The measured results and the ground-water flow models elaborated by the Bundesanstalt fuer Geowissenschaften und Rohstoffe, Hannover, show that the ground-water flows relatively fast in the high-permeability quaternary sands and gravels of the Gorleben channel down to the caprock. This accounts for the current subrosion rate which has been determined to be up to 1 mm per annum, so that a subrosion volume of up to 10.000 m 3 a year is to be expected, which means that ground-water flow from the channel bottom to the soil surface is expected to take a period of only 600 up to 3700 years. These quaternary-hydrogeological results give reason to doubt whether the model of the geologic multi-barriers, according to which a protective function is attributed to the ''caprock barrier'', can really be applied. The results show that the ''salt-bed barrier'' at the Gorleben site is geologically unstable and endangered by subrosion, which is reason enough to likewise question the protective effect of this salt formation in the long run. (orig./HP) [de

  20. Acceptable knowledge summary report for combustible/noncombustible, metallic, and HEPA filter waste resulting from 238Pu fabrication activities

    International Nuclear Information System (INIS)

    Rogers, P.S.Z.; Foxx, C.L.

    1998-01-01

    All transuranic (TRU) waste must be sufficiently characterized and certified before it is shipped to the Waste Isolation Pilot Plant (WIPP). The US Environmental Protection Agency (EPA) allows use of acceptable knowledge (AK) for waste characterization. EPA uses the term AK in its guidance document and defines AK and provides guidelines on how acceptable knowledge should be obtained and documented. This AK package has been prepared in accordance with Acceptable Knowledge Documentation (TWCP-QP-1.1-021,R.2). This report covers acceptable knowledge information for five waste streams generated at TA-55 during operations to fabricate various heat sources using feedstock 238 Pu supplied by the Savannah River Site (SRS). The 238 Pu feedstock itself does not contain quantities of RCRA-regulated constituents above regulatory threshold limits, as known from process knowledge at SRS and as confirmed by chemical analysis. No RCRA-regulated chemicals were used during 238 Pu fabrication activities at TA-55, and all 238 Pu activities were physically separated from other plutonium processing activities. Most of the waste generated from the 238 Pu fabrication activities is thus nonmixed waste, including waste streams TA-55-43, 45, and 47. The exceptions are waste streams TA-55-44, which contains discarded lead-lined rubber gloves used in the gloveboxes that contained the 238 Pu material, and TA-55-46, which may contain pieces of discarded lead. These waste streams have been denoted as mixed because of the presence of the lead-containing material

  1. Supplemental results of the human health risk analysis for the U.S. Department of Energy draft waste management programmatic environmental impact statement

    International Nuclear Information System (INIS)

    1995-08-01

    This report is intended as an information supplement to the human health risk analysis performed for the US Department of Energy's Draft Waste Management Programmatic Environmental Impact Statement for Managing Treatment, Storage, and Disposal of Radioactive and Hazardous Waste, hereinafter called the PEIS. This report provides the installation-by-installation human health risk analysis results from which the risk estimate summaries for the PEIS were drawn. Readers should bear in mind that the risk estimates presented here are the result of a program-wide (as opposed to site-specific) study. They are based on best available data; systematically applied assumptions; and professional judgment about DOE waste inventories, waste volumes generated annually, currently available treatment and disposal technologies, technical limitations of treatment, and facility capacities across the numerous installations in the DOE complex

  2. Extraction, scrub, and strip test results for the salt waste processing facility caustic side solvent extraction solvent example

    Energy Technology Data Exchange (ETDEWEB)

    Peters, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-08-01

    An Extraction, Scrub, and Strip (ESS) test was performed on a sample of Salt Waste Processing Facility (SWPF) Caustic-Side Solvent Extraction (CSSX) solvent and salt simulant to determine cesium distribution ratios (D(Cs)), and cesium concentration in the strip effluent (SE) and decontaminated salt solution (DSS) streams; this data will be used by Parsons to help determine if the solvent is qualified for use at the SWPF. The ESS test showed acceptable performance of the solvent for extraction, scrub, and strip operations. The extraction D(Cs) measured 12.9, exceeding the required value of 8. This value is consistent with results from previous ESS tests using similar solvent formulations. Similarly, scrub and strip cesium distribution ratios fell within acceptable ranges.

  3. Incorporation of high-level wastes in SYNROC: results from recent process-engineering studies at Lawrence Livermore National Laboratory

    International Nuclear Information System (INIS)

    Campbell, J.H.; Hoenig, C.L.; Ackerman, F.J.; Peters, P.E.; Grens, J.Z.

    1982-01-01

    In this paper, highlights from recent engineering research and development, in particular, results from fluidized bed calcination studies of SYNROC slurry are summarized. A schematic diagram of the envisioned SYNROC process (at this stage of development) is also presented. It shows the use of a fluidized bed calciner to prepare SYNROC powder that is then fed to a storage hopper. Bellows-type canisters are filled, evacuated, sealed and preheated. The preheated canisters are loaded into a hot isotactic pressing unit where they are densified, then removed and cooled and finally loaded into a waste storage container. After sealing, this container is decontaminated and transferred to the interim storage facility and then, ultimately, to an underground repository

  4. The International Conference on Radioactive Waste Management

    International Nuclear Information System (INIS)

    1983-01-01

    The IAEA has been concerned with radioactive waste management since its inception. Its programme in this area was expanded in the mid 1970s as questions related to the management and disposal of radioactive wastes came into focus in conjunction with the further industrial development of nuclear power. The objectives of the Agency's wastes management programme are to assist its Member States in the safe and effective management of wastes by organizing the exchange and dissemination of information, providing guidance and technical assistance and supporting research. The current programme addresses all aspects of the industrial use of nuclear power under the aspects (a) technology of handling and treatment of wastes, (b) underground disposal of wastes, (c) environmental aspects of nuclear energy, including sea disposal of radioactive wastes. Systematic reviews have been made and publications issued concerning the technology of handling, treating, conditioning, and storing various categories of wastes, including liquid and gaseous wastes, wastes from nuclear power plants, spent fuel reprocessing and mining and milling of uranium ores, as well as wastes from decommissioning of nuclear facilities. As waste disposal is the current issue of highest interest, an Agency programme was set up in 1977 to develop a set of guidelines on the safe underground disposal of low-, intermediate- and high-level wastes in shallow ground, rock cavities or deep geological repositories. This programme will continue until 1990. Eleven Safety Series and Technical documents and reports have been published under this programme so far, which also addresses safety and other criteria for waste disposal. The environmental part of the waste management programme is concerned with the assessment of radiological and non-radiological consequences of discharges from nuclear facilities, including de minimis concepts in waste disposal and environmental models and data for radionuclide releases. The Agency

  5. Distribution of Rare Earth Metals in Technogenic Wastes of Energy Enterprises (Results of the Laboratory Studies)

    OpenAIRE

    Alexandr Ivanovich Khanchuk; Aleksandr Alekseevich Yudakov; Mikhail Azaryevich Medkov; Leonid Nikolayevich Alekseyko; Andrey Vasilyevich Taskin; Sergey Igorevich Ivannikov

    2016-01-01

    The results of the research interaction between ash and slag samples from Vladivostok TPP’s landfills saturated with underburning and ammonium hydrodifluoride were given. It was found out that the reactions of the main components of a concentrate with NH4HF2 are flowing with creation of complex ammonium fluoro-metalate. It is shown that the distribution of REM (rare earth metals) between foam and heavier products is going during the flotation process of carbon-containing ash and slag samples ...

  6. Development of grout formulations for 106-AN waste: Mixture-experiment results and analysis

    International Nuclear Information System (INIS)

    Spence, R.D.; McDaniel, E.W.; Anderson, C.M.; Lokken, R.O.; Piepel, G.F.

    1993-09-01

    Twenty potential ingredients were identified for use in developing a 106-AN grout formulation, and 18 were subsequently obtained and tested. Four ingredients: Type II-LA (moderate heat of hydration) Portland cement, Class F fly ash, attapulgite 150 drilling clay, and ground air-cooled blast-furnace slag (GABFS) -- were selected for developing the 106-AN grout formulations. A mixture experiment was designed and conducted around the following formulation: 2.5 lb of cement per gallon, 1.2 lb of fly ash per gallon, 0.8 lb of attapulgite per gallon, and 3.5 lb of GABFS per gallon. Reduced empirical models were generated from the results of the mixture experiment. These models were used to recommend several grout formulations for 106-AN. Westinghouse Hanford Company selected one of these formulations to be verified for use with 106-AN and a backup formulation in case problems arise with the first choice. This report presents the mixture-experimental results and leach data

  7. Vapor space characterization of waste Tank 241-TX-118 (in situ): Results from samples collected on 9/7/94

    International Nuclear Information System (INIS)

    Thomas, B.L.; Clauss, T.W.; Ligotke, M.W.; Pool, K.H.; McVeety, B.D.; Olsen, K.B.; Fruchter, J.S.; Goheen, S.C.

    1995-10-01

    This report describes inorganic and organic analyses results from in situ samples obtained from the headspace of the Hanford waste storage Tank 241-TX-118 (referred to as Tank TX-118). The results described here were obtained to support safety and toxicological evaluations. A summary of the results for inorganic and organic analytes is listed in Table 1. Detailed descriptions of the results appear in the text. Quantitative results were obtained for the inorganic compounds ammonia (NH 3 ), nitrogen dioxide (NO 2 ), nitric oxide (NO), hydrogen cyanide (CHN), and water (H 2 O). Sampling for sulfur oxides (SO x ) was not requested. In addition, quantitative results were obtained for the 39 TO-14 compounds plus an additional 13 analytes. Hexane, normally included in the additional analytes, was removed because a calibration standard was not available during analysis of Tank TX-118 SUMMA trademark canisters. Of these, 12 were observed above the 5-ppbv reporting cutoff. Fourteen tentatively identified compounds (TICs) were observed above the reporting cutoff of (ca.) 10 ppbv and are reported with concentrations that are semiquantitative estimates based on internal-standard response factors. The 10 organic analytes with the highest estimated concentrations are listed in Table 1 and account for approximately 86% of the total organic components in Tank TX-118. Permanent gas analysis was not conducted on the tank-headspace samples. Tank TX-118 is on both the Ferrocyanide and Organic Watch List

  8. Vapor space characterization of Waste Tank 241-TY-104: Results from samples collected on 4/27/95

    International Nuclear Information System (INIS)

    Klinger, G.S.; Olsen, K.B.; Clauss, T.W.

    1995-10-01

    This report describes inorganic and organic analyses results from samples obtained from the headspace of the Hanford waste storage Tank 241-TY-104 (referred to as Tank TY-104). The results described here were obtained to support safety and toxicological evaluations. A summary of the results for inorganic and organic analytes is listed in Table 1. Detailed descriptions of the results appear in the text. Quantitative results were obtained for the inorganic compounds ammonia (NH 3 ), nitrogen dioxide (NO 2 ), nitric oxide (NO), and water (H 2 O). Sampling for hydrogen cyanide (HCN) and sulfur oxides (SO x ) was not requested. In addition, quantitative results were obtained for the 39 TO-14 compounds plus an additional 14 analytes. Of these, 8 were observed above the 5-ppbv reporting cutoff. Five tentatively identified compounds (TICs) were observed above the reporting cutoff of (ca.) 10 ppbv and are reported with concentrations that are semiquantitative estimates based on internal-standard response factors. The 10 organic analytes with the highest estimated concentrations are listed in Table 1 and account for approximately 94% of the total organic components in Tank TY-104. Nitrous oxide (N 2 O) was the only permanent gas detected in the tank-headspace samples. Tank TY-104 is on the Ferrocyanide Watch List

  9. Vapor space characterization of waste tank 241-TY-103: Results from samples collected on 4/11/95

    International Nuclear Information System (INIS)

    Ligotke, M.W.; Clauss, T.W.; Pool, K.H.

    1995-10-01

    This report describes inorganic and organic analyses results from samples obtained from the headspace of the Hanford waste storage Tank 241-TY-103 (referred to as Tank TY-103). The results described here were obtained to support safety and toxicological evaluations. A summary of the results for inorganic and organic analytes is listed in Table 1. Detailed descriptions of the results appear in the text. Quantitative results were obtained for the inorganic compounds ammonia (NH 3 ), nitrogen dioxide (NO 2 ), nitric oxide (NO), and water (H 2 O). Sampling for hydrogen cyanide (HCN) and sulfur oxides (SO x ) was not requested. In addition, quantitative results were obtained for the 39 TO-14 compounds plus an additional 14 analytes. Of these, 16 were observed above the 5-ppbv reporting cutoff. Sixteen tentatively identified compounds (TICs) were observed above the reporting cutoff of (ca.) 10 ppbv and are reported with concentrations that are semiquantitative estimates based on internal-standard response factors. The 10 organic analytes with the highest estimated concentrations are listed in Table 1 and account for approximately 95% of the total organic components in Tank TY-103. Two permanent gases, carbon dioxide (CO 2 ) and nitrous oxide (N 2 O), were also detected

  10. Vapor space characterization of Waste Tank 241-S-111: Results from samples collected on 3/21/95

    International Nuclear Information System (INIS)

    Klinger, G.S.; Clauss, T.W.; Ligotke, M.W.

    1995-10-01

    This report describes inorganic and organic analyses results from samples obtained from the headspace of the Hanford waste storage Tank 241-S-111 (referred to as Tank S-111). The results described here were obtained to support safety and toxicological evaluations. A summary of the results for inorganic and organic analytes is listed in Table 1. Detailed descriptions of the results appear in the text. Quantitative results were obtained for the inorganic compounds ammonia (NH 3 ), nitrogen dioxide (NO 2 ), nitric oxide (NO), and water (H 2 O). Sampling for hydrogen cyanide (HCN) and sulfur oxides (SO x ) was not requested. In addition, quantitative results were obtained for the 39 TO-14 compounds plus an additional 14 analytes. Of these, seven were observed above the 5-ppbv reporting cutoff. Five tentatively identified compounds (TICs) were observed above the reporting cutoff of (ca.) 10 ppbv and are reported with concentrations that are semiquantitative estimates based on internal-standard response factors. The 10 organic analytes with the highest estimated concentrations are listed in Table 1 and account for approximately 98% of the total organic components in Tank S-111. Two permanent gases, hydrogen (H 2 ) and nitrous oxide (N 2 O), were also detected. Tank S-111 is on the Hydrogen Watch List

  11. Vapor space characterization of Waste Tank 241-U-105: Results from samples collected on 2/24/95

    International Nuclear Information System (INIS)

    Pool, K.H.; Clauss, T.W.; Ligotke, M.W.

    1995-10-01

    This report describes inorganic and organic analyses results from samples obtained from the headspace of the Hanford waste storage Tank 241-U-105 (referred to as Tank U-105). The results described here were obtained to support safety and toxicological evaluations. A summary of the results for inorganic and organic analytes is listed in Table 1. Detailed descriptions of the results appear in the text. Quantitative results were obtained for the inorganic compounds ammonia (NH 3 ), nitrogen dioxide (NO 2 ), nitric oxide (NO), and water (H 2 O). Sampling for hydrogen cyanide (HCN) and sulfur oxides (SO x ) was not requested. In addition, quantitative results were obtained for the 39 TO-14 compounds plus an additional 14 analytes. Of these, six were observed above the 5-ppbv reporting cutoff. Three tentatively identified compounds (TICs) were observed above the reporting cutoff of (ca.) 10 ppbv and are reported with concentrations that are semiquantitative estimates based on internal-standard response factors. All nine of the organic analytes identified are listed in Table 1 and account for 100% of the total organic components in Tank U-105. Nitrous oxide (N 2 O) was the only permanent gas detected in the tank-headspace sample. Tank U-105 is on the Hydrogen Watch List

  12. Vapor space characterization of waste Tank 241-U-103: Results from samples collected on 2/15/95

    International Nuclear Information System (INIS)

    Ligotke, M.W.; Pool, K.H.; Clauss, T.W.; McVeety, B.D.; Klinger, G.S.; Olsen, K.B.; Bredt, O.P.; Fruchter, J.S.; Goheen, S.C.

    1995-11-01

    This report describes inorganic and organic analyses results from samples obtained from the headspace of the Hanford waste storage Tank 241-U-103 (referred to as Tank U-103). The results described her were obtained to support safety and toxicological evaluations. A summary of the results for inorganic and organic analytes is listed in Table 1. Detailed descriptions of the results appear in the text. Quantitative results were obtained for the inorganic compounds ammonia (NH 3 ), nitrogen dioxide (NO 2 ), nitric oxide (NO), and water vapor (H 2 O). Sampling for hydrogen cyanide (HCN) and sulfur oxides (SO x ) was not requested. In addition, quantitative results were obtained for the 39 TO-14 compounds plus an additional 14 analytes. Of these, 11 were observed above the 5-ppbv reporting cutoff. Eleven tentatively identified compounds (TICs) were observed above the reporting cutoff of (ca.) 10 ppbv and are reported with concentrations that are semiquantitative estimates based on internal-standard response factors. The 10 organic analytes with the highest estimated concentrations are listed in Table 1 and account for approximately 90% of the total organic components in Tank U-103. Two permanent gases, hydrogen (H 2 ) and nitrous oxide (N 2 O), were also detected. Tank U-103 is on the Hydrogen Watch List

  13. Vapor space characterization of waste Tank 241-SX-106: Results from samples collected on 3/24/95

    International Nuclear Information System (INIS)

    Klinger, G.S.; Clauss, T.W.; Litgotke, M.W.

    1995-11-01

    This report describes inorganic and organic analyses results from samples obtained from the headspace of the Hanford waste storage Tank 241-SX-106 (referred to as Tank SX-106). The results described here were obtained to support safety and toxicological evaluations. A summary of the results for inorganic and organic analytes is listed in Table 1. Detailed descriptions of the results appear in the text. Quantitative results were obtained for the inorganic compounds ammonia (NH 3 ), nitrogen dioxide (NO 2 ), nitric oxide (NO), and water (H 2 O). Sampling for hydrogen cyanide (HCN) and sulfur oxides (SO x ) was not requested. In addition, quantitative results were obtained for the 39 TO-14 compounds plus an additional 14 analytes. Of these, 4 were observed above the 5-ppbv reporting cutoff. Three tentatively identified compounds (TICs) were observed above the reporting cutoff of (ca.) 10 ppbv and are reported with concentrations that are semiquantitative estimates based on internal-standard response factors. The 7 organic analytes identified are listed in Table 1 and account for approximately 100% of the total organic components in Tank SX-106. Carbon dioxide (CO 2 ) was the only permanent gas detected. Tank SX-106 is on the Ferrocyanide Watch List

  14. Vapor space characterization of waste tank 241-TX-118: Results from samples collected on 12/16/94

    International Nuclear Information System (INIS)

    Lucke, R.B.; Ligotke, M.W.; McVeety, B.D.

    1995-10-01

    This report describes inorganic and organic analyses results from samples obtained from the headspace of the Hanford waste storage Tank 241-TX-118 (referred to as Tank TX-118). The results described here were obtained to support safety and toxicological evaluations. A summary of the results for inorganic and organic analytes is listed in Table 1. Detailed descriptions of the results appear in the text. Quantitative results were obtained for the inorganic compounds ammonia (NH 3 ), nitrogen dioxide (NO 2 ), nitric oxide (NO), and water (H 2 O). Sampling for hydrogen cyanide (HCN) and sulfur oxides (SO x ) was not requested. In addition, quantitative results were obtained for the 39 TO-14 compounds plus an additional 14 analytes. Of these, 3 were observed above the 5-ppbv reporting cutoff. Twenty three organic tentatively identified compounds (TICs) were observed above the reporting cutoff of (ca.) 10 ppbv, and are reported with concentrations that are semiquantitative estimates based on internal-standard response factors. The 10 organic analytes with the highest estimated concentrations are listed in Table 1 and account for approximately 84% of the total organic components in Tank TX-118. Two permanent gases, carbon dioxide (CO 2 ) and nitrous oxide (N 2 O), were also detected

  15. Vapor space characterization of waste tank 241-S-102: Results from samples collected on 3/14/95

    International Nuclear Information System (INIS)

    Pool, K.H.; McVeety, B.D.; Clauss, T.W.

    1995-10-01

    This report describes inorganic and organic analyses results from samples obtained from the headspace of the Hanford waste storage Tank 241-S-102 (referred to as Tank S-102). The results described here were obtained to support safety and toxicological evaluations. A summary of the results for inorganic and organic analytes is listed in Table 1. Detailed descriptions of the results appear in the text. Quantitative results were obtained for the inorganic compounds ammonia (NH 3 ), nitrogen dioxide (NO 2 ), nitric oxide (NO), and water (H 2 O). Sampling for hydrogen cyanide (HCN) and sulfur oxides (SO x ) was not requested. In addition, quantitative results were obtained for the 39 TO-14 compounds plus an additional 14 analytes. Of these, 11 were observed above the 5-ppbv reporting cutoff. Eleven tentatively identified compounds (TICs) were observed above the reporting cutoff of (ca.) 10 ppbv and are reported with concentrations that are semiquantitative estimates based on internal-standard response factors. The 10 organic analytes with the highest estimated concentrations are listed in Table 1 and account for approximately 95% of the total organic components in Tank S-102. Two permanent gases, hydrogen (H 2 ) and nitrous oxide (N 2 O), were also detected

  16. Vapor space characterization of Waste Tank 241-U-107: Results from samples collected on 2/17/95

    International Nuclear Information System (INIS)

    McVeety, B.D.; Clauss, T.W.; Ligotke, M.W.

    1995-10-01

    This report describes inorganic and organic analyses results from samples obtained from the headspace of the Hanford waste storage Tank 241-U-107 (referred to as Tank U-107). The results described here were obtained to support safety and toxicological evaluations. A summary of the results for inorganic and organic analytes is listed in Table 1. Detailed descriptions of the results appear in the text. Quantitative results were obtained for the inorganic compounds ammonia (NH 3 ), nitrogen dioxide (NO 2 ), nitric oxide (NO), and water (H 2 O). Sampling for hydrogen cyanide (HCN) and sulfur oxides (SO x ) was not requested. In addition, quantitative results were obtained for the 39 TO-14 compounds plus an additional 14 analytes. Of these, 10 were observed above the 5-ppbv reporting cutoff. Sixteen organic tentatively identified compounds (TICs) were observed above the reporting cutoff of (ca.) 10 ppbv, and are reported with concentrations that are semiquantitative estimates based on internal-standard response factors. The 10 organic analytes with the highest estimated concentrations are listed in Table 1 and account for approximately 88% of the total organic components in Tank U-107. Nitrous oxide (N 2 O) was the only permanent gas detected in the tank-headspace samples. Tank U-107 is on the Organic and the Hydrogen Watch Lists

  17. Vapor space characterization of waste Tank 241-SX-103: Results from samples collected on 3/23/95

    International Nuclear Information System (INIS)

    Ligotke, M.W.; Clauss, T.W.; Pool, K.H.; McVeety, B.D.; Klinger, G.S.; Olsen, K.B.; Bredt, O.P.; Fruchter, J.S.; Goheen, S.C.

    1995-11-01

    This report describes inorganic and organic analyses results from samples obtained from the headspace of the Hanford waste storage tank 241-SX-103 (referred to as Tank SX-103). The results described here were obtained to support safety and toxicological evaluations. A summary of the results for inorganic and organic analytes is listed in Table 1. Detailed descriptions of the results appear in the text. Quantitative results were obtained for the inorganic compounds ammonia (NH 3 ), nitrogen dioxide (NO 2 ), nitric oxide (NO), and water vapor (H 2 O). Sampling for hydrogen cyanide (HCN) and sulfur oxides (SO x ) was not requested. In addition, quantitative results were obtained for the 39 TO-14 compounds plus an additional 14 analytes. Of these, two were observed above the 5-ppbv reporting cutoff. Two tentatively identified compounds (TICs) were observed above the reporting cutoff of (ca.) 10 ppbv and are reported with concentrations that are semiquantitative estimates based on internal-standard response factors. The four organic analytes identified are listed in Table 1 and account for approximately 100% of the total organic components in Tank SX-103. Carbon dioxide (CO 2 ) was the only permanent gas detected in the tank-headspace samples. Tank SX-103 is on the Hydrogen Watch List

  18. Vapor space characterization of waste Tank 241-TY-101: Results from samples collected on 4/6/95

    International Nuclear Information System (INIS)

    Klinger, G.S.; Clauss, T.W.; Ligotke, M.W.; Pool, K.H.; McVeety, B.D.; Olsen, K.B.; Bredt, O.P.; Fruchter, J.S.; Goheen, S.C.

    1995-11-01

    This report describes inorganic and organic analyses results from samples obtained from the headspace of the Hanford waste storage Tank 241-TY-101 (referred to as Tank TY-101). The results described here were obtained to support safety and toxicological evaluations. A summary of the results for inorganic and organic analytes is listed in Table 1. Detailed descriptions of the results appear in the text. Quantitative results were obtained for the inorganic compounds ammonia (NH 3 ), nitrogen dioxide (NO 2 ), nitric oxide (NO), and water vapor (H 2 O). Sampling for hydrogen cyanide (HCN) and sulfur oxides (SO x ) was not requested. In addition, quantitative results were obtained for the 39 TO-14 compounds plus an additional 14 analytes. Off these, 5 were observed above the 5-ppbv reporting cutoff. One tentatively identified compound (TIC) was observed above the reporting cutoff of (ca.) 10 ppbv and are reported with concentrations that are semiquantitative estimates based on internal-standard response factors. The six organic analyses identified are listed in Table 1 and account for approximately 100% of the total organic components in Tank TY-101. Two permanent gases, carbon dioxide (CO 2 ) and nitrous oxide (N 2 O), were also detected. Tank TY-101 is on the Ferrocyanide Watch List

  19. Vapor space characterization of waste tank 241-BY-105 (in situ): Results from samples collected on May 9, 1994

    International Nuclear Information System (INIS)

    McVeety, B.D.; Pool, K.H.; Ligotke, M.W.; Clauss, T.W.; Lucke, R.B.; Sharma, A.K.; McCulloch, M.; Fruchter, J.S.; Goheen, S.C.

    1995-05-01

    This report describes inorganic and organic analyses results from in situ samples obtained from the tank headspace of the Hanford waste storage Tank 241-BY-105 (referred to as Tank BY-105). The results described here were obtained to support safety and toxicological evaluations. A summary of the results for inorganic and organic analytes is listed. Detailed descriptions of the results appear in the text. Quantitative results were obtained for the inorganic compounds NH 3 , NO 2 , NO, HCN, and H 2 O. Sampling for sulfur oxides was not requested. Results of the inorganic samples were affected by sampling errors that led to an undefined uncertainty in sample volume. Consequently, tank-headspace concentrations are estimated only. Thirty-nine tentatively identified organic analytes were observed above the detection limit of (ca.) 10 ppbv, but standards for most of these were not available at the time of analysis, and their quantitation is beyond the scope of this study. In addition, we looked for the 41 standard TO-14 analytes. Of these, only a few were observed above the 2-ppbv detection limit. The 16 organic analytes with the highest estimated concentrations are listed. These 16 analytes account for approximately 68% of the total or organic components in Tank BY-105

  20. Vapor space characterization of waste tank 241-BX-104: Results from samples collected on 12/30/94

    International Nuclear Information System (INIS)

    Pool, K.H.; Ligotke, M.W.; McVeety, B.D.

    1995-10-01

    This report describes inorganic and organic analyses results from samples obtained from the headspace of the Hanford waste storage Tank 241-BX-104 (referred to as Tank BX-104). The results described here were obtained to support safety and toxicological evaluations. A summary of the results for inorganic and organic analytes is listed in Table 1. Detailed descriptions of the results appear in the text. Quantitative results were obtained. for the inorganic compounds ammonia (NH 3 ), nitrogen dioxide (NO 2 ), nitric oxide (NO), and water (H 2 O). Sampling for hydrogen cyanide (HCN) and sulfur oxides (SOx) was not requested. In addition, quantitative results were obtained for the 39 TO-14 compounds plus an additional 14 analytes. Of these, 13 were observed above the 5-ppbv reporting cutoff. Sixty-six organic tentatively identified compounds (TICs) were observed above the reporting cutoff of (ca.) 10 ppbv and are reported with concentrations that are semiquantitative estimates based on internal-standard response factors. The 10 organic analytes, with the highest estimated concentrations are listed in Table 1 and account for approximately 70% of the total organic components in Tank BX-104. Two permanent gases, carbon dioxide (CO 2 ) and nitrous oxide (N 2 O), were also detected

  1. Development of grout formulations for 106-AN waste: Mixture-experiment results and analysis

    International Nuclear Information System (INIS)

    Spence, R.D.; McDaniel, E.W.; Anderson, C.M.; Lokken, R.O.; Piepel, G.F.

    1993-09-01

    Twenty potential ingredients were identified for use in developing a 106-AN grout formulation, and 18 were subsequently obtained and tested. Four ingredients-Type II-LA (moderate heat of hydration) Portland cement, Class F fly ash, attapulgite 150 drilling clay, and ground air-cooled blast-furnace slag (GABFS) were selected for developing the 106-AN grout formulations. A mixture experiment was designed and conducted around the following formulation: 2.5 lb of cement per gallon, 1.2 lb of fly ash per gallon, 0.8 lb of attapulgite per gallon, and 3.5 lb of GABFS per gallon. Reduced empirical models were generated from the results of the mixture experiment. These models were used to recommend several grout formulations for 106-AN. Westinghouse Hanford Company selected one of these formulations to be verified for use with 106-AN and a backup formulation in case problems arise with the first choice

  2. Evaluation of hearings. Results from reviews of the nuclear waste issue in the Swedish site candidate municipalities

    International Nuclear Information System (INIS)

    Drottz-Sjoeberg, B.M.

    2001-10-01

    The purpose of this report was to present an evaluation of the public hearings that took place in February of 2001 in the Swedish municipalities of Oesthammar, Tierp, and Aelvkarleby in Norduppland, Hultsfred and Oskarshamn in Smaaland, and Nykoeping in Soedermanland. These municipalities had participated in feasibility studies conducted by the Swedish Nuclear Fuel and Waste Management Co (SKB). A company report on the results of these studies had been published shortly before the hearings (FUD-K). The regulatory authorities, i.e. the Swedish Nuclear Power Inspectorate (SKI) and the Swedish Radiation Protection Institute (SSI), organized the hearings for additional information and aid in their ongoing evaluation of the SKB report. Representatives of the municipalities participated in the planning of the events, and a large meeting in Tierp in January 2001, that also involved the authorities, consultants and interested parties, agreed on the aims and practical arrangements. The authorities furthermore ordered a report for a summary and evaluation of the events, and the results are presented here. The aim of, and the preparations for, the hearings were based on a theoretical model developed within the RISCOM project, i.e. the RISCOM-model of transparency, which postulates three basic elements, i.e. technical/scientific issues, normative issues and authenticity. These elements combine to achieve an optimal clarification on the interaction between scientific and value-laden components in decision-making. An assumption is that the quality of decisions would improve given that transparency can be increased. The hearings were designed to 'stretch' the implementer by means of asking essential questions and to clarify what was achieved and known so far in the process, as well as to clarify what matters required further attention. The content covered technical, legal and social aspects on issues of nuclear waste management and the choices involved in the process towards

  3. Uncertainty and Sensitivity Analysis Results Obtained in the 1996 Performance Assessment for the Waste Isolation Pilot Plant

    Energy Technology Data Exchange (ETDEWEB)

    Bean, J.E.; Berglund, J.W.; Davis, F.J.; Economy, K.; Garner, J.W.; Helton, J.C.; Johnson, J.D.; MacKinnon, R.J.; Miller, J.; O' Brien, D.G.; Ramsey, J.L.; Schreiber, J.D.; Shinta, A.; Smith, L.N.; Stockman, C.; Stoelzel, D.M.; Vaughn, P.

    1998-09-01

    The Waste Isolation Pilot Plant (WPP) is located in southeastern New Mexico and is being developed by the U.S. Department of Energy (DOE) for the geologic (deep underground) disposal of transuranic (TRU) waste. A detailed performance assessment (PA) for the WIPP was carried out in 1996 and supports an application by the DOE to the U.S. Environmental Protection Agency (EPA) for the certification of the WIPP for the disposal of TRU waste. The 1996 WIPP PA uses a computational structure that maintains a separation between stochastic (i.e., aleatory) and subjective (i.e., epistemic) uncertainty, with stochastic uncertainty arising from the many possible disruptions that could occur over the 10,000 yr regulatory period that applies to the WIPP and subjective uncertainty arising from the imprecision with which many of the quantities required in the PA are known. Important parts of this structure are (1) the use of Latin hypercube sampling to incorporate the effects of subjective uncertainty, (2) the use of Monte Carlo (i.e., random) sampling to incorporate the effects of stochastic uncertainty, and (3) the efficient use of the necessarily limited number of mechanistic calculations that can be performed to support the analysis. The use of Latin hypercube sampling generates a mapping from imprecisely known analysis inputs to analysis outcomes of interest that provides both a display of the uncertainty in analysis outcomes (i.e., uncertainty analysis) and a basis for investigating the effects of individual inputs on these outcomes (i.e., sensitivity analysis). The sensitivity analysis procedures used in the PA include examination of scatterplots, stepwise regression analysis, and partial correlation analysis. Uncertainty and sensitivity analysis results obtained as part of the 1996 WIPP PA are presented and discussed. Specific topics considered include two phase flow in the vicinity of the repository, radionuclide release from the repository, fluid flow and radionuclide

  4. Uncertainty and Sensitivity Analysis Results Obtained in the 1996 Performance Assessment for the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Bean, J.E.; Berglund, J.W.; Davis, F.J.; Economy, K.; Garner, J.W.; Helton, J.C.; Johnson, J.D.; MacKinnon, R.J.; Miller, J.; O'Brien, D.G.; Ramsey, J.L.; Schreiber, J.D.; Shinta, A.; Smith, L.N.; Stockman, C.; Stoelzel, D.M.; Vaughn, P.

    1998-01-01

    The Waste Isolation Pilot Plant (WPP) is located in southeastern New Mexico and is being developed by the U.S. Department of Energy (DOE) for the geologic (deep underground) disposal of transuranic (TRU) waste. A detailed performance assessment (PA) for the WIPP was carried out in 1996 and supports an application by the DOE to the U.S. Environmental Protection Agency (EPA) for the certification of the WIPP for the disposal of TRU waste. The 1996 WIPP PA uses a computational structure that maintains a separation between stochastic (i.e., aleatory) and subjective (i.e., epistemic) uncertainty, with stochastic uncertainty arising from the many possible disruptions that could occur over the 10,000 yr regulatory period that applies to the WIPP and subjective uncertainty arising from the imprecision with which many of the quantities required in the PA are known. Important parts of this structure are (1) the use of Latin hypercube sampling to incorporate the effects of subjective uncertainty, (2) the use of Monte Carlo (i.e., random) sampling to incorporate the effects of stochastic uncertainty, and (3) the efficient use of the necessarily limited number of mechanistic calculations that can be performed to support the analysis. The use of Latin hypercube sampling generates a mapping from imprecisely known analysis inputs to analysis outcomes of interest that provides both a display of the uncertainty in analysis outcomes (i.e., uncertainty analysis) and a basis for investigating the effects of individual inputs on these outcomes (i.e., sensitivity analysis). The sensitivity analysis procedures used in the PA include examination of scatterplots, stepwise regression analysis, and partial correlation analysis. Uncertainty and sensitivity analysis results obtained as part of the 1996 WIPP PA are presented and discussed. Specific topics considered include two phase flow in the vicinity of the repository, radionuclide release from the repository, fluid flow and radionuclide

  5. Vapor space characterization of waste tank 241-BY-109 (in situ): Results from samples collected on 9/22/94

    International Nuclear Information System (INIS)

    Pool, K.H.; Clauss, T.W.; Ligotke, M.W.

    1995-06-01

    This report describes inorganic and organic analyses results from in situ samples obtained from the headspace of the Hanford waste storage Tank 241-BY-109 (referred to as Tank BY-109). The results described here were obtained to support safety and toxicological evaluations. A summary of the results for inorganic and organic analytes is listed in Summary Table 1. Detailed descriptions of the results appear in the text. Quantitative results were obtained for the inorganic compounds ammonia (NH 3 ), nitrogen dioxide (NO 2 ), nitric oxide (NO), and water (H 2 O). Sampling for hydrogen cyanide (HCN) and sulfur oxides (SO x ) was not requested. Organic compounds were also quantitatively determined. Twenty-three organic tentatively identified compounds (TICs) were observed above the detection limit of (ca.) 10 ppbv, but standards for most of these were not available at the time of analysis, and the reported concentrations are semiquantitative estimates. In addition, we looked for the 40 standard TO-14 analytes. We observed 38. Of these, only a few were observed above the 2-ppbv calibrated instrument detection limit. The ten organic analytes with the highest estimated concentrations are listed in Summary Table 1. The ten analytes account for approximately 84% of the total organic components in Tank BY-109

  6. Headspace vapor characterization of Hanford waste Tank 241-BX-110: Results from samples collected on 04/30/96

    International Nuclear Information System (INIS)

    Evans, J.C.; Pool, K.H.; Thomas, B.L.; Olsen, K.B.; Fruchter, J.S.; Silvers, K.L.

    1997-01-01

    This report describes the analytical results of vapor samples taken from the headspace of the waste storage tank 241-BX-110 (Tank BX-110) at the Hanford Site in Washington State. The results described in this report were obtained to characterize the vapors present in the tank headspace and to support safety evaluations and tank farm operations. The results include air concentrations of selected inorganic and organic analytes and grouped compounds from samples obtained by Westinghouse Hanford Company (WHC) and provided for analysis to Pacific Northwest National Laboratory (PNNL). Analyses were performed by the Vapor Analytical Laboratory (VAL) at PNNL. Analyte concentrations were based on analytical results and, where appropriate, sample volumes provided by WHC. A summary of the inorganic analytes, permanent gases, and total non-methane organic compounds is listed in a table. The three highest concentration analytes detected in SUMMA trademark canister and triple sorbent trap samples are also listed in the table. Detailed descriptions of the analytical results appear in the appendices

  7. Vapor space characterization of Waste Tank 241-TY-104 (in situ): Results from samples collected on 8/5/94

    International Nuclear Information System (INIS)

    Ligotke, M.W.; Pool, K.H.; Lucke, R.B.

    1995-10-01

    This report describes inorganic and organic analyses results from in situ samples obtained from the headspace of the Hanford waste storage Tank 241-TY-104 (referred to as Tank TY-104). The results described here were obtained to support safety and toxicological evaluations. A summary of the results for inorganic and organic analytes is listed in Table 1. Detailed descriptions of the results appear in the text. Quantitative results were obtained for the inorganic compounds ammonia (NH 3 ), nitrogen dioxide (NO 2 ), nitric oxide (NO), and water (H 2 O). Sampling for hydrogen cyanide (HCN) and sulfur oxides (SO x ) was not performed. In addition, the authors looked for the 39 TO-14 compounds plus an additional 14 analytes. Of these, eight were observed above the 5-ppbv reporting cutoff. Twenty-four organic tentatively identified compounds (TICs) were observed above the reporting cutoff of (ca.) 10 ppbv and are reported with concentrations that are semiquantitative estimates based on internal standard response factors. The 10 organic analytes with the highest estimated concentrations are listed in Table 1 and account for approximately 86% of the total organic components in Tank TY-104. Tank TY-104 is on the Ferrocyanide Watch List

  8. Vapor space characterization of Waste Tank 241-U-106 (in situ): Results from samples collected on 8/25/94

    International Nuclear Information System (INIS)

    Ligotke, M.W.; Lucke, R.B.; Pool, K.H.

    1995-10-01

    This report describes inorganic and organic analyses results from in situ samples obtained from the headspace of the Hanford waste storage Tank 241-U-106 (referred to as Tank U-106). The results described here were obtained to support safety and toxicological evaluations. A summary of the results for inorganic and organic analytes is listed in Table 1. Detailed descriptions of the results appear in the text. Quantitative results were obtained for the inorganic compounds ammonia (NH 3 ), nitrogen dioxide (NO 2 ), nitric oxide (NO), and water (H 2 O). Sampling for hydrogen cyanide (HCN) and sulfur oxides (SO x ) was not performed. In addition, the authors looked for the 39 TO-14 compounds plus an additional 14 target analytes. Of these, six were observed above the 5-ppbv reporting cutoff. Ten organic tentatively identified compounds (TICs) were observed above the reporting cutoff of (ca.) 10 ppbv in two or more of the three samples collected and are reported with concentrations that are semiquantitative estimates based on internal standard response factors. The 10 organic analytes with the highest estimated concentrations are listed in Table 1 and account for approximately 89% of the total organic components in Tank U-106. Methyl isocyanate, a compound of possible concern in Tank U-106, was not detected. Tank U-106 is on the Organic Watch List

  9. Vapor space characterization of waste Tank 241-BY-108: Results from samples collected on 10/27/94

    International Nuclear Information System (INIS)

    McVeety, B.D.; Clauss, T.W.; Ligotke, M.W.

    1995-10-01

    This report describes inorganic and organic analyses results from samples obtained from the headspace of the Hanford waste storage Tank 241-BY-108 (referred to as Tank BY-108). The results described here were obtained to support safety and toxicological evaluations. A summary of the results for inorganic and organic analytes is listed in Table 1. Detailed descriptions of the results appear in the text. Quantitative results were obtained for the inorganic compounds ammonia (NH 3 ), nitrogen dioxide (NO 2 ), nitric oxide (NO), and water vapor (H 2 O). Trends in NH 3 and H 2 O samples indicated a possible sampling problem. Sampling for hydrogen cyanide (HCN) and sulfur oxides (SO x ) was not requested. In addition, the authors looked for the 40 TO-14 compounds plus an additional 15 analytes. Of these, 17 were observed above the 5-ppbv reporting cutoff. Also, eighty-one organic tentatively identified compounds (TICs) were observed above the reporting cutoff (ca.) 10 ppbv, and are reported with concentrations that are semiquantitative estimates based on internal standard response factors. The nine organic analytes with the highest estimated concentrations are listed in Summary Table 1 and account for approximately 48% of the total organic components in the headspace of Tank BY-108. Three permanent gases, hydrogen (H 2 ), carbon dioxide (CO 2 ), and nitrous oxide (N 2 O) were also detected. Tank BY-108 is on the Ferrocyanide Watch List

  10. A treatment strategy for waste waters resulting from uranium mine decommissioning in Romania

    International Nuclear Information System (INIS)

    Georgescu, D.P.; Vacariu, V.T.; Popa, N.

    2000-01-01

    The exploitation activities in two important uranium mining areas in Romania are foreseen to be closed down in correlation with the national energy policy and nuclear strategy. This close down activity involves a number of technical decisions for environmental restoration. Reducing the contamination due to radioactive water of these areas, during the operation period and after the close down period, is one of the main components of the environment rehabilitation strategy. In this paper, the current situation and the program foreseen for ground and surface water treatment at an uranium mining unit situated in the S-W of Romania are presented. This program was established on the base of the results of our research carried out in order to decrease the content of radioactive elements. After closing down the mining facility, naturally flooding waters should be evacuated at the surface by a pump system and properly treated. A station for water decontamination is under construction. The underground water decontamination is based on two methods: ion exchange for uranium and adsorption on active coal for Ra-226. The technological flow chart of the treatment installation is realized on the basis of laboratory and industrial research and it will output treated water with less than 60 mg solid/l, 0.021 mg U/l and 0.088 Bq Ra-226/l. The installation is able to treat contaminated water flow rates between 10 and 30 l/s at a cost of about 0.1 USD/m 3 . The total investment cost is estimated to be 9.7 - 12.6 billions RO Lei (USD 500.000 - 650.000), depending of the treatment capacity. (authors)

  11. Review of research results for the photocatalytic oxidation of hazardous wastes in air

    Energy Technology Data Exchange (ETDEWEB)

    Nimlos, M R; Wolfrum, E J; Gratson, D A; Watt, A S; Jacoby, W A; Turchi, C

    1995-01-01

    Laboratory experiments of gas-phase photocatalytic oxidation (PCO) at NREL have focused on measurements that can help commercialize this technology for treating gaseous air streams. This effort proceeds earlier NREL work and studies conducted elsewhere which demonstrated the general applicability of PCO. The more recent work has concentrated on: (1) the kinetics of the PCO process; (2) the formation and destruction of intermediates; and (3) possible enhancements to improve the destruction rates. The results from these studies will be used to help design large scale PCO equipment and they will be used to evaluate the economics of the PCO process. For trichloroethylene and ethanol, extensive studies of the rates of destruction have yielded kinetic parameters for the destruction of intermediates as well as the substrate. The kinetics of intermediates is essential for sizing a large scale reactor, as complete conversion to carbon dioxide is often desired. The kinetic data from these laboratory studies has been used for analyzing IT`s pilot PCO reactor and has been used to suggest modifications to this unit. For compounds that are more difficult to destroy (such as the components of BTEX), rate enhancement experiments have been conducted. These compounds represent a very large market for this technology and improvement of the rate of the process should make it competitive. Towards this goal, the enhancement of the destruction of BTEX components have been studied. Experiments have demonstrated that there is a significant increase in the rates of destruction of BTEX with the addition of ozone. Preliminary economic assessments have shown that PCO with ozone may be cost competitive. Future laboratory experiments of PCO will focus on refinements of what has been learned. Rate measurements will also be expanded to include other compounds representing significant markets for the PCO technology.

  12. Consequences of a radioactive surface pool resulting from waste transfer operations between tanks 214-C-106 and 241-AY-102

    Energy Technology Data Exchange (ETDEWEB)

    Van Vleet, R.J.

    1997-08-05

    This document contains supporting calculations for quantifying the dose consequences from a pool formed from an underground leak or a-leak from an above grade structure for the Waste Retrieval Sluicing System (Project W-320), i.e., sluicing the contents of Tank 241-C-106 (high heat, SST) into Tank 241-AY-102 (aging waste, DST).

  13. Consequences of a radioactive surface pool resulting from waste transfer operations between tanks 214-C-106 and 241-AY-102

    International Nuclear Information System (INIS)

    Van Vleet, R.J.

    1997-01-01

    This document contains supporting calculations for quantifying the dose consequences from a pool formed from an underground leak or a-leak from an above grade structure for the Waste Retrieval Sluicing System (Project W-320), i.e., sluicing the contents of Tank 241-C-106 (high heat, SST) into Tank 241-AY-102 (aging waste, DST)

  14. The KS-KT-100 plant for two-stage vitrification of radioactive waste: results of tests with simulators

    International Nuclear Information System (INIS)

    Davydov, V.I.; Dobrygin, P.G.; Dolgov, V.V.; Sergeev, G.A.

    1976-01-01

    The Soviet Union has developed a two-stage process for phosphate vitrification of liquid radioactive waste involving the use, at the initial stage, of calcination in the pseudo-liquefied layer, followed by melting of the calcinate in a ceramic crucible (second stage). On the basis of the laboratory studies and bench tests using experimental equipment, the authors have developed and tried out an enlarged plant - the KS-KT-100. The plant includes units for preparing the solution, evaporation, calcination, melting and gas purification. The initial solution containing 240 g/litre of aluminium nitrate, 125 g/litre of sodium nitrate, 120 to 130 g/litre of orthophosphoric acid, and 90 to 150 g/litre of industrial molasses simulated fluxed nitrate waste. The tests have shown that the various units operate satisfactorily. The authors have determined the technological parameters for evaporation, calcination of the solution and melting of the calcinate. The presence of molasses in the solution (150 g/litre) makes it possible to decompose and distil 40% of the nitrate ion during evaporation. The calcination temperature is 350 to 400 0 C, and the fluidization rate 1.5 m/s. The capacity of the plant for the initial solution is 100 litres/h, for the evaporated solution 65 litres/h, and for the glass 20 kg/h. The efficiency of the gas purification system ranges between 10 7 and 10 9 . The test results show the feasibility of the two-stage method of vitrification in actual practice. (author)

  15. Vapor space characterization of waste tank 241-C-101: Results from samples collected on 9/1/94

    International Nuclear Information System (INIS)

    Lucke, R.B.; Clauss, T.W.; Ligotke, M.W.

    1995-11-01

    This report describes results of the analyses of tank-headspace samples taken from the Hanford waste Tank 241-C-101 (referred to as Tank C-101) and the ambient air collected - 30 ft upwind near the tank and through the VSS near the tank. Pacific Northwest Laboratory (PNL) contracted with Westinghouse Hanford Company (WHC) to provide sampling devices and to analyze inorganic and organic analytes collected from the tank headspace and ambient air near the tank. The sample job was designated S4056, and samples were collected by WHC on September 1, 1994, using the vapor sampling system (VSS). The samples were inspected upon delivery to the 326/23B laboratory and logged into PNL record book 55408 before implementation of PNL Technical Procedure PNL-TVP-07. Custody of the sorbent traps was transferred to PNL personnel performing the inorganic analysis and stored at refrigerated (≤ 10 degrees C) temperature until the time of analysis. The canisters were stored in the 326/23B laboratory at ambient (25 degrees C) temperature until the time of the analysis. Access to the 326/23B laboratory is limited to PNL personnel working on the waste-tank safety program. Analyses described in this report were performed at PNL in the 300 area of the Hanford Reservation. Analytical methods that were used are described in the text. In summary, sorbent traps for inorganic analyses containing sample materials were either weighed (for water analysis) or desorbed with the appropriate aqueous solutions (for NH 3 , NO 2 , and NO analyses). The aqueous extracts were analyzed either by selective electrode or by ion chromatography (IC). Organic analyses were performed using cryogenic preconcentration followed by gas chromatography/mass spectrometry (GC/MS)

  16. Dumping of radioactive waste and investigation of contamination in the Kara Sea. Results from 3 years of investigations (1992-1994) in the Kara Sea

    Energy Technology Data Exchange (ETDEWEB)

    Strand, P [Statens Straalevern, Oesteraas (Norway); Foeyn, L [Norsk Inst. for Vannforskning, Oslo (Norway); Nikitin, A I [SPA ` ` Typhoon` ` , Roshydromet (Russian Federation); and others

    1996-03-01

    The report summarises the results obtained from the joint Russian-Norwegian investigation concerning the consequences of dumping of radioactive waste in the Kara Sea. Three expeditions were undertaken to the Kara Sea and the present dumping sites for radioactive waste. Samples of water, sediments and biota were collected and analysed. An impact and risk assessment was performed, based on the information provided through the joint cooperation. Enhanced levels and artificially produced radionuclides in the sediments collected in the very close vicinity of almost all localised dumped objects, demonstrate that leakage occur. No contribution from dumped radioactive waste was observed in the open Kara Sea. Due to the potential for leakage from the dumped waste in the future and the presence of other potential sources in the area, a regular monitoring programme is highly recommended. 65 refs., 42 figs., 16 tabs.

  17. Dumping of radioactive waste and investigation of contamination in the Kara Sea. Results from 3 years of investigations (1992-1994) in the Kara Sea

    International Nuclear Information System (INIS)

    Strand, P.; Foeyn, L.; Nikitin, A.I.

    1996-03-01

    The report summarises the results obtained from the joint Russian-Norwegian investigation concerning the consequences of dumping of radioactive waste in the Kara Sea. Three expeditions were undertaken to the Kara Sea and the present dumping sites for radioactive waste. Samples of water, sediments and biota were collected and analysed. An impact and risk assessment was performed, based on the information provided through the joint cooperation. Enhanced levels and artificially produced radionuclides in the sediments collected in the very close vicinity of almost all localised dumped objects, demonstrate that leakage occur. No contribution from dumped radioactive waste was observed in the open Kara Sea. Due to the potential for leakage from the dumped waste in the future and the presence of other potential sources in the area, a regular monitoring programme is highly recommended. 65 refs., 42 figs., 16 tabs

  18. MATRIX 2 RESULTS OF THE FY07 ENHANCED DOE HIGH-LEVEL WASTE MELTER THROUGHPUT STUDIES AT SRNL

    Energy Technology Data Exchange (ETDEWEB)

    Raszewski, F; Tommy Edwards, T; David Peeler, D

    2008-10-23

    High-level waste (HLW) throughput (i.e., the amount of waste processed per unit time) is a function of two critical parameters: waste loading (WL) and melt rate. For the Waste Treatment and Immobilization Plant (WTP) at the Hanford Site and the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS), increasing HLW throughput would significantly reduce the overall mission life cycle costs for the Department of Energy (DOE). The objective of this study was to generate supplemental validation data that could be used to determine the applicability of the current liquidus temperature (TL) model to expanded DWPF glass composition regions of interest based on higher WLs. Two specific flowsheets were used in this study to provide such insight: (1) Higher WL glasses (45 and 50%) based on future sludge batches that have (and have not) undergone the Al-dissolution process. (2) Coupled operations supported by the Salt Waste Processing Facility (SWPF), which increase the TiO{sub 2} concentration in glass to greater than 2 wt%. Glasses were also selected to address technical issues associated with Al{sub 2}O{sub 3} solubility, nepheline formation, and homogeneity issues for coupled operations. A test matrix of 28 glass compositions was developed to provide insight into these issues. The glasses were fabricated and characterized using chemical composition analysis, X-ray Diffraction (XRD), TL measurement and the Product Consistency Test (PCT). The results of this study are summarized below: (1) TiO{sub 2} concentrations up to {approx} 3.5 wt% were retained in DWPF type glasses, where retention is defined as the absence of crystalline TiO{sub 2} (i.e., unreacted or undissolved) in the as-fabricated glasses. Although this TiO{sub 2} content does not bound the projected SWPF high output flowsheet (up to 6 wt% TiO{sub 2} may be required in glass), these data demonstrate the potential for increasing the TiO{sub 2} limit in glass above the current limit of 2 wt

  19. The IAEA's activities in the field of radioactive waste management

    International Nuclear Information System (INIS)

    Semenov, B.A.

    1984-01-01

    The IAEA has been concerned with radioactive waste management since its inception. Its programme in this area was expanded in the mid 1970s as questions related to the management and disposal of radioactive wastes came into focus in conjunction with the further industrial development of nuclear power. The objectives of the Agency's waste management programme are to assist its Member States in the safe and effective management of wastes by organizing the exchange and dissemination of information, providing guidance and technical assistance and supporting research. The current programme addresses all aspects of the industrial use of nuclear power under the aspects (a) technology of handling and treatment of wastes, (b) underground disposal of wastes, (c) environmental aspects of nuclear energy, including sea disposal of radioactive wastes. Systematic reviews have been made and publications issued concerning the technology of handling, treating, conditioning, and storing various categories of wastes, including liquid and gaseous wastes, wastes from nuclear power plants, spent fuel reprocessing and mining and milling of uranium ores, as well as wastes from decommissioning of nuclear facilities. As waste disposal is the current issue of highest interest, an Agency programme was set up in 1977 to develop a set of guidelines on the safe underground disposal of low-, intermediate- and high-level wastes in shallow ground, rock cavities or deep geological repositories. This programme will continue until 1990. Eleven Safety Series and Technical Documents and Reports have been published under this programme so far, which also addresses safety and other criteria for waste disposal. The environmental part of the waste management programme is concerned with the assessment of radiological and non-radiological consequences of discharges from nuclear facilities, including de minimis concepts in waste disposal and environmental models and data for radionuclide releases

  20. Biosphere transport and radiation dose calculations resulting from radioactive waste stored in deep salt formation (PACOMA-project)

    International Nuclear Information System (INIS)

    Jong, E.J. de; Koester, H.W.; Vries, W.J. de; Lembrechts, J.F.

    1990-03-01

    Parts are presented of the results of a safety-assessment study of disposal of medium and low level radioactive waste in salt formations in the Netherlands. The study concerns several disposal concepts for 2 kinds of salt formation, a deep dome and a shallow dome. 7 cases were studied with the same Dutch inventory and 1 with a reference inventory R, in order to compare results with those of other PACOMA participants. The total activity of the reference inventory R is 30 percent lower than the Dutch inventory, but some long living nuclides such as I-129, Np-237 and U-238 have a considerably higher activity. This reference inventor R has been combined with the disposal concept of mined cavities in a shallow salt dome. In each case. the released fraction of stored radio-nuclides moves gradually with water through the geosphere to the bio-sphere where it enters a river. River water is used for sprinkler irrigation and for drinking by man and livestock. The dispersal of the radionuclides into the biosphere is calculated with the BIOS program of the NRPB. Subroutines linked to the program add doses via different pathways to obtain a maximum individual dose, a collective dose and an integrated collective dose. This study presents results of these calculations. (author). 11 refs.; 39 figs.; 111 tabs

  1. Vapor space characterization of waste Tank 241-C-109 (in situ): Results from samples collected on 6/23/94

    International Nuclear Information System (INIS)

    Clauss, T.W.; Ligotke, M.W.; Pool, K.H.; Lucke, R.B.; McVeety, B.D.; Sharma, A.K.; McCulloch, M.; Fruchter, J.S.; Goheen, S.C.

    1995-10-01

    This report describes organic analyses results from in situ samples obtained from the headspace of the Hanford waste storage Tank 241-C-109 (referred to as Tank C-109). The results described here were obtained to support safety and toxicological evaluations. Organic compounds were quantitatively determined. Thirteen organic tentatively identified compounds (TICs) were observed above the detection limit of (ca.) 10 ppbv, but standards for most of these were not available at the time of analysis, and the reported concentrations are semiquantitative estimates. In addition, the authors looked for the 40 standard TO-14 analytes. Of these, only one was observed above the 2-ppbv calibrated instrumental detection limit. However, it is believed, even though the values for dichlorodifluoromethane and trichlorofluoromethane are below the instrumental detection limit, they are accurate at these low concentrations. The six analytes account for approximately 100% of the total organic components in Tank C-109. These six organic analytes with the highest estimated concentrations are listed in Summary Table 1. Detailed descriptions of the results appear in the text

  2. Vapor space characterization of waste Tank 241-BY-107: Results from in situ sample collected on 3/25/94

    International Nuclear Information System (INIS)

    Sharma, A.K.; Lucke, R.B.; Clauss, T.W.; McVeety, B.D.; Fruchter, J.S.; Goheen, S.C.

    1995-06-01

    This report describes organic results from vapors of the Hanford single-shell waste storage Tank 241-BY-107 (referred to as Tank BY-107). Samples for selected inorganic compounds were obtained but not anlayzed (Section 2.0). Quantitative results were obtained for several organic analytes, but quantities of analytes not listed in US Environmental Protection Agency (EPA) compendium Method TO-14 were estimated. Approximately 80 tentatively identified organic analytes were observed above the detection limit of (ca.) 10 ppbv, but standards for most of these were not available at the time of analysis, and their quantitative determination is beyond the scope of this study. The SUMMATM canister samples were also analyzed for the 41 organic compounds listed in EPA compendium Method TO-14. Of these, only a few were observed above the 2-ppbv detection limits. These are summarized in Table 3.1. Estimated quantities were determined of tentatively identified compounds (TICs). A summary of these results shows quantities of all TICs above the concentration of ca. 10 ppbv. This consists of more than 80 organic analytes. The 12 organic analytes with the highest estimated concentrations are shown

  3. Efficiency of energy recovery from municipal solid waste and the resultant effect on the greenhouse gas balance.

    Science.gov (United States)

    Gohlke, Oliver

    2009-11-01

    Global warming is a focus of political interest and life-cycle assessment of waste management systems reveals that energy recovery from municipal solid waste is a key issue. This paper demonstrates how the greenhouse gas effects of waste treatment processes can be described in a simplified manner by considering energy efficiency indicators. For evaluation to be consistent, it is necessary to use reasonable system boundaries and to take the generation of electricity and the use of heat into account. The new European R1 efficiency criterion will lead to the development and implementation of optimized processes/systems with increased energy efficiency which, in turn, will exert an influence on the greenhouse gas effects of waste management in Europe. Promising technologies are: the increase of steam parameters, reduction of in-plant energy consumption, and the combined use of heat and power. Plants in Brescia and Amsterdam are current examples of good performance with highly efficient electricity generation. Other examples of particularly high heat recovery rates are the energy-from-waste (EfW) plants in Malmö and Gothenburg. To achieve the full potential of greenhouse gas reduction in waste management, it is necessary to avoid landfilling combustible wastes, for example, by means of landfill taxes and by putting incentives in place for increasing the efficiency of EfW systems.

  4. Validation of activity determination codes and nuclide vectors by using results from processing of retired components and operational waste

    International Nuclear Information System (INIS)

    Lundgren, Klas; Larsson, Arne

    2012-01-01

    Decommissioning studies for nuclear power reactors are performed in order to assess the decommissioning costs and the waste volumes as well as to provide data for the licensing and construction of the LILW repositories. An important part of this work is to estimate the amount of radioactivity in the different types of decommissioning waste. Studsvik ALARA Engineering has performed such assessments for LWRs and other nuclear facilities in Sweden. These assessments are to a large content depending on calculations, senior experience and sampling on the facilities. The precision in the calculations have been found to be relatively high close to the reactor core. Of natural reasons the precision will decline with the distance. Even if the activity values are lower the content of hard to measure nuclides can cause problems in the long term safety demonstration of LLW repositories. At the same time Studsvik is processing significant volumes of metallic and combustible waste from power stations in operation and in decommissioning phase as well as from other nuclear facilities such as research and waste treatment facilities. Combining the unique knowledge in assessment of radioactivity inventory and the large data bank the waste processing represents the activity determination codes can be validated and the waste processing analysis supported with additional data. The intention with this presentation is to highlight how the European nuclear industry jointly could use the waste processing data for validation of activity determination codes. (authors)

  5. Interim recommendations concerning the risks to the Dutch population resulting from the use of radioactive wastes in building materials

    International Nuclear Information System (INIS)

    1985-01-01

    The present report, drawn up at the request of the former Minister of Public Health and Environmental Affairs, discusses the potential radiological consequences for the population of the Netherlands of using waste materials as building materials in housing construction. In his request the Minister points to the growing need to use various waste products as building materials. The highest increase of the effective dose equivalent for the foreseeable use of waste products in building construction implies that the annual exposure, averaged over the entire population, could eventually be increased by a maximum of 0.05 mSv per caput. (Auth.)

  6. Impact of Waste Materials Resulting from the Refining of Crude Oil on Some Soil Physico-Chemical Properties

    Directory of Open Access Journals (Sweden)

    pari asadi alasvand

    2017-02-01

    Full Text Available Introduction: Soil and ground water pollution with organic matter and toxic materials is an ordinary environmental problem. In this case, oil compounds are among the most important environmental pollutants. Tehran oil refinery is one of the largest and oldest refineries in Iran located south of Tehran city. Since the beginning of its activity in 1968, its waste materials (solid, semisolid and liquid have been disposed in large lagoons next to the refinery site. During this long period, considerable changes in soil properties have occurred, which are of great research interests for soil and environmental scientists. Materials and Methods: The studied area (about 60 ha was located in the south of Tehran (latitude: 35°30.299' to 35°30.814' N and longitude: 51°25.682' to 51°26.296' E. Six pedons, including four Technosols developed on the oil refinery waste materials (pedons no. 1, 3, 4 and 6 and two reference pedons (pedons no. 2 and 5 were fully described and sampled. Particle size distribution (PSD of gypsiferous samples was determined by the specific method for gypsiferous soils (Hesse, 1976. PSD of non-gypsiferous samples were determined according to the standard hydrometer method (Gee and Bauder, 1986, but the oil-polluted samples were analyzed according to the standard ultrasound method (Sawhny, 1996. Organic carbon content was determined by Walkley and Black (1934. pH and EC were measured in soil saturation extracts using EC and pH meter (Jenway. Gypsum and CaCO3 contents were determined using acetone (Sparks, 1996, and calcimetery methods, respectively. Mineralogical analysis was done by Decarreau (1990. Micromorphological descriptions were carried out using the terminology of Stoops (2003. Diagnostic horizons were identified and finally the studied pedons were classified according to the Keys to Soil Taxonomy (Soil Survey Staff, 2014 and the World Reference Base (FAO, 2014. Results and Discussion: Horizons of both polluted and

  7. Assessment of biological effects resulting from large scale applications of coal power plant wastes in building technology in Poland

    International Nuclear Information System (INIS)

    Pensko, J.; Geisler, J.

    1980-01-01

    Some of the building materials commonly used in Poland contain natural radioactive elements and some contain radioactive industrial wastes. It has been shown that these building materials could induce additional annual doses to the inhabitants of the order of 0.4 mGy gamma radiation to the whole body and about 13 mSv alpha radiation to the critical tissues of the respiratory tract. On the basis of these dosimetric data and demographic and forecasting data, the number of severe genetic effects and cancer deaths caused by the additional radiation doses in dwellings were assessed for the population of Poland for the period 1951-2010. It was estimated that additional somatic effects in six consecutive decades will result in approximately 31,200 cancer deaths, including about 26,300 deaths caused by lung cancer. The expected number of severe genetic effects resulting from additional doses of ionizing radiation absorbed by parents indoors will amount to about 260 cases in the first generation and about 7500 cases in succeeding generations. (H.K.)

  8. Preliminary investigation results as applied to utilization of Ukrainian salt formations for disposal of high-level radioactive waste

    International Nuclear Information System (INIS)

    Shekhunova, S.B.; Khrushchov, D.P.; Petrichenko, O.I.

    1994-01-01

    The salt-bearing formations have been investigated in five regions of Ukraine. Upper Devonian and Lower Permian evaporite formations in Dnieper-Donets Depression and in the NW part of Donets basin are considered to be promising for disposal of high-level radioactive waste (HLRW). Rock salt occurs there either as bedded salts or as salt pillows and salt diapirs. Preliminary studies have resulted in selection of several candidate sites that show promise for construction of a subsurface pilot lab. Ten salt domes and two sites in bedded salts have been proposed for further exploration. Based on microstructural studies it is possible to separate the body of a salt structure and to locate within its limits the rock salt structure and to locate within its limits the rock salt blocks of different genesis, i.e.: (a) blocks characteristic of initial undisturbed sedimentary structure; (b) flow zones; (c) sliding planes; (d) bodies of loose or uncompacted rock salt. Ultramicrochemical examination of inclusions in halite have shown that they are composed of more than 40 minerals. It is emphasized that to assess suitability of a structure for construction of subsurface lab, and also the potential construction depth intervals, account should be taken of the results of ultra microchemical and microstructural data

  9. Results of Waste Transfer and Back-Dilution in Tanks 241-SY-101 and 241-SY-102

    International Nuclear Information System (INIS)

    Mahoney, L.A.; Antoniak, Z.I.; Barton, W.B.; Conner, J.M.; Kirch, N.W.; Stewart, C.W.; Wells, B.E.

    2000-01-01

    This report chronicles the process of remediation of the flammable gas hazard in Tank 241-SY-101 (SY-101) by waste transfer and back-dilution from December 18, 1999 through April 2, 2000. A brief history is given of the development of the flammable gas retention and release hazard in this tank, and the transfer and dilution systems are outlined. A detailed narrative of each of the three transfer and dilution campaigns is given to provide structure for the balance of the report. Details of the behavior of specific data are then described, including the effect of transfer and dilution on the waste levels in Tanks SY-101 and SY-102, data from strain gauges on equipment suspended from the tank dome, changes in waste configuration as inferred from neutron and gamma logs, headspace gas concentrations, waste temperatures, and the mixerpump operating performance. Operating data and performance of the transfer pump in SY-101 are also discussed

  10. SRTC Spreadsheet to Determine Relative Percent Difference (RPD) for Duplicate Waste Assay Results and to Perform the RPD Acceptance Test

    International Nuclear Information System (INIS)

    Casella, V.R.

    2002-01-01

    This report documents the calculations and logic used for the Microsoft(R) Excel spreadsheet that is used at the 773-A Solid Waste Assay Facility for evaluating duplicate analyses, and validates that the spreadsheet is performing these functions correctly

  11. ORNL results for Test Case 1 of the International Atomic Energy Agency's research program on the safety assessment of Near-Surface Radioactive Waste Disposal Facilities

    International Nuclear Information System (INIS)

    Thorne, D.J.; McDowell-Boyer, L.M.; Kocher, D.C.; Little, C.A.; Roemer, E.K.

    1993-01-01

    The International Atomic Energy Agency (IAEA) started the Coordinated Research Program entitled '''The Safety Assessment of Near-Surface Radioactive Waste Disposal Facilities.'' The program is aimed at improving the confidence in the modeling results for safety assessments of waste disposal facilities. The program has been given the acronym NSARS (Near-Surface Radioactive Waste Disposal Safety Assessment Reliability Study) for ease of reference. The purpose of this report is to present the ORNL modeling results for the first test case (i.e., Test Case 1) of the IAEA NSARS program. Test Case 1 is based on near-surface disposal of radionuclides that are subsequently leached to a saturated-sand aquifer. Exposure to radionuclides results from use of a well screened in the aquifer and from intrusion into the repository. Two repository concepts were defined in Test Case 1: a simple earth trench and an engineered vault

  12. A methodology for assessing social considerations in transport of low and intermediate level radioactive waste

    International Nuclear Information System (INIS)

    Allsop, R.E.; Banister, D.J.; Holden, D.J.; Bird, J.; Downe, H.E.

    1986-05-01

    A methodology is proposed for taking into account non-radiological social aspects of the transport of low and intermediate level radioactive waste when considering the location of disposal facilities and the transport of waste to such facilities from the sites where it arises. As part of a data acquisition programme, an attitudinal survey of a sample of people unconnected with any suggested site or transport route is proposed in order to estimate levels of concern felt by people of different kinds about waste transport. Probabilities of accident occurrence during transport by road and rail are also discussed, and the limited extent of quantified information about consequences of accidents is reviewed. The scope for malicious interference with consignments of waste in transit is considered. (author)

  13. Results of HWVP transuranic process waste treatment laboratory and pilot-scale filtration tests using specially ground zeolite

    International Nuclear Information System (INIS)

    Eakin, D.E.

    1996-03-01

    Process waste streams from the Hanford Waste Vitrification Plant (HWVP) may require treatment for cesium, strontium, and transuranic (TRU) element removal in order to meet criteria for incorporation in grout. The approach planned for cesium and strontium removal is ion exchange using a zeolite exchanger followed by filtration. Filtration using a pneumatic hydropulse filter is planned to remove TRU elements which are associated with process solids and to also remove zeolite bearing the cesium and strontium. The solids removed during filtration are recycled to the melter feed system to be incorporated into the HWVP glass product. Fluor Daniel, Inc., the architect-engineering firm for HWVP, recommended a Pneumatic Hydropulse (PHP) filter manufactured by Mott Metallurgical Corporation for use in the HWVP. The primary waste streams considered for application of zeolite contact and filtration are melter off-gas condensate from the submerged bed scrubber (SBS), and equipment decontamination solutions from the Decontamination Waste Treatment Tank (DWTT). Other waste streams could be treated depending on TRU element and radionuclide content. Laboratory and pilot-scale filtration tests were conducted to provide a preliminary assessment of the adequacy of the recommended filter for application to HWVP waste treatment

  14. An assessment of potential risk resulting from a maximum credible accident scenario at the proposed explosive waste storage facility (EWSF)

    International Nuclear Information System (INIS)

    Otsuki, K.; Harrach, R.; Berger, R.

    1992-10-01

    Lawrence Livermore National Laboratory (LLNL) proposes to build, permit, and operate a storage facility for explosive wastes at LLNL's Explosive Test Site, Site 300. The facility would consist of four existing magazines, four new magazettes (small concrete vaults), and a new prefabricated metal building. Ash from on-site treatment of explosive waste would also be stored in the prefabricated metal building prior to sampling analysis, and shipment. The magazettes would be installed at each magazine-and would provide segregated storage for explosive waste types including detonators, actuators, and other initiating devices. The proposed facility would be used to store explosive wastes generated by the Hydrotest and Explosive Development Programs at LLNL prior to treatment on-site or shipment to permitted, commercial, off-site treatment facilities. Explosive wastes to be stored in the proposed facility represent a full spectrum of Department of Energy (DOE) and LLNL explosive wastes. This document identifies and evaluates the risk to human health and the environment associated with the operation of the proposed EWSF

  15. Methodology for Safety Assessment Applied to Predisposal Waste Management. Report of the Results of the International Project on Safety Assessment Driving Radioactive Waste Management Solutions (SADRWMS) 2004–2010)

    International Nuclear Information System (INIS)

    2015-12-01

    Report of the Results of the International Project on Safety Assessment Driving Radioactive Waste Management Solutions (SADRWMS) (2004–2010) The IAEA’s progamme on Safety Assessment Driving Radioactive Waste Management Solutions (SADRWMS) focused on approaches and mechanisms for application of safety assessment methodologies for the predisposal management of radioactive waste. The initial outcome of the SADRWMS Project was achieved through the development of flowcharts, which have since been incorporated into IAEA Safety Standards Series No. GSG-3, Safety Case and Safety Assessment for Predisposal Management of Radioactive Waste. In 2005, an initial specification was developed for the Safety Assessment Framework (SAFRAN) software tool to apply the SADRWMS flowcharts. In 2008, an in-depth application of the SAFRAN tool and the SADRWMS methodology was carried out on the predisposal management facilities of the Thailand Institute of Nuclear Technology Radioactive Waste Management Centre (TINT Facility). This publication summarizes the content and outcomes of the SADRWMS programme. The Chairman’s Report of the SADRWMS Project and the Report of the TINT test case are provided on the CD-ROM which accompanies this report

  16. Results of hydrologic research at a low-level radioactive-waste disposal site near Sheffield, Illinois

    Science.gov (United States)

    Ryan, Barbara J.

    1989-01-01

    Ten years of hydrologic research have been conducted by the U.S. Geological Survey at a commercial low-level radioactive-waste disposal site near Sheffield, Illinois. Research included studies of microclimate, evapotranspiration, and tritium release by plants; runoff and land modification; water movement through a trench cover; water and tritium movement in the unsaturated zone; gases in the unsaturated zone; water and tritium movement in the saturated zone; and water chemistry. Implications specific to each research topic and those based on overlapping research topics are summarized as to their potential effect on the selection, characterization, design, operation, and decommissioning processes of future low-level radioactive-waste disposal sites. Unconsolidated deposits at the site are diverse in lithologic character and are spatially and stratigraphically complex. Thickness of these Quaternary deposits ranges from 3 to 27 meters and averages 17 meters. The unconsolidated deposits overlay 140 meters of Pennsylvanian shale, mudstone, siltstone, and coal. Approximately 90,500 cubic meters of waste were buried from August 1967 through August 1978, in 21 trenches that were constructed in glacial materials by using a cut-and-fill process. Trenches generally were constructed below grade and ranged from 11 to 180 meters long, 2.4 to 21 meters wide, and 2.4 to about 7.9 meters deep. Research on microclimate and evapotranspiration at the site was conducted from July 1982 through June 1984. Continuous measurements were made of precipitation, incoming and reflected solar (shortwave) radiation, incoming and emitted terrestrial (longwave) radiation, horizontal windspeed and direction, wet- and dry-bulb air temperature, barometric pressure, soil-heat fluxes, and soil temperature. Soil-moisture content, for this research phase, was measured approximately biweekly. Evapotranspiration rates were estimated by using three techniques--energy budget, aerodynamic profile, and water

  17. Liquid and Gaseous Waste Operations Department annual operating report, CY 1992

    Energy Technology Data Exchange (ETDEWEB)

    Gillespie, M.A.; Maddox, J.J.; Scott, C.B.

    1993-03-01

    A total of 6.05 x 10{sup 7} gal of liquid waste was decontaminated by the Process Waste Treatment Plant (PWTP) ion exchange system during CY 1992. This averaged to 115 gpm throughout the year. When necessary, a wastewater sidestream of 50--80 gpm was treated through the use of a natural zeolite treatment system. An additional 8.00 x 10{sup 6} gal (average of 15 gpm throughout the year) were treated by the zeolite system. Therefore, the average total flow treated at the PWTP for CY 1992 was 130 gpm. In mid-June, the zeolite system was repiped to allow it the capability to treat the ion exchange system`s discharge due to rising Cs problems in the wastewater. While being used to treat the ion exchange system`s discharge, it cannot treat a sidestream of wastewater. During the year, the regeneration of the cation exchange resins resulted in the generation of 7.83 x 10{sup 3} gal of liquid low-level waste (LLLW) concentrate and 1.15 x 10{sup 4} gal of LLLW evaporator feed. The head-end softening process (precipitation/clarification) generated 604 drums (4.40 x 10{sup 3} ft{sup 3}) of solid low-level waste sludge. The zeolite treatment system generated approximately 8.40 x 10{sup 2} ft{sup 3} of spent zeolite resin, which was turned over to the Solid Waste Operations Department for disposal. See Table 1 for a monthly summary of activities at the PWTP. Figures 1, 2, 3, and 4 show a comparison of operations at the PWTP in 1992 with previous years. Figure 5 shows a comparison of annual rainfall at Oak Ridge National Laboratory (ORNL) since 1987. A total of 1.55 x 10{sup 8} gal of liquid waste (average of 294 gpm throughout the year) was treated at the Nonradiological Wastewater Treatment Plant (NRWTP). Of this amount, 1.40 x 10{sup 7} gal were treated by the precipitation/clarification process for removal of heavy metals. Twenty-five boxes (1.60 x 10{sup 3} ft{sup 3}) of solid sludge generated by the precipitation/clarification process were removed from the filter press room.

  18. Liquid and Gaseous Waste Operations Department annual operating report, CY 1992

    Energy Technology Data Exchange (ETDEWEB)

    Gillespie, M.A.; Maddox, J.J.; Scott, C.B.

    1993-03-01

    A total of 6.05 x 10[sup 7] gal of liquid waste was decontaminated by the Process Waste Treatment Plant (PWTP) ion exchange system during CY 1992. This averaged to 115 gpm throughout the year. When necessary, a wastewater sidestream of 50--80 gpm was treated through the use of a natural zeolite treatment system. An additional 8.00 x 10[sup 6] gal (average of 15 gpm throughout the year) were treated by the zeolite system. Therefore, the average total flow treated at the PWTP for CY 1992 was 130 gpm. In mid-June, the zeolite system was repiped to allow it the capability to treat the ion exchange system's discharge due to rising Cs problems in the wastewater. While being used to treat the ion exchange system's discharge, it cannot treat a sidestream of wastewater. During the year, the regeneration of the cation exchange resins resulted in the generation of 7.83 x 10[sup 3] gal of liquid low-level waste (LLLW) concentrate and 1.15 x 10[sup 4] gal of LLLW evaporator feed. The head-end softening process (precipitation/clarification) generated 604 drums (4.40 x 10[sup 3] ft[sup 3]) of solid low-level waste sludge. The zeolite treatment system generated approximately 8.40 x 10[sup 2] ft[sup 3] of spent zeolite resin, which was turned over to the Solid Waste Operations Department for disposal. See Table 1 for a monthly summary of activities at the PWTP. Figures 1, 2, 3, and 4 show a comparison of operations at the PWTP in 1992 with previous years. Figure 5 shows a comparison of annual rainfall at Oak Ridge National Laboratory (ORNL) since 1987. A total of 1.55 x 10[sup 8] gal of liquid waste (average of 294 gpm throughout the year) was treated at the Nonradiological Wastewater Treatment Plant (NRWTP). Of this amount, 1.40 x 10[sup 7] gal were treated by the precipitation/clarification process for removal of heavy metals. Twenty-five boxes (1.60 x 10[sup 3] ft[sup 3]) of solid sludge generated by the precipitation/clarification process were removed from the filter

  19. Liquid and Gaseous Waste Operations Department annual operating report, CY 1992

    International Nuclear Information System (INIS)

    Gillespie, M.A.; Maddox, J.J.; Scott, C.B.

    1993-03-01

    A total of 6.05 x 10 7 gal of liquid waste was decontaminated by the Process Waste Treatment Plant (PWTP) ion exchange system during CY 1992. This averaged to 115 gpm throughout the year. When necessary, a wastewater sidestream of 50--80 gpm was treated through the use of a natural zeolite treatment system. An additional 8.00 x 10 6 gal (average of 15 gpm throughout the year) were treated by the zeolite system. Therefore, the average total flow treated at the PWTP for CY 1992 was 130 gpm. In mid-June, the zeolite system was repiped to allow it the capability to treat the ion exchange system's discharge due to rising Cs problems in the wastewater. While being used to treat the ion exchange system's discharge, it cannot treat a sidestream of wastewater. During the year, the regeneration of the cation exchange resins resulted in the generation of 7.83 x 10 3 gal of liquid low-level waste (LLLW) concentrate and 1.15 x 10 4 gal of LLLW evaporator feed. The head-end softening process (precipitation/clarification) generated 604 drums (4.40 x 10 3 ft 3 ) of solid low-level waste sludge. The zeolite treatment system generated approximately 8.40 x 10 2 ft 3 of spent zeolite resin, which was turned over to the Solid Waste Operations Department for disposal. See Table 1 for a monthly summary of activities at the PWTP. Figures 1, 2, 3, and 4 show a comparison of operations at the PWTP in 1992 with previous years. Figure 5 shows a comparison of annual rainfall at Oak Ridge National Laboratory (ORNL) since 1987. A total of 1.55 x 10 8 gal of liquid waste (average of 294 gpm throughout the year) was treated at the Nonradiological Wastewater Treatment Plant (NRWTP). Of this amount, 1.40 x 10 7 gal were treated by the precipitation/clarification process for removal of heavy metals. Twenty-five boxes (1.60 x 10 3 ft 3 ) of solid sludge generated by the precipitation/clarification process were removed from the filter press room

  20. Environmental and other evaluations of alternatives for management of defense transuranic waste at the Idaho National Engineering Laboratory

    International Nuclear Information System (INIS)

    1982-04-01

    Appendices to this report contain the following information: INEL history of Waste Management; text of communications between Idaho and the federal government on long-term management; agency and public response to a proposed environmental impact statement; updated estimates on radiological releases from the slagging-pyrolysis incinerator; modeling studies of subsurface migration of radionuclides; nonradiological emissions and their environmental effects; methods for calculating radiological consequences; analysis of abnormal events in conceptual retrieval and processing operations; environmental contamination by accidental releases; hazards to waste management workers; environmental and other effects of rail and truck shipment of wastes; effects of hypothetical worst-case shipping accidents in urban areas; environmental and other effects of processing INEL transuranic waste at the offsite geological repository; and regulations applicable to INEL TRU waste management

  1. Current status of the waste identification program at AECL's Chalk River Laboratories

    International Nuclear Information System (INIS)

    Csullog, G.W.; Edwards, N.W.; TerHuurne, M.A.

    1998-01-01

    The management of routine operating waste by Waste Management and Decommissioning (WM and D) at Atomic Energy of Canada Limited's (AECL) Chalk River Laboratories (CRL) is supported by the Waste Identification (WI) Program. The principal purpose of the WI Program is to minimize the cost and the effort associated with waste characterization and waste tracking, which are needed to optimize waste handling, storage and disposal. The major steps in the WI Program are: (1) identify and characterize the processes that generate the routine radioactive wastes accepted by WM and D - radioisotope production, radioisotope use, reactor operation, fuel fabrication, et cetera (2) identify and characterize the routine blocks of waste generated by each process or activity - the initial characterization is based on inference (process knowledge) (3) prepare customized, template data sheets for each routine waste block - templates contain information such as package type, waste material, waste type, solidifying agent, the average non-radiological contaminant inventory, the average radiological contaminant inventory, and the waste class (4) ensure generators 'use the right piece of paper with the right waste' when they transfer waste to WM and D - that is they use the correct template data sheets to transfer routine wastes, by: identifying and marking waste collection points in the generator's facility; ensuring that generators implement effective waste collection/segregation procedures; implementing standard procedures to transfer waste to WM and D; and, auditing waste collection and segregation within a generator's facility (5) determine any additional waste block characterization requirements (is anything needed beyond the original characterization by process knowledge?) This paper describes the WI Program, it provides an example of its implementation, and it summarizes the current status of its implementation for both CRL and non-CRL waste generators. (author)

  2. Initial results from the canistered waste forms produced during the first campaign of the DWPF Startup Test Program

    International Nuclear Information System (INIS)

    Harbour, J.R.

    1995-01-01

    As part of the Defense Waste Processing Facility (DWPF) Startup Test Program, approximately 90 canisters will be filled with glass containing simulated radioactive waste during five separate campaigns. The first campaign is a facility acceptance test to demonstrate the operability of the facility and to collect initial data on the glass and the canistered waste forms. During the next four campaigns (the waste qualification campaigns) data will be obtained which will be used to demonstrate that the DWPF product meets DOE's Waste Acceptance Product Specifications (WAPS). Currently 12 of the 16 canisters have been filled with glass during the first campaign (FA-13). This paper describes the tests that have been carried out on these 12 glass-filled canisters and presents the data with reference to the acceptance criteria of the WAPS. These tests include measurement of canister dimensions prior to and after glass filling. dew point, composition, and pressure of the gas within the free volume of the canister, fill height, free volume, weight, leak rates of welds and temporary seals, and weld parameters

  3. Study of optimal transformation of liquid effluents resulting from the destruction of radioactive sodium by water into ultimate solid wastes

    International Nuclear Information System (INIS)

    Rodriguez, G.; Camaro, S.; Fiquet, O.; Bernard, A.; Le Bescop, P.

    1997-01-01

    In the framework of sodium waste processing, it has been proposed to retain only processes that treat the sodium using water, thus generating the same by-products: hydrogen and sodium hydroxide. As the objective is to minimise radioactive liquid releases and as, moreover, the authorizations with respect to sodium salt releases are highly restrictive, several solutions have been envisaged for transforming the active sodium hydroxide coming from sodium destruction processes into ultimate solid wastes that can be stored on the surface in a storage site approved by the ANDRA (National Radioactive Waste Management Agency): the Aube Storage Site (CSA). Two processes have been considered and compared: immobilisation in concrete (cementation) and immobilisation in ceramic (ceramisation). These two processes are evaluated according to several criteria: the state of advancement of the process, the quantity of sodium hydroxide (and therefore of sodium) that can be treated per package. (author)

  4. Scoring methods and results for qualitative evaluation of public health impacts from the Hanford high-level waste tanks. Integrated Risk Assessment Program

    International Nuclear Information System (INIS)

    Buck, J.W.; Gelston, G.M.; Farris, W.T.

    1995-09-01

    The objective of this analysis is to qualitatively rank the Hanford Site high-level waste (HLW) tanks according to their potential public health impacts through various (groundwater, surface water, and atmospheric) exposure pathways. Data from all 149 single-shell tanks (SSTs) and 23 of the 28 double-shell tanks (DSTs) in the Tank Waste Remediation System (TWRS) Program were analyzed for chemical and radiological carcinogenic as well as chemical noncarcinogenic health impacts. The preliminary aggregate score (PAS) ranking system was used to generate information from various release scenarios. Results based on the PAS ranking values should be considered relative health impacts rather than absolute risk values

  5. Results concerning a clean co-combustion technology of waste biomass with fossil fuel, in a pilot fluidised bed combustion facility

    Energy Technology Data Exchange (ETDEWEB)

    Ionel, Ioana; Trif-Tordai, Gavril; Ungureanu, Corneliu; Popescu, Francisc; Lontis, Nicolae [Politehnica Univ. Timisoara (Romania). Faculty for Mechanical Engineering

    2008-07-01

    The research focuses on a facility, the experimental results, interpretation and future plans concerning a new developed technology of using waste renewable energy by applying the cocombustion of waste biomass with coal, in a fluidised bed system. The experimental facility is working entirely in accordance to the allowed limits for the exhaust flue gas concentration, with special concern for typical pollutants. The experiments conclude that the technology is cleaner, has as main advantage the possibility to reduce both the SO{sub 2} and CO{sub 2} exhaust in comparison to standard fossil fuel combustion, under comparable circumstances. The combustion is occurring in a stable fluidised bed. (orig.)

  6. Results and prospects of personnel protection against radiation in a low and medium level waste storage center

    International Nuclear Information System (INIS)

    Scheidhauer, J.; Lasseur, C.

    1982-01-01

    The national low- and intermediate-level waste storage center (CM) located near the Channel is presented, together with its history, the various regulations governing it and their incidence on its operation. A description is given of the radiation sources and workplaces to be found in this kind of waste storage center. The radiation protection problems associated with the respective workplaces are analysed: organization, prevention, collective and individual protection, dosimetry. Occupational radiation exposures from 1970 to 1980 are presented. The various means of reducing operating workers' exposures are discussed, especially the incidence of transportation [fr

  7. Result Summary for the Area 5 Radioactive Waste Management Site Performance Assessment Model Version 4.110

    International Nuclear Information System (INIS)

    2011-01-01

    Results for Version 4.110 of the Area 5 Radioactive Waste Management Site (RWMS) performance assessment (PA) model are summarized. Version 4.110 includes the fiscal year (FY) 2010 inventory estimate, including a future inventory estimate. Version 4.110 was implemented in GoldSim 10.11(SP4). The following changes have been implemented since the last baseline model, Version 4.105: (1) Updated the inventory and disposal unit configurations with data through the end of FY 2010. (1) Implemented Federal Guidance Report 13 Supplemental CD dose conversion factors (U.S. Environmental Protection Agency, 1999). Version 4.110 PA results comply with air pathway and all-pathways annual total effective dose (TED) performance objectives (Tables 2 and 3, Figures 1 and 2). Air pathways results decrease moderately for all scenarios. The time of the maximum for the air pathway open rangeland scenario shifts from 1,000 to 100 years (y). All-pathways annual TED increases for all scenarios except the resident scenario. The maximum member of public all-pathways dose occurs at 1,000 y for the resident farmer scenario. The resident farmer dose was predominantly due to technetium-99 (Tc-99) (82 percent) and lead-210 (Pb-210) (13 percent). Pb-210 present at 1,000 y is produced predominantly by radioactive decay of uranium-234 (U-234) present at the time of disposal. All results for the postdrilling and intruder-agriculture scenarios comply with the performance objectives (Tables 4 and 5, Figures 3 and 4). The postdrilling intruder results are similar to Version 4.105 results. The intruder-agriculture results are similar to Version 4.105, except for the Pit 6 Radium Disposal Unit (RaDU). The intruder-agriculture result for the Shallow Land Burial (SLB) disposal units is a significant fraction of the performance objective and exceeds the performance objective at the 95th percentile. The intruder-agriculture dose is due predominantly to Tc-99 (75 percent) and U-238 (9.5 percent). The acute

  8. Analysis the potential gas production of old municipal solid waste landfill as an alternative energy source: Preliminary results

    Science.gov (United States)

    Hayati, A. P.; Emalya, N.; Munawar, E.; Schwarzböck, T.; Lederer, J.; Fellner, J.

    2018-03-01

    The MSW landfill produces gas which is represent the energy resource that lost and polluted the ambient air. The objective of this study is to evaluate the potential gas production of old landfill as an alternative energy source. The study was conducted by using 10 years old waste in landfill simulator reactor (LSR). Four Landfills Simulator Reactors (LSR) were constructed for evaluate the gas production of old MSW landfilled. The LSR was made of high density poly ethylene (HDPE) has 50 cm outside diameter and 150 cm of high. The 10 years old waste was excavated from closed landfill and subsequently separated from inorganic fraction and sieved to maximum 50 mm size particle prior emplaced into the LSR. Although quite small compare to the LSR containing fresh waste has been reported, the LRS containing 10 years old waste still produce much landfill gas. The landfill gas produced of LSR operated with and without leachate recirculation were about 29 and 21 litter. The composition of landfill gas produced was dominated by CO2 with the composition of CH4 and O2 were around 12.5% and 0.2 %, respectively.

  9. First results of in-can microwave processing experiments for radioactive liquid wastes at the Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    White, T.L.; Youngblood, E.L.; Berry, J.B.; Mattus, A.J.

    1990-01-01

    The Oak Ridge National Laboratory (ORNL) Waste Handling and Packaging Plant is developing a microwave process to reduce and solidify remote-handled transuranic (RH-TRU) liquids and sludges presently stored in large tanks at ORNL. Testing has recently begun on an in drum microwave process using nonradioactive RH-TRU surrogates. The microwave process development effort has focused on an in-drum process to dry the RH-TRU liquids and sludges in the final storage container and then melt the salt residues to form a solid monolith. A 1/3-scale proprietary microwave applicator was designed, fabricated, and tested to demonstrate the essential features of the microwave design and to provide input into the design of the full-scale applicator. Conductivity cell measurements suggest that the microwave energy heats near the surface of the surrogate over a wide range of temperatures. The final wasteform meets the waste acceptance criteria for the Waste Isolation Pilot Plant, a federal repository for defense transuranic wastes near Carlsbad, New Mexico. 7 refs., 3 figs., 1 tab

  10. Evaluation of the water hazard potential of solid wastes. Pt. 1. Experimental results; Untersuchung von Abfaellen mit biologischen Testverfahren zur Bewertung der Wassergefaehrdung. T. 1. Experimentelle Ergebnisse

    Energy Technology Data Exchange (ETDEWEB)

    Brackemann, H.; Hahn, J.; Vogel, U. [Umweltbundesamt, Berlin (Germany); Hagendorf, U. [Umweltbundesamt, Langen (Germany)

    2000-07-01

    Wastes from three different types of waste treatment facilities (slag from a municipal waste incineration plant, slag granules from a pilot plant combining carbonization and incineration, mechanical and biological treated wastes) were examined to determine their hazard potential to different waters sites. The process temperature is seen to be the main difference between the three treatment processes. The wastes were extracted with water according to the German standard DIN 38414 S 4 and additionally at a constant pH value of 4. The leachates were investigated in a battery of aquatic bioassays and characterised physically and chemically. Every leachate revealed in a toxic effect at least in one test. The toxicity of the leachates prepared at a pH of 4 was significantly higher than the toxicity of the leachates prepared by extraction with water without pH adjustment. The leachates of the slag granules showed the lowest toxicity. On the basis of these experimental results, a scheme to derive Water Hazard Classes of wastes, which is presented in part II of this publication, was developed. (orig.) [German] Zur Bestimmung der wassergefaehrdenden Eigenschaften wurden die Eluate von Abfaellen/Rueckstaenden aus drei verschiedenen Abfallbehandlungsanlagen untersucht (Schlacke aus einer Abfallverbrennungsanlage (Rostfeuerung), Schmelzgranulat aus einer Versuchsanlage mit kombinierter Verschwelung und Hochtemperaturverbrennung sowie biologisch-mechanisch behandelter Abfall). Ein wesentlicher Unterschied dieser drei Verfahren liegt in der Behandlungstemperatur. Die Rueckstaende wurden nach DIN 38414, Teil 4 sowie bei einem konstant eingestellten pH-Wert von 4 eluiert. Die Eluate wurden mit verschiedenen aquatischen Biotests untersucht sowie physikalisch-chemisch charakterisiert. Dabei zeigte sich, dass jedes untersuchte Eluat in mindestens einem Test eine toxische Wirkung aufwies; die Toxizitaet der bei saurem pH-Wert durchgefuehrten Elutionen war deutlich erhoeht. Die Eluate

  11. Management of radioactive waste

    International Nuclear Information System (INIS)

    Neerdael, B.; Marivoet, J.; Put, M.; Van Iseghem, P.; Volckaert, G.; Wacquier, W.

    1998-09-01

    The document gives an overview of of different aspects of radioactive waste management in Belgium. The document discusses the radioactive waste inventory in Belgium, the treatment and conditioning of radioactive waste as well as activities related to the characterisation of different waste forms. A separate chapter is dedicated to research and development regarding deep geological disposal of radioactive waste. In the Belgian waste management programme, particular emphasis is on studies for disposal in clay. Main results of these studies are highlighted and discussed

  12. Radon emission from uranium mining waste rock dumps and resulting radon immission; Radonemissionsverhalten von Halden des Uranbergbaus und daraus resultierende Radonemissionen

    Energy Technology Data Exchange (ETDEWEB)

    Regner, J.; Hinz, W.; Schmidt, P. [Wismut GmbH, Chemnitz (Germany)

    2016-07-01

    Since more than 20 years, Wismut GmbH has been investigating the radon situation at uranium mining waste rock dumps. In the present paper the results of 19 complex studies at uranium mining dumps in the Erzgebirge (Ore Mountains) are reported. Although the mean specific activity of Ra-226 of the waste rock material was on a rather low level of about 0.5 Bq/g, the mean radon concentration in free atmosphere at the public exposure sites in the immediate vicinity of the dumps reached a value of about 1000 Bq/m{sup 3} for a half-year exposition and of about 600 Bq/m{sup 3} for a one-year exposition. Certain geometries and structures of waste rock dumps and the occurrence of convective airflows in the dumps are main reasons for the high radon emission despite of the relatively low specific Ra-226 activity. A case study for two buildings directly on the top of a waste rock dump in the town Johanngeorgenstadt is presented. The hypothetical interpolation of the results for Ra-226-activity to a value below the threshold value of 0.2 Bq/g leads to the assumption that problematic radon situations may also occur outside the areas of legacies of uranium mining. Considering the aspects mentioned, a clearance level for NORM of 1 Bq/g is questionable.

  13. Westinghouse Hanford Company effluent discharges and solid waste management report for calendar year 1989: 200/600 Areas

    International Nuclear Information System (INIS)

    Brown, M.J.; P'Pool, R.K.; Thomas, S.P.

    1990-05-01

    This report presents calendar year 1989 radiological and nonradiological effluent discharge data from facilities in the 200 Areas and the 600 Area of the Hanford Site. Both summary and detailed effluent data are presented. In addition, radioactive and nonradioactive solid waste storage and disposal data for calendar year 1989 are furnished. Where appropriate, comparisons to previous years are made. The intent of the report is to demonstrate compliance of Westinghouse Hanford Company-operated facilities with administrative control values for radioactive constituents and applicable guidelines and standards (including Federal permit limits) for nonradioactive constituents. 11 refs., 20 tabs

  14. 2005 dossier. ANDRA's researches on the geological disposal of high-level and long-lived radioactive wastes. Results and perspectives

    International Nuclear Information System (INIS)

    2005-06-01

    This document makes a status of the researches carried out by the French national agency of radioactive wastes (ANDRA) about the geologic disposal of high-level and long-lived radioactive wastes in deep geologic formations (argilites and granites). Content: 1 - Research on deep disposal of radioactive waste: general interest task: Legislative framework, ANDRA scientific objectives, Inspections and assessments; 2 - Designing a safe and reversible disposal system: Repository safety, Reversibility: an essential requirement; 3 - Clay Research on a repository in a clay formation, A long research programme, Dossier 2005 Argile; 4 - Meuse/Haute-Marne site clay: Expected properties of the rock formation, Choice of argillite, Meuse/Haute-Marne site, Conclusions from 10 years of research at the Meuse/Haute-Marne site; 5 - Repository installations: Safe and reversible architecture, Disposal of B waste, Disposal of C waste, Possible disposal of spent fuel (CU); 6 - The disposal facility in operation: From waste packages reception to their disposal in cells, Stages of the progressive closure of engineered structures; 7 - Reversible management: Freedom of choice for future generations, Various closure stages; 8 - Long-term evolution of the repository: Apprehending the repository complexity Main evolutions expected, Slow and limited release of radioactive substances; 9 - Repository safety and impact on man: Several evolution scenarios, Normal evolution, Altered evolution; 10 - Granite Research on a repository in a granite formation: A global approach, Scientific co-operations, Dossier 2005 Granite; 11 - Characteristics of French granite formations: What properties are required for a repository?, Different types of granite formations; 12 - Repository installations: Repository design adapted to granite fractures, Clay seals to prevent water flows, Waste disposal packages ensuring long-term leak-tightness, Physical and chemical environment favourable for waste packages, Architecture

  15. Change of carcinogenic chrysotile fibers in the asbestos cement (eternit) to harmless waste by artificial carbonatization: Petrological and technological results

    Energy Technology Data Exchange (ETDEWEB)

    Radvanec, Martin; Tuček, Ľubomír; Derco, Ján; Čechovská, Katarína [State Geological Institute of Dionýz Štúr, Mlynská dolina 1, SK-817 04 Bratislava (Slovakia); Németh, Zoltán, E-mail: zoltan.nemeth@geology.sk [State Geological Institute of Dionýz Štúr, Mlynská dolina 1, SK-817 04 Bratislava (Slovakia)

    2013-05-15

    Highlights: ► Carcinogenic chrysotile fibers in asbestos cement (eternit) are liquidated. ► Thermally modified eternit grist (at 650 °C, 1 h) reacts with CO{sub 2} + water. ► Carbonates hydromagnesite and magnesite are the newly formed products of artificial carbonatization. ► Neutralizing of extreme pH values (around 12) at large eternit dumps. ► An alternative methodology for permanent liquidation of a part of CO{sub 2} emissions. -- Abstract: Asbestos cement materials, mainly the eternit roof ceiling, being widely applied in the past, represent a serious environmental load. The solar radiation, rain and frost cause the deliberation of cement from the eternit roofing and consequently the wind contaminates the surrounding area by the asbestos (chrysotile) fibers. In combination with other carcinogens (e.g. smoking), or at reduced immunity of a man, they may cause serious respiratory diseases and lung cancer. The article presents the procedure and experimental results of artificial carbonatization, applied in the asbestos cement (eternit). The wet crushed and pulverized asbestos cement was thermally modified at 650 °C and then the chrysotile fibers easily and completely reacted with the mixture of CO{sub 2} and water, producing new Mg-rich carbonates – hydromagnesite and magnesite: 2Mg{sub 3}Si{sub 2}O{sub 5}(OH){sub 3thermally} {sub modified} {sub chrysotile}+5CO{sub 2}+nH{sub 2}O→Mg{sub 5}(CO{sub 3}){sub 4}(OH){sub 2}⋅4H{sub 2}O{sub hydromagnesite}+MgCO{sub 3magnesite}+4SiO{sub 2} · nH{sub 2}O{sub in} a{sub morphous} {sub phase};n=3÷9 Applying this methodology, the asbestos-bearing waste can be stabilized and environmentally friendly permanently deposited. Finding a way of neutralizing of extreme pH values (around 12) at large eternit dumps represents also an asset of presented research. Simultaneously, the artificial carbonatization of chrysotile asbestos, applying CO{sub 2}, offers an alternative way for permanent liquidation of a part of

  16. Change of carcinogenic chrysotile fibers in the asbestos cement (eternit) to harmless waste by artificial carbonatization: Petrological and technological results

    International Nuclear Information System (INIS)

    Radvanec, Martin; Tuček, Ľubomír; Derco, Ján; Čechovská, Katarína; Németh, Zoltán

    2013-01-01

    Highlights: ► Carcinogenic chrysotile fibers in asbestos cement (eternit) are liquidated. ► Thermally modified eternit grist (at 650 °C, 1 h) reacts with CO 2 + water. ► Carbonates hydromagnesite and magnesite are the newly formed products of artificial carbonatization. ► Neutralizing of extreme pH values (around 12) at large eternit dumps. ► An alternative methodology for permanent liquidation of a part of CO 2 emissions. -- Abstract: Asbestos cement materials, mainly the eternit roof ceiling, being widely applied in the past, represent a serious environmental load. The solar radiation, rain and frost cause the deliberation of cement from the eternit roofing and consequently the wind contaminates the surrounding area by the asbestos (chrysotile) fibers. In combination with other carcinogens (e.g. smoking), or at reduced immunity of a man, they may cause serious respiratory diseases and lung cancer. The article presents the procedure and experimental results of artificial carbonatization, applied in the asbestos cement (eternit). The wet crushed and pulverized asbestos cement was thermally modified at 650 °C and then the chrysotile fibers easily and completely reacted with the mixture of CO 2 and water, producing new Mg-rich carbonates – hydromagnesite and magnesite: 2Mg 3 Si 2 O 5 (OH) 3thermally modified chrysotile +5CO 2 +nH 2 O→Mg 5 (CO 3 ) 4 (OH) 2 ⋅4H 2 O hydromagnesite +MgCO 3magnesite +4SiO 2 · nH 2 O in a morphous phase ;n=3÷9 Applying this methodology, the asbestos-bearing waste can be stabilized and environmentally friendly permanently deposited. Finding a way of neutralizing of extreme pH values (around 12) at large eternit dumps represents also an asset of presented research. Simultaneously, the artificial carbonatization of chrysotile asbestos, applying CO 2 , offers an alternative way for permanent liquidation of a part of industrial CO 2 emissions, contributing to multiple benefit of this methodology

  17. Draft environmental impact statement. High-level waste repository site suitability criteria

    International Nuclear Information System (INIS)

    1978-01-01

    The purpose of HLWRSSC is to present guidelines which will help in the development of safe waste management schemes. Current regulations require solidification of all high-level waste within 5 years of their generation and transfer to a Federal waste repository within 10 years. Development of the proposed HLWRSSC is part of the overall NRC program to close the ''back end'' of the commercial LWR fuel cycle. In this document, the need for the HLWRSSC is reviewed, and the national energy policy, the need for electrical energy, and the nuclear fuel cycle are discussed. Considerations for HLWRSSC are presented, including the nature of the repository, important site-related factors, and radiological risk assessment methodology. Radiological and nonradiological environment impacts associated with the HLWRSSC are defined. Alternatives to the criteria are presented, and the cost-benefit-risk evaluation is reviewed

  18. Cost effectiveness of below-threshold waste disposal at DOE sites

    International Nuclear Information System (INIS)

    Wickham, L.E.; Smith, C.F.; Cohen, J.J.

    1986-01-01

    Previous study has indicated the feasibility of establishing a threshold of concentration below which certain low-level (radioactive wastes) (LLW) could be safely handled and disposed of by conventional means such as landfills. Such below-threshold wastes have been synonymously termed de minimis or below regulatory concern (BRC) and can be deemed appropriate for management according to their nonradiological characteristics. The objective of this study was to determine the cost effectiveness for management and disposal of below-threshold waste at certain US Department of Energy sites. The sites selected for this study were the Idaho National Engineering Laboratory and Savannah River Laboratory. Cost-benefit analysis was used to determine the impacts, benefits, and potential cost advantages of establishing and implementing a threshold limit

  19. Corrective Measures Study Modeling Results for the Southwest Plume - Burial Ground Complex/Mixed Waste Management Facility

    International Nuclear Information System (INIS)

    Harris, M.K.

    1999-01-01

    Groundwater modeling scenarios were performed to support the Corrective Measures Study and Interim Action Plan for the southwest plume of the Burial Ground Complex/Mixed Waste Management Facility. The modeling scenarios were designed to provide data for an economic analysis of alternatives, and subsequently evaluate the effectiveness of the selected remedial technologies for tritium reduction to Fourmile Branch. Modeling scenarios assessed include no action, vertical barriers, pump, treat, and reinject; and vertical recirculation wells

  20. Hydrogeologic studies at Yucca Mountain, Nevada, USA. An interpretation of results for radioactive waste disposal site characterization

    International Nuclear Information System (INIS)

    Dudley, W.W.

    1984-02-01

    Of nine potential nuclear-waste repository sites being investigated in the United States, Yucca Mountain is the only one for which disposal above the water table is proposed. The host rock is a fractured, permeable welded tuff more than 300 m beneath the surface. The principal factors contributing to the isolation of waste include: a small recharge flux, estimated to be about 5 mm/yr; free drainage in the host rock and little opportunity for contact of water with the waste; near-neutral water of low ionic and organic content; unsaturated-zone and saturated-zone flowpaths through altered tuffs that are rich in sorptive zeolites and clays; and very deep regional ground-water flow that terminates in a closed basin. Hydraulic testing of the saturated zone has demonstrated that fractures cause the observed high transmissivity, and seepage velocities in major fracture zones may be as high as 0.01 to 0.1 km/yr. Diffusion of radionuclides from water in fractures to that in the porous rock matrix, however, would attenuate their migration and allow sorptive processes to operate if a release from the repository were to occur. Psychrometers, heat-dissipation probes, pressure transducers, and sampling tubes that were recently installed in a 380-m drill hole are still undergoing stabilization. Data from this hole and other planned experiments will allow definition of recharge flux, frequency, and flowpaths for statistical treatment in models

  1. Results of Phase I groundwater quality assessment for single-shell tank waste management areas T and TX-TY at the Hanford Site

    International Nuclear Information System (INIS)

    Hodges, F.N.

    1998-01-01

    Pacific Northwest National Laboratory (PNNL) conducted a Phase I, Resource Conservation and Recovery Act of 1976 (RCRA) groundwater quality assessment for the Richland Field Office of the U.S. Department of Energy (DOE-RL) under the requirements of the Federal Facility Compliance Agreement. The purpose of the investigation was to determine if the Single-Shell Tank Waste Management Areas (WMAs) T and TX-TY have impacted groundwater quality. Waste Management Areas T and TX-TY, located in the northern part of the 200 West Area of the Hanford Site, contain the 241-T, 241-TX, and 241-TY tank farms and ancillary waste systems. These two units are regulated under RCRA interim-status regulations (under 40 CFR 265.93) and were placed in assessment groundwater monitoring because of elevated specific conductance in downgradient wells. Anomalous concentrations of technetium-99, chromium, nitrate, iodine-129, and cobalt-60 also were observed in some downgradient wells. Phase I assessment, allowed under 40 CFR 265, provides the owner-operator of a facility with the opportunity to show that the observed contamination has a source other than the regulated unit. For this Phase I assessment, PNNL evaluated available information on groundwater chemistry and past waste management practices in the vicinity of WMAs T and TX-TY. Background contaminant concentrations in the vicinity of WMAs T and TX-TY are the result of several overlapping contaminant plumes resulting from past-practice waste disposal operations. This background has been used as baseline for determining potential WMA impacts on groundwater

  2. Nutritional value content, biomass production and growth performance of Daphnia magna cultured with different animal wastes resulted from probiotic bacteria fermentation

    Science.gov (United States)

    Endar Herawati, Vivi; Nugroho, R. A.; Pinandoyo; Hutabarat, Johannes

    2017-02-01

    Media culture is an important factor for the growth and quality of Daphnia magna nutrient value. This study has purpose to find the increasing of nutritional content, biomass production and growth performance of D. magna using different animal wastes fermented by probiotic bacteria. This study conducted using completely randomized experimental design with 10 treatments and 3 replicates. Those media used different animal manures such as chicken manure, goat manure and quail manure mixed by rejected bread and tofu waste fermented by probiotic bacteria then cultured for 24 days. The results showed that the media which used 50% chicken manure, 100% rejected bread and 50% tofu waste created the highest biomass production, population and nutrition content of D.magna about 2111788.9 ind/L for population; 342 grams biomass production and 68.85% protein content. The highest fatty acid profile is 6.37% of linoleic and the highest essential amino acid is 22.8% of lysine. Generally, the content of ammonia, DO, temperature, and pH during the study were in the good range of D. magna’s life. This research has conclusion that media used 50% chicken manure, 100% rejected bread and 50% tofu waste created the highest biomass production, population and nutrition content of D. magna.

  3. results

    Directory of Open Access Journals (Sweden)

    Salabura Piotr

    2017-01-01

    Full Text Available HADES experiment at GSI is the only high precision experiment probing nuclear matter in the beam energy range of a few AGeV. Pion, proton and ion beams are used to study rare dielectron and strangeness probes to diagnose properties of strongly interacting matter in this energy regime. Selected results from p + A and A + A collisions are presented and discussed.

  4. Methods and results of the investigation of the thermomechanical behaviour of rock salt with regard to the final disposal of high-level radioactive wastes

    International Nuclear Information System (INIS)

    Wieczorek, K.; Klarr, K.

    1993-01-01

    This report summarizes the knowledge about thermal and mechanical behaviour of rock salt that has been accumulated by various R and D institutions in Germany from laboratory and in situ investigations. An important objective is to give a comprehensive overview of the investigation methods and instruments available and to discuss these methods and instruments with regard to their applicability and reliability for the investigation of the thermomechanical effects of high level radioactive waste emplacement in rock salt formations. The report is focused on the activities of the GSF-Institut fur Tieflagerung in the Asse mine regarding the disposal of high and intermediate level radioactive waste during the last decades. The design and the results of the most important in situ experiments are presented and discussed in detail. The results are compared to model calculations in order to evaluate the reliability of both the measurements and the calculation results. The relevance of the results for the situation in Spain is discussed in a separate chapter. As the investigations in Germany have been performed in domal salt, while the Spanish concept is based on waste disposal in bedded salt, significant differences in the thermomechanical behaviour cannot be excluded. The investigation methods, however, will be applicable. (Author)

  5. Chemical composition analysis and product consistency tests to support enhanced Hanford waste glass models. Results for the third set of high alumina outer layer matrix glasses

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K. M. [Savannah River Site (SRS), Aiken, SC (United States); Edwards, T. B. [Savannah River Site (SRS), Aiken, SC (United States)

    2015-12-01

    In this report, the Savannah River National Laboratory provides chemical analyses and Product Consistency Test (PCT) results for 14 simulated high level waste glasses fabricated by the Pacific Northwest National Laboratory. The results of these analyses will be used as part of efforts to revise or extend the validation regions of the current Hanford Waste Treatment and Immobilization Plant glass property models to cover a broader span of waste compositions. The measured chemical composition data are reported and compared with the targeted values for each component for each glass. All of the measured sums of oxides for the study glasses fell within the interval of 96.9 to 100.8 wt %, indicating recovery of all components. Comparisons of the targeted and measured chemical compositions showed that the measured values for the glasses met the targeted concentrations within 10% for those components present at more than 5 wt %. The PCT results were normalized to both the targeted and measured compositions of the study glasses. Several of the glasses exhibited increases in normalized concentrations (NCi) after the canister centerline cooled (CCC) heat treatment. Five of the glasses, after the CCC heat treatment, had NCB values that exceeded that of the Environmental Assessment (EA) benchmark glass. These results can be combined with additional characterization, including X-ray diffraction, to determine the cause of the higher release rates.

  6. Performance evaluation of the technical capabilities of DOE sites for disposal of mixed low-level waste. Volume 2: Technical basis and discussion of results

    International Nuclear Information System (INIS)

    Waters, R.D.; Gruebel, M.M.; Hospelhorn, M.B.

    1996-03-01

    A team of analysts designed and conducted a performance evaluation to estimate the technical capabilities of fifteen Department of Energy sites for disposal of mixed low-level waste (i.e., waste that contains both low-level radioactive materials and hazardous constituents). Volume 1 summarizes the process for selecting the fifteen sites, the methodology used in the evaluation, and the conclusions derived from the evaluation. Volume 2 first describes the screening process used to determine the sites to be considered in the PEs. This volume then provides the technical details of the methodology for conducting the performance evaluations. It also provides a comparison and analysis of the overall results for all sites that were evaluated. Volume 3 contains detailed evaluations of the fifteen sites and discussions of the results for each site

  7. Waste management

    International Nuclear Information System (INIS)

    Soule, H.F.

    1975-01-01

    Current planning for the management of radioactive wastes, with some emphasis on plutonium contaminated wastes, includes the provision of re-positories from which the waste can be safely removed to permanent disposal. A number of possibilities for permanent disposal are under investigation with the most favorable, at the present time, apparently disposal in a stable geological formation. However, final choice cannot be made until all studies are completed and a pilot phase demonstrates the adequacy of the chosen method. The radioactive wastes which result from all portions of the fuel cycle could comprise an important source of exposure to the public if permitted to do so. The objectives of the AEC waste management program are to provide methods of treating, handling and storing these wastes so that this exposure will not occur. This paper is intended to describe some of the problems and current progress of waste management programs, with emphasis on plutonium-contaminated wastes. Since the technology in this field is advancing at a rapid pace, the descriptions given can be regarded only as a snapshot at one point in time. (author)

  8. Processing of Irradiated Graphite to Meet Acceptance Criteria for Waste Disposal. Results of a Coordinated Research Project

    International Nuclear Information System (INIS)

    2016-05-01

    Graphite is widely used in the nuclear industry and in research facilities and this has led to increasing amounts of irradiated graphite residing in temporary storage facilities pending disposal. This publication arises from a coordinated research project (CRP) on the processing of irradiated graphite to meet acceptance criteria for waste disposal. It presents the findings of the CRP, the general conclusions and recommendations. The topics covered include, graphite management issues, characterization of irradiated graphite, processing and treatment, immobilization and disposal. Included on the attached CD-ROM are formal reports from the participants

  9. Effect of absorbing impurities on the accuracy of the optical method for the detection of the iodine-containing substances resulting from the processing of waste nuclear fuel

    Science.gov (United States)

    Kireev, S. V.; Simanovsky, I. G.; Shnyrev, S. L.

    2010-12-01

    The study is aimed at an increase in the accuracy of the optical method for the detection of the iodine-containing substances in technological liquids resulting form the processing of the waste nuclear fuel. It is demonstrated that the accuracy can be increased owing to the measurements at various combinations of wavelengths depending on the concentrations of impurities that are contained in the sample under study and absorb in the spectral range used for the detection of the iodine-containing substances.

  10. Tank Vapor Characterization Project: Headspace vapor characterization of Hanford waste tank 241-S-101: Results from samples collected on 06/06/96

    International Nuclear Information System (INIS)

    Thomas, B.L.; Evans, J.C.; Pool, K.H.; Olsen, K.B.; Fruchter, J.S.; Silvers, K.L.

    1997-01-01

    This report describes the analytical results of vapor samples taken from the headspace of the waste storage tank 241-S-101. The results described in this report were obtained to characterize the vapors present in the tank headspace and to support safety evaluations and tank farm operations. The results include air concentrations of selected inorganic and organic analytes and grouped compounds from samples obtained. Analyte concentrations were based on analytical results and sample volumes provided by WHC. A summary of the inorganic analytes, permanent gases, and total non-methane organic compounds is listed

  11. Food Waste Generation at Household Level: Results of a Survey among Employees of Two European Research Centers in Italy and Germany

    Directory of Open Access Journals (Sweden)

    Juliane Jörissen

    2015-03-01

    Full Text Available There is a broad consensus in literature that private households are significant contributors to the total amount of food waste in the EU. Thus, any strategy to meaningfully combat food wastage must put the end consumer in the center of prevention activities. This requires deeper insights into people’s motivations to discard still edible food and knowledge about potential barriers to reduce wasting. This paper reports on results of an online survey among two European research centers in Italy (JRC/Ispra and Germany (KIT/Karlsruhe. The focus of the survey was on households’ behaviors (shopping, eating, and food preparation habits and its influence on the generation of food waste. Furthermore, reasons for the disposal of food as well as measures and technologies most needed to prevent wastage were discussed. The results of the survey are analyzed, especially with regard to two questions: (1 Are there considerable differences between Ispra and Karlsruhe? (2 Are there considerable similarities or inconsistencies with the results of previous studies?

  12. Results of screening activities in salt states prior to the enactment of the Nationall Waste Policy Act

    International Nuclear Information System (INIS)

    Carbiener, W.A.

    1983-01-01

    The identification of potential sites for a nuclear waste repository through screening procedures in the salt states is a well-established, deliberate process. This screening process has made it possible to carry out detailed studies of many of the most promising potential sites, and general studies of all the sites, in anticipation of the siting guidelines specified in the Nuclear Waste Policy Act. The screening work completed prior to the passage of the Act allowed the Secretary of Energy to identify seven salt sites as potentially acceptable under the provisions of Section 116(a) of the Act. These sites were formally identified by letters from Secretary Hodel to the states of Texas, Utah, Mississippi, and Louisiana on February 2, 1983. The potentially acceptable salt sites were in Deaf Smith and Swisher Counties in Texas; Davis and Lavender Canyons in the Gibson Dome location in Utah; Richton and Cypress Creek Domes in Mississippi; and Vacherie Dome in Louisiana. Further screening will include comparison of each potentially acceptable site against disqualification factors and selection of a preferred site in each of the three geohydrologic settings from those remaining, in accordance with the siting guidelines. These steps will be documented in statutory Environmental Assessments prepared for each site to be nominated for detailed characterization. 9 references

  13. Results From an International Simulation Study on Coupled Thermal, Hydrological, and Mechanical (THM) Processes Near Geological Nuclear Waste Repositories

    International Nuclear Information System (INIS)

    J. Rutqvist; D. Barr; J.T. Birkholzer; M. Chijimatsu; O. Kolditz; Q. Liu; Y. Oda; W. Wang; C. Zhang

    2006-01-01

    As part of the ongoing international DECOVALEX project, four research teams used five different models to simulate coupled thermal, hydrological, and mechanical (THM) processes near waste emplacement drifts of geological nuclear waste repositories. The simulations were conducted for two generic repository types, one with open and the other with back-filled repository drifts, under higher and lower postclosure temperatures, respectively. In the completed first model inception phase of the project, a good agreement was achieved between the research teams in calculating THM responses for both repository types, although some disagreement in hydrological responses is currently being resolved. In particular, good agreement in the basic thermal-mechanical responses was achieved for both repository types, even though some teams used relatively simplified thermal-elastic heat-conduction models that neglected complex near-field thermal-hydrological processes. The good agreement between the complex and simplified process models indicates that the basic thermal-mechanical responses can be predicted with a relatively high confidence level

  14. Annual report, spring 2015. Alternative chemical cleaning methods for high level waste tanks-corrosion test results

    Energy Technology Data Exchange (ETDEWEB)

    Wyrwas, R. B. [Savannah River Site (SRS), Aiken, SC (United States)

    2015-07-06

    The testing presented in this report is in support of the investigation of the Alternative Chemical Cleaning program to aid in developing strategies and technologies to chemically clean radioactive High Level Waste tanks prior to tank closure. The data and conclusions presented here were the examination of the corrosion rates of A285 carbon steel and 304L stainless steel when interacted with the chemical cleaning solution composed of 0.18 M nitric acid and 0.5 wt. % oxalic acid. This solution has been proposed as a dissolution solution that would be used to remove the remaining hard heel portion of the sludge in the waste tanks. This solution was combined with the HM and PUREX simulated sludge with dilution ratios that represent the bulk oxalic cleaning process (20:1 ratio, acid solution to simulant) and the cumulative volume associated with multiple acid strikes (50:1 ratio). The testing was conducted over 28 days at 50°C and deployed two methods to invest the corrosion conditions; passive weight loss coupon and an active electrochemical probe were used to collect data on the corrosion rate and material performance. In addition to investigating the chemical cleaning solutions, electrochemical corrosion testing was performed on acidic and basic solutions containing sodium permanganate at room temperature to explore the corrosion impacts if these solutions were to be implemented to retrieve remaining actinides that are currently in the sludge of the tank.

  15. Qualification and characterization programmes for disposal of a glass product resulting from high level waste vitrification in the PAMELA installation of BELGOPROCESS

    International Nuclear Information System (INIS)

    Goeyse, A. de; De, A.K.; Demonie, M.; Iseghem, P. van

    1993-01-01

    In the framework of a general quality assurance and quality control (QA/QC) programme, the quality of a conditioned waste product is achieved in two phases. The first phase is the design of a process and facility which will ensure the required quality of the product. In the second phase the conformance of the product with the preset requirements is verified. NIRAS/ONDRAF, as the agency responsible for the management of all radioactive waste in Belgium (including treatment, conditioning, storage and disposal), controls compliance with the quality requirements during both phases. The purpose of the paper is to describe the different phases of this general procedure in the case of a vitrified HLW product resulting from a vitrification campaign in the PAMELA facility at the BELGOPROCESS site. The active glass product of type SM527 produced during the vitrification of highly enriched waste concentrate (HEWC) (resulting from the reprocessing of highly enriched uranium fuel) has been selected for illustration. During the process qualification phase, the Deutsche Gesellschaft fuer Wiederaufarbeitung von Kernbrennstoffen mbH, responsible for the development of the vitrification process of PAMELA, defined and performed and R and D programmed for each glass product originating from the vitrification of the different HEWC solutions stored at the BELGOPROCESS site. At the end of this qualification phase a data catalogue was prepared. In order to ensure that the active glass product corresponds with the selected product from the data catalogue, the QA/QC handbook for the vitrification process describes all measures to be taken by the waste producer, BELGOPROCESS, during the vitrification. Finally, verification analyses are performed by the characterization of inactive and active samples by an independent laboratory. This phase is called the product quality verification phase. The details of the characterization programmes performed during the different phases and their results

  16. The long-term behaviour of cemented research reactor waste under the geological disposal conditions of the Boom Clay Formation: results from leach experiments

    International Nuclear Information System (INIS)

    Sneyers, A.; Fays, J.; Iseghem, P. van

    2001-01-01

    The Belgian Nuclear Research Centre SCK-CEN has carried out a number of studies to evaluate the long-term behaviour of cemented research reactor waste under the geological disposal conditions of the Boom Clay Formation. Static leach experiments in synthetic clay water were performed on active samples of cemented research reactor waste. The leach experiments were carried out under anaerobic conditions at two testing temperatures (23 and 85 o C). Leach rates of seven radionuclides ( 60 Co, 90 Sr, 134 Cs, 137 Cs, 144 Ce, 154 Eu and 241 Am) were measured. Most investigated radionuclides are well retained within the cement matrix over a 280 days testing period. Results on the source term of radionuclides were complemented with data on the leaching behaviour of cement matrix constituents as Ca, Si, Al, Na, K, Mg and SO 4 as well as with data from performance assessment calculations and in situ tests. Despite limitations inherent to short-term experiments, combined results from these investigations indicate only limited interactions of disposed research reactor waste with the near field of a geological repository in clay. (author)

  17. Rice Husk Ash to Stabilize Heavy Metals Contained in Municipal Solid Waste Incineration Fly Ash: First Results by Applying New Pre-treatment Technology

    Directory of Open Access Journals (Sweden)

    Laura Benassi

    2015-10-01

    Full Text Available A new technology was recently developed for municipal solid waste incineration (MSWI fly ash stabilization, based on the employment of all waste and byproduct materials. In particular, the proposed method is based on the use of amorphous silica contained in rice husk ash (RHA, an agricultural byproduct material (COSMOS-RICE project. The obtained final inert can be applied in several applications to produce “green composites”. In this work, for the first time, a process for pre-treatment of rice husk, before its use in the stabilization of heavy metals, based on the employment of Instant Pressure Drop technology (DIC was tested. The aim of this work is to verify the influence of the pre-treatment on the efficiency on heavy metals stabilization in the COSMOS-RICE technology. DIC technique is based on a thermomechanical effect induced by an abrupt transition from high steam pressure to a vacuum, to produce changes in the material. Two different DIC pre-treatments were selected and thermal annealing at different temperatures were performed on rice husk. The resulting RHAs were employed to obtain COSMOS-RICE samples, and the stabilization procedure was tested on the MSWI fly ash. In the frame of this work, some thermal treatments were also realized in O2-limiting conditions, to test the effect of charcoal obtained from RHA on the stabilization procedure. The results of this work show that the application of DIC technology into existing treatment cycles of some waste materials should be investigated in more details to offer the possibility to stabilize and reuse waste.

  18. Rice Husk Ash to Stabilize Heavy Metals Contained in Municipal Solid Waste Incineration Fly Ash: First Results by Applying New Pre-treatment Technology

    Science.gov (United States)

    Benassi, Laura; Franchi, Federica; Catina, Daniele; Cioffi, Flavio; Rodella, Nicola; Borgese, Laura; Pasquali, Michela; Depero, Laura E.; Bontempi, Elza

    2015-01-01

    A new technology was recently developed for municipal solid waste incineration (MSWI) fly ash stabilization, based on the employment of all waste and byproduct materials. In particular, the proposed method is based on the use of amorphous silica contained in rice husk ash (RHA), an agricultural byproduct material (COSMOS-RICE project). The obtained final inert can be applied in several applications to produce “green composites”. In this work, for the first time, a process for pre-treatment of rice husk, before its use in the stabilization of heavy metals, based on the employment of Instant Pressure Drop technology (DIC) was tested. The aim of this work is to verify the influence of the pre-treatment on the efficiency on heavy metals stabilization in the COSMOS-RICE technology. DIC technique is based on a thermomechanical effect induced by an abrupt transition from high steam pressure to a vacuum, to produce changes in the material. Two different DIC pre-treatments were selected and thermal annealing at different temperatures were performed on rice husk. The resulting RHAs were employed to obtain COSMOS-RICE samples, and the stabilization procedure was tested on the MSWI fly ash. In the frame of this work, some thermal treatments were also realized in O2-limiting conditions, to test the effect of charcoal obtained from RHA on the stabilization procedure. The results of this work show that the application of DIC technology into existing treatment cycles of some waste materials should be investigated in more details to offer the possibility to stabilize and reuse waste. PMID:28793605

  19. Results of the IAEA CRP on studies of advanced reactor technology options for effective incineration of radioactive waste

    International Nuclear Information System (INIS)

    Maschek, W.; Stanculescu, A.; ); Gopalakrishnan, V.

    2007-01-01

    The IAEA has initiated a Coordinated Research Project (CRP) on 'Studies of Advanced Reactor Technology Options for Effective Incineration of Radioactive Waste'. The overall objective of the CRP, performed within the framework of IAEA's Nuclear Power Technology Development Section's Technical Working Group on Fast Reactors (TWG-FR), is to increase the capability of Member States in developing and applying advanced technologies in the area of long-lived radioactive waste utilization and transmutation. More specifically, the final goal of the CRP is to deepen the understanding of the dynamics of transmutation systems, especially systems with high minor actinide content. Currently, 20 institutions from 15 member states and one international organization are participating in this CRP. The current author list comprises the participants of the last CRP Vienna meeting. The CRP concentrates on the assessment of the transient behaviour of various transmutation systems. For a sound assessment of the transient and accident behaviour, neutron kinetics and dynamics methods and codes have to be qualified, especially as the margins for the safety relevant neutronics parameters are generally becoming small in a transmutation system. Hence, the availability of adequate and qualified methods for the analysis of the various systems is an important point of the exercise. A benchmarking effort between the codes and nuclear data used for the analyses has been performed, which will help specifying the range of validity of methods, and also formulate requirements for future theoretical and experimental research. Should transient experiments become available during the course of the CRP, experimental benchmarking work will also be pursued

  20. Fiscal 1999 report on result of the model project for waste heat recovery in hot blast stove; 1999 nendo netsufuro hainetsu kaishu model jigyo seika hokokusho

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-07-01

    For the purpose of curtailing energy consumption of the steel industry, a heavy energy consuming industry in China, a model project was carried out for waste heat recovery in a hot blast stove, with the fiscal 1999 results reported. In the process of this project, a heat exchanger for recovering heat is installed in the exhaust gas flue of a hot blast stove in ironworks, with sensible heat recovered through a heating medium. The heat exchanger for recovering heat and the preheating heat exchanger, which was installed in the main pipe for blast furnace gas and for combustion air, were connected by pressure piping, with the blast furnace gas and the combustion air preheated. In addition, a heating medium circulating pump for transporting the heating medium is installed, as are an expansion tank for absorbing expansion/contraction due to change in temperature, a heating medium storage tank for accepting the entire heating medium in the system for the maintenance of the equipment, and heating medium feeding pump, for example. This year, on the basis of the 'Agreement Annex', basic designs and detailed designs were performed for each equipment in the waste heat recovering equipment for the hot blast stove. Further, procurement and manufacturing were implemented for various component parts and devices of the waste heat recovering equipment. (NEDO)

  1. Radioactive wastes

    International Nuclear Information System (INIS)

    Dupuis, M.C.

    2007-01-01

    Managing radioactive wastes used to be a peripheral activity for the French atomic energy commission (Cea). Over the past 40 years, it has become a full-fledged phase in the fuel cycle of producing electricity from the atom. In 2005, the national radioactive waste management agency (ANDRA) presented to the government a comprehensive overview of the results drawn from 15 years of research. This landmark report has received recognition beyond France's borders. By broadening this agency's powers, an act of 28 June 2006 acknowledges the progress made and the quality of the results. It also sets an objective for the coming years: work out solutions for managing all forms of radioactive wastes. The possibility of recovering wastes packages from the disposal site must be assured as it was asked by the government in 1998. The next step will be the official demand for the creation of a geological disposal site in 2016

  2. Analysis of the total system life cycle cost for the Civilian Radioactive Waste Management Program: Volume 1, The analysis and its results

    International Nuclear Information System (INIS)

    1987-06-01

    This report provides cost estimates for the fifth evaluation of the adequacy of the fee and is consistent with the program strategy and plans. The total-system cost for the reference cases in the improved-performance system is estimated at $32.1 to $38.2 billion (expressed in constant 1986 dollars) over the entire life of the system...or $1.5 to $1.6 billion more than that of the authorized system (i.e., the system without an MRS facility). The current estimate of the total-system cost for the reference cases in the improved-performance system is $3.8 to $5.4 billion higher than the estimate for the same system in the 1986 TSLCC analysis. In the case with the maximum increase, nearly all of the higher cost is due to a $5.2-billion increase in the costs of development and evaluation (D and E); all other system costs are essentially unchanged. The cost difference between the improved-performance system and the authorized system is smaller than the difference estimated in last year's TSLCC analysis. Volume 2 presents the detailed results for the 1987 analysis of the total-system life cycle cost (TSLCC). It consists of four sections: Section A presents the yearly flows of waste between waste-management facilities for the 12 aggregate logistics cases that were studied; Section B presents the annual total-system costs for each of the 30 TSLCC cases by major cost category; Section C presents the annual costs for the disposal of 16,000 canisters of defense high-level waste (DHLW) by major cost category for each of the 30 TSLCC cases; and Section D presents a summary of the cost-allocation factors that were calculated to determine the defense waste share of the total-system costs

  3. Vapor space characterization of Waste Tank 241-C-103: Inorganic results from sample Job 7B (May 12-25, 1994)

    International Nuclear Information System (INIS)

    Ligotke, M.W.; Pool, K.H.; Lerner, B.D.

    1994-10-01

    This report is to provide analytical results for use in safety and toxicological evaluations of the vapor space of Hanford single-shell waste storage tanks C-103. Samples were analysed to determine concentrations of ammonia, nitric oxide, nitrogen dioxide, sulfur oxides, and hydrogen cyanide. In addition to the samples, controls were analyzed that included blanks, spiked blanks, and spiked samples. These controls provided information about the suitability of sampling and analytical methods. Also included are the following: information describing the methods and sampling procedures used; results of sample analyses; and Conclusions and recommendations

  4. Reflecting socio-technical combinations in radioactive waste management. Results from the InSOTEC European research project

    International Nuclear Information System (INIS)

    Kallenbach-Herbert, Beate; Bergmans, Anne; Martell, Meritxell; Schroeder, Jantine

    2015-01-01

    InSOTEC is a three-year collaborative social sciences research project funded under the European Atomic Energy Community's 7th Framework Programme FP7. The project aims to generate a better understanding of the complex interplay between the technical and the social in the context of geological disposal of radioactive waste. In doing so, InSOTEC has moved beyond the social and technical division that is frequently being found in this context by - investigating the consideration of social sciences and the recognition of socio-technical combinations in research programs on geological disposal, - analyzing the socio-technical entanglement in selected contexts like siting, reversibility and retrievability, demonstrating safety and technology transfer on the basis of case studies, and - exploring the integration of diverse stakeholders in technology oriented networks. The analyses reveal that activities in the context of geological disposal, whether related to research, planning, siting etc., rather support the divide of social and technical aspects than fostering the consideration of their entanglement. Reasons identified for this are manifold. The wish to reduce complexity by focusing stakeholder involvement on social questions and fixing the technical part ''when acceptance is reached'' is only one of them. However, the analyses also show that over the long timescales of repository planning and implementation, robust management strategies must provide the flexibility to adapt to both technical and social developments and demands. Understanding the socio-technical interplay and creating structures for its consideration provides the basis for dealing with this challenge. This presentation will focus on the main findings of the InSOTEC project with regard to the consideration of socio-technical combinations in practice. These insights are currently under development and will be finalized at the end of the project in June 2014. We will reflect on

  5. EDF operational experience of primary circuit filter usage. Analysis of results and strategy for optimizing filtration and reducing solid wastes

    International Nuclear Information System (INIS)

    Mascarenhas, Darren; Moleiro, Edgar; Bancelin, Estelle; Bretelle, Jean-Luc

    2014-01-01

    Pleated fibreglass media filter cartridges are used throughout the auxiliary systems at nuclear power plants across the 58 reactors of EDF fleet. The main role of these filters is to remove suspended solids from coolant to prevent them accumulating in circuits or in equipments. In the primary circuit, these filters therefore limit the deposition of solids that are active or could become active if allowed to recirculate throughout the primary circuit, avoiding potential consequences such as an increase in dose rates, axial offset anomalies, demineralisers fouling, higher pressure losses in primary loop, and clogging of the primary pumps. Since 2008, a steady increase in the consumption of filters has been noticed, and therefore an increase in the amount of solid waste to treat. Preliminary studies have identified the primary circuit high-flow filters of the 1300/1450 MWe reactors as the main source of this increase. Not only has this stretched of solid waste containers production to the limit, as well as strained site resources and increased risks of operational errors during periods of frequent filter changes; it has also suggested that there is an underlying problem that could pose a serious risk to the primary circuit if untreated. Further studies have been carried out to identify more precisely the impact of possible causes, including increased quality surveillance of the filters, correlation of consumption data with the concentrations of various conditioning products and typical pollutants, and an impact analysis of events such as steam generator replacements or new practices like zinc injection. Work has been done with the filter manufacturer to improve their service lifetime and a simulation tool has been developed in order to understand and optimise filtration. We are also working with sites on creating good practices and avoiding bad ones. These actions should reduce the consumption in the short term while still assuring a high quality of filtration and

  6. Reflecting socio-technical combinations in radioactive waste management. Results from the InSOTEC European research project

    Energy Technology Data Exchange (ETDEWEB)

    Kallenbach-Herbert, Beate [Oeko-Institut e.V., Darmstadt (Germany); Bergmans, Anne [Antwerp Univ. (Belgium); Martell, Meritxell [Merience Strategic Thinking, Olerdola (Spain); Schroeder, Jantine [Antwerp Univ. (Belgium); SCK - CEN, Mol (Belgium)

    2015-07-01

    InSOTEC is a three-year collaborative social sciences research project funded under the European Atomic Energy Community's 7th Framework Programme FP7. The project aims to generate a better understanding of the complex interplay between the technical and the social in the context of geological disposal of radioactive waste. In doing so, InSOTEC has moved beyond the social and technical division that is frequently being found in this context by - investigating the consideration of social sciences and the recognition of socio-technical combinations in research programs on geological disposal, - analyzing the socio-technical entanglement in selected contexts like siting, reversibility and retrievability, demonstrating safety and technology transfer on the basis of case studies, and - exploring the integration of diverse stakeholders in technology oriented networks. The analyses reveal that activities in the context of geological disposal, whether related to research, planning, siting etc., rather support the divide of social and technical aspects than fostering the consideration of their entanglement. Reasons identified for this are manifold. The wish to reduce complexity by focusing stakeholder involvement on social questions and fixing the technical part ''when acceptance is reached'' is only one of them. However, the analyses also show that over the long timescales of repository planning and implementation, robust management strategies must provide the flexibility to adapt to both technical and social developments and demands. Understanding the socio-technical interplay and creating structures for its consideration provides the basis for dealing with this challenge. This presentation will focus on the main findings of the InSOTEC project with regard to the consideration of socio-technical combinations in practice. These insights are currently under development and will be finalized at the end of the project in June 2014. We will reflect on

  7. Radioactive mixed waste disposal

    International Nuclear Information System (INIS)

    Jasen, W.G.; Erpenbeck, E.G.

    1993-02-01

    Various types of waste have been generated during the 50-year history of the Hanford Site. Regulatory changes in the last 20 years have provided the emphasis for better management of these wastes. Interpretations of the Atomic Energy Act of 1954 (AEA), the Resource Conservation and Recovery Act of 1976 (RCRA), and the Hazardous and Solid Waste Amendments (HSWA) have led to the definition of radioactive mixed wastes (RMW). The radioactive and hazardous properties of these wastes have resulted in the initiation of special projects for the management of these wastes. Other solid wastes at the Hanford Site include low-level wastes, transuranic (TRU), and nonradioactive hazardous wastes. This paper describes a system for the treatment, storage, and disposal (TSD) of solid radioactive waste

  8. Disposal of waste from the cleanup of large areas contaminated as a result of a nuclear accident

    International Nuclear Information System (INIS)

    1992-01-01

    The report provides an overview of the methodology and technology available to load, transport and dispose of large volumes of contaminated material arising from the cleanup of areas after a nuclear accident and includes data on the planning, implementation, management and costing of such activities. To demonstrate the use of this information, three cleanup and disposal scenarios are examined, ranging from disposal in many small mounds or trenches within the contaminated area to disposal in a large facility away from the plant. As in the two companion reports, it is assumed that the population has been evacuated from the affected area. The report reviews the generic types of low level radioactive waste which are likely to arise from such a cleanup. The report does not deal with the recovery and disposal of intermediate and high level radioactive material on or near the plant site. This material will have to be recovered, packaged, transported and stored on-site or disposed of at an appropriate facility. These operations should be done by specialist teams using shielded or remotely operated equipment. Also not included are methods of in situ stabilization of contamination, for example ploughing to bury the top contaminated layer at a suitable depth. These techniques, which are likely to be widely used in part of the evacuated are, are discussed in IAEA Technical Reports Series No. 300, Vienna, 1989. 50 refs, 18 figs, 4 tabs

  9. Results from an International Simulation Study on Couples Thermal, Hydrological, and Mechanical (THM) Processes Near Geological Nuclear Waste Repositories

    International Nuclear Information System (INIS)

    J. Rutqvist; J.T. Birkholzer; M. Chijimatsu; O. Kolditz; Q.S. Liu; Y. Oda; W. Wang; C.Y. Zhang

    2006-01-01

    As part of the ongoing international code comparison project DECOVALEX, four research teams used five different models to simulate coupled thermal, hydrological, and mechanical (THM) processes near underground waste emplacement drifts. The simulations were conducted for two generic repository types with open or back-filled repository drifts under higher and lower post-closure temperature, respectively. In the completed first model inception phase of the project, a good agreement was achieved between the research teams in calculating THM responses for both repository types, although some disagreement in hydrological responses are currently being resolved. Good agreement in the basic thermal-mechanical responses was achieved for both repository types, even with some teams using relatively simplified thermal-elastic heat-conduction models that neglect complex near-field thermal-hydrological processes. The good agreement between the complex and simplified (and well-known) process models indicates that the basic thermal-mechanical responses can be predicted with a relatively high confidence level. The research teams have now moved on to the second phase of the project, the analysis of THM-induced permanent (irreversible) changes and the impact of those changes on the fluid flow field near an emplacement drift

  10. Results of geophysical surveys of glacial deposits near a former waste-disposal site, Nashua, New Hampshire

    Science.gov (United States)

    Ayotte, Joseph D.; Dorgan, Tracy H.

    1995-01-01

    Geophysical investigations were done near a former waste-disposal site in Nashua, New Hampshire to determine the thickness and infer hydraulic characteristics of the glacial sediments that underlie the area. Approximately 5 miles of ground- penetrating radar (GPR) data were collected in the study area by use of dual-80 Megahertz antennas. Three distinct radar-reflection signatures were evident from the data and are interpreted to represent (1) glacial lake-bottom sediments, (2) coarse sand and gravel and (or) sandy glacial till, and (3) bedrock. The GPR signal penetrated as much as 70 feet of sediment in coarse-grained areas, but penetration depth was generally less than 40 feet in extensive areas of fine-grained deposits. Geologic features were evident in many of the profiles. Glacial-lake-bottom sediments were the most common features identified. Other features include deltas deposited in glacial Lake Nashua and lobate fans of sediment deposited subaqueously at the distal end of deltaic sediments. Cross-bedded sands were often identifiable in the deltaic sediments. Seismic-refraction data were also collected at five of the GPR data sites. In most cases, depths to the water table and to the till and (or) bedrock surface indicated by the seismic-refraction data compared favorably with depths calculated from the GPR data. Test holes were drilled at three locations to determine the true depths to radar reflectors and to determine the types of geologic material represented by the various reflectors.

  11. Waste Isolation Pilot Plant site environmental report for calendar year 1990

    International Nuclear Information System (INIS)

    1990-01-01

    The US Department of Energy (DOE) Waste Isolation Pilot Plant (WIPP) Operational Environmental Monitoring Plan (OEMP) monitors a comprehensive set of parameters in order to detect any potential environmental impacts and establish baselines for future quantitative environmental impact evaluations. Surface water and groundwater, soil, and biotics are measured for background radiation. Nonradiological environmental monitoring activities include meteorological, air quality, soil properties, and the status of the local biological community. Ecological studies focus on the immediate area surrounding the site with emphasis on the salt storage pile, whereas baseline radiological surveillance covers a broader geographic area including nearby ranches, villages, and cities. Since the WIPP is still in a preoperational state, no waste has been received; therefore, certain elements required by Order DOE 5400.1 are not presented in this report. 15 figs. 19 tabs

  12. Waste Isolation Pilot Plant site environmental report for calendar year 1990

    Energy Technology Data Exchange (ETDEWEB)

    1990-01-01

    The US Department of Energy (DOE) Waste Isolation Pilot Plant (WIPP) Operational Environmental Monitoring Plan (OEMP) monitors a comprehensive set of parameters in order to detect any potential environmental impacts and establish baselines for future quantitative environmental impact evaluations. Surface water and groundwater, soil, and biotics are measured for background radiation. Nonradiological environmental monitoring activities include meteorological, air quality, soil properties, and the status of the local biological community. Ecological studies focus on the immediate area surrounding the site with emphasis on the salt storage pile, whereas baseline radiological surveillance covers a broader geographic area including nearby ranches, villages, and cities. Since the WIPP is still in a preoperational state, no waste has been received; therefore, certain elements required by Order DOE 5400.1 are not presented in this report. 15 figs. 19 tabs.

  13. Tank Vapor Characterization Project: Headspace vapor characterization of Hanford Waste Tank U-203, Results from samples collected on August 8, 1995

    International Nuclear Information System (INIS)

    Pool, K.H.; Clauss, T.W.; Evans, J.C.; McVeety, B.D.; Thomas, B.L.; Olsen, K.B.; Fruchter, J.S.; Ligotke, M.W.

    1995-11-01

    This report describes the analytical results of vapor samples taken from the headspace of the waste storage tank 241-U-203 (Tank U-203) at the Hanford Site in Washington State. The results described in this report were obtained to characterize the vapors present in the tank headspace and to support safety evaluations and tank-farm operations. The results include air concentrations of selected inorganic and organic analytes and grouped compounds from samples obtained by Westinghouse Hanford Company (WHC) and provided for analysis to Pacific Northwest Laboratory (PNL). Analyses were performed by the Vapor Analytical Laboratory (VAL) at PNL. Analyte concentrations were based on analytical results and, where appropriate, sample volumes provided by WHC. A summary of the results is listed. Detailed descriptions of the analytical results appear in the text

  14. Tank Vapor Characterization Project: Headspace vapor characterization of Hanford Waste Tank U-204, Results from samples collected on August 8, 1995

    International Nuclear Information System (INIS)

    Clauss, T.W.; Evans, J.C.; McVeety, B.D.; Pool, K.H.; Thomas, B.L.; Olsen, K.B.; Fruchter, J.S.; Ligotke, M.W.

    1995-11-01

    This report describes the analytical results of vapor samples taken from the headspace of the waste storage tank 241-U-204 (Tank U-204) at the Hanford Site in Washington State. The results described in this report were obtained to characterize the vapors present in the tank headspace and to support safety evaluations and tank-farm operations. The results include air concentrations of selected inorganic and organic analytes and grouped compounds from samples obtained by Westinghouse Hanford Company (WHC) and provided for analysis to Pacific Northwest National Laboratory (PNL). Analyses were performed by the Vapor Analytical Laboratory (VAL) at PNL. Analyte concentrations were based on analytical results and, where appropriate, sample volumes provided by WHC. A summary of the results is listed. Detailed descriptions of the analytical results appear in the text

  15. Non-radiological contrast agents (MRI)

    International Nuclear Information System (INIS)

    Bonnemain, B.; Lautrou, J.; Meyer, D.; Doucet, D.

    1987-01-01

    Over the past few years, extensive research has been carried out in an attempt to develop contrast agents that could help improve both the performance (acquisition times) and the diagnostic efficacy of Magnetic Resonance Imaging (MRI) techniques. On the basis of physicochemical and pharmacological criteria discussed in this presentation, a few efficacious, well-tolerated compounds could be developed. Two of them, the gadolinium complexes Gd-DOTA and Gd-DTPA, are currently being tried in man. This first generation of contrast agents, which are aspecific markers of the intravascular space, has been shown to have good diagnostic potential in conventional MRI procedures. The diagnostic contribution of these contrast agents will probably be a most essential factor in new MRI techniques using low field strengh or fast imaging sequences [fr

  16. Tank Vapor Characterization Project. Headspace vapor characterization of Hanford Waste Tank AX-102: Results from samples collected on June 27, 1995

    International Nuclear Information System (INIS)

    Clauss, T.W.; Pool, K.H.; Evans, J.C.; McVeety, B.D.; Thomas, B.L.; Olsen, K.B.; Fruchter, J.S.; Ligotke, M.W.

    1995-11-01

    This report describes the analytical results of vapor samples taken from the headspace of the waste storage tank 241-AX-102 (Tank AX-102) at the Hanford Site in Washington State. The results described in this report were obtained to characterize the vapors present in the tank headspace and to support safety evaluations and tank-farm operations. The results include air concentrations of selected inorganic and organic analytes and grouped compounds from samples obtained by Westinghouse Hanford Company (WHC) and provided for analysis to Pacific Northwest Laboratory (PNL). Analyses were performed by the Vapor Analytical Laboratory (VAL) at PNL. Analyte concentrations were based on analytical results and, where appropriate, sample volumes provided by WHC. Detailed descriptions of the analytical results appear in the text

  17. Waste Isolation Pilot Plant site environmental report, for calendar year 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

    The U.S. Department of Energy (DOE) Order 5400.1 General Environmental Protection Program, requires DOE facilities, that conduct environmental protection programs, to annually prepare a Site Environmental Report (SER). The purpose of the SER is to provide an abstract of environmental assessments conducted in order to characterize site environmental management performance, to confirm compliance with environmental standards and requirements, and to highlight significant programs and efforts of environmental merit. The content of this SER is not restricted to a synopsis of the required data, in addition, information pertaining to new and continued monitoring and compliance activities during the 1995 calendar year are also included. Data contained in this report are derived from those monitoring programs directed by the Waste Isolation Pilot Plant (WIPP) Environmental Monitoring Plan (EMP). The EMP provides inclusive guidelines implemented to detect potential impacts to the environment and to establish baseline measurements for future environmental evaluations. Surface water, groundwater. air, soil, and biotic matrices are monitored for an array of radiological and nonradiological factors. The baseline radiological surveillance program encompasses a broader geographic area that includes nearby ranches, villages, and cities. Most elements of nonradiological assessments are conducted within the geographic vicinity of the WIPP site.

  18. Waste Isolation Pilot Plant site environmental report for calendar year 1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-06-01

    US Department of Energy (DOE) Order 5400.1 General Environmental Protection Program, requires each DOE facility that conducts significant environmental protection programs to prepare an Annual Site Environmental Report (ASER). The purpose of the ASER is to summarize environmental data in order to characterize site environmental management performance, to confirm compliance with environmental standards and requirements, and to highlight significant programs and efforts. This ASER not only documents the required data, it also documents new and continued monitoring and compliance activities during the 1994 calendar year. Data contained in this report are derived from those monitoring programs directed by the Waste Isolation Pilot Plant (WIPP) Environmental Monitoring Plan (EMP) (DOE/WIPP 94-024). The EMP defines a comprehensive set of parameters that must be monitored to detect potential impacts to the environment and to establish baseline measurements for future environmental evaluations. Surface water, groundwater, air, soil, and biotics are monitored for radiological and nonradiological activity levels. The baseline radiological surveillance program covers the broader geographic area that encompasses nearby ranches, villages, and cities. Nonradiological studies focus on the area immediately surrounding the WIPP site.

  19. Waste Isolation Pilot Plant site environmental report for calendar year 1994

    International Nuclear Information System (INIS)

    1995-06-01

    US Department of Energy (DOE) Order 5400.1 General Environmental Protection Program, requires each DOE facility that conducts significant environmental protection programs to prepare an Annual Site Environmental Report (ASER). The purpose of the ASER is to summarize environmental data in order to characterize site environmental management performance, to confirm compliance with environmental standards and requirements, and to highlight significant programs and efforts. This ASER not only documents the required data, it also documents new and continued monitoring and compliance activities during the 1994 calendar year. Data contained in this report are derived from those monitoring programs directed by the Waste Isolation Pilot Plant (WIPP) Environmental Monitoring Plan (EMP) (DOE/WIPP 94-024). The EMP defines a comprehensive set of parameters that must be monitored to detect potential impacts to the environment and to establish baseline measurements for future environmental evaluations. Surface water, groundwater, air, soil, and biotics are monitored for radiological and nonradiological activity levels. The baseline radiological surveillance program covers the broader geographic area that encompasses nearby ranches, villages, and cities. Nonradiological studies focus on the area immediately surrounding the WIPP site

  20. Waste Isolation Pilot Plant site environmental report, for calendar year 1995

    International Nuclear Information System (INIS)

    1996-01-01

    The U.S. Department of Energy (DOE) Order 5400.1 General Environmental Protection Program, requires DOE facilities, that conduct environmental protection programs, to annually prepare a Site Environmental Report (SER). The purpose of the SER is to provide an abstract of environmental assessments conducted in order to characterize site environmental management performance, to confirm compliance with environmental standards and requirements, and to highlight significant programs and efforts of environmental merit. The content of this SER is not restricted to a synopsis of the required data, in addition, information pertaining to new and continued monitoring and compliance activities during the 1995 calendar year are also included. Data contained in this report are derived from those monitoring programs directed by the Waste Isolation Pilot Plant (WIPP) Environmental Monitoring Plan (EMP). The EMP provides inclusive guidelines implemented to detect potential impacts to the environment and to establish baseline measurements for future environmental evaluations. Surface water, groundwater. air, soil, and biotic matrices are monitored for an array of radiological and nonradiological factors. The baseline radiological surveillance program encompasses a broader geographic area that includes nearby ranches, villages, and cities. Most elements of nonradiological assessments are conducted within the geographic vicinity of the WIPP site

  1. TSA waste stream and final waste form composition

    International Nuclear Information System (INIS)

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

    1993-01-01

    A final vitrified waste form composition, based upon the chemical compositions of the input waste streams, is recommended for the transuranic-contaminated waste stored at the Transuranic Storage Area of the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory. The quantities of waste are large with a considerable uncertainty in the distribution of various waste materials. It is therefore impractical to mix the input waste streams into an ''average'' transuranic-contaminated waste. As a result, waste stream input to a melter could vary widely in composition, with the potential of affecting the composition and properties of the final waste form. This work examines the extent of the variation in the input waste streams, as well as the final waste form under conditions of adding different amounts of soil. Five prominent Rocky Flats Plant 740 waste streams are considered, as well as nonspecial metals and the ''average'' transuranic-contaminated waste streams. The metals waste stream is the most extreme variation and results indicate that if an average of approximately 60 wt% of the mixture is soil, the final waste form will be predominantly silica, alumina, alkaline earth oxides, and iron oxide. This composition will have consistent properties in the final waste form, including high leach resistance, irrespective of the variation in waste stream. For other waste streams, much less or no soil could be required to yield a leach resistant waste form but with varying properties

  2. Tank vapor characterization project - headspace vapor characterization of Hanford Waste Tank 241-C-107: Second comparison study results from samples collected on 3/26/96

    International Nuclear Information System (INIS)

    Evans, J.C.; Pool, K.H.; Thomas, B.L.

    1997-01-01

    This report describes the analytical results of vapor samples taken from the headspace of waste storage tank 241-C-107 (Tank C-107) at the Hanford Site in Washington State. The results described in this report is the second in a series comparing vapor sampling of the tank headspace using the Vapor Sampling System (VSS) and In Situ Vapor Sampling (ISVS) system without high efficiency particulate air (HEPA) prefiltration. The results include air concentrations of water (H 2 O) and ammonia (NH 3 ), permanent gases, total non-methane organic compounds (TO-12), and individual organic analytes collected in SUMMA trademark canisters and on triple sorbent traps (TSTs). Samples were collected by Westinghouse Hanford Company (WHC) and analyzed by Pacific Northwest National Laboratory (PNNL). Analyses were performed by the Vapor Analytical Laboratory (VAL) at PNNL. Analyte concentrations were based on analytical results and, where appropriate, sample volume measurements provided by WHC

  3. Tank vapor characterization project. Headspace vapor characterization of Hanford waste tank 241-BY-108: Second comparison study results from samples collected on 3/28/96

    International Nuclear Information System (INIS)

    Thomas, B.L.; Pool, K.H.; Evans, J.C.

    1997-01-01

    This report describes the analytical results of vapor samples taken from the headspace of waste storage tank 241-BY-108 (Tank BY-108) at the Hanford Site in Washington State. The results described in this report is the second in a series comparing vapor sampling of the tank headspace using the Vapor Sampling System (VSS) and In Situ Vapor Sampling (ISVS) system without high efficiency particulate air (HEPA) prefiltration. The results include air concentrations of water (H 2 O) and ammonia (NH 3 ), permanent gases, total non-methane organic compounds (TO-12), and individual organic analytes collected in SUMMA trademark canisters and on triple sorbent traps (TSTs). Samples were collected by Westinghouse Hanford Company (WHC) and analyzed by Pacific Northwest National Laboratory (PNNL). Analyses were performed by the Vapor Analytical Laboratory (VAL) at PNNL. Analyte concentrations were based on analytical results and, where appropriate, sample volume measurements provided by WHC

  4. Ground-water flow near two radioactive-waste-disposal areas at the Western New York Nuclear Service Center, Cattaraugus County, New York; results of flow simulation

    Science.gov (United States)

    Bergeron, M.P.; Bugliosi, E.F.

    1988-01-01

    Two adjacent burial areas were excavated in a clay-rich till at a radioactive waste disposal site near West Valley in Cattaraugus County, N.Y.: (1) which contains mainly low-level radioactive wastes generated onsite by a nuclear fuel reprocessing plant, has been in operation since 1966; and (2) which contains commercial low-level radioactive wastes, was operated during 1963-75. Groundwater below the upper 3 meters of till generally moves downward through a 20- to 30-meter thick sequence of tills underlain by lacustrine and kame-delta deposits of fine sand and silt. Groundwater in the weathered, upper 3 meters of till can move laterally for several meters before either moving downward into the kame-delta deposits or discharging to the land surface. A two-dimensional finite-element model that simulates two vertical sections was used to evaluate hydrologic factors that control groundwater flow in the till. Conditions observed during March 1983 were reproduced accurately in steady-state simulations that used four isotropic units of differing hydraulic conductivity to represent two fractured and weathered till units near land surfaces, an intermediate group of isolated till zones that contain significant amounts of fine sand and silt, and a sequence of till units at depths that have been consolidated by overburden pressure. Recharge rates used in the best-fit simulation ranged from 1.4 cm/yr along smooth, sloping or compacted surfaces to 3.8 cm/yr near swampy areas. Values of hydraulic conductivity and infiltration used in the calibrated best-fit model were nearly identical to values used in a previous model analysis of the nearby commercial-waste burial area. Results of the model simulations of a burial pit assumed to be filled with water indicate that water near the bottom of the burial pit would migrate laterally in the shallow, weathered till for 5 to 6 meters before moving downward into the unweathered till, and water near the top of the pit would move laterally

  5. Thermal, chemical, and mass-transport processes induced in abyssal sediments by the emplacement of nuclear waste: experimental and modeling results

    International Nuclear Information System (INIS)

    McVey, D.F.; Erickson, K.L.; Seyfried, W.

    1980-01-01

    This paper discusses heat and mass transport studies of marine red clay sediments being considered as a nuclear waste isolation medium. Numerical models indicate that for a maximum allowable sediment/canister interface temperature of 200 to 250 0 C, the sediment can absorb about 1.5 kW initial power from waste in a 3 m long by 0.3 m dia canister buried 30 m in the sediment. Fluid displacement due to convection is found to be less than 1 m. Laboratory studies of the geochemical effects induced by heating sediment/seawater mixtures indicate that the canister and waste form must be designed to resist a hot, acid (pH 3 to 4) oxidizing environment. Since the thermally altered sediment volume of about 5.5 m 3 is small relative to the sediment volume overlying the canister, the acid and oxidizing conditions are not anticipated to effect the properties of the far field. Using sorption coefficient correlations, the migration of four nuclides 239 Pu, 137 Cs, 129 I, and 99 Tc were computer for a canister buried 30 m deep in a 60 m thick red clay sediment layer. It was found that the 239 Pu and 137 Cs are essentially completely contained in the sediments, while 129 I and 99 Tc broke through the 30 m of sediment in about 5000 years. The resultant peak injection rates of 4.6 x 10 -5 μCi/year-m 2 for 129 I and 1.6 x 10 -2 μCi/year-m 2 for 99 Tc were less than the natural radioactive flux of 226 Ra (3.5 to 8.8 x 10 -4 μCi/year-m 2 ) and 222 Rn

  6. The Water Reuse project: Sustainable waste water re-use technologies for irrigated land in NIS and southern European states; project overview and results.

    Science.gov (United States)

    van den Elsen, E.; Doerr, S.; Ritsema, C. J.

    2009-04-01

    In irrigated areas in the New Independent States (NIS) and southern European States, inefficient use of conventional water resources occurs through incomplete wetting of soils, which causes accelerated runoff and preferential flow, and also through excessive evaporation associated with unhindered capillary rise. Furthermore, a largely unexploited potential exists to save conventional irrigation water by supplementation with organic-rich waste water, which, if used appropriately, can also lead to improvements to soil physical properties and soil nutrient and organic matter content. This project aims to (a) reduce irrigation water losses by developing, evaluating and promoting techniques that improve the wetting properties of soils, and (b) investigate the use of organic-rich waste water as a non-conventional water resource in irrigation and, in addition, as a tool in improving soil physical properties and soil nutrient and organic matter content. Key activities include (i) identifying, for the NIS and southern European partner countries, the soil type/land use combinations, for which the above approaches are expected to be most effective and their implementation most feasible, using physical and socio-economic research methods, and (ii) examining the water saving potential, physical, biological and chemical effects on soils of the above approaches, and also their impact on performance. Expected outputs include techniques for sustainable improvements in soil wettability management as a novel approach in water saving, detailed evaluation of the prospects and effects of using supplemental organic-rich waste waters in irrigation, an advanced process-based numerical hydrological model, fully adapted to quantify and upscale resulting water savings and nutrient and potential contaminant fluxes for irrigated areas, and identification of suitable areas in the NIS and Mediterranean (in soil, land use, legislative and socio-economic terms) for implementation.

  7. Radioactive waste management

    International Nuclear Information System (INIS)

    Kawakami, Yutaka

    2008-01-01

    Radioactive waste generated from utilization of radioisotopes and each step of the nuclear fuel cycle and decommissioning of nuclear facilities are presented. On the safe management of radioactive waste management, international safety standards are established such as ''The Principles of Radioactive Waste Management (IAEA)'' and T he Joint Convention on the Safety of Radioactive Waste Management . Basic steps of radioactive waste management consist of treatment, conditioning and disposal. Disposal is the final step of radioactive waste management and its safety is confirmed by safety assessment in the licensing process. Safety assessment means evaluation of radiation dose rate caused by radioactive materials contained in disposed radioactive waste. The results of the safety assessment are compared with dose limits. The key issues of radioactive waste disposal are establishment of long term national strategies and regulations for safe management of radioactive waste, siting of repository, continuity of management activities and financial bases for long term, and security of human resources. (Author)

  8. Experience gained in the management of radioactive waste from maintenance, decontamination and partial decommissioning of a reprocessing plant and conclusions resulting for the management of radioactive wastes from nuclear power plants

    International Nuclear Information System (INIS)

    Hild, W.

    1983-01-01

    After a short description of the historical background of Eurochemic, its main tasks and the various operational phases, a detailed description of the waste management principles applied is presented. The practical experience in the waste treatment is reported for both the operational phase of the reprocessing plant and its decontamination and partial decommissioning after shutdown. Based on this experience and the presented data, an assessment of the practical operations is made and conclusions are drawn. Finally, recommendations are formulated both for the general waste management policy and the practical waste treatment processes in nuclear power reactors. (author)

  9. Human waste

    NARCIS (Netherlands)

    Amin, Md Nurul; Kroeze, Carolien; Strokal, Maryna

    2017-01-01

    Many people practice open defecation in south Asia. As a result, lot of human waste containing nutrients such as nitrogen (N) and phosphorus (P) enter rivers. Rivers transport these nutrients to coastal waters, resulting in marine pollution. This source of nutrient pollution is, however, ignored in

  10. Experimental results on salt concrete for barrier elements made of salt concrete in a repository for radioactive waste in a salt mine

    International Nuclear Information System (INIS)

    Gutsch, Alex-W.; Preuss, Juergen; Mauke, Ralf

    2012-01-01

    The Bartensleben rock salt mine in Germany was used as a repository for low and intermediate level radioactive waste from 1971 to 1991 and from 1994 to 1998. The repository with an overall volume of about 6 million m 3 has to be closed. Salt concrete is used for the refill of the voids of the repository. The concrete mixtures contain crushed salt instead of natural aggregates as the void filling material should be as similar to the salt rock as possible. Very high requirements regarding low heat development and little or even no cracking during concrete hardening had to be fulfilled even for the barrier elements made from salt concrete which separate the radioactive waste from the environment. Requirements for the salt concrete were set up with regard to the fluidity of the fresh concrete during the hardening process and its durability. In the view of a comprehensive numerical calculations of the temperature development and thermal stresses in the massive salt concrete elements of the backfill of the voids, experimental results for material properties of the salt concrete are presented: mixture of the salt concrete, thermodynamic properties (adiabatic heat release, thermal dilatation, thermal conductivity and heat capacity), mechanical short term properties, creep (under tension, under compression), autogenous shrinkage

  11. Hazard Ranking System evaluation of CERCLA [Comprehensive Environmental Response, Compensation, and Liability Act] inactive waste sites at Hanford: Volume 1, Evaluation methods and results

    International Nuclear Information System (INIS)

    Stenner, R.D.; Cramer, K.H.; Higley, K.A.; Jette, S.J.; Lamar, D.A.; McLaughlin, T.J.; Sherwood, D.R.; Van Houten, N.C.

    1988-10-01

    The purpose of this report is to formally document the individual site Hazard Ranking System (HRS) evaluations conducted as part of the preliminary assessment/site inspection (PA/SI) activities at the US Department of Energy (DOE) Hanford Site. These activities were carried out pursuant to the DOE orders that describe the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) Program addressing the cleanup of inactive waste sites. These orders incorporate the US Environmental Protection Agency methodology, which is based on the Superfund Amendments and Reauthorization Act of 1986 (SARA). The methodology includes six parts: PA/SI, remedial investigation/feasibility study, record of decision, design and implementation of remedial action, operation and monitoring, and verification monitoring. Volume 1 of this report discusses the CERCLA inactive waste-site evaluation process, assumptions, and results of the HRS methodology employed. Volume 2 presents the data on the individual CERCLA engineered-facility sites at Hanford, as contained in the Hanford Inactive Site Surveillance (HISS) Data Base. Volume 3 presents the data on the individual CERCLA unplanned-release sites at Hanford, as contained in the HISS Data Base. 34 refs., 43 figs., 47 tabs

  12. Pilot-Scale Test Results Of A Thin Film Evaporator System For Management Of Liquid High-Level Wastes At The Hanford Site Washington USA -11364

    International Nuclear Information System (INIS)

    Corbett, J.E.; Tedesch, A.R.; Wilson, R.A.; Beck, T.H.; Larkin, J.

    2011-01-01

    A modular, transportable evaporator system, using thin film evaporative technology, is planned for deployment at the Hanford radioactive waste storage tank complex. This technology, herein referred to as a wiped film evaporator (WFE), will be located at grade level above an underground storage tank to receive pumped liquids, concentrate the liquid stream from 1.1 specific gravity to approximately 1.4 and then return the concentrated solution back into the tank. Water is removed by evaporation at an internal heated drum surface exposed to high vacuum. The condensed water stream will be shipped to the site effluent treatment facility for final disposal. This operation provides significant risk mitigation to failure of the aging 242-A Evaporator facility; the only operating evaporative system at Hanford maximizing waste storage. This technology is being implemented through a development and deployment project by the tank farm operating contractor, Washington River Protection Solutions (WRPS), for the Office of River Protection/Department of Energy (ORPIDOE), through Columbia Energy and Environmental Services, Inc. (Columbia Energy). The project will finalize technology maturity and install a system at one of the double-shell tank farms. This paper summarizes results of a pilot-scale test program conducted during calendar year 2010 as part of the ongoing technology maturation development scope for the WFE.

  13. Hazard Ranking System evaluation of CERCLA (Comprehensive Environmental Response, Compensation, and Liability Act) inactive waste sites at Hanford: Volume 1, Evaluation methods and results

    Energy Technology Data Exchange (ETDEWEB)

    Stenner, R.D.; Cramer, K.H.; Higley, K.A.; Jette, S.J.; Lamar, D.A.; McLaughlin, T.J.; Sherwood, D.R.; Van Houten, N.C.

    1988-10-01

    The purpose of this report is to formally document the individual site Hazard Ranking System (HRS) evaluations conducted as part of the preliminary assessment/site inspection (PA/SI) activities at the US Department of Energy (DOE) Hanford Site. These activities were carried out pursuant to the DOE orders that describe the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) Program addressing the cleanup of inactive waste sites. These orders incorporate the US Environmental Protection Agency methodology, which is based on the Superfund Amendments and Reauthorization Act of 1986 (SARA). The methodology includes six parts: PA/SI, remedial investigation/feasibility study, record of decision, design and implementation of remedial action, operation and monitoring, and verification monitoring. Volume 1 of this report discusses the CERCLA inactive waste-site evaluation process, assumptions, and results of the HRS methodology employed. Volume 2 presents the data on the individual CERCLA engineered-facility sites at Hanford, as contained in the Hanford Inactive Site Surveillance (HISS) Data Base. Volume 3 presents the data on the individual CERCLA unplanned-release sites at Hanford, as contained in the HISS Data Base. 34 refs., 43 figs., 47 tabs.

  14. U.S. Department of Energy's initiatives for proliferation prevention in Russia: results of radioactive liquid waste treatment project, year 2

    International Nuclear Information System (INIS)

    Pokhitonov, Y.; Kamachev, V.; Kelley, D.

    2010-10-01

    The objective of the project is to engage weapons scientists with training and research programs at selected nuclear sites in Russia and apply high technology polymers to immobilize legacy ILW and HLW liquids that have posed environmental challenges over the years. One compelling advantage of the projects is that V.G. Khlopin Radium Institute and Pacific Nuclear Solutions have been engaged in applied research for seven years to validate the performance and effectiveness of the polymer technology for use with radioactive liquids. With conclusive results of the research work on sixty active and simulant waste streams, the project can focus on actual applications of the technology at Ozersk (Mayak), Sever sk (SCC) Zheleznogorsk (MCC) and Gatchyna rather that on pure research. The long term objective of the project is find viable waste management solutions for serious radioactive and chemical contamination that has existed in Russia and the U. S. for several decades. The polymer technologies may be applied to all radioactive liquid. This paper summarizes the experimental work of the immobilization process and data definition of the most effective polymer compositions in addition to determining the optimum polymer to liquid ratios for economic considerations. (Author)

  15. PILOT-SCALE TEST RESULTS OF A THIN FILM EVAPORATOR SYSTEM FOR MANAGEMENT OF LIQUID HIGH-LEVEL WASTES AT THE HANFORD SITE WASHINGTON USA -11364

    Energy Technology Data Exchange (ETDEWEB)

    CORBETT JE; TEDESCH AR; WILSON RA; BECK TH; LARKIN J

    2011-02-14

    A modular, transportable evaporator system, using thin film evaporative technology, is planned for deployment at the Hanford radioactive waste storage tank complex. This technology, herein referred to as a wiped film evaporator (WFE), will be located at grade level above an underground storage tank to receive pumped liquids, concentrate the liquid stream from 1.1 specific gravity to approximately 1.4 and then return the concentrated solution back into the tank. Water is removed by evaporation at an internal heated drum surface exposed to high vacuum. The condensed water stream will be shipped to the site effluent treatment facility for final disposal. This operation provides significant risk mitigation to failure of the aging 242-A Evaporator facility; the only operating evaporative system at Hanford maximizing waste storage. This technology is being implemented through a development and deployment project by the tank farm operating contractor, Washington River Protection Solutions (WRPS), for the Office of River Protection/Department of Energy (ORPIDOE), through Columbia Energy and Environmental Services, Inc. (Columbia Energy). The project will finalize technology maturity and install a system at one of the double-shell tank farms. This paper summarizes results of a pilot-scale test program conducted during calendar year 2010 as part of the ongoing technology maturation development scope for the WFE.

  16. Follow-up and control of analytical results from environmental monitoring program of the Radioactive Waste Disposal Facility - Abadia de Goias

    International Nuclear Information System (INIS)

    Peixoto, Claudia Marques; Jacomino, Vanusa Maria Feliciano

    2000-01-01

    The analytical results for the 12 month period (August/1997 to July/1998) of the Environmental Monitoring Program operational phase of the radioactive waste disposal facility 'Abadia de Goias' (DIGOI), located in the District of Goiania, are summarized in this report. A statistical treatment of the data using control graphs is also presented. The use of these graphs allows the arrangement of the data in a way that facilitates process control and visualization of data trends and periodicity organized according to temporal variation. A comparison is made of these results vs. those obtained during the pre-operational phase. Moreover, the effective equivalent dose received by the public individuals for different critical pathways is estimated. (author)

  17. Wastes options

    International Nuclear Information System (INIS)

    Maes, M.

    1992-01-01

    After a description of the EEC environmental policy, some wastes families are described: bio-contaminant wastes (municipal and industrial), hospitals wastes, toxic wastes in dispersed quantities, nuclear wastes (radioactive and thermal), plastics compounds wastes, volatiles organic compounds, hydrocarbons and used solvents. Sources, quantities and treatments are given. (A.B.). refs., figs., tabs

  18. Recommendations to the NRC for review criteria for alternative methods of low-level radioactive waste disposal: Environmental monitoring and surveillance programs

    International Nuclear Information System (INIS)

    Denham, D.H.; Stenner, R.D.; Eddy, P.A.; Jaquish, R.E.; Ramsdell, J.V. Jr.

    1988-07-01

    Licensing of a facility for low-level radioactive waste disposal requires the review of the environmental monitoring and surveillance programs. A set of review criteria is recommended for the US Nuclear Regulatory Commission (NRC) staff to use in each monitoring phase---preoperational, operational, and post operational---for evaluating radiological and selected nonradiological parameters in proposed environmental monitoring and surveillance programs at low-level waste disposal facilities. Applicable regulations, industry standards, and technical guidance on low-level radioactive waste are noted throughout the document. In the preoperational phase, the applicant must demonstrate that the environmental monitoring program identifies radiation levels and radionuclide concentrations at the site and also provides adequate basic data on the disposal site. Data recording and statistical analyses for this phase are addressed

  19. Utilization of waste waters in fish production: preliminary results from fish culture studies in floating cages in a sewage pond, New Bussa, Nigeria

    OpenAIRE

    Otubusin, S.O.; Olatunde, A.A.

    1993-01-01

    The utilization of waste waters in aquaculture were briefly reviewed. At the National Institute for Freshwater Fisheries Research (NIFFR), stocking density (20 to 160 fish/m super(3)) experiments using Sarotherodon galilaeus (without supplementary feeding) in floating cages were carried out in a sewage pond (0.4ha surface area). Cage culture of S. galilaeus was observed to have potentials in waste waters aquaculture. Recommendations were made on the execution of an intergrated waste water ...

  20. Waste Sites - Municipal Waste Operations

    Data.gov (United States)

    NSGIC Education | GIS Inventory — A Municipal Waste Operation is a DEP primary facility type related to the Waste Management Municipal Waste Program. The sub-facility types related to Municipal Waste...

  1. The biomonitoring and bioremediation of toxic water resulting from municipal waste storage of Somârd, Sibiu county

    OpenAIRE

    Ioan C. Oprea; Dana Malschi; Liviu O. Muntean

    2013-01-01

    This paper presents information from the specialty literature and laboratory experimental results on biomonitoring, phytoextraction, biodegradation, and biotransformation of toxic water pollutants at the biotechnology laboratory of the Faculty of Environmental Science and Engineering. The study was conducted in laboratory micro tanks with contaminated water from the municipal landfill water storage pit with toxic bund of Somârd/Medias, Sibiu County, in order to research and develo...

  2. Mechanisms of gas bubble retention and release: results for Hanford Waste Tanks 241-S-102 and 241-SY-103 and single-shell tank simulants

    International Nuclear Information System (INIS)

    Gauglitz, P.A.; Rassat, S.D.; Bredt, P.R.; Konynenbelt, J.H.; Tingey, S.M.; Mendoza, D.P.

    1996-09-01

    Research at Pacific Northwest National Laboratory (PNNL) has probed the physical mechanisms and waste properties that contribute to the retention and release of flammable gases from radioactive waste stored in underground tanks at Hanford. This study was conducted for Westinghouse Hanford Company as part of the PNNL Flammable Gas Project. The wastes contained in the tanks are mixes of radioactive and chemical products, and some of these wastes are known to generate mixtures of flammable gases, including hydrogen, nitrous oxide, and ammonia. Because these gases are flammable, their retention and episodic release pose a number of safety concerns

  3. Mechanisms of gas bubble retention and release: results for Hanford Waste Tanks 241-S-102 and 241-SY-103 and single-shell tank simulants

    Energy Technology Data Exchange (ETDEWEB)

    Gauglitz, P.A.; Rassat, S.D.; Bredt, P.R.; Konynenbelt, J.H.; Tingey, S.M.; Mendoza, D.P.

    1996-09-01

    Research at Pacific Northwest National Laboratory (PNNL) has probed the physical mechanisms and waste properties that contribute to the retention and release of flammable gases from radioactive waste stored in underground tanks at Hanford. This study was conducted for Westinghouse Hanford Company as part of the PNNL Flammable Gas Project. The wastes contained in the tanks are mixes of radioactive and chemical products, and some of these wastes are known to generate mixtures of flammable gases, including hydrogen, nitrous oxide, and ammonia. Because these gases are flammable, their retention and episodic release pose a number of safety concerns.

  4. Radioactive wastes and discharges

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-07-01

    The guide sets out the radiation safety requirements and limits for the treatment of radioactive waste. They shall be observed when discharging radioactive substances into the atmosphere or sewer system, or when delivering solid, low-activity waste to a landfill site without a separate waste treatment plan. The guide does not apply to the radioactive waste resulting from the utilisation of nuclear energy or natural resources.

  5. Radioactive wastes and discharges

    International Nuclear Information System (INIS)

    2000-01-01

    The guide sets out the radiation safety requirements and limits for the treatment of radioactive waste. They shall be observed when discharging radioactive substances into the atmosphere or sewer system, or when delivering solid, low-activity waste to a landfill site without a separate waste treatment plan. The guide does not apply to the radioactive waste resulting from the utilisation of nuclear energy or natural resources

  6. Radioecological aspects of at-sea dumping of nuclear wastes resulting from the FSU nuclear fleet activities: Reliability of packings and necessity of rehabilitation measures

    International Nuclear Information System (INIS)

    Lavkovsky, S.; Kvasha, N.; Kobzev, V.; Sadovnikov, V.; Goltsev, V.

    2002-01-01

    The practice of radioactive waste treatment in the former USSR was that prior to at-sea dumping of objects with spent nuclear fuel (SNF) a set of design and technological measures was undertaken with a view to form packings with additional barriers to prevent radionuclide release in the environment. Based upon the results of most conservative evaluations of the protective barrier corrosion resistance it was concluded, that till Year 2300 there will be no grounds to worry about a possibility of the loss of tightness of the majority of packings. However, should unfavourable external natural factors combine, the loss of sealing of the packing with the screening assembly of the nuclear icebreaker 'Lenin' can occur at any moment. (author)

  7. Headspace vapor characterization of Hanford Waste Tank 241-S-112: Results from samples collected on July 11, 1995. Tank Vapor Characterization Project

    International Nuclear Information System (INIS)

    Clauss, T.W.; Pool, K.H.; Evans, J.C.

    1996-05-01

    This report describes the results of vapor samples taken from the headspace of waste storage Tank 241-S-112 (Tank S-112) at the Hanford. Pacific Northwest National Laboratory (PNNL) is contracted with Westinghouse Hanford Company (WHC) to provide sampling devices and analyze samples for inorganic and organic analytes collected from the tank headspace and ambient air near the tank. The analytical work was performed by the PNNL Vapor Analytical Laboratory (VAL) by the Tank Vapor Characterization Project. Work performed was based on a sample and analysis plan (SAP) prepared by WHC. The SAP provided job-specific instructions for samples, analyses, and reporting. The SAP for this sample job was open-quotes Vapor Sampling and Analysis Planclose quotes, and the sample job was designated S5044. Samples were collected by WHC on July 11, 1995, using the Vapor Sampling System (VSS), a truck-based sampling method using a heated probe inserted into the tank headspace

  8. Headspace vapor characterization of Hanford Waste Tank SX-102: Results from samples collected on July 19, 1995. Tank Vapor Characterization Project

    International Nuclear Information System (INIS)

    McVeety, B.D.; Evans, J.C.; Clauss, T.W.; Pool, K.H.

    1996-05-01

    This report describes the results of vapor samples taken from the headspace of waste storage tank 241-SX-102 (Tank SX-102) at the Hanford Site in Washington State. Pacific Northwest National Laboratory (PNNL) contracted with Westinghouse Hanford Company (WHC) to provide sampling devices and analyze samples for inorganic and organic analytes collected from the tank headspace and ambient air near the tank. The analytical work was performed under the PNNL Vapor Analytical Laboratory (VAL) by the Tank Vapor Characterization Project. Work performed was based on a sample and analysis plan (SAP) prepared by WHC. The SAP provided job-specific instructions for samples, analyses, and reporting. The SAP for this sample job was open-quotes Vapor Sampling and Analysis Planclose quotes, and the sample job was designated S5046. Samples were collected by WHC on July 19, 1995, using the vapor sampling system (VSS), a truck-based sampling method using a heated probe inserted into the tank headspace

  9. Headspace vapor characterization of Hanford Waste Tank 241-T-110: Results from samples collected on August 31, 1995. Tank Vapor Characterization Project

    International Nuclear Information System (INIS)

    McVeety, B.D.; Thomas, B.L.; Evans, J.C.

    1996-05-01

    This report describes the results of vapor samples taken from the headspace of waste storage tank 241-T-110 (Tank T-110) at the Hanford Site in Washington State. Pacific Northwest National Laboratory (PNNL) contracted with Westinghouse Hanford Company (WHC) to provide sampling devices and analyze samples for inorganic and organic analytes collected from the tank headspace and ambient air near the tank. The analytical work was performed by the PNNL Vapor Analytical Laboratory (VAL) by the Tank Vapor Characterization Project. Work performed was based on a sample and analysis plan (SAP) prepared by WHC. The SAP provided job-specific instructions for samples, analyses, and reporting. The SAP for this sample job was open-quotes Vapor Sampling and Analysis Planclose quotes, and the sample job was designated S5056. Samples were collected by WHC on August 31, 1995, using the Vapor Sampling System (VSS), a truck-based sampling method using a heated probe inserted into the tank headspace

  10. Headspace vapor characterization of Hanford Waste Tank 241-TX-111: Results from samples collected on October 12, 1995. Tank Vapor Characterization Project

    International Nuclear Information System (INIS)

    Pool, K.H.; Clauss, T.W.; Evans, J.C.

    1996-06-01

    This report describes the results of vapor samples taken from the headspace of waste storage tank 241-TX-111 (Tank TX-111) at the Hanford Site in Washington State. Pacific Northwest National Laboratory (PNNL) contracted with Westinghouse Hanford Company (WHC) to provide sampling devices and analyze samples for inorganic and organic analytes collected from the tank headspace and ambient air near the tank. The analytical work was performed by the PNNL Vapor Analytical Laboratory (VAL) by the Tank Vapor Characterization Project. Work performed was based on a sample and analysis plan (SAP) prepared by WHC. The SAP provided job-specific instructions for samples, analyses, and reporting. The SAP for this sample job was open-quotes Vapor Sampling and Analysis Planclose quotes, and the sample job was designated S5069. Samples were collected by WHC on October 12, 1995, using the Vapor Sampling System (VSS), a truck-based sampling method using a heated probe inserted into the tank headspace

  11. Headspace vapor characterization of Hanford Waste Tank AX-103: Results from samples collected on June 21, 1995. Tank Vapor Characterization Project

    International Nuclear Information System (INIS)

    Ligotke, M.W.; Pool, K.H.; Clauss, T.W.

    1996-05-01

    This report describes the results of vapor samples taken from the headspace of waste storage tank 241-AX-103 (Tank AX-103) at the Hanford Site in Washington State. Pacific Northwest National Laboratory (PNNL) contracted with Westinghouse Hanford Company (WHC) to provide sampling devices and analyze samples for inorganic and organic analytes collected from the tank headspace and ambient air near the tank. The analytical work was performed by the PNNL Vapor Analytical Laboratory (VAL) by the Tank Vapor Characterization Project. Work performed was based on a sample and analysis plan (SAP) prepared by WHC. The SAP provided job-specific instructions for samples, analyses, and reporting. The SAP for this sample job was open-quotes Vapor Sampling and Analysis Planclose quotes, and the sample job was designated S5029. Samples were collected by WHC on June 21, 1995, using the Vapor Sampling System (VSS), a truck-based sampling method using a heated probe inserted into the tank headspace

  12. Headspace vapor characterization of Hanford Waste Tank AX-101: Results from samples collected on June 15, 1995. Tank Vapor Characterization Project

    International Nuclear Information System (INIS)

    Pool, K.H.; Clauss, T.W.; Evans, J.C.; McVeety, B.D.

    1996-05-01

    This report describes the results of vapor samples taken from the headspace of waste storage tank 241-AX-101 (Tank AX-101) at the Hanford Site in Washington State. Pacific Northwest National Laboratory (PNNL) contracted with Westinghouse Hanford Company (WHC) to provide sampling devices and analyze samples for inorganic and organic analytes collected from the tank headspace and ambient air near the tank. The analytical work was performed by the PNNL Vapor Analytical Laboratory (VAL) under the Tank Vapor Characterization Project. Work performed was based on a sample and analysis plan (SAP) prepared by WHC. The SAP provided job-specific instructions for samples, analyses, and reporting. The SAP for this sample job was open-quotes Vapor Sampling and Analysis Planclose quotes, and the sample job was designated S5028. Samples were collected by WHC on June 15, 1995, using the Vapor Sampling System (VSS), a truck-based sampling method using a heated probe inserted into the tank headspace

  13. Headspace vapor characterization of Hanford Waste Tank 241-SX-109: Results from samples collected on August 1, 1995. Tank Vapor Characterization Project

    International Nuclear Information System (INIS)

    Pool, K.H.; Clauss, T.W.; Evans, J.C.

    1996-05-01

    This report describes the results of vapor samples taken from the headspace of waste storage tank 241-SX-109 (Tank SX-109) at the Hanford Site in Washington State. Pacific Northwest National Laboratory (PNNL) contracted with Westinghouse Hanford Company (WHC) to provide sampling devices and analyze samples for inorganic and organic analytes collected from the tank headspace and ambient air near the tank. The analytical work was performed by the PNNL Vapor Analytical Laboratory (VAL) by the Tank Vapor Characterization Project. Work performed was based on a sample and analysis plan (SAP) prepared by WHC. The SAP provided job-specific instructions for samples, analyses, and reporting. The SAP for this sample job was open-quotes Vapor Sampling and Analysis Planclose quotes, and the sample job was designated S5048. Samples were collected by WHC on August 1, 1995, using the Vapor Sampling System (VSS), a truck-based sampling method using a heated probe inserted into the tank headspace

  14. Headspace vapor characterization of Hanford Waste Tank 241-SX-104: Results from samples collected on July 25, 1995. Tank Vapor Characterization Project

    International Nuclear Information System (INIS)

    Thomas, B.L.; Clauss, T.W.; Evans, J.C.

    1996-05-01

    This report describes the results of vapor samples taken from the headspace of waste storage tank 241-SX-104 (Tank SX-104) at the Hanford Site in Washington State. Pacific Northwest National Laboratory (PNNL) contracted with Westinghouse Hanford Company (WHC) to provide sampling devices and analyze samples for inorganic and organic analytes collected from the tank headspace and ambient air near the tank. The analytical work was performed by the PNNL Vapor Analytical Laboratory (VAL) by the Tank Vapor Characterization Project. Work performed was based on a sample and analysis plan (SAP) prepared by WHC. The SAP provided job-specific instructions for samples, analyses, and reporting. The SAP for this sample job was open-quotes Vapor Sampling and Analysis Planclose quotes, and the sample job was designated S5049. Samples were collected by WHC on July 25, 1995, using the Vapor Sampling System (VSS), a truck-based sampling method using a heated probe inserted into the tank headspace

  15. Uses of the potassium permanganate to eliminate copper cyanide from waste water resulting from a lixiviation plant in a gold mine (I)

    International Nuclear Information System (INIS)

    Sancho, J. P.; Fernandez, B.; Ayala, J.; Garcia, M. P.; Lavandeira, A.

    2009-01-01

    The use of cyanide in the hydrometallurgical and chemical industries has led to the emergence of a major environmental problem due to its high toxicity. Te wastewater generated at these plants is hazardous to the environment and therefore must be managed properly. For this purpose, they undergo detoxification processes after lodes from the plant are accumulated in waste-resistant containment ponds that mast be waterproof to prevent environmental disasters from leakages or massive flood. This work shows the results obtained in laboratory tests carried out with plant waters and demonstrates the efficacy of potassium permanganate as an oxidant of cyanide wastewater from a gold hydrometallurgical plant. In the process the destruction of the copper cyanide complexes is solution is achieved and copper metal ions are eliminated through precipitation mostly as hydroxide. (Author) 28 refs.

  16. Headspace vapor characterization of Hanford Waste Tank 241-SX-105: Results from samples collected on July 26, 1995. Tank Vapor Characterization Project

    International Nuclear Information System (INIS)

    Pool, K.H.; Clauss, T.W.; Evans, J.C.

    1996-05-01

    This report describes the results of vapor samples taken from the headspace of waste storage tank 241-SX-105 (Tank SX-105) at the Hanford Site in Washington State. Pacific Northwest National Laboratory (PNNL) contracted with Westinghouse Hanford Company (WHC) to provide sampling devices and analyze samples for inorganic and organic analytes collected from the tank headspace and ambient air near the tank. The analytical work was performed by the PNNL Vapor Analytical Laboratory (VAL) by the Tank Vapor Characterization Project. Work performed was based on a sample and analysis plan (SAP) prepared by WHC. The SAP provided job-specific instructions for samples, analyses, and reporting. The SAP for this sample job was open-quotes Vapor Sampling and Analysis Planclose quotes, and the sample job was designated S5047. Samples were collected by WHC on July 26, 1995, using the Vapor Sampling System (VSS), a truck-based sampling method using a heated probe inserted into the tank headspace

  17. Nuclear waste

    International Nuclear Information System (INIS)

    1988-01-01

    As required by the Nuclear Waste Policy Act of 1982, the Department of Energy is to annually determine whether the waste disposal fee will produce sufficient revenues to offset the total estimated costs of the waste disposal program. In its June 1987 assessment, DOE recommended that the fee remain unchanged even though its analysis showed that at an inflation rate of 4 percent the current fee would result in end-of-program deficits ranging from $21 billion to $76 billion in 2085. The 1988 assessment calls for reduced total costs because of program changes. Thus, DOE may be able to begin using a realistic inflation rate in determining fee adequacy in 1988 without proposing a major fee increase

  18. Investigations of subterranean microorganisms and their importance for performance assessment of radioactive waste disposal. Results and conclusions achieved during the period 1995 to 1997

    International Nuclear Information System (INIS)

    Pedersen, K.

    1997-11-01

    In 1987, microbiology became a part of the Swedish scientific program for the safe disposal of high level nuclear waste (HLW). The goal of the microbiology sub-program is to understand how subterranean microorganisms will interact with the performance of a future HLW repository. The Swedish research program on subterranean microbiology has mainly been performed at two sites in granitic rock aquifers at depths ranging from 70 m down to 1240 m; the Stripa research mine in the middle of Sweden and the Aespoe hard rock laboratory (HRL) situated on the south eastern coast of Sweden. Some work has also been performed in co-operation with other national or international research groups in Sweden, Canada and at the natural analogue sites in Oklo in Gabon and Maqarin in Jordan. The most recent report in the SKB technical report series on microbiology and performance assessment, SKB-TR--95-10, gave the state of the art regarding microorganisms and their importance for performance assessment. That report is recommended as a source of knowledge about basic microbiology, microbial ecology of subterranean environments and the nuclear waste disposal concept in a microbiological perspective. The present report summarises results and conclusions achieved during the period 1995 to 1997 and is a continuation of SKB TR 95-10. The report is structured as summary which explains and analyses the obtained results and conclusions in a performance assessment perspective. The scientific basis for the summary is an enclosed series of eleven papers of which eight have gone through an international peer review process for publication in international scientific journals and reports and papers published earlier

  19. Investigations of subterranean microorganisms and their importance for performance assessment of radioactive waste disposal. Results and conclusions achieved during the period 1995 to 1997

    Energy Technology Data Exchange (ETDEWEB)

    Pedersen, K [Goeteborg Univ. (Sweden). Dept. of General and Marine Microbiology

    1997-11-01

    In 1987, microbiology became a part of the Swedish scientific program for the safe disposal of high level nuclear waste (HLW). The goal of the microbiology sub-program is to understand how subterranean microorganisms will interact with the performance of a future HLW repository. The Swedish research program on subterranean microbiology has mainly been performed at two sites in granitic rock aquifers at depths ranging from 70 m down to 1240 m; the Stripa research mine in the middle of Sweden and the Aespoe hard rock laboratory (HRL) situated on the south eastern coast of Sweden. Some work has also been performed in co-operation with other national or international research groups in Sweden, Canada and at the natural analogue sites in Oklo in Gabon and Maqarin in Jordan. The most recent report in the SKB technical report series on microbiology and performance assessment, SKB-TR--95-10, gave the state of the art regarding microorganisms and their importance for performance assessment. That report is recommended as a source of knowledge about basic microbiology, microbial ecology of subterranean environments and the nuclear waste disposal concept in a microbiological perspective. The present report summarises results and conclusions achieved during the period 1995 to 1997 and is a continuation of SKB TR 95-10. The report is structured as summary which explains and analyses the obtained results and conclusions in a performance assessment perspective. The scientific basis for the summary is an enclosed series of eleven papers of which eight have gone through an international peer review process for publication in international scientific journals and reports and papers published earlier. 413 refs, 56 figs, 39 tabs.

  20. Radioactive Waste in Perspective

    International Nuclear Information System (INIS)

    2011-01-01

    Large volumes of hazardous wastes are produced each year, however only a small proportion of them are radioactive. While disposal options for hazardous wastes are generally well established, some types of hazardous waste face issues similar to those for radioactive waste and also require long-term disposal arrangements. The objective of this NEA study is to put the management of radioactive waste into perspective, firstly by contrasting features of radioactive and hazardous wastes, together with their management policies and strategies, and secondly by examining the specific case of the wastes resulting from carbon capture and storage of fossil fuels. The study seeks to give policy makers and interested stakeholders a broad overview of the similarities and differences between radioactive and hazardous wastes and their management strategies. Contents: - Foreword; - Key Points for Policy Makers; - Executive Summary; - Introduction; - Theme 1 - Radioactive and Hazardous Wastes in Perspective; - Theme 2 - The Outlook for Wastes Arising from Coal and from Nuclear Power Generation; - Risk, Perceived Risk and Public Attitudes; - Concluding Discussion and Lessons Learnt; - Strategic Issues for Radioactive Waste; - Strategic Issues for Hazardous Waste; - Case Studies - The Management of Coal Ash, CO 2 and Mercury as Wastes; - Risk and Perceived Risk; - List of Participants; - List of Abbreviations. (authors)

  1. Evaluation of the potentialities to reduce greenhouse gases (GHG) emissions resulting from various treatments of municipal solid wastes (MSW) in moist tropical climates: application to Yaounde.

    Science.gov (United States)

    Ngnikam, Emmanuel; Tanawa, Emile; Rousseaux, Patrick; Riedacker, Arthur; Gourdon, Rémy

    2002-12-01

    The authors here analyse the emission of greenhouse gases (GHG) resulting from the various treatment of municipal solid waste found in the town of Yaounde. Four management systems have been taken as the basis for analyses. System 1 is the traditional collection and landfill disposal, while in system 2 the hiogas produced in the landfill is recuperated to produce electricity. In systems 3 and 4, in addition to the collection, we have introduced a centralised composting or biogas plant before the landfilling disposal of refuse. A Life Cycle Inventory (LCI) of the four systems was made; this enable us to quantify the flux of matter and of energy, consumed or produced by the systems. Following this, only the greenhouse effect was taken into account to evaluate the ecological consequences of the MSW management systems. The method used to evaluate this impact takes into consideration on the one hand, GHG emissions or avoided emission following the substitution of fuel with methane recovered from landfills or produced in the digesters, and on the other hand, sequestrated carbon in the soil following the regular deposit of compost. Landfilling without recuperation of methane is the most emitting solution for greenhouse gas: it leads to the emission of 1.7 ton of carbon dioxide equivalent (tCO2E) per ton of household waste. Composting and methanisation allow one to have a comparable level of emission reduction, either respectively 1.8 and 2 tCO2E/t of MSW. In order to reduce the emission of GHG in the waste management systems, it is advisable to avoid first of all the emissions of methane coming from the landfills. System 2 seems to be a solution that would reduce the emissions of GHG at low cost (2.2 to 4 $/tCO2E). System 2 is calculated as the most effective at the environmental and economic level in the context of Yaounde. Therefore traditional collection, landfill disposal and biogas recuperation to produce electricity is preferable in moist tropical climates.

  2. TACIS Belarus - an overview of results and planned activities in the field of radiation protection, emergency preparedness and waste management

    Energy Technology Data Exchange (ETDEWEB)

    Ackermann, L. [Gesellschaft fur Anlagen- und Reaktorsicherheit (GRS) mbH, Berlin (Germany)

    2006-07-01

    Since 1996 the nuclear safety authorities of the Republic of Belarus were assisted with TACIS activities entitled: 'Transfer of Western European Regulatory Methodology and Practices to the Nuclear Safety Authorities of Belarus'. Considering the results of the Exploratory Mission which was arranged in 1996 the Regulatory Assistance Projects BE/RA/01 and BE/RA/02 were successful realised in 1998/1999 and in 2003/2004, respectively. These projects were financed by the Commission of the European Communities (EC) and implemented by a consortium of Technical Support Organisations (TSOs) from France (IRSN (former IPSN)), Germany (GRS) and Sweden (SSI) led by Riskaudit IRSN/GRS International. Beneficiary of the projects were Promatomnadzor at the beginning and later the Ministry for Emergency Situations of the Republic of Belarus each in connection with the Republican Centre of Radiation Control and Monitoring (RCRCM). The actual project BE/RA/03 'Regulatory Assistance to Belarus in the Field of Nuclear Safety and Radiation Protection including Radiological Emergency Preparedness' was started by the end of August 2006. (author)

  3. TACIS Belarus - an overview of results and planned activities in the field of radiation protection, emergency preparedness and waste management

    International Nuclear Information System (INIS)

    Ackermann, L.

    2006-01-01

    Since 1996 the nuclear safety authorities of the Republic of Belarus were assisted with TACIS activities entitled: 'Transfer of Western European Regulatory Methodology and Practices to the Nuclear Safety Authorities of Belarus'. Considering the results of the Exploratory Mission which was arranged in 1996 the Regulatory Assistance Projects BE/RA/01 and BE/RA/02 were successful realised in 1998/1999 and in 2003/2004, respectively. These projects were financed by the Commission of the European Communities (EC) and implemented by a consortium of Technical Support Organisations (TSOs) from France (IRSN (former IPSN)), Germany (GRS) and Sweden (SSI) led by Riskaudit IRSN/GRS International. Beneficiary of the projects were Promatomnadzor at the beginning and later the Ministry for Emergency Situations of the Republic of Belarus each in connection with the Republican Centre of Radiation Control and Monitoring (RCRCM). The actual project BE/RA/03 'Regulatory Assistance to Belarus in the Field of Nuclear Safety and Radiation Protection including Radiological Emergency Preparedness' was started by the end of August 2006. (author)

  4. Tank Vapor Characterization Project: Headspace vapor characterization of Hanford Waste Tank 241-S-103: Results from samples collected on 06/12/96

    International Nuclear Information System (INIS)

    Evans, J.C.; Pool, K.H.; Thomas, B.L.

    1997-01-01

    This report describes the analytical results of vapor samples taken from the headspace of the waste storage tank 241-S-103 (Tank S-103) at the Hanford Site in Washington State. The results described in this report were obtained to characterize the vapors present in the tank headspace and to support safety evaluations and tank farm operations. The results include air concentrations of selected inorganic and organic analytes and grouped compounds from samples obtained by Westinghouse Hanford Company (WHC) and provided for analysis to Pacific Northwest National Laboratory (PNNL). Analyses were performed by the Vapor Analytical Laboratory (VAL) at PNNL. Analyte concentrations were based on analytical results and, where appropriate, sample volumes provided by WHC. A summary of the inorganic analytes, permanent gases, and total non-methane organic compounds is listed in Table S.1. The three highest concentration analytes detected in SUMMA trademark canister and triple sorbent trap samples are also listed in Table S.1. Detailed descriptions of the analytical results appear in the appendices

  5. Tank Vapor Characterization Project: Headspace vapor characterization of Hanford Waste Tank 241-C-204: Results from samples collected on 07/02/96

    International Nuclear Information System (INIS)

    Thomas, B.L.; Evans, J.C.; Pool, K.H.

    1997-01-01

    This report describes the analytical results of vapor samples taken from the headspace of the waste storage tank 241-C-204 (Tank C-204) at the Hanford Site in Washington State. The results described in this report were obtained to characterize the vapors present in the tank headspace and to support safety evaluations and tank farm operations. The results include air concentrations of selected inorganic and organic analytes and grouped compounds from samples obtained by Westinghouse Hanford Company (WHC) and provided for analysis to Pacific Northwest National Laboratory (PNNL). Analyses were performed by the Vapor Analytical Laboratory (VAL) at PNNL. Analyte concentrations were based on analytical results and, where appropriate, sample volumes provided by WHC. A summary of the inorganic analytes, permanent gases, and total non-methane organic compounds is listed in Table S.1. The three highest concentration analytes detected in SUMMA trademark canister and triple sorbent trap samples are also listed in Table S.1. Detailed descriptions of the analytical results appear in the appendices

  6. Vapor space characterization of waste Tank 241-C-104: Results from samples collected on 2/17/94 and 3/3/94

    International Nuclear Information System (INIS)

    Lucke, R.B.; McVeety, B.D.; Clauss, T.W.; Pool, K.H.; Young, J.S.; McCulloch, M.; Ligotke, M.W.; Fruchter, J.S.; Goheen, S.C.

    1995-10-01

    This report describes inorganic and organic analyses results from samples obtained from the headspace of the Hanford waste storage Tank 241-C-104 (referred to as Tank C-104). The results described here were obtained to support safety and toxicological evaluations. A summary of the results for inorganic and organic analytes is listed in Summary Table 1. Detailed descriptions of the results appear in the text. Quantitative results were obtained for the inorganic compounds ammonia (NH 3 ), nitrogen dioxide (NO 2 ), nitric oxide (NO), sulfur oxides (SO x ), and water vapor (H 2 O). Organic compounds were also quantitatively determined. Occupational Safety and Health Administration (OSHA) versatile sampler (OVS) tubes were analyzed for tributyl phosphate. Twenty-four organic tentatively identified compounds (TICs) were observed above the detection limit of (ca.) 10 ppbv, but standards for most of these were not available at the time of analysis, and the reported concentrations are semiquantitative estimates. In addition, the authors looked for the 40 standard TO-14 analytes. Of these, two were observed above the 2-ppbv calibrated instrument detection limit. The 10 organic analytes with the highest estimated concentrations are listed in Summary Table 1. These 10 analytes account for approximately 88% of the total organic components in Tank C 104. Tank C-104 is not on any of the Watch Lists

  7. Solid waste

    International Nuclear Information System (INIS)

    1995-01-01

    The article drawn up within the framework of 'the assessment of the state of the environment in Lebanon' provides an overview of solid waste management, and assesses future wastes volume and waste disposal issues.In particular it addresses the following concerns: - Long term projections of solid waste arisings (i.e. domestic, industrial, such commercial wastes, vehicle types, construction waste, waste oils, hazardous toxic wastes and finally hospital and clinical wastes) are described. - Appropriate disposal routes, and strategies for reducing volumes for final disposal - Balance between municipal and industrial solid waste generation and disposal/treatment and - environmental impacts (aesthetics, human health, natural environment )of existing dumps, and the potential impact of government plans for construction of solid waste facilities). Possible policies for institutional reform within the waste management sector are proposed. Tables provides estimations of generation rates and distribution of wastes in different regions of Lebanon. Laws related to solid waste management are summarized

  8. Spent fuel and high level waste: Chemical durability and performance under simulated repository conditions. Results of a coordinated research project 1998-2004. Part 2: Results of a previously unpublished CRP: Performance of high level waste forms and packages under repository conditions. Results of a co-ordinated research project 1991-1998

    International Nuclear Information System (INIS)

    2007-07-01

    The objective of the CRP (Coordinated Research Projekt) on the 'Performance of High Level Waste Forms and Packages under Repository Conditions' was to contribute to the development and implementation of proper and sound technologies for HLW and spent fuel management. Special emphasis was given to the identification of various waste form properties and the study of their long term durability in simulated repository conditions. Another objective was to promote the co-operation and exchange of information between Member States on experimental concerning behaviour of the waste form. The CRP was composed of research contracts and agreements with Argentina, Australia, Belgium, Canada, China, Czech Republic, Finland, France, Germany, India, Japan, Russia, and the United States of America. The publication includes 14 individual contributions of the participants to the CRP, which are indexed separately.

  9. Environmental impact assessment of the Swedish high-level radioactive waste disposal system - examples of likely considerations

    International Nuclear Information System (INIS)

    1994-01-01

    Sweden is investigating the feasibility of establishing a high-level radioactive waste (HLW) disposal system consisting of three components as follows: (1) Encapsulation facility, (2) system for transporting waste and (3) geologic repository. Swedish law requires that an Environmental Impact Assessment (EIA) be written for any planned action expected to have a significant impact on the environment. Before embarking on construction and operation of a HLW disposal system, the Swedish government will evaluate the expected environmental impacts to assure that the Swedish people and environmental will not be unduly affected by the disposal system. The EIA process requires that reasonable alternatives to the proposed action, including the 'zero' or 'no action' alternative, be considered so that the final approved plan for disposal will have undergone scrutiny and comparison of alternatives to arrive at a plan which is the best achievable given reasonable physical and monetary constraints. This report has been prepared by the Center for Nuclear Waste Regulatory Analyses (CNWRA) for use by the Swedish Radiation Protection Institute (SSI). The purpose of this report is to establish a document which outlines the types of information which would be in an EIA for a three part disposal system like that envisioned by the Swedish Nuclear Fuel and Waste Management Company (SKB) for the disposal of Sweden's HLW. Technical information that would normally be included in an EIA is outlined in this document. The SSI's primary interest is in radiological impacts. However, for the sake of completeness and also to evaluate all environmental impacts in a single document, non-radiological impacts are also included. Swedish authorities other than the SSI may have interest in the non-radiological parts of the document. 26 refs

  10. Analysis of the total system life cycle cost for the Civilian Radioactive Waste Management Program. Volume 1. The analysis and its results

    International Nuclear Information System (INIS)

    1986-04-01

    The total-system life-cycle cost (TSLCC) analysis for the Department of Energy's (DOE) Civilian Radioactive Waste Management Program is an ongoing activity that helps determine whether the revenue-producing mechanism established by the Nuclear Waste Policy Act of 1982 is sufficient to cover the cost of the program. This report provides cost estimates for the fourth evaluation of the adequacy of the fee. The total-system cost for the reference authorized-system program is estimated to be 24 to 32 billion (1985) dollars. The total-system cost for the reference improved-performance system is estimated to be 26 to 34 billion dollars. A number of sensitivity cases were analyzed. For the authorized system, the costs for the sensitivity cases studied range from 21 to 39 billion dollars. For the improved-performance system, which includes a facility for monitored retrievable storage, the total-system cost in the sensitivity cases is estimated to be as high as 41 billion dollars. The factors that affect costs more than any other single factor for both the authorized and the improved-performance systems are delays in repository startup. A preliminary analysis of the impact of extending the burnup of nuclear fuel in the reactor was also performed; its results indicate that the impact is insignificant: the total-system cost is essentially unchanged from the comparable constant-burnup cases. The current estimate of the the total-system cost for the reference authorized system is zero to 3 billion dollars (9%) higher than the estimate for the reference system in the January 1985 TSLCC analysis

  11. Waste management - sewage - special wastes

    International Nuclear Information System (INIS)

    1987-01-01

    The 27 papers represent a cross-section of the subject waste management. Particular attention is paid to the following themes: waste avoidance, waste product utilization, household wastes, dumping technology, sewage sludge treatments, special wastes, seepage from hazardous waste dumps, radioactive wastes, hospital wastes, purification of flue gas from waste combustion plants, flue gas purification and heavy metals, as well as combined sewage sludge and waste product utilization. The examples given relate to plants in Germany and other European countries. 12 papers have been separately recorded in the data base. (DG) [de

  12. Analytical results and effective dose estimation of the operational Environmental Monitoring Program for the radioactive waste repository in Abadia de Goias from 1998 to 2008

    Energy Technology Data Exchange (ETDEWEB)

    Ribeiro, Edison, E-mail: edison@cnen.gov.b [Centro Regional de Ciencias Nucleares do Centro-Oeste, Comissao Nacional de Energia Nuclear- Br 060 km 174, 5-Abadia de Goias- Goias, CEP 75345-000 (Brazil); Tauhata, Luiz, E-mail: tauhata@ird.gov.b [Instituto de Radioprotecao e Dosimetria, Comissao Nacional de Energia Nuclear, Recreio dos Bandeirantes, Rio de Janeiro, RJ, CEP 22780-160 (Brazil); Eugenia dos Santos, Eliane, E-mail: esantos@cnen.gov.b [Centro Regional de Ciencias Nucleares do Centro-Oeste, Comissao Nacional de Energia Nuclear- Br 060 km 174, 5-Abadia de Goias- Goias, CEP 75345-000 (Brazil); Silveira Correa, Rosangela da, E-mail: rcorrea@cnen.gov.b [Centro Regional de Ciencias Nucleares do Centro-Oeste, Comissao Nacional de Energia Nuclear- Br 060 km 174, 5-Abadia de Goias- Goias, CEP 75345-000 (Brazil)

    2011-02-15

    This paper presents the results of the Environmental Monitoring Program for the Radioactive waste repository of Abadia de Goias, which was originated from the accident of Goiania, conducted by the Regional Center of Nuclear Sciences (CRCN-CO) of the National Commission on Nuclear Energy (CNEN), from 1998 to 2008. The results are related to the determination of {sup 137}Cs activity per unit of mass or volume of samples from surface water, ground water, depth sediments of the river, soil and vegetation, and also the air-kerma rate estimation for gamma exposure in the monitored site. In the phase of operational Environmental Monitoring Program, the values of the geometric mean and standard deviation obtained for {sup 137}Cs activity per unit of mass or volume in the analyzed samples were (0.08 {+-} 1.16) Bq.L{sup -1} for surface and underground water, (0.22 {+-} 2.79) Bq.kg{sup -1} for soil, and (0.19 {+-} 2.72) Bq.kg{sup -1} for sediment, and (0.19 {+-} 2.30) Bq.kg{sup -1} for vegetation. These results were similar to the values of the pre-operational Environmental Monitoring Program. With these data, estimations for effective dose were evaluated for public individuals in the neighborhood of the waste repository, considering the main possible way of exposure of this population group. The annual effective dose obtained from the analysis of these results were lower than 0.3 mSv.y{sup -1}, which is the limit established by CNEN for environmental impact in the public individuals indicating that the facility is operating safely, without any radiological impact to the surrounding environment. - Research highlights: {yields} A stolen capsule of Cesium 137 was opened in the city of Goiania, generating some 6000 tons of debris which were stored in the Repository area built for this purpose. {yields} The activity of cesium 137 of the surface water, underground water, depth sediments of river, soil, vegetation, and air, inside and surround the Repository area. {yields

  13. TRU Waste Sampling Program: Volume I. Waste characterization

    International Nuclear Information System (INIS)

    Clements, T.L. Jr.; Kudera, D.E.

    1985-09-01

    Volume I of the TRU Waste Sampling Program report presents the waste characterization information obtained from sampling and characterizing various aged transuranic waste retrieved from storage at the Idaho National Engineering Laboratory and the Los Alamos National Laboratory. The data contained in this report include the results of gas sampling and gas generation, radiographic examinations, waste visual examination results, and waste compliance with the Waste Isolation Pilot Plant-Waste Acceptance Criteria (WIPP-WAC). A separate report, Volume II, contains data from the gas generation studies

  14. 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)

    2004-06-01

    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

  15. Treatability tests on water from a low-level waste burial ground

    International Nuclear Information System (INIS)

    Taylor, P.A.

    1990-01-01

    Lab-scale treatability tests on trench water from a low-level waste burial ground have shown that the water can be successfully treated by existing wastewater treatment plants at Oak Ridge National Laboratory. Water from the four most highly contaminated trenches that had been identified to date was used in the treatability tests. The softening and ion exchange processes used in the Process Wastewater Treatment Plant removed Sr-90 from the trench water, which was the only radionuclide present at above the discharge limits. The air stripping and activated carbon adsorption processes used in the Nonradiological Wastewater Treatment Plant removed volatile and semi-volatile organics, which were the main contaminants in the trench water, to below detection limits. 6 refs., 2 figs., 7 tabs

  16. Tank waste treatment science

    International Nuclear Information System (INIS)

    LaFemina, J.P.; Blanchard, D.L.; Bunker, B.C.; Colton, N.G.; Felmy, A.R.; Franz, J.A.; Liu, J.; Virden, J.W.

    1994-01-01

    Remediation efforts at the U.S. Department of Energy's Hanford Site require that many technical and scientific principles be combined for effectively managing and disposing the variety of wastes currently stored in underground tanks. Based on these principles, pretreatment technologies are being studied and developed to separate waste components and enable the most suitable treatment methods to be selected for final disposal of these wastes. The Tank Waste Treatment Science Task at Pacific Northwest Laboratory is addressing pretreatment technology development by investigating several aspects related to understanding and processing the tank contents. The experimental work includes evaluating the chemical and physical properties of the alkaline wastes, modeling sludge dissolution, and evaluating and designing ion exchange materials. This paper gives some examples of results of this work and shows how these results fit into the overall Hanford waste remediation activities. This work is part of series of projects being conducted for the Tank Waste Remediation System

  17. FY 1998 report on the results of the development for an advanced application system for glass waste with CO2 emission reduction; 1998 nendo CO{sub 2} haishutsu yokuseigata hai glass kodo riyo system no kenkyu kaihatsu seika hokokusho

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-03-01

    For the realization of the cycling recycle society, it is important to promote recycling using glass waste for CO2 reduction and effective use of resources. The paper described the FY 1998 results of the development. It included the classification by which the composition of the particle sizes required for recycled products will be made possible, basic studies for low-cost pulverization/separation of foreign substances, and tests on heightening of process accuracy. Relating to the artificial synthesis of feldspar/pottery stone using glass waste/alumina waste as starting materials, vitrification behavior was studied to find out the optimum synthesis conditions. Glass waste of 60% and aluminum hydroxide of 40% were mixed at calcination temperature of 1,100 degrees C for more than 1 hour. Properties of the artificial feldspar/pottery stone were similar to those of the natural one. As to the utilization of glass waste to hydrothermally solidified materials, high-strength solidified matter without crack was obtained in the three-component system of G powder/hydrated lime/gypsum. Concerning the production of eco-glass block, a study was made on the size and mixing ratio of the cullet waste which is very soluble and has no effects on performance. The paper touched also on the R and D of the environmentally friendly type system using multi-functional hybrid materials. (NEDO)

  18. Removal of strontium and transuranics from Hanford tank waste via addition of metal cations and chemical oxidant: FY 1995 test results

    International Nuclear Information System (INIS)

    Orth, R.J.; Zacher, A.H.; Schmidt, A.J.; Elmore, M.R.; Elliott, K.R.; Neuenschwander, G.G.; Gano, S.R.

    1995-09-01

    Chelating organics and some of their degradation products in the Hanford tank waste, such as EDTA, HEDTA, and NTA act to solubilize strontium and transuranics (TRU) in the tank waste supernatant. Displacement of strontium and TRU will facilitate the removal of these radionuclides via precipitation/filtration, ion exchange, or solvent extraction so that low-level waste feed specifications can be met. Pacific Northwest Laboratory has investigated two methods for releasing organic-complexed strontium and TRU components to allow for effective pretreatment of tank waste supernatant: metal cation addition (to promote displacement and flocculation) and chemical oxidant (pennanganate) addition (to promote chelator destruction/defunctionalization and possibly flocculation). These methods, which can be conducted at near-ambient. temperatures and pressures, could be deployed as intank processes

  19. Development and demonstration of a stabilization system for buried mixed waste tanks: Initital results of the tank V-9 hot demonstration

    International Nuclear Information System (INIS)

    Matthern, G.E.; Kuhns, D.J.; Meservey, R.H.; Farnsworth, R.K.

    1996-01-01

    This paper describes a systematic approach for the stabilization of buried mixed waste tanks and presents the status of an application of this approach to a specific hot waste tank demonstration to be performed in FY-96. The approach uses the cradle-to-grave concept and includes technical, health and safety, and regulatory considerations and requirements. It starts with the identification of the tank and continues to the final disposition and monitoring of the tank

  20. Radionuclide content of simulated and fully radioactive SRLLL waste glasses: comparison of results from ICP-MS, gamma spectrometry and alpha spectrometry

    International Nuclear Information System (INIS)

    Wolf, S.F.; Bates, J.K.

    1995-01-01

    We have measured the transuranic content of two transuranic=doped, simulated waste glasses, using inductively coupled plasma-mass spectrometry (ICP-MS), γ-spectrometry, and α-spectrometry. Average concentrations measured by each technique were within ± 10% of the as-doped concentrations. We also report the transuranic content of three fully radioactive SRL waste glasses that were determined using γ- and α-spectrometry measurements to deconvolute isobaric interferences present in the ICP-MS analyses

  1. Waste management

    International Nuclear Information System (INIS)

    Chmielewska, E.

    2010-01-01

    In this chapter formation of wastes and basic concepts of non-radioactive waste management are explained. This chapter consists of the following parts: People in Peril; Self-regulation of nature as a guide for minimizing and recycling waste; The current waste management situation in the Slovak Republic; Categorization and determination of the type of waste in legislative of Slovakia; Strategic directions waste management in the Slovak Republic.

  2. Business unusual - Waste Act implementation: solid waste

    CSIR Research Space (South Africa)

    Oelofse, Suzanna HH

    2013-08-01

    Full Text Available The preamble to the Waste Act (2008) is very clear that, as a result of this legislation, waste management in South Africa will never be the same again. This should send a clear message that ‘business as usual’ will no longer be sufficient....

  3. Radioactive wastes and discharges

    International Nuclear Information System (INIS)

    1993-01-01

    According to the Section 24 of the Finnish Radiation Decree (1512/91), the Finnish Centre for Radiation and Nuclear Safety shall specify the concentration and activity limits and principles for the determination whether a waste can be defined as a radioactive waste or not. The radiation safety requirements and limits for the disposal of radioactive waste are given in the guide. They must be observed when discharging radioactive waste into the atmosphere or sewer system, or when delivering solid low-activity waste to a landfill site without a separate waste disposal plan. The guide does not apply to the radioactive waste resulting from the utilization of nuclear energy of natural resources. (4 refs., 1 tab.)

  4. General requirements applicable to the production, inspection, processing, packaging and storage of various types of waste resulting from the reprocessing of fuels irradiated in pressurized light water reactors

    International Nuclear Information System (INIS)

    1982-09-01

    The Fundamental Safety Rules applicable to certain types of nuclear installation are intended to clarify the conditions of which observance, for the type of installation concerned and for the subject that they deal with, is considered as equivalent to compliance with regulatory French technical practice. These Rules should facilitate safety analysises and the clear understanding between persons interested in matters related to nuclear safety. They in no way reduce the operator's liability and pose no obstacle to statutory provisions in force. For any installation to which a Fundamental Safety Rule applies according to the foregoing paragraph, the operator may be relieved from application of the Rule if he shows proof that the safety objectives set by the Rule are attained by other means that he proposes within the framework of statutory procedures. Furthermore, the Central Service for the Safety of Nuclear Installations reserves the right at all times to alter any Fundamental Safety Rule, as required, should it deem this necessary, while specifying the applicability conditions. This rule is intended to define the general provisions applicable to the production, inspection, processing, packaging and storage of the different types of wastes resulting from the reprocessing of fuels irradiated in a PWR

  5. A comparison of geostatistically based inverse techniques for use in performance assessment analysis at the Waste Isolation Pilot Plant Site: Results from Test Case No. 1

    International Nuclear Information System (INIS)

    Zimmerman, D.A.; Gallegos, D.P.

    1993-10-01

    The groundwater flow pathway in the Culebra Dolomite aquifer at the Waste Isolation Pilot Plant (WIPP) has been identified as a potentially important pathway for radionuclide migration to the accessible environment. Consequently, uncertainties in the models used to describe flow and transport in the Culebra need to be addressed. A ''Geostatistics Test Problem'' is being developed to evaluate a number of inverse techniques that may be used for flow calculations in the WIPP performance assessment (PA). The T