WorldWideScience

Sample records for surface waste handling

  1. Waste Handling Building Conceptual Study

    International Nuclear Information System (INIS)

    G.W. Rowe

    2000-01-01

    The objective of the ''Waste Handling Building Conceptual Study'' is to develop proposed design requirements for the repository Waste Handling System in sufficient detail to allow the surface facility design to proceed to the License Application effort if the proposed requirements are approved by DOE. Proposed requirements were developed to further refine waste handling facility performance characteristics and design constraints with an emphasis on supporting modular construction, minimizing fuel inventory, and optimizing facility maintainability and dry handling operations. To meet this objective, this study attempts to provide an alternative design to the Site Recommendation design that is flexible, simple, reliable, and can be constructed in phases. The design concept will be input to the ''Modular Design/Construction and Operation Options Report'', which will address the overall program objectives and direction, including options and issues associated with transportation, the subsurface facility, and Total System Life Cycle Cost. This study (herein) is limited to the Waste Handling System and associated fuel staging system

  2. Radioactive wastes handling facility

    International Nuclear Information System (INIS)

    Hirose, Emiko; Inaguma, Masahiko; Ozaki, Shigeru; Matsumoto, Kaname.

    1997-01-01

    There are disposed an area where a conveyor is disposed for separating miscellaneous radioactive solid wastes such as metals, on area for operators which is disposed in the direction vertical to the transferring direction of the conveyor, an area for receiving the radioactive wastes and placing them on the conveyor and an area for collecting the radioactive wastes transferred by the conveyor. Since an operator can conduct handling while wearing a working cloth attached to a partition wall as he wears his ordinary cloth, the operation condition can be improved and the efficiency for the separating work can be improved. When the area for settling conveyors and the area for the operators is depressurized, cruds on the surface of the wastes are not released to the outside and the working clothes can be prevented from being involved. Since the wastes are transferred by the conveyor, the operator's moving range is reduced, poisonous materials are fallen and moved through a sliding way to an area for collecting materials to be separated. Accordingly, the materials to be removed can be accumulated easily. (N.H.)

  3. Solid waste handling

    International Nuclear Information System (INIS)

    Parazin, R.J.

    1995-01-01

    This study presents estimates of the solid radioactive waste quantities that will be generated in the Separations, Low-Level Waste Vitrification and High-Level Waste Vitrification facilities, collectively called the Tank Waste Remediation System Treatment Complex, over the life of these facilities. This study then considers previous estimates from other 200 Area generators and compares alternative methods of handling (segregation, packaging, assaying, shipping, etc.)

  4. Handling of radioactive waste

    International Nuclear Information System (INIS)

    Sanhueza Mir, Azucena

    1998-01-01

    Based on characteristics and quantities of different types of radioactive waste produced in the country, achievements in infrastructure and the way to solve problems related with radioactive waste handling and management, are presented in this paper. Objectives of maintaining facilities and capacities for controlling, processing and storing radioactive waste in a conditioned form, are attained, within a great range of legal framework, so defined to contribute with safety to people and environment (au)

  5. Handling of waste in ports

    International Nuclear Information System (INIS)

    Olson, P.H.

    1994-01-01

    The regulations governing the handling of port-generated waste are often national and/or local legislation, whereas the handling of ship-generated waste is governed by the MARPOL Convention in most parts of the world. The handling of waste consists of two main phases -collection and treatment. Waste has to be collected in every port and on board every ship, whereas generally only some wastes are treated and to a certain degree in ports and on board ships. This paper considers the different kinds of waste generated in both ports and on board ships, where and how it is generated, how it could be collected and treated. The two sources are treated together to show how some ship-generated waste may be treated in port installations primarily constructed for the treatment of the port-generated waste, making integrated use of the available treatment facilities. (author)

  6. Handling and disposing of radioactive waste

    International Nuclear Information System (INIS)

    Trauger, D.B.

    1983-01-01

    Radioactive waste has been separated by definition into six categories. These are: commercial spent fuel; high-level wastes; transuranium waste; low-level wastes; decommissioning and decontamination wastes; and mill tailings and mine wastes. Handling and disposing of these various types of radioactive wastes are discussed briefly

  7. Handling construction waste of building demolition

    Directory of Open Access Journals (Sweden)

    Vondráčková Terezie

    2018-01-01

    Full Text Available Some building defects lead to their demolition. What about construction and demolition waste? According to the Waste Act 185/2001 Coll. and its amendment 223/2015 Coll., which comes into force on January 1, 2017, the production of waste has to be reduced because, as already stated in the amendment to Act No. 229/2014 Coll., the ban on landfilling of waste will apply from 2024 onwards. The main goals of waste management can thus be considered: Preventing or minimizing waste; Waste handling to be used as a secondary raw material - recycling, composting, combustion and the remaining waste to be dumped. Company AZS 98 s. r. o. was established, among other activities, also for the purpose of recycling construction and demolition waste. It operates 12 recycling centers throughout the Czech Republic and therefore we have selected it for a demonstration of the handling of construction and demolition waste in addressing the defects of the buildings.

  8. Certification Plan, low-level waste Hazardous Waste Handling Facility

    International Nuclear Information System (INIS)

    Albert, R.

    1992-01-01

    The purpose of this plan is to describe the organization and methodology for the certification of low-level radioactive waste (LLW) handled in the Hazardous Waste Handling Facility (HWHF) at Lawrence Berkeley Laboratory (LBL). This plan also incorporates the applicable elements of waste reduction, which include both up-front minimization and end-product treatment to reduce the volume and toxicity of the waste; segregation of the waste as it applies to certification; an executive summary of the Waste Management Quality Assurance Implementing Management Plan (QAIMP) for the HWHF and a list of the current and planned implementing procedures used in waste certification. This plan provides guidance from the HWHF to waste generators, waste handlers, and the Waste Certification Specialist to enable them to conduct their activities and carry out their responsibilities in a manner that complies with the requirements of WHC-WAC. Waste generators have the primary responsibility for the proper characterization of LLW. The Waste Certification Specialist verifies and certifies that LBL LLW is characterized, handled, and shipped in accordance with the requirements of WHC-WAC. Certification is the governing process in which LBL personnel conduct their waste generating and waste handling activities in such a manner that the Waste Certification Specialist can verify that the requirements of WHC-WAC are met

  9. Certification plan transuranic waste: Hazardous Waste Handling Facility

    International Nuclear Information System (INIS)

    1992-06-01

    The purpose of this plan is to describe the organization and methodology for the certification of transuranic (TRU) waste handled in the Hazardous Waste Handling Facility at Lawrence Berkeley Laboratory (LBL). The plan incorporates the applicable elements of waste reduction, which include both up-front minimization and end-product treatment to reduce the volume and toxicity of the waste; segregation of the waste as it applies to certification; an executive summary of the Quality Assurance Implementing Management Plan (QAIMP) for the HWBF; and a list of the current and planned implementing procedures used in waste certification

  10. FFTF radioactive solid waste handling and transport

    International Nuclear Information System (INIS)

    Thomson, J.D.

    1982-01-01

    The equipment necessary for the disposal of radioactive solid waste from the Fast Flux Test Facility (FFTF) is scheduled to be available for operation in late 1982. The plan for disposal of radioactive waste from FFTF will utilize special waste containers, a reusable Solid Waste Cask (SWC) and a Disposable Solid Waste Cask (DSWC). The SWC will be used to transport the waste from the Reactor Containment Building to a concrete and steel DSWC. The DSWC will then be transported to a burial site on the Hanford Reservation near Richland, Washington. Radioactive solid waste generated during the operation of the FFTF consists of activated test assembly hardware, reflectors, in-core shim assemblies and control rods. This radioactive waste must be cleaned (sodium removed) prior to disposal. This paper provides a description of the solid waste disposal process, and the casks and equipment used for handling and transport

  11. Waste handling for isotope users

    International Nuclear Information System (INIS)

    1967-01-01

    Aimed at institutes and laboratories involved in the use of radioisotopes, this film emphasizes simple storage and disposal methods but also gives a background of more detailed treatment and final disposal of wastes

  12. Waste handling for isotope users

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1968-12-31

    Aimed at institutes and laboratories involved in the use of radioisotopes, this film emphasizes simple storage and disposal methods but also gives a background of more detailed treatment and final disposal of wastes

  13. Preoperational checkout of the remote-handled transuranic waste handling at the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    1987-09-01

    This plan describes the preoperational checkout for handling Remote-Handled Transuranic (RH-TRU) Wastes from their receipt at the Waste Isolation Pilot Plant (WIPP) to their emplacement underground. This plan identifies the handling operations to be performed, personnel groups responsible for executing these operations, and required equipment items. In addition, this plan describes the quality assurance that will be exercised throughout the checkout, and finally, it establishes criteria by which to measure the success of the checkout. 7 refs., 5 figs

  14. Automated system for handling tritiated mixed waste

    International Nuclear Information System (INIS)

    Dennison, D.K.; Merrill, R.D.; Reitz, T.C.

    1995-03-01

    Lawrence Livermore National Laboratory (LLNL) is developing a semi system for handling, characterizing, processing, sorting, and repackaging hazardous wastes containing tritium. The system combines an IBM-developed gantry robot with a special glove box enclosure designed to protect operators and minimize the potential release of tritium to the atmosphere. All hazardous waste handling and processing will be performed remotely, using the robot in a teleoperational mode for one-of-a-kind functions and in an autonomous mode for repetitive operations. Initially, this system will be used in conjunction with a portable gas system designed to capture any gaseous-phase tritium released into the glove box. This paper presents the objectives of this development program, provides background related to LLNL's robotics and waste handling program, describes the major system components, outlines system operation, and discusses current status and plans

  15. Liberalisation of municipal waste handling

    DEFF Research Database (Denmark)

    Busck, Ole Gunni

    2006-01-01

    for improved performance of municipal waste management. The study stresses the need for training and guidance of municipal administrators. Highlighting ‘best practice’ examples the study shows, however, that it is perfectly possible to end up with quality service on contract. It takes a mixture of careful...... of price reductions in stead of quality demands in both environmental and working environmental terms. A recent study showed major deficits in the capacities of the municipalities to administer qualitative requirements in the tender process and to manage the contracts as an integral part of a scheme...... forces and low quality performance. By assuming responsibility, setting and following up on high quality standards the tender instrument presents an additional instrument to legislation and market based means to institutionalize more sustainable practices in waste management...

  16. 340 Waste Handling Facility interim safety basis

    International Nuclear Information System (INIS)

    Bendixsen, R.B.

    1995-01-01

    This document establishes the interim safety basis (ISB) for the 340 Waste Handling Facility (340 Facility). An ISB is a documented safety basis that provides a justification for the continued operation of the facility until an upgraded final safety analysis report is prepared that complies with US Department of Energy (DOE) Order 5480.23, Nuclear Safety Analysis Reports. The ISB for the 340 Facility documents the current design and operation of the facility. The 340 Facility ISB (ISB-003) is based on a facility walkdown and review of the design and operation of the facility, as described in the existing safety documentation. The safety documents reviewed, to develop ISB-003, include the following: OSD-SW-153-0001, Operating Specification Document for the 340 Waste Handling Facility (WHC 1990); OSR-SW-152-00003, Operating Limits for the 340 Waste Handling Facility (WHC 1989); SD-RE-SAP-013, Safety Analysis Report for Packaging, Railroad Liquid Waste Tank Cars (Mercado 1993); SD-WM-TM-001, Safety Assessment Document for the 340 Waste Handling Facility (Berneski 1994a); SD-WM-SEL-016, 340 Facility Safety Equipment List (Berneski 1992); and 340 Complex Fire Hazard Analysis, Draft (Hughes Assoc. Inc. 1994)

  17. Remote-handled transuranic waste study

    International Nuclear Information System (INIS)

    1995-10-01

    The Waste Isolation Pilot Plant (WIPP) was developed by the US Department of Energy (DOE) as a research and development facility to demonstrate the safe disposal of transuranic (TRU) radioactive wastes generated from the Nation's defense activities. The WIPP disposal inventory will include up to 250,000 cubic feet of TRU wastes classified as remote handled (RH). The remaining inventory will include contact-handled (CH) TRU wastes, which characteristically have less specific activity (radioactivity per unit volume) than the RH-TRU wastes. The WIPP Land Withdrawal Act (LWA), Public Law 102-579, requires a study of the effect of RH-TRU waste on long-term performance. This RH-TRU Waste Study has been conducted to satisfy the requirements defined by the LWA and is considered by the DOE to be a prudent exercise in the compliance certification process of the WIPP repository. The objectives of this study include: conducting an evaluation of the impacts of RH-TRU wastes on the performance assessment (PA) of the repository to determine the effects of Rh-TRU waste as a part of the total WIPP disposal inventory; and conducting a comparison of CH-TRU and RH-TRU wastes to assess the differences and similarities for such issues as gas generation, flammability and explosiveness, solubility, and brine and geochemical interactions. This study was conducted using the data, models, computer codes, and information generated in support of long-term compliance programs, including the WIPP PA. The study is limited in scope to post-closure repository performance and includes an analysis of the issues associated with RH-TRU wastes subsequent to emplacement of these wastes at WIPP in consideration of the current baseline design. 41 refs

  18. 340 waste handling facility interim safety basis

    Energy Technology Data Exchange (ETDEWEB)

    VAIL, T.S.

    1999-04-01

    This document presents an interim safety basis for the 340 Waste Handling Facility classifying the 340 Facility as a Hazard Category 3 facility. The hazard analysis quantifies the operating safety envelop for this facility and demonstrates that the facility can be operated without a significant threat to onsite or offsite people.

  19. 340 waste handling facility interim safety basis

    International Nuclear Information System (INIS)

    VAIL, T.S.

    1999-01-01

    This document presents an interim safety basis for the 340 Waste Handling Facility classifying the 340 Facility as a Hazard Category 3 facility. The hazard analysis quantifies the operating safety envelop for this facility and demonstrates that the facility can be operated without a significant threat to onsite or offsite people

  20. Certification Plan, Radioactive Mixed Waste Hazardous Waste Handling Facility

    International Nuclear Information System (INIS)

    Albert, R.

    1992-01-01

    The purpose of this plan is to describe the organization and methodology for the certification of radioactive mixed waste (RMW) handled in the Hazardous Waste Handling Facility at Lawrence Berkeley Laboratory (LBL). RMW is low-level radioactive waste (LLW) or transuranic (TRU) waste that is co-contaminated with dangerous waste as defined in the Westinghouse Hanford Company (WHC) Solid Waste Acceptance Criteria (WAC) and the Washington State Dangerous Waste Regulations, 173-303-040 (18). This waste is to be transferred to the Hanford Site Central Waste Complex and Burial Grounds in Hanford, Washington. This plan incorporates the applicable elements of waste reduction, which include both up-front minimization and end-product treatment to reduce the volume and toxicity of the waste; segregation of the waste as it applies to certification; an executive summary of the Waste Management Quality Assurance Implementing Management Plan (QAIMP) for the HWHF (Section 4); and a list of the current and planned implementing procedures used in waste certification

  1. Repository waste-handling operations, 1998

    International Nuclear Information System (INIS)

    Cottam, A.E.; Connell, L.

    1986-04-01

    The Civilian Radioactive Waste Management Program Mission Plan and the Generic Requirements for a Mined Geologic Disposal System state that beginning in 1998, commercial spent fuel not exceeding 70,000 metric tons of heavy metal, or a quantity of solidified high-level radioactive waste resulting from the reprocessing of such a quantity of spent fuel, will be shipped to a deep geologic repository for permanent storage. The development of a waste-handling system that can process 3000 metric tons of heavy metal annually will require the adoption of a fully automated approach. The safety and minimum exposure of personnel will be the prime goals of the repository waste handling system. A man-out-of-the-loop approach will be used in all operations including the receipt of spent fuel in shipping casks, the inspection and unloading of the spent fuel into automated hot-cell facilities, the disassembly of spent fuel assemblies, the consolidation of fuel rods, and the packaging of fuel rods into heavy-walled site-specific containers. These containers are designed to contain the radionuclides for up to 1000 years. The ability of a repository to handle more than 6000 pressurized water reactor spent-fuel rods per day on a production basis for approximately a 23-year period will require that a systems approach be adopted that combines space-age technology, robotics, and sophisticated automated computerized equipment. New advanced inspection techniques, maintenance by robots, and safety will be key factors in the design, construction, and licensing of a repository waste-handling facility for 1998

  2. Radioactive waste treatment and handling in France

    International Nuclear Information System (INIS)

    Sivintsev, Yu.V.

    1984-01-01

    Classification of radioactive wastes customary in France and the program of radiation protection in handling them are discussed. Various methods of radioactive waste processing and burial are considered. The French classification of radioactive wastes differs from one used in the other countries. Wastes are classified under three categories: A, B and C. A - low- and intermediate-level radioactive wastes with short-lived radionuclides (half-life - less than 30 years, negligible or heat release, small amount of long-lived radionuclides, especially such as plutonium, americium and neptunium); B - low- and intermediate-level radioactive wastes with long-lived radionuclides (considerable amounts of long-lived radionuclides including α-emitters, low and moderate-level activity of β- and γ-emitters, low and moderate heat release); C - high-level radioactive wastes with long-lived radionuclides (high-level activity of β- and γ-emitters, high heat release, considerable amount of long-lived radionuclides). Volumetric estimations of wastes of various categories and predictions of their growth are given. It is noted that the concept of closed fuel cycle with radiochemical processing of spent fuel is customary in France

  3. System for handling and storing radioactive waste

    Science.gov (United States)

    Anderson, John K.; Lindemann, Paul E.

    1984-01-01

    A system and method for handling and storing spent reactor fuel and other solid radioactive waste, including canisters to contain the elements of solid waste, storage racks to hold a plurality of such canisters, storage bays to store these racks in isolation by means of shielded doors in the bays. This system also includes means for remotely positioning the racks in the bays and an access tunnel within which the remotely operated means is located to position a rack in a selected bay. The modular type of these bays will facilitate the construction of additional bays and access tunnel extension.

  4. Adaptive control of manipulators handling hazardous waste

    International Nuclear Information System (INIS)

    Colbaugh, R.; Glass, K.

    1994-01-01

    This article focuses on developing a robot control system capable of meeting hazardous waste handling application requirements, and presents as a solution an adaptive strategy for controlling the mechanical impedance of kinematically redundant manipulators. The proposed controller is capable of accurate end-effector impedance control and effective redundancy utilization, does not require knowledge of the complex robot dynamic model or parameter values for the robot or the environment, and is implemented without calculation of the robot inverse transformation. Computer simulation results are given for a four degree of freedom redundant robot under adaptive impedance control. These results indicate that the proposed controller is capable of successfully performing important tasks in robotic waste handling applications. (author) 3 figs., 39 refs

  5. Waste Handling in SVAFO's Hot Cell

    International Nuclear Information System (INIS)

    Moeller, Jennifer; Ekenborg, Fredrik; Hellsten, Erik

    2016-01-01

    The decommissioning and dismantling of nuclear installations entails the generation of significant quantities of radioactive waste that must be accepted for disposal. In order to optimise the use of the final repositories for radioactive waste it is important that the waste be sent to the correct repository; that is, that waste containing short-lived radionuclides not be designated as long-lived due to conservative characterisation procedures. The disposal of short-lived waste in a future Swedish repository for long-lived waste will result in increased costs, due to the higher volumetric cost of the disposal as well as costs associated with decades of interim storage before disposal can occur. SVAFO is a non-profit entity that is responsible for the decommissioning of nuclear facilities from historical research and development projects in Sweden. They provide interim storage for radioactive waste arising from research activities until the final repository for long-lived waste is available. SVAFO's offices and facilities are located on the Studsvik site on the east coast of Sweden near the town of Nykoeping. Some of the retired facilities that SVAFO is in the process of decommissioning are located elsewhere in Sweden. The HM facility is a small waste treatment plant owned and operated by SVAFO. The plant processes both liquid and solid radioactive wastes. The facility includes a hot cell equipped with a compactor, a saw and other tools as well as manipulators for the handling and packaging of waste with high dose rates. The cell is fitted with special systems for transporting waste in and passing it out in drums. As with most hot cells there has been an accumulation of surface contamination on the walls, floor and other surfaces during decades of operation. Until recently there has been no attempt to quantify or characterize this contamination. Current practices dictate that after waste is handled in the hot cell it is conservatively designated as long

  6. WASTE HANDLING BUILDING SHIELD WALL ANALYSIS

    International Nuclear Information System (INIS)

    Padula, D.

    2000-01-01

    The scope of this analysis is to estimate the shielding wall, ceiling or equivalent door thicknesses that will be required in the Waste Handling Building to maintain the radiation doses to personnel within acceptable limits. The shielding thickness calculated is the minimum required to meet administrative limits, and not necessarily what will be recommended for the final design. The preliminary evaluations will identify the areas which have the greatest impact on mechanical and facility design concepts. The objective is to provide the design teams with the necessary information to assure an efficient and effective design

  7. Handling of tritium-bearing wastes

    International Nuclear Information System (INIS)

    1981-01-01

    The generation of nuclear power and reprocessing of nuclear fuel results in the production of tritium and the possible need to control the release of tritium-contaminated effluents. In assessing the need for controls, it is necessary to know the production rates of tritium at different nuclear facilities, the technologies available for separating tritium from different gaseous and liquid streams, and the methods that are satisfactory for storage and disposal of tritiated wastes. The intention in applying such control technologies and methods is to avoid undesirable effects on the environment, and to reduce the radiation burden on operational personnel and the general population. This technical report is a result of the IAEA Technical Committee Meeting on Handling of Tritium-bearing Effluents and Wastes, which was held in Vienna, 4 - 8 December 1978. It summarizes the main topics discussed at the meeting and appends the more detailed reports on particular aspects that were prepared for the meeting by individual participants

  8. Waste Handling Equipment Development Test and Evaluation Study

    International Nuclear Information System (INIS)

    R.L. Tome

    1998-01-01

    The purpose of this study is to identify candidate Monitored Geologic Repository (MGR) surface waste handling equipment for development testing. This study will also identify strategies for performing the development tests. Development testing shall be implemented to support detail design and reduce design risks. Development testing shall be conducted to confirm design concepts, evaluate alternative design concepts, show the availability of needed technology, and provide design documentation. The candidate equipment will be selected from MGR surface waste handling equipment that is the responsibility of the Management and Operating Contractor (M and O) Surface Design Department. The equipment identified in this study is based on Viability Assessment (VA) design. The ''Monitored Geologic Repository Test and Evaluation Plan'' (MGR T and EP), Reference 5.1, was used as a basis for this study. The MGR T and EP reflects the extent of test planning and analysis that can be conducted, given the current status of the MGR requirements and latest VA design information. The MGR T and EP supports the appropriate sections in the license application (LA) in accordance with 10 CFR 60.2 1(c)(14). The MGR T and EP describes the following test activities: site characterization to confirm, by test and analysis, the suitability of the Yucca Mountain site for housing a geologic repository; development testing to investigate and document design concepts to reduce risk; qualification testing to verify equipment compliance with design requirements, specifications, and regulatory requirements; system testing to validate compliance with MGR requirements, which include the receipt, handling, retrieval, and disposal of waste; periodic performance testing to verify preclosure requirements and to demonstrate safe and reliable MGR operation; and performance confirmation modeling, testing, and analysis to verify adherence to postclosure regulatory requirements. Development test activities can be

  9. Remote waste handling and feed preparation for Mixed Waste Management

    International Nuclear Information System (INIS)

    Couture, S.A.; Merrill, R.D.; Densley, P.J.

    1995-05-01

    The Mixed Waste Management Facility (MWMF) at the Lawrence Livermore National Laboratory (LLNL) will serve as a national testbed to demonstrate mature mixed waste handling and treatment technologies in a complete front-end to back-end --facility (1). Remote operations, modular processing units and telerobotics for initial waste characterization, sorting and feed preparation have been demonstrated at the bench scale and have been selected for demonstration in MWMF. The goal of the Feed Preparation design team was to design and deploy a robust system that meets the initial waste preparation flexibility and productivity needs while providing a smooth upgrade path to incorporate technology advances as they occur. The selection of telerobotics for remote handling in MWMF was made based on a number of factors -- personnel protection, waste generation, maturity, cost, flexibility and extendibility. Modular processing units were selected to enable processing flexibility and facilitate reconfiguration as new treatment processes or waste streams are brought on line for demonstration. Modularity will be achieved through standard interfaces for mechanical attachment as well as process utilities, feeds and effluents. This will facilitate reconfiguration of contaminated systems without drilling, cutting or welding of contaminated materials and with a minimum of operator contact. Modular interfaces also provide a standard connection and disconnection method that can be engineered to allow convenient remote operation

  10. Handling and treatment of radioactive aqueous wastes

    International Nuclear Information System (INIS)

    1992-07-01

    This report aims to provide essential guidance to developing Member States without a nuclear power programme regarding selection, design and operation of cost effective treatment processes for radioactive aqueous liquids arising as effluents from small research institutions, hospitals and industries. The restricted quantities and low activity associated with the relevant wastes will generally permit contact-handling and avoid the need for shielding requirements. The selection of liquid waste treatment involves: Characterization of arising with the possibility of segregation; Discharge requirements for decontaminated liquors, both radioactive and non-radioactive; Available technologies and costs; Conditioning of the concentrates resulting from the treatment; Storage and disposal of the conditioned concentrates. The report will serve as a technical manual providing reference material and direct step-by-step know-how to staff in radioisotope user establishments and research centres in the developing Member States without nuclear power generation. Therefore, emphasis is limited to the simpler treatment facilities, which will be included with only the robust, well-established waste management processes carefully chosen as appropriate to developing countries. 20 refs, 12 figs, 7 tabs

  11. Handling nuclear waste over long periods

    International Nuclear Information System (INIS)

    Ancelin, B.; Chenevier, E.

    1983-01-01

    The handling of nuclear waste over long periods throws up new problems, such as the safety for a very long term and the employment of economic logic in order to justify choices involving extended time scales. The result is a very great difficulty of apprehension of the problem by the specialists as well as by the public. A clear policy decision, associated with a coherent administrative organization, will therefore have to make up for an impossible technical-economical optimization of the various possible options. The difficulty of simple technical choices is only going to reinforce this wish; the absence of a global and comparative measuring system is responsible for the fact that in this field the passions often override many of the scientific truths [fr

  12. WASTE HANDLING BUILDING FIRE PROTECTION SYSTEM DESCRIPTION DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    J. D. Bigbee

    2000-06-21

    The Waste Handling Building Fire Protection System provides the capability to detect, control, and extinguish fires and/or mitigate explosions throughout the Waste Handling Building (WHB). Fire protection includes appropriate water-based and non-water-based suppression, as appropriate, and includes the distribution and delivery systems for the fire suppression agents. The Waste Handling Building Fire Protection System includes fire or explosion detection panel(s) controlling various detectors, system actuation, annunciators, equipment controls, and signal outputs. The system interfaces with the Waste Handling Building System for mounting of fire protection equipment and components, location of fire suppression equipment, suppression agent runoff, and locating fire rated barriers. The system interfaces with the Waste Handling Building System for adequate drainage and removal capabilities of liquid runoff resulting from fire protection discharges. The system interfaces with the Waste Handling Building Electrical Distribution System for power to operate, and with the Site Fire Protection System for fire protection water supply to automatic sprinklers, standpipes, and hose stations. The system interfaces with the Site Fire Protection System for fire signal transmission outside the WHB as needed to respond to a fire emergency, and with the Waste Handling Building Ventilation System to detect smoke and fire in specific areas, to protect building high-efficiency particulate air (HEPA) filters, and to control portions of the Waste Handling Building Ventilation System for smoke management and manual override capability. The system interfaces with the Monitored Geologic Repository (MGR) Operations Monitoring and Control System for annunciation, and condition status.

  13. WASTE HANDLING BUILDING FIRE PROTECTION SYSTEM DESCRIPTION DOCUMENT

    International Nuclear Information System (INIS)

    J. D. Bigbee

    2000-01-01

    The Waste Handling Building Fire Protection System provides the capability to detect, control, and extinguish fires and/or mitigate explosions throughout the Waste Handling Building (WHB). Fire protection includes appropriate water-based and non-water-based suppression, as appropriate, and includes the distribution and delivery systems for the fire suppression agents. The Waste Handling Building Fire Protection System includes fire or explosion detection panel(s) controlling various detectors, system actuation, annunciators, equipment controls, and signal outputs. The system interfaces with the Waste Handling Building System for mounting of fire protection equipment and components, location of fire suppression equipment, suppression agent runoff, and locating fire rated barriers. The system interfaces with the Waste Handling Building System for adequate drainage and removal capabilities of liquid runoff resulting from fire protection discharges. The system interfaces with the Waste Handling Building Electrical Distribution System for power to operate, and with the Site Fire Protection System for fire protection water supply to automatic sprinklers, standpipes, and hose stations. The system interfaces with the Site Fire Protection System for fire signal transmission outside the WHB as needed to respond to a fire emergency, and with the Waste Handling Building Ventilation System to detect smoke and fire in specific areas, to protect building high-efficiency particulate air (HEPA) filters, and to control portions of the Waste Handling Building Ventilation System for smoke management and manual override capability. The system interfaces with the Monitored Geologic Repository (MGR) Operations Monitoring and Control System for annunciation, and condition status

  14. Waste Handling Shaft concrete liner degradation conclusions and recommendations

    International Nuclear Information System (INIS)

    1992-10-01

    The primary function of the Waste Handling Shaft (WHS) at the Waste Isolation Pilot Plant (WIPP) is to permit the transfer of radioactive waste from the surface waste handling building to the underground storage area. It also serves as an intake shaft for small volumes of air during normal storage operations and as an emergency escape route. Part of the construction was the placement of a concrete liner and steel reinforced key in 1984. During a routine shaft inspection in May 1990, some degradation of the WHS concrete liner was observed between the depths of 800 and 900 feet below the ground surface. Detailed investigations of the liner had been carried out by Sandia National Laboratories and by Westinghouse Electric Corporation Waste Isolation Division (WID) through Lankard Materials Laboratory. Observations, reports, and data support the conclusion that the concrete degradation, resulting from attack by chemically aggressive brine, is a localized phenomena. It is the opinion of the WID that the degradation is not considered an immediate or near term concern; this is supported by technical experts. WID recommendations have been made which, when implemented, will ensure an extended liner life. Based on the current assessment of available data and the proposed shaft liner monitoring program described in this report, it is reasonable to assume that the operational life of the concrete shaft liner can safely support the 25-year life of the WIPP. Analysis of data indicates that degradation of the shaft's concrete liner is attributed to chemically aggressive brine seeping through construction joints and shrinkage cracks from behind the liner in and around the 834-foot depth. Chemical and mechanical components of concrete degradation have been identified. Chemical attack is comprised of several stages of concrete alteration. The other component, mechanical degradation, results from the expansive forces of crystals forming in the concrete pore space

  15. Full scale tests on remote handled FFTF fuel assembly waste handling and packaging

    International Nuclear Information System (INIS)

    Allen, C.R.; Cash, R.J.; Dawson, S.A.; Strode, J.N.

    1986-01-01

    Handling and packaging of remote handled, high activity solid waste fuel assembly hardware components from spent FFTF reactor fuel assemblies have been evaluated using full scale components. The demonstration was performed using FFTF fuel assembly components and simulated components which were handled remotely using electromechanical manipulators, shielding walls, master slave manipulators, specially designed grapples, and remote TV viewing. The testing and evaluation included handling, packaging for current and conceptual shipping containers, and the effects of volume reduction on packing efficiency and shielding requirements. Effects of waste segregation into transuranic (TRU) and non-transuranic fractions also are discussed

  16. Handling, treating and injecting of oilfield wastes

    International Nuclear Information System (INIS)

    Pippard, D.K.

    1997-01-01

    The waste management practices of the oil and gas industry in British Columbia were reviewed. Newalta is a waste management company that offers services to both the upstream (oilfield) and downstream (refined products) petroleum industries. The measures that this company has taken to comply with the new regulations in British Columbia were discussed. Issues regarding deep well disposal, oil and gas waste regulation, and liquid waste streams not authorized for disposal, were addressed. Oil and gas production waste liquids generated in British Columbia can be transported into Alberta for treatment and disposal under Alberta's hazardous wastes disposal legislation. The treatment of crude oil wastes, oilfield waste solids were also addressed. Solid wastes can be disposed of by in-situ treatment, by ex-situ, on-site treatment such as biodegradation and thermal treatment, and by ex-situ, off-site treatment

  17. Handling and storage of conditioned high-level wastes

    International Nuclear Information System (INIS)

    1983-01-01

    This report deals with certain aspects of the management of one of the most important wastes, i.e. the handling and storage of conditioned (immobilized and packaged) high-level waste from the reprocessing of spent nuclear fuel and, although much of the material presented here is based on information concerning high-level waste from reprocessing LWR fuel, the principles, as well as many of the details involved, are applicable to all fuel types. The report provides illustrative background material on the arising and characteristics of high-level wastes and, qualitatively, their requirements for conditioning. The report introduces the principles important in conditioned high-level waste storage and describes the types of equipment and facilities, used or studied, for handling and storage of such waste. Finally, it discusses the safety and economic aspects that are considered in the design and operation of handling and storage facilities

  18. Certification document for newly generated contact-handled transuranic waste

    International Nuclear Information System (INIS)

    Box, W.D.; Setaro, J.

    1984-01-01

    The US Department of Energy has requested that all national laboratories handling defense waste develop and augment a program whereby all newly generated contact-handled transuranic (TRU) waste be contained, stored, and then shipped to the Waste Isolation Pilot Plant (WIPP) in accordance with the requirements set forth in WIPP-DOE-114. The program described in this report delineates how Oak Ridge National Laboratory intends to comply with these requirements and lists the procedures used by each generator to ensure that their TRU wastes are certifiable for shipment to WIPP

  19. Centralized processing of contact-handled TRU waste feasibility analysis

    International Nuclear Information System (INIS)

    1986-12-01

    This report presents work for the feasibility study of central processing of contact-handled TRU waste. Discussion of scenarios, transportation options, summary of cost estimates, and institutional issues are a few of the subjects discussed

  20. Management of remote-handled defense transuranic wastes

    International Nuclear Information System (INIS)

    Ebra, M.A.; Pierce, G.D.; Carson, P.H.

    1988-01-01

    Transuranic (TRU) wastes generated by defense-related activities are scheduled for emplacement at the Waste Isolation Pilot Plant (WIPP) in New Mexico beginning in October 1988. After five years of operation as a research and development facility, the WIPP may be designated as a permanent repository for these wastes, if it has been demonstrated that this deep, geologically stable formation is a safe disposal option. Defense TRU wastes are currently stored at various Department of Energy (DOE) sites across the nation. Approximately 2% by volume of currently stored TRU wastes are defined, on the basis of dose rates, as remote-handled (RH). RH wastes continue to be generated at various locations operated by DOE contractors. They require special handling and processing prior to and during emplacement in the WIPP. This paper describes the strategy for managing defense RH TRU wastes

  1. Unresolved issues for the disposal of remote-handled transuranic waste in the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Silva, M.K.; Neill, R.H.

    1994-09-01

    The purpose of the Waste Isolation Pilot Plant (WIPP) is to dispose of 176,000 cubic meters of transuranic (TRU) waste generated by the defense activities of the US Government. The envisioned inventory contains approximately 6 million cubic feet of contact-handled transuranic (CH TRU) waste and 250,000 cubic feet of remote handled transuranic (RH TRU) waste. CH TRU emits less than 0.2 rem/hr at the container surface. Of the 250,000 cubic feet of RH TRU waste, 5% by volume can emit up to 1,000 rem/hr at the container surface. The remainder of RH TRU waste must emit less than 100 rem/hr. These are major unresolved problems with the intended disposal of RH TRU waste in the WIPP. (1) The WIPP design requires the canisters of RH TRU waste to be emplaced in the walls (ribs) of each repository room. Each room will then be filled with drums of CH TRU waste. However, the RH TRU waste will not be available for shipment and disposal until after several rooms have already been filled with drums of CH TRU waste. RH TRU disposal capacity will be loss for each room that is first filled with CH TRU waste. (2) Complete RH TRU waste characterization data will not be available for performance assessment because the facilities needed for waste handling, waste treatment, waste packaging, and waste characterization do not yet exist. (3) The DOE does not have a transportation cask for RH TRU waste certified by the US Nuclear Regulatory Commission (NRC). These issues are discussed along with possible solutions and consequences from these solutions. 46 refs

  2. Remote automated material handling of radioactive waste containers

    International Nuclear Information System (INIS)

    Greager, T.M.

    1994-09-01

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

  3. The handling and disposal of fusion wastes

    International Nuclear Information System (INIS)

    Broden, K.; Hultgren, Aa.; Olsson, G.

    1985-02-01

    The radioactive wastes from fusion reactor operation will include spent components, wastes from repair operations, and decontamination waste. Various disposal routes may be considered depending on i.a. the contents of tritium and of long-lived nuclides, and on national regulations. The management philosophy and disposal technology developed in Sweden for light water reactor wastes has been studied at STUDSVIK during 1983--84 and found to be applicable also to fusion wastes, provided a detritiation stage is included. These studies will continue during 1985 and include experimental work on selected fusion activation nuclides. The work presented is associated to the CEC fusion research programme. Valuable discussions and contacts with people working in this programme at Saclay, Ispra and Garching are deeply appreciated. (author)

  4. Advanced robotics handling and controls applied to Mixed Waste characterization, segregation and treatment

    International Nuclear Information System (INIS)

    Grasz, E.; Huber, L.; Horvath, J.; Roberson, P.; Wilhelmsen, K.; Ryon, R.

    1994-11-01

    At Lawrence Livermore National Laboratory under the Mixed Waste Operations program of the Department of Energy Robotic Technology Development Program (RTDP), a key emphasis is developing a total solution to the problem of characterizing, handling and treating complex and potentially unknown mixed waste objects. LLNL has been successful at looking at the problem from a system perspective and addressing some of the key issues including non-destructive evaluation of the waste stream prior to the materials entering the handling workcell, the level of automated material handling required for effective processing of the waste stream objects (both autonomous and tele-operational), and the required intelligent robotic control to carry out the characterization, segregation, and waste treating processes. These technologies were integrated and demonstrated in a prototypical surface decontamination workcell this past year

  5. SITE GENERATED RADIOLOGICAL WASTE HANDLING SYSTEM DESCRIPTION DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    S. C. Khamankar

    2000-06-20

    The Site Generated Radiological Waste Handling System handles radioactive waste products that are generated at the geologic repository operations area. The waste is collected, treated if required, packaged for shipment, and shipped to a disposal site. Waste streams include low-level waste (LLW) in solid and liquid forms, as-well-as mixed waste that contains hazardous and radioactive constituents. Liquid LLW is segregated into two streams, non-recyclable and recyclable. The non-recyclable stream may contain detergents or other non-hazardous cleaning agents and is packaged for shipment. The recyclable stream is treated to recycle a large portion of the water while the remaining concentrated waste is packaged for shipment; this greatly reduces the volume of waste requiring disposal. There will be no liquid LLW discharge. Solid LLW consists of wet solids such as ion exchange resins and filter cartridges, as-well-as dry active waste such as tools, protective clothing, and poly bags. Solids will be sorted, volume reduced, and packaged for shipment. The generation of mixed waste at the Monitored Geologic Repository (MGR) is not planned; however, if it does come into existence, it will be collected and packaged for disposal at its point of occurrence, temporarily staged, then shipped to government-approved off-site facilities for disposal. The Site Generated Radiological Waste Handling System has equipment located in both the Waste Treatment Building (WTB) and in the Waste Handling Building (WHB). All types of liquid and solid LLW are processed in the WTB, while wet solid waste from the Pool Water Treatment and Cooling System is packaged where received in the WHB. There is no installed hardware for mixed waste. The Site Generated Radiological Waste Handling System receives waste from locations where water is used for decontamination functions. In most cases the water is piped back to the WTB for processing. The WTB and WHB provide staging areas for storing and shipping LLW

  6. SITE GENERATED RADIOLOGICAL WASTE HANDLING SYSTEM DESCRIPTION DOCUMENT

    International Nuclear Information System (INIS)

    S. C. Khamankar

    2000-01-01

    The Site Generated Radiological Waste Handling System handles radioactive waste products that are generated at the geologic repository operations area. The waste is collected, treated if required, packaged for shipment, and shipped to a disposal site. Waste streams include low-level waste (LLW) in solid and liquid forms, as-well-as mixed waste that contains hazardous and radioactive constituents. Liquid LLW is segregated into two streams, non-recyclable and recyclable. The non-recyclable stream may contain detergents or other non-hazardous cleaning agents and is packaged for shipment. The recyclable stream is treated to recycle a large portion of the water while the remaining concentrated waste is packaged for shipment; this greatly reduces the volume of waste requiring disposal. There will be no liquid LLW discharge. Solid LLW consists of wet solids such as ion exchange resins and filter cartridges, as-well-as dry active waste such as tools, protective clothing, and poly bags. Solids will be sorted, volume reduced, and packaged for shipment. The generation of mixed waste at the Monitored Geologic Repository (MGR) is not planned; however, if it does come into existence, it will be collected and packaged for disposal at its point of occurrence, temporarily staged, then shipped to government-approved off-site facilities for disposal. The Site Generated Radiological Waste Handling System has equipment located in both the Waste Treatment Building (WTB) and in the Waste Handling Building (WHB). All types of liquid and solid LLW are processed in the WTB, while wet solid waste from the Pool Water Treatment and Cooling System is packaged where received in the WHB. There is no installed hardware for mixed waste. The Site Generated Radiological Waste Handling System receives waste from locations where water is used for decontamination functions. In most cases the water is piped back to the WTB for processing. The WTB and WHB provide staging areas for storing and shipping LLW

  7. Radioactive wastes handling problems in Venezuela

    International Nuclear Information System (INIS)

    Ramirez, R.; Venegas, R.

    1984-07-01

    A brief description of the radioactive wastes problem in Venezuela is presented. The origins of the problem are shown in a squematic form. The requirements for its solution are divided into three parts: information system, control system, radioactive wastes hadling system. A questionnaire summarizing factors to be considered when looking for a solution to the problem in Venezuela is included, as well as conclusions and recomendations for further discussion

  8. Solid waste handling and decontamination facility

    International Nuclear Information System (INIS)

    Lampton, R.E.

    1979-01-01

    The Title 1 design of the decontamination part of the SWH and D facility is underway. Design criteria are listed. A flowsheet is given of the solid waste reduction. The incinerator scrubber is described. Design features of the Gunite Tank Sludge Removal and a schematic of the sluicer, TV camera, and recirculating system are given. 9 figures

  9. Remote-Handled Low Level Waste Disposal Project Alternatives Analysis

    Energy Technology Data Exchange (ETDEWEB)

    David Duncan

    2010-10-01

    This report identifies, evaluates, and compares alternatives for meeting the U.S. Department of Energy’s mission need for management of remote-handled low-level waste generated by the Idaho National Laboratory and its tenants. Each alternative identified in the Mission Need Statement for the Remote-Handled Low-Level Waste Treatment Project is described and evaluated for capability to fulfill the mission need. Alternatives that could meet the mission need are further evaluated and compared using criteria of cost, risk, complexity, stakeholder values, and regulatory compliance. The alternative for disposal of remote-handled low-level waste that has the highest confidence of meeting the mission need and represents best value to the government is to build a new disposal facility at the Idaho National Laboratory Site.

  10. An analysis of repository waste-handling operations

    International Nuclear Information System (INIS)

    Dennis, A.W.

    1990-09-01

    This report has been prepared to document the operational analysis of waste-handling facilities at a geologic repository for high-level nuclear waste. The site currently under investigation for the geologic repository is located at Yucca Mountain, Nye County, Nevada. The repository waste-handling operations have been identified and analyzed for the year 2011, a steady-state year during which the repository receives spent nuclear fuel containing the equivalent of 3000 metric tons of uranium (MTU) and defense high-level waste containing the equivalent of 400 MTU. As a result of this analysis, it has been determined that the waste-handling facilities are adequate to receive, prepare, store, and emplace the projected quantity of waste on an annual basis. In addition, several areas have been identified where additional work is required. The recommendations for future work have been divided into three categories: items that affect the total waste management system, operations within the repository boundary, and the methodology used to perform operational analyses for repository designs. 7 refs., 48 figs., 11 tabs

  11. Uncertainty Regarding Waste Handling in Everyday Life

    Directory of Open Access Journals (Sweden)

    Susanne Ewert

    2010-09-01

    Full Text Available According to our study, based on interviews with households in a residential area in Sweden, uncertainty is a cultural barrier to improved recycling. Four causes of uncertainty are identified. Firstly, professional categories not matching cultural categories—people easily discriminate between certain categories (e.g., materials such as plastic and paper but not between others (e.g., packaging and “non-packaging”. Thus a frequent cause of uncertainty is that the basic categories of the waste recycling system do not coincide with the basic categories used in everyday life. Challenged habits—source separation in everyday life is habitual, but when a habit is challenged, by a particular element or feature of the waste system, uncertainty can arise. Lacking fractions—some kinds of items cannot be left for recycling and this makes waste collection incomplete from the user’s point of view and in turn lowers the credibility of the system. Missing or contradictory rules of thumb—the above causes seem to be particularly relevant if no motivating principle or rule of thumb (within the context of use is successfully conveyed to the user. This paper discusses how reducing uncertainty can improve recycling.

  12. Handling and Treatment of Poultry Hatchery Waste: A Review

    Directory of Open Access Journals (Sweden)

    Belinda Rodda

    2011-01-01

    Full Text Available A literature review was undertaken to identify methods being used to handle and treat hatchery waste. Hatchery waste can be separated into solid waste and liquid waste by centrifuging or by using screens. Potential methods for treating hatchery waste on site include use of a furnace to heat the waste to produce steam to run a turbine generator or to use an in line composter to stabilise the waste. There is also potential to use anaerobic digestion at hatcheries to produce methane and fertilisers. Hatcheries disposing wastewater into lagoons could establish a series of ponds where algae, zooplankton and fish utilise the nutrients using integrated aquaculture which cleans the water making it more suitable for irrigation. The ideal system to establish in a hatchery would be to incorporate separation and handling equipment to separate waste into its various components for further treatment. This would save disposal costs, produce biogas to reduce power costs at plants and produce a range of value added products. However the scale of operations at many hatcheries is too small and development of treatment systems may not be viable.

  13. Overview of DOE LLWMP waste treatment, packaging, and handling activities

    International Nuclear Information System (INIS)

    Pechin, W.H.

    1982-01-01

    The program objective is to develop the best available technology for waste treatment, packaging, and handling to meet the needs of shallow land burial disposal and for greater confinement than shallow land burial. The program has reviewed many of the hardware options for appropriate usage with low-level waste, but promising options remain to be evaluated. The testing of treatment technologies with actual radioactive process wastes has been initiated. The analysis of the interaction of treatment, solidification and disposal needs to be completed

  14. Mixed waste certification plan for the Lawrence Berkeley Laboratory Hazardous Waste Handling Facility. Revision 1

    International Nuclear Information System (INIS)

    1995-01-01

    The purpose of this plan is to describe the organization and methodology for the certification of mixed waste handled in the Hazardous Waste Handling Facility (HWHF) at Lawrence Berkeley Laboratory (LBL). This plan is composed to meet the requirements found in the Westinghouse Hanford Company (WHC) Solid Waste Acceptance Criteria (WAC) and follows the suggested outline provided by WHC in the letter of April 26, 1990, to Dr. R.H. Thomas, Occupational Health Division, LBL. Mixed waste is to be transferred to the WHC Hanford Site Central Waste Complex and Burial Grounds in Hanford, Washington

  15. Handling and processing of radioactive waste from nuclear applications

    International Nuclear Information System (INIS)

    2001-01-01

    The main objective of this report is to provide technical information and reference material on different steps and components of radioactive waste management for staff in establishments that use radionuclides and in research centres in Member States. It provides technical information on the safe handling, treatment, conditioning and storage of waste arising from the various activities associated with the production and application of radioisotopes in medical, industrial, educational and research facilities. The technical information cited in this report consists mainly of processes that are commercialised or readily available, and can easily be applied as they are or modified to solve specific waste management requirements. This report covers the sources and characteristics of waste and approaches to waste classification, and describes the particular processing steps from pretreatment until storage of conditioned packages

  16. Handling and storage of conditioned high-level wastes

    International Nuclear Information System (INIS)

    Heafield, W.

    1984-01-01

    This paper deals with certain aspects of the management of one of the most important radioactive wastes arising from the nuclear fuel cycle, i.e. the handling and storage of conditioned high-level wastes. The paper is based on an IAEA report of the same title published during 1983 in the Technical Reports Series. The paper provides illustrative background material on the characteristics of high-level wastes and, qualitatively, their requirements for conditioning. The principles important in the storage of high-level wastes are reviewed in conjunction with the radiological and socio-political considerations involved. Four fundamentally different storage concepts are described with reference to published information and the safety aspects of particular storage concepts are discussed. Finally, overall conclusions are presented which confirm the availability of technology for constructing and operating conditioned high-level waste storage facilities for periods of at least several decades. (author)

  17. Plans for Managing Hanford Remote Handled Transuranic (TRU) Waste

    International Nuclear Information System (INIS)

    MCKENNEY, D.E.

    2001-01-01

    The current Hanford Site baseline and life-cycle waste forecast predicts that approximately 1,000 cubic meters of remote-handled transuranic (RH-TRU) waste will be generated by waste management and environmental restoration activities at Hanford. These 1,000 cubic meters, comprised of both transuranic and mixed transuranic (TRUM) waste, represent a significant portion of the total estimated inventory of RH-TRU to be disposed of at the Waste Isolation Pilot Plant (WIPP). A systems engineering approach is being followed to develop a disposition plan for each RH-TRU/TRUM waste stream at Hanford. A number of significant decision-making efforts are underway to develop and finalize these disposition plans, including: development and approval of a RH-TRU/TRUM Waste Project Management Plan, revision of the Hanford Waste Management Strategic Plan, the Hanford Site Options Study (''Vision 2012''), the Canyon Disposal Initiative Record-of-Decision, and the Hanford Site Solid (Radioactive and Hazardous) Waste Program Environmental Impact Statement (SW-EIS). Disposition plans may include variations of several options, including (1) sending most RH-TRU/TRUM wastes to WIPP, (2) deferrals of waste disposal decisions in the interest of both efficiency and integration with other planned decision dates and (3) disposition of some materials in place consistent with Department of Energy Orders and the regulations in the interest of safety, risk minimization, and cost. Although finalization of disposition paths must await completion of the aforementioned decision documents, significant activities in support of RH-TRU/TRUM waste disposition are proceeding, including Hanford participation in development of the RH TRU WIPP waste acceptance criteria, preparation of T Plant for interim storage of spent nuclear fuel sludge, sharing of technology information and development activities in cooperation with the Mixed Waste Focus Area, RH-TRU technology demonstrations and deployments, and

  18. Defense Remote Handled Transuranic Waste Cost/Schedule Optimization Study

    International Nuclear Information System (INIS)

    Pierce, G.D.; Wolaver, R.W.; Carson, P.H.

    1986-11-01

    The purpose of this study is to provide the DOE information with which it can establish the most efficient program for the long management and disposal, in the Waste Isolation Pilot Plant (WIPP), of remote handled (RH) transuranic (TRU) waste. To fulfill this purpose, a comprehensive review of waste characteristics, existing and projected waste inventories, processing and transportation options, and WIPP requirements was made. Cost differences between waste management alternatives were analyzed and compared to an established baseline. The result of this study is an information package that DOE can use as the basis for policy decisions. As part of this study, a comprehensive list of alternatives for each element of the baseline was developed and reviewed with the sites. The principle conclusions of the study follow. A single processing facility for RH TRU waste is both necessary and sufficient. The RH TRU processing facility should be located at Oak Ridge National Laboratory (ORNL). Shielding of RH TRU to contact handled levels is not an economic alternative in general, but is an acceptable alternative for specific waste streams. Compaction is only cost effective at the ORNL processing facility, with a possible exception at Hanford for small compaction of paint cans of newly generated glovebox waste. It is more cost effective to ship certified waste to WIPP in 55-gal drums than in canisters, assuming a suitable drum cask becomes available. Some waste forms cannot be packaged in drums, a canister/shielded cask capability is also required. To achieve the desired disposal rate, the ORNL processing facility must be operational by 1996. Implementing the conclusions of this study can save approximately $110 million, compared to the baseline, in facility, transportation, and interim storage costs through the year 2013. 10 figs., 28 tabs

  19. Low-level radioactive wastes: Their treatment, handling, disposal

    Energy Technology Data Exchange (ETDEWEB)

    Straub, Conrad P [Robert A. Taft Sanitary Engineering Center, Radiological Health Research Activities, Cincinnati, OH(United States)

    1964-07-01

    The release of low level wastes may result in some radiation exposure to man and his surroundings. This book describes techniques of handling, treatment, and disposal of low-level wastes aimed at keeping radiation exposure to a practicable minimum. In this context, wastes are considered low level if they are released into the environment without subsequent control. This book is concerned with practices relating only to continuous operations and not to accidental releases of radioactive materials. It is written by use for those interested in low level waste disposal problems and particularly for the health physicist concerned with these problems in the field. It should be helpful also to water and sewage works personnel concerned with the efficiency of water and sewage treatment processes for the removal of radioactive materials; the personnel engaged in design, construction, licensing, and operation of treatment facilities; and to student of nuclear technology. After an introduction the following areas are discussed: sources, quantities and composition of radioactive wastes; collection, sampling and measurement; direct discharge to the water, soil and air environment; air cleaning; removal of radioactivity by water-treatment processes and biological processes; treatment on site by chemical precipitation , ion exchange and absorption, electrodialysis, solvent extraction and other methods; treatment on site including evaporation and storage; handling and treatment of solid wastes; public health implications. Appendices include a glossary; standards for protection against radiation; federal radiation council radiation protection guidance for federal agencies; site selection criteria for nuclear energy facilities.

  20. Hazardous Waste Cerification Plan: Hazardous Waste Handling Facility, Lawrence Berkeley Laboratory

    International Nuclear Information System (INIS)

    1992-02-01

    The purpose of this plan is to describe the organization and methodology for the certification of hazardous waste (HW) handled in the Lawrence Berkeley Laboratory (LBL) Hazardous Waste Handling Facility (HWHF). The plan also incorporates the applicable elements of waste reduction, which include both up-front minimization and end- product treatment to reduce the volume and toxicity of the waste; segregation of the waste as it applies to certification; and executive summary of the Quality Assurance Program Plan (QAPP) for the HWHF and a list of the current and planned implementing procedures used in waste certification. The plan provides guidance from the HWHF to waste generators, waste handlers, and the Systems Group Manager to enable them to conduct their activities and carry out their responsibilities in a manner that complies with several requirements of the Federal Resource Conservation and Resource Recovery Act (RCRA), the Federal Department of Transportation (DOT), and the State of California, Code of Regulations (CCR), Title 22

  1. Radioactive and mixed waste management plan for the Lawrence Berkeley Laboratory Hazardous Waste Handling Facility

    International Nuclear Information System (INIS)

    1995-01-01

    This Radioactive and Mixed Waste Management Plan for the Hazardous Waste Handling Facility at Lawrence Berkeley Laboratory is written to meet the requirements for an annual report of radioactive and mixed waste management activities outlined in DOE Order 5820.2A. Radioactive and mixed waste management activities during FY 1994 listed here include principal regulatory and environmental issues and the degree to which planned activities were accomplished

  2. CLASSIFICATION OF THE MGR WASTE HANDLING BUILDING ELECTRICAL SYSTEM

    International Nuclear Information System (INIS)

    S.E. Salzman

    1999-01-01

    The purpose of this analysis is to document the Quality Assurance (QA) classification of the Monitored Geologic Repository (MGR) waste handling building electrical system structures, systems and components (SSCs) performed by the MGR Safety Assurance Department. This analysis also provides the basis for revision of YMP/90-55Q, Q-List (YMP 1998). The Q-List identifies those MGR SSCs subject to the requirements of DOE/RW-0333P, ''Quality Assurance Requirements and Description'' (QARD) (DOE 1998)

  3. Transportation system (TRUPACT) for contact-handled transuranic wastes

    International Nuclear Information System (INIS)

    Romesberg, L.E.; Pope, R.B.; Burgoyne, R.M.

    1982-04-01

    Contact-handled transuranic defense waste is being, and will continue to be, moved between a number of locations in the United States. The DOE is sponsoring development of safe, efficient, licensable, and cost-effective transportation systems to handle this waste. The systems being developed have been named TRUPACT which stands for TRansUranic PACkage Transporter. The system will be compatible with Type A packagings used by waste generators, interim storage facilities, and repositories. TRUPACT is required to be a Type B packaging since larger than Type A quantities of some radionuclides (particularly plutonium) may be involved in the collection of Type A packagings. TRUPACT must provide structural and thermal protection to the waste in hypothetical accident environments specified in DOT regulations 49CFR173 and NRC regulations 10CFR71. Preliminary design of the systems has been completed and final design for a truck system is underway. The status of the development program is reviewed in this paper and the reference design is described. Tests that have been conducted are discussed and long-term program objectives are reviewed

  4. Potential applications of advanced remote handling and maintenance technology to future waste handling facilities

    International Nuclear Information System (INIS)

    Kring, C.T.; Herndon, J.N.; Meacham, S.A.

    1987-01-01

    The Consolidated Fuel Reprocessing Program (CFRP) at the Oak Ridge National Laboratory (ORNL) has been advancing the technology in remote handling and remote maintenance of in-cell systems planned for future US nuclear fuel reprocessing plants. Much of the experience and technology developed over the past decade in this endeavor are directly applicable to the in-cell systems being considered for the facilities of the Federal Waste Management System (FWMS). The ORNL developments are based on the application of teleoperated force-reflecting servomanipulators controlled by an operator completely removed from the hazardous environment. These developments address the nonrepetitive nature of remote maintenance in the unstructured environments encountered in a waste handling facility. Employing technological advancements in dexterous manipulators, as well as basic design guidelines that have been developed for remotely maintained equipment and processes, can increase operation and maintenance system capabilities, thereby allowing the attainment of two Federal Waste Management System major objectives: decreasing plant personnel radiation exposure and increasing plant availability by decreasing the mean-time-to-repair in-cell maintenance and process equipment

  5. Potential applications of advanced remote handling and maintenance technology to future waste handling facilities

    International Nuclear Information System (INIS)

    Kring, C.T.; Herndon, J.N.; Meacham, S.A.

    1987-01-01

    The Consolidated Fuel Reprocessing Program (CFRP) at the Oak Ridge National Laboratory (ORNL) has been advancing the technology in remote handling and remote maintenance of in-cell systems planned for future U.S. nuclear fuel reprocessing plants. Much of the experience and technology developed over the past decade in this endeavor are directly applicable to the in-cell systems being considered for the facilities of the Federal Waste Management System (FWMS). The ORNL developments are based on the application of teleoperated force-reflecting servomanipulators controlled by an operator completely removed from the hazardous environment. These developments address the nonrepetitive nature of remote maintenance in the unstructured environments encountered in a waste handling facility. Employing technological advancements in dexterous manipulators, as well as basic design guidelines that have been developed for remotely maintained equipment and processes, can increase operation and maintenance system capabilities, thereby allowing the attainment of two Federal Waste Management System major objectives: decreasing plant personnel radiation exposure and increasing plant availability by decreasing the mean-time-to-repair in-cell maintenance and process equipment

  6. Radioactive waste management decommissioning spent fuel storage. V. 3. Waste transport, handling and disposal spent fuel storage

    International Nuclear Information System (INIS)

    1985-01-01

    As part of the book entitled Radioactive waste management decommissioning spent fuel storage, vol. 3 dealts with waste transport, handling and disposal, spent fuel storage. Twelve articles are presented concerning the industrial aspects of nuclear waste management in France [fr

  7. Remote-Handled Transuranic Waste Content Codes (RH-Trucon)

    International Nuclear Information System (INIS)

    2006-01-01

    The Remote-Handled Transuranic (RH-TRU) Content Codes (RH-TRUCON) document describes the inventory of RH-TRU waste within the transportation parameters specified by the Remote-Handled Transuranic Waste Authorized Methods for Payload Control (RH-TRAMPAC). The RH-TRAMPAC defines the allowable payload for the RH-TRU 72-B. This document is a catalog of RH-TRU 72-B authorized contents by site. A content code is defined by the following components: (1) A two-letter site abbreviation that designates the physical location of the generated/stored waste (e.g., ID for Idaho National Laboratory [INL]). The site-specific letter designations for each of the sites are provided in Table 1. (2) A three-digit code that designates the physical and chemical form of the waste (e.g., content code 317 denotes TRU Metal Waste). For RH-TRU waste to be transported in the RH-TRU 72-B, the first number of this three-digit code is ''3''. The second and third numbers of the three-digit code describe the physical and chemical form of the waste. Table 2 provides a brief description of each generic code. Content codes are further defined as subcodes by an alpha trailer after the three-digit code to allow segregation of wastes that differ in one or more parameter(s). For example, the alpha trailers of the subcodes ID 322A and ID 322B may be used to differentiate between waste packaging configurations. As detailed in the RH-TRAMPAC, compliance with flammable gas limits may be demonstrated through the evaluation of compliance with either a decay heat limit or flammable gas generation rate (FGGR) limit per container specified in approved content codes. As applicable, if a container meets the watt*year criteria specified by the RH-TRAMPAC, the decay heat limits based on the dose-dependent G value may be used as specified in an approved content code. If a site implements the administrative controls outlined in the RH-TRAMPAC and Appendix 2.4 of the RH-TRU Payload Appendices, the decay heat or FGGR

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

    International Nuclear Information System (INIS)

    Hauptmann, J.P.

    1985-09-01

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

  9. Waste Handling and Emplacement Options for Disposal of Radioactive Waste in Deep Boreholes.

    Energy Technology Data Exchange (ETDEWEB)

    Cochran, John R.; Hardin, Ernest

    2015-11-01

    Traditional methods cannot be used to handle and emplace radioactive wastes in boreholes up to 16,400 feet (5 km) deep for disposal. This paper describes three systems that can be used for handling and emplacing waste packages in deep borehole: (1) a 2011 reference design that is based on a previous study by Woodward–Clyde in 1983 in which waste packages are assembled into “strings” and lowered using drill pipe; (2) an updated version of the 2011 reference design; and (3) a new concept in which individual waste packages would be lowered to depth using a wireline. Emplacement on coiled tubing was also considered, but not developed in detail. The systems described here are currently designed for U.S. Department of Energy-owned high-level waste (HLW) including the Cesium- 137/Strontium-90 capsules from the Hanford Facility and bulk granular HLW from fuel processing in Idaho.

  10. Waste Handling and Emplacement Options for Disposal of Radioactive Waste in Deep Boreholes

    International Nuclear Information System (INIS)

    Cochran, John R.; Hardin, Ernest

    2015-01-01

    Traditional methods cannot be used to handle and emplace radioactive wastes in boreholes up to 16,400 feet (5 km) deep for disposal. This paper describes three systems that can be used for handling and emplacing waste packages in deep borehole: (1) a 2011 reference design that is based on a previous study by Woodward-Clyde in 1983 in which waste packages are assembled into ''strings'' and lowered using drill pipe; (2) an updated version of the 2011 reference design; and (3) a new concept in which individual waste packages would be lowered to depth using a wireline. Emplacement on coiled tubing was also considered, but not developed in detail. The systems described here are currently designed for U.S. Department of Energy-owned high-level waste (HLW) including the Cesium- 137/Strontium-90 capsules from the Hanford Facility and bulk granular HLW from fuel processing in Idaho.

  11. Qualitative comparisons of fusion reactor materials for waste handling and disposal

    International Nuclear Information System (INIS)

    Maninger, R.C.

    1985-01-01

    The activation of five structural materials and seven coolant/breeder/multiplier materials in a common reference neutron environment was calculated with the FORIG activation code. The reference environment was the neutron flux and spectrum at the first wall of the mirror advanced reactor study (MARS) reactor. Qualitative comparison of these activated materials were made with respect to worker protection requirements for gamma radiation in handling the materials and with respect to their classifications for near-surface disposal of radioactive waste

  12. Handling 78,000 drums of mixed-waste sludge

    International Nuclear Information System (INIS)

    Berry, J.B.; Gilliam, T.M.; Harrington, E.S.; Youngblood, E.L.; Baer, M.B.

    1991-01-01

    The Oak Ridge Gaseous Diffusion Plant (now know as the Oak Ridge K-25 Site) prepared two mixed-waste surface impoundments for closure by removing the sludge and contaminated pond-bottom clay and attempting to process it into durable, nonleachable, concrete monoliths. Interim, controlled, above-ground storage of the stabilized waste was planned until final disposition. The strategy for disposal included delisting the stabilized pond sludge from hazardous to nonhazardous and disposing of the delisted monoliths as radioactive waste. Because of schedule constraints and process design and control deficiencies, ∼46,000 drums of material in various stages of solidification and ∼32,000 drums of unprocessed sludge are presently being stored. In addition, the abandoned treatment facility still contains ∼16,000 gal of raw sludge. Such conditions do not comply with the requirements set forth by the Resource Conservation and Recovery Act (RCRA) for the storage of listed waste. Various steps are being taken to bring the storage of ∼78,000 drums of mixed waste into compliance with RCRA. This paper (1) reviews the current situation, (2) discusses the plan for remediation of regulatory noncompliances, including decanting liquid from stabilized waste and dewatering untreated waste, and (3) provides an assessment of alternative raw-waste treatment processes. 1 ref., 6 figs., 2 tabs

  13. 340 waste handling complex: Deactivation project management plan

    International Nuclear Information System (INIS)

    Stordeur, R.T.

    1998-01-01

    This document provides an overview of the strategy for deactivating the 340 Waste Handling Complex within Hanford's 300 Area. The plan covers the period from the pending September 30, 1998 cessation of voluntary radioactive liquid waste (RLW) transfers to the 340 Complex, until such time that those portions of the 340 Complex that remain active beyond September 30, 1998, specifically, the Retention Process Sewer (RPS), can also be shut down and deactivated. Specific activities are detailed and divided into two phases. Phase 1 ends in 2001 after the core RLW systems have been deactivated. Phase 2 covers the subsequent interim surveillance of deactivated and stand-by components during the period of continued RPS operation, through the final transfer of the entire 340 Complex to the Environmental Restoration Contractor. One of several possible scenarios was postulated and developed as a budget and schedule planning case

  14. B cell remote-handled waste shipment cask alternatives study

    International Nuclear Information System (INIS)

    RIDDELLE, J.G.

    1999-01-01

    The decommissioning of the 324 Facility B Cell includes the onsite transport of grouted remote-handled radioactive waste from the 324 Facility to the 200 Areas for disposal. The grouted waste has been transported in the leased ATG Nuclear Services 3-82B Radioactive Waste Shipping Cask (3-82B cask). Because the 3-82B cask is a U.S. Nuclear Regulatory Commission (NRC)-certified Type B shipping cask, the lease cost is high, and the cask operations in the onsite environment may not be optimal. An alternatives study has been performed to develop cost and schedule information on alternative waste transportation systems to assist in determining which system should be used in the future. Five alternatives were identified for evaluation. These included continued lease of the 3-82B cask, fabrication of a new 3-82B cask, development and fabrication of an onsite cask, modification of the existing U.S. Department of Energy-owned cask (OH-142), and the lease of a different commercially available cask. Each alternative was compared to acceptance criteria for use in the B Cell as an initial screening. Only continued leasing of the 3-82B cask, fabrication of a new 3-82B cask, and the development and fabrication of an onsite cask were found to meet all of the B Cell acceptance criteria

  15. Engineered Multifunctional Surfaces for Fluid Handling

    Science.gov (United States)

    Thomas, Chris; Ma, Yonghui; Weislogel, Mark

    2012-01-01

    Designs incorporating variations in capillary geometry and hydrophilic and/or antibacterial surface properties have been developed that are capable of passive gas/liquid separation and passive water flow. These designs can incorporate capillary grooves and/or surfaces arranged to create linear and circumferential capillary geometry at the micro and macro scale, radial fin configurations, micro holes and patterns, and combinations of the above. The antibacterial property of this design inhibits the growth of bacteria or the development of biofilm. The hydrophilic property reduces the water contact angle with a treated substrate such that water spreads into a thin layer atop the treated surface. These antibacterial and hydrophilic properties applied to a thermally conductive surface, combined with capillary geometry, create a novel heat exchanger capable of condensing water from a humid, two-phase water and gas flow onto the treated heat exchanger surfaces, and passively separating the condensed water from the gas flow in a reduced gravity application. The overall process to generate the antibacterial and hydrophilic properties includes multiple steps to generate the two different surface properties, and can be divided into two major steps. Step 1 uses a magnetron-based sputtering technique to implant the silver atoms into the base material. A layer of silver is built up on top of the base material. Completion of this step provides the antibacterial property. Step 2 uses a cold-plasma technique to generate the hydrophilic surface property on top of the silver layer generated in Step 1. Completion of this step provides the hydrophilic property in addition to the antibacterial property. Thermally conductive materials are fabricated and then treated to create the antibacterial and hydrophilic surface properties. The individual parts are assembled to create a condensing heat exchanger with antibacterial and hydrophilic surface properties and capillary geometry, which is

  16. 77 FR 58416 - Comparative Environmental Evaluation of Alternatives for Handling Low-Level Radioactive Waste...

    Science.gov (United States)

    2012-09-20

    ... for Handling Low-Level Radioactive Waste Spent Ion Exchange Resins From Commercial Nuclear Power... Radioactive Waste Spent Ion Exchange Resins from Commercial Nuclear Power Reactors. DATES: Please submit... Evaluation of Alternatives for Handling Low-Level Radioactive Waste Spent Ion Exchange Resins from Commercial...

  17. CLASSIFICATION OF THE MGR WASTE HANDLING BUILDING VENTILATION SYSTEM

    International Nuclear Information System (INIS)

    J.A. Ziegler

    2000-01-01

    The purpose of this analysis is to document the Quality Assurance (QA) classification of the Monitored Geologic Repository (MGR) waste handling building ventilation system structures, systems and components (SSCs) performed by the MGR Preclosure Safety and Systems Engineering Section. This analysis also provides the basis for revision of YMP/90-55Q, Q-List (YMP 2000). The Q-List identifies those MGR SSCs subject to the requirements of DOE/RW-0333P, ''Quality Assurance Requirements and Description'' (QARD) (DOE 2000). This QA classification incorporates the current MGR design and the results of the ''Design Basis Event Frequency and Dose Calculation for Site Recommendation'' (CRWMS M andO 2000a) and ''Bounding Individual Category 1 Design Basis Event Dose Calculation to Support Quality Assurance Classification'' (Gwyn 2000)

  18. 324 Building liquid waste handling and removal system project plan

    Energy Technology Data Exchange (ETDEWEB)

    Ham, J.E.

    1998-07-29

    This report evaluates the modification options for handling radiological liquid waste generated during decontamination and cleanout of the 324 Building. Recent discussions indicate that the Hanford site railroad system will be closed by the end of FY 1998 necessitating the need for an alternate transfer method. The issue of handling of Radioactive Liquid Waste (RLW) from the 324 Building (assuming the 340 Facility is not available to accept the RLW) has been examined in at least two earlier engineering studies (Parsons 1997a and Hobart 1997). Each study identified a similar preferred alternative that included modifying the 324 Building RLWS to allow load-out of wastewater to a truck tanker, while making maximum use of existing piping, tanks, instrumentation, controls and other features to minimize costs and physical changes to the building. This alternative is accepted as the basis for further discussion presented in this study. The goal of this engineering study is to verify the path forward presented in the previous studies and assure that the selected alternative satisfies the 324 Building deactivation goals and objectives as currently described in the project management plan. This study will also evaluate options available to implement the preferred alternative and select the preferred option for implementation of the entire system. Items requiring further examination will also be identified. Finally, the study will provide a conceptual design, schedule and cost estimate for the required modifications to the 324 Building to allow removal of RLW. Attachment 5 is an excerpt from the project baseline schedule found in the Project Management Plan.

  19. 327 Building liquid waste handling options modification project plan

    International Nuclear Information System (INIS)

    Ham, J.E.

    1998-01-01

    This report evaluates the modification options for handling radiological liquid waste (RLW) generated during decontamination and cleanout of the 327 Building. The overall objective of the 327 Facility Stabilization Project is to establish a passively safe and environmentally secure configuration of the 327 Facility. The issue of handling of RLW from the 327 Facility (assuming the 34O Facility is not available to accept the RLW) has been conceptually examined in at least two earlier engineering studies (Parsons 1997a and Hobart l997). Each study identified a similar preferred alternative that included modifying the 327 Facility RLWS handling systems to provide a truck load-out station, either within the confines of the facility or exterior to the facility. The alternatives also maximized the use of existing piping, tanks, instrumentation, controls and other features to minimize costs and physical changes. An issue discussed in each study involved the anticipated volume of the RLW stream. Estimates ranged between 113,550 and 387,500 liters in the earlier studies. During the development of the 324/327 Building Stabilization/Deactivation Project Management Plan, the lower estimate of approximately 113,550 liters was confirmed and has been adopted as the baseline for the 327 Facility RLW stream. The goal of this engineering study is to reevaluate the existing preferred alternative and select a new preferred alternative, if appropriate. Based on the new or confirmed preferred alternative, this study will also provide a conceptual design and cost estimate for required modifications to the 327 Facility to allow removal of RLWS and treatment of the RLW generated during deactivation

  20. Disposal and handling of nuclear steam generator chemical cleaning wastes

    International Nuclear Information System (INIS)

    Larrick, A.P.; Schneidmiller, D.

    1978-01-01

    A large number of pressurized water nuclear reactor electrical generating plants have experienced a corrosion-related problem with their steam generators known as denting. Denting is a mechanical deformation of the steam generator tubes that occurs at the tube support plates. Corrosion of the tube support plates occurs within the annuli through which the tubes pass and the resulting corrosion oxides, which are larger in volume than the original metal, compress and deform the tubes. In some cases, the induced stresses have been severe enough to cause tube and/or support cracking. The problem was so severe at the Turkey Point and Surrey plants that the tubing is being replaced. For less severe cases, chemical cleaning of the oxides, and other materials which deposit in the annuli from the water, is being considered. A Department of Energy-sponsored program was conducted by Consolidated Edison Co. of New York which identified several suitable cleaning solvents and led to in-plant chemical cleaning pilot demonstrations in the Indian Point Unit 1 steam generators. Current programs to improve the technology are being conducted by the Electric Power Research Institute, and the three PWR NSSS vendors with the assistance of numerous consultants, vendors, and laboratories. These programs are expected to result in more effective, less corrosive solvents. However, after a chemical cleaning is conducted, a large problem still remains- that of disposing of the spent wastes. The paper summarizes some of the methods currently available for handling and disposal of the wastes

  1. Waste Isolation Pilot Plant contact-handled transuranic waste preoperational checkout: Final report

    International Nuclear Information System (INIS)

    1988-07-01

    This report documents the results of the WIPP CH TRU Preoperational Checkout which was completed between June 8 and June 14, 1988 during which period, a total of 10 TRUPACT shipping containers were processed from site receipt through emplacement of the simulated waste packages in the underground storage area. Since the design of WIPP includes provisions to unload an internally contaminated TRUPACT, in the controlled environment of the Overpack and Repair Room, one TRUPACT was partially processed through this sequence of operations to verify this portion of the waste handling process as part of the checkout. The successful completion of the CH TRU Preoperational Checkout confirmed the acceptability of WIPP operating procedures, personnel, equipment, and techniques. Extrapolation of time-line data using a computer simulation model of the waste handling process has confirmed that WIPP operations can achieve the design throughput capability of 500,000 ft 3 /year, if required, using two waste handling shifts. The single shift throughput capability of 273,000 ft 3 /year exceeds the anticipated operating receival rate of about 230,000 ft 3 /year. At the 230,000 ft 3 /year rate, the combined CH TRU annual operator dose and the average individual dose (based on minimum crew size) is projected to be 13.7 rem and 0.7 rem, respectively. 6 refs., 27 figs., 3 tabs

  2. Handling 78,000 drums of mixed-waste sludge

    International Nuclear Information System (INIS)

    Berry, J.B.; Harrington, E.S.; Mattus, A.J.

    1991-01-01

    The Oak Ridge Gaseous Diffusion Plant (now known as the Oak Ridge K-25 Site) closed two mixed-waste surface impoundments by removing the sludge and contaminated pond-bottom clay and attempting to process it into durable, nonleachable, concrete monoliths. Interim, controlled, above-ground storage included delisting the stabilized sludge from hazardous to nonhazardous and disposing of the delisted monoliths as Class 1 radioactive waste. Because of schedule constraints and process design and control deficiencies, ∼46,000 drums of material in various stages of solidification and ∼32,000 barrels of unprocessed sludge are stored. The abandoned treatment facility still contains ∼16,000 gal of raw sludge. Such storage of mixed waste does not comply with the Resource Conservation and Recovery Act (RCRA) guidelines. This paper describes actions that are under way to bring the storage of ∼78,000 drums of mixed waste into compliance with RCRA. Remediation of this problem by treatment to meet regulatory requirements is the focus of the discussion. 3 refs., 2 figs., 4 tabs

  3. REMOTE MATERIAL HANDLING IN THE YUCCA MOUNTAIN WASTE PACKAGE CLOSURE CELL AND SUPPORT AREA GLOVEBOX

    International Nuclear Information System (INIS)

    K.M. Croft; S.M. Allen; M.W. Borland

    2005-01-01

    The Yucca Mountain Waste Package Closure System (WPCS) cells provide for shielding of highly radioactive materials contained in unsealed waste packages. The purpose of the cells is to provide safe environments for package handling and sealing operations. Once sealed, the packages are placed in the Yucca Mountain Repository. Closure of a typical waste package involves a number of remote operations. Those involved typically include the placement of matched lids onto the waste package. The lids are then individually sealed to the waste package by welding. Currently, the waste package includes three lids. One lid is placed before movement of the waste package to the closure cell; the final two are placed inside the closure cell, where they are welded to the waste package. These and other important operations require considerable remote material handling within the cell environment. This paper discusses the remote material handling equipment, designs, functions, operations, and maintenance, relative to waste package closure

  4. 340 Waste handling Facility Hazard Categorization and Safety Analysis

    International Nuclear Information System (INIS)

    Rodovsky, T.J.

    2010-01-01

    The analysis presented in this document provides the basis for categorizing the facility as less than Hazard Category 3. The final hazard categorization for the deactivated 340 Waste Handling Facility (340 Facility) is presented in this document. This hazard categorization was prepared in accordance with DOE-STD-1 027-92, Change Notice 1, Hazard Categorization and Accident Analysis Techniques for Compliance with Doe Order 5480.23, Nuclear Safety Analysis Reports. The analysis presented in this document provides the basis for categorizing the facility as less than Hazard Category (HC) 3. Routine nuclear waste receiving, storage, handling, and shipping operations at the 340 Facility have been deactivated, however, the facility contains a small amount of radioactive liquid and/or dry saltcake in two underground vault tanks. A seismic event and hydrogen deflagration were selected as bounding accidents. The generation of hydrogen in the vault tanks without active ventilation was determined to achieve a steady state volume of 0.33%, which is significantly less than the lower flammability limit of 4%. Therefore, a hydrogen deflagration is not possible in these tanks. The unmitigated release from a seismic event was used to categorize the facility consistent with the process defined in Nuclear Safety Technical Position (NSTP) 2002-2. The final sum-of-fractions calculation concluded that the facility is less than HC 3. The analysis did not identify any required engineered controls or design features. The Administrative Controls that were derived from the analysis are: (1) radiological inventory control, (2) facility change control, and (3) Safety Management Programs (SMPs). The facility configuration and radiological inventory shall be controlled to ensure that the assumptions in the analysis remain valid. The facility commitment to SMPs protects the integrity of the facility and environment by ensuring training, emergency response, and radiation protection. The full scale

  5. Performance Assessment for the Idaho National Laboratory Remote-Handled Low-Level Waste Disposal Facility

    Energy Technology Data Exchange (ETDEWEB)

    Annette L. Schafer; A. Jeffrey Sondrup; Arthur S. Rood

    2012-05-01

    This performance assessment for the Remote-Handled Low-Level Radioactive Waste Disposal Facility at the Idaho National Laboratory documents the projected radiological dose impacts associated with the disposal of low-level radioactive waste at the facility. This assessment evaluates compliance with the applicable radiological criteria of the U.S. Department of Energy and the U.S. Environmental Protection Agency for protection of the public and the environment. The calculations involve modeling transport of radionuclides from buried waste to surface soil and subsurface media, and eventually to members of the public via air, groundwater, and food chain pathways. Projections of doses are calculated for both offsite receptors and individuals who inadvertently intrude into the waste after site closure. The results of the calculations are used to evaluate the future performance of the low-level radioactive waste disposal facility and to provide input for establishment of waste acceptance criteria. In addition, one-factor-at-a-time, Monte Carlo, and rank correlation analyses are included for sensitivity and uncertainty analysis. The comparison of the performance assessment results to the applicable performance objectives provides reasonable expectation that the performance objectives will be met

  6. Waste Isolation Pilot Plant remote-handled transuranic waste disposal strategy

    International Nuclear Information System (INIS)

    1995-01-01

    The remote-handled transuranic (RH-TRU) waste disposal strategy described in this report identifies the process for ensuring that cost-effective initial disposal of RH-TRU waste will begin in Fiscal Year 2002. The strategy also provides a long-term approach for ensuring the efficient and sustained disposal of RH-TRU waste during the operating life of WIPP. Because Oak Ridge National Laboratory stores about 85 percent of the current inventory, the strategy is to assess the effectiveness of modifying their facilities to package waste, rather than constructing new facilities. In addition, the strategy involves identification of ways to prepare waste at other sites to supplement waste from Oak Ridge National Laboratory. DOE will also evaluate alternative packagings, modes of transportation, and waste emplacement configurations, and will select preferred alternatives to ensure initial disposal as scheduled. The long-term strategy provides a systemwide planning approach that will allow sustained disposal of RH-TRU waste during the operating life of WIPP. The DOE's approach is to consider the three relevant systems -- the waste management system at the generator/storage sites, the transportation system, and the WIPP disposal system -- and to evaluate the system components individually and in aggregate against criteria for improving system performance. To ensure full implementation, in Fiscal Years 1996 and 1997 DOE will: (1) decide whether existing facilities at Oak Ridge National Laboratory or new facilities to package and certify waste are necessary; (2) select the optimal packaging and mode of transportation for initial disposal; and (3) select an optimal disposal configuration to ensure that the allowable limits of RH-TRU waste can be disposed. These decisions will be used to identify funding requirements for the three relevant systems and schedules for implementation to ensure that the goal of initial disposal is met

  7. Preliminary Safety Design Report for Remote Handled Low-Level Waste Disposal Facility

    Energy Technology Data Exchange (ETDEWEB)

    Timothy Solack; Carol Mason

    2012-03-01

    A new onsite, remote-handled low-level waste disposal facility has been identified as the highest ranked alternative for providing continued, uninterrupted remote-handled low-level waste disposal for remote-handled low-level waste from the Idaho National Laboratory and for nuclear fuel processing activities at the Naval Reactors Facility. Historically, this type of waste has been disposed of at the Radioactive Waste Management Complex. Disposal of remote-handled low-level waste in concrete disposal vaults at the Radioactive Waste Management Complex will continue until the facility is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). This preliminary safety design report supports the design of a proposed onsite remote-handled low-level waste disposal facility by providing an initial nuclear facility hazard categorization, by discussing site characteristics that impact accident analysis, by providing the facility and process information necessary to support the hazard analysis, by identifying and evaluating potential hazards for processes associated with onsite handling and disposal of remote-handled low-level waste, and by discussing the need for safety features that will become part of the facility design.

  8. Handling of quarry waste from schist production at Oppdal, Norway

    Science.gov (United States)

    Willy Danielsen, Svein; Alnæs, Lisbeth; Azrague, Kamal; Suleng, Jon

    2017-04-01

    Handling of quarry waste from schist production at Oppdal, Norway Svein Willy Danielsen1), Lisbeth Alnæs2), Kamal Azrague2), Jon Suleng3) 1) Geomaterials Consultant, Trondheim Norway, 2) SINTEF, Trondheim, Norway, 3) AF Gruppen AS, Oppdal, Norway A significant amount of aggregate research in Norway has been focused on the recovery and use of surplus sizes from hard rock aggregate quarries. The use of sand sized quarry waste (QW) from crushing/processing has been motivated by the rapid depletion of traditional sand/gravel resources, increasing land-use conflicts, and the need to minimise QW deposits which for some quarries are becoming a critical factor for economy as well as for environmental reasons. With an annual aggregate production of 77 million tons, out of which approximately 83 % comes from hard rock, the annual volume of size market, the economic - and also environmental - potential will be considerable. Understanding the geological conditions and petrographic properties of the rock is vital. This is a quartz-feldspar rich metamorphic rock - a meta-arkose - containing rhythmically distributed planar lamina (less than 2 mm thick) or scattered occurrence of mica, separated by layers composed predominately of quartz and feldspar. The rock can be split along the lamina to slabs varying from 0.5 cm to more than 10 cm in thickness, and the microstructure can be characterized as being granoblastic to gneissic. . This makes it possible by well designed crushing process and careful selection of the in-going rock particles, to obtain well shaped aggregates up to at least 20 mm. The on-going project will also study the total cost situation depending on the QW utilisation, discuss the environmental and sustainability issues with a societal perspective, and also consider the market opportunities.

  9. The presence and leachability of antimony in different wastes and waste handling facilities in Norway.

    Science.gov (United States)

    Okkenhaug, G; Almås, Å R; Morin, N; Hale, S E; Arp, H P H

    2015-11-01

    The environmental behaviour of antimony (Sb) is gathering attention due to its increasingly extensive use in various products, particularly in plastics. Because of this it may be expected that plastic waste is an emission source for Sb in the environment. This study presents a comprehensive field investigation of Sb concentrations in diverse types of waste from waste handling facilities in Norway. The wastes included waste electrical and electronic equipment (WEEE), glass, vehicle fluff, combustibles, bottom ash, fly ash and digested sludge. The highest solid Sb concentrations were found in WEEE and vehicle plastic (from 1238 to 1715 mg kg(-1)) and vehicle fluff (from 34 to 4565 mg kg(-1)). The type of acid used to digest the diverse solid waste materials was also tested. It was found that HNO3:HCl extraction gave substantially lower, non-quantitative yields compared to HNO3:HF. The highest water-leachable concentration for wastes when mixed with water at a 1 : 10 ratio were observed for plastic (from 0.6 to 2.0 mg kg(-1)) and bottom ash (from 0.4 to 0.8 mg kg(-1)). For all of the considered waste fractions, Sb(v) was the dominant species in the leachates, even though Sb(iii) as Sb2O3 is mainly used in plastics and other products, indicating rapid oxidation in water. This study also presents for the first time a comparison of Sb concentrations in leachate at waste handling facilities using both active grab samples and DGT passive samples. Grab samples target the total suspended Sb, whereas DGT targets the sum of free- and other chemically labile species. The grab sample concentrations (from 0.5 to 50 μg L(-1)) were lower than the predicted no-effect concentration (PNEC) of 113 μg L(-1). The DGT concentrations were substantially lower (from 0.05 to 9.93 μg L(-1)) than the grab samples, indicating much of the Sb is present in a non-available colloidal form. In addition, air samples were taken from the chimney and areas within combustible waste incinerators, as

  10. Contact-Handled Transuranic Waste Acceptance Criteria for the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    2005-01-01

    The purpose of this document is to summarize the waste acceptance criteria applicable to the transportation, storage, and disposal of contact-handled transuranic (CH-TRU) waste at the Waste Isolation Pilot Plant (WIPP). These criteria serve as the U.S. Department of Energy's (DOE) primary directive for ensuring that CH-TRU waste is managed and disposed of in a manner that protects human health and safety and the environment.The authorization basis of WIPP for the disposal of CH-TRU waste includes the U.S.Department of Energy National Security and Military Applications of Nuclear EnergyAuthorization Act of 1980 (reference 1) and the WIPP Land Withdrawal Act (LWA;reference 2). Included in this document are the requirements and associated criteriaimposed by these acts and the Resource Conservation and Recovery Act (RCRA,reference 3), as amended, on the CH-TRU waste destined for disposal at WIPP.|The DOE TRU waste sites must certify CH-TRU waste payload containers to thecontact-handled waste acceptance criteria (CH-WAC) identified in this document. Asshown in figure 1.0, the flow-down of applicable requirements to the CH-WAC istraceable to several higher-tier documents, including the WIPP operational safetyrequirements derived from the WIPP CH Documented Safety Analysis (CH-DSA;reference 4), the transportation requirements for CH-TRU wastes derived from theTransuranic Package Transporter-Model II (TRUPACT-II) and HalfPACT Certificates ofCompliance (references 5 and 5a), the WIPP LWA (reference 2), the WIPP HazardousWaste Facility Permit (reference 6), and the U.S. Environmental Protection Agency(EPA) Compliance Certification Decision and approval for PCB disposal (references 7,34, 35, 36, and 37). The solid arrows shown in figure 1.0 represent the flow-down of allapplicable payload container-based requirements. The two dotted arrows shown infigure 1.0 represent the flow-down of summary level requirements only; i.e., the sitesmust reference the regulatory source

  11. Contact-Handled Transuranic Waste Acceptance Criteria for the Waste Isolation Pilot Plant

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions LLC

    2005-12-29

    The purpose of this document is to summarize the waste acceptance criteria applicable to the transportation, storage, and disposal of contact-handled transuranic (CH-TRU) waste at the Waste Isolation Pilot Plant (WIPP). These criteria serve as the U.S. Department of Energy's (DOE) primary directive for ensuring that CH-TRU waste is managed and disposed of in a manner that protects human health and safety and the environment.The authorization basis of WIPP for the disposal of CH-TRU waste includes the U.S.Department of Energy National Security and Military Applications of Nuclear EnergyAuthorization Act of 1980 (reference 1) and the WIPP Land Withdrawal Act (LWA;reference 2). Included in this document are the requirements and associated criteriaimposed by these acts and the Resource Conservation and Recovery Act (RCRA,reference 3), as amended, on the CH-TRU waste destined for disposal at WIPP.|The DOE TRU waste sites must certify CH-TRU waste payload containers to thecontact-handled waste acceptance criteria (CH-WAC) identified in this document. Asshown in figure 1.0, the flow-down of applicable requirements to the CH-WAC istraceable to several higher-tier documents, including the WIPP operational safetyrequirements derived from the WIPP CH Documented Safety Analysis (CH-DSA;reference 4), the transportation requirements for CH-TRU wastes derived from theTransuranic Package Transporter-Model II (TRUPACT-II) and HalfPACT Certificates ofCompliance (references 5 and 5a), the WIPP LWA (reference 2), the WIPP HazardousWaste Facility Permit (reference 6), and the U.S. Environmental Protection Agency(EPA) Compliance Certification Decision and approval for PCB disposal (references 7,34, 35, 36, and 37). The solid arrows shown in figure 1.0 represent the flow-down of allapplicable payload container-based requirements. The two dotted arrows shown infigure 1.0 represent the flow-down of summary level requirements only; i.e., the sitesmust reference the regulatory source

  12. Demonstration of remotely operated TRU waste size reduction and material handling equipment

    International Nuclear Information System (INIS)

    Looper, M.G.; Charlesworth, D.L.

    1988-01-01

    The Savannah River Laboratory (SRL) is developing remote size reduction and material handling equipment to prepare 238 Pu contaminated waste for permanent disposal at the Waste Isolation Pilot Plant (WIPP) in New Mexico. The waste is generated at the Savannah River Plant (SRP) from normal operation and decommissioning activity and is retrievably stored onsite. A Transuranic Waste Facility for preparing, size-reducing, and packaging this waste for disposal is scheduled for completion in 1995. A cold test facility for demonstrating the size reduction and material handling equipment was built, and testing began in January 1987. 9 figs., 1 tab

  13. Logistics of Transport and Handling with the Waste in the Upper Gemer region

    Directory of Open Access Journals (Sweden)

    Ján Spišák

    2005-11-01

    Full Text Available In the future, not any society (even the most advanced society can exists without waste formed by production processes or by any human activity. Increasing of the waste volume as well as its structure influences the living space of the mankind in a negative way. Therefore, the production, disposal or the exploitation of the waste is not only ecological but also the economical problem for the whole society. New methods of handling and disposal of the waste are preferred. This contribution is oriented on the application of micrologistics proceedings in order to reach a more effective system of transporting and handling with the waste.

  14. 76 FR 33277 - Proposed Approval of the Central Characterization Project's Remote-Handled Transuranic Waste...

    Science.gov (United States)

    2011-06-08

    ... disposal of TRU radioactive waste. As defined by the WIPP Land Withdrawal Act (LWA) of 1992 (Pub. L. 102... certification of the WIPP's compliance with disposal regulations for TRU radioactive waste [63 Federal Register... radioactive remote-handled (RH) transuranic (TRU) waste characterization program implemented by the Central...

  15. Construction and operation of replacement hazardous waste handling facility at Lawrence Berkeley Laboratory

    International Nuclear Information System (INIS)

    1992-09-01

    The US Department of Energy (DOE) has prepared an environmental assessment (EA), DOE/EA-0423, for the construction and operation of a replacement hazardous waste handling facility (HWHF) and decontamination of the existing HWHF at Lawrence Berkeley Laboratory (LBL), Berkeley, California. The proposed facility would replace several older buildings and cargo containers currently being used for waste handling activities and consolidate the LBL's existing waste handling activities in one location. The nature of the waste handling activities and the waste volume and characteristics would not change as a result of construction of the new facility. Based on the analysis in the EA, DOE has determined that the proposed action would not constitute a major Federal action significantly affecting the quality of the human environment within the meaning of the National Environmental Policy Act (NEPA) of 1969, 42 USC. 4321 et seq. Therefore, an environmental impact statement is not required

  16. Construction and operation of replacement hazardous waste handling facility at Lawrence Berkeley Laboratory. Environmental Assessment

    Energy Technology Data Exchange (ETDEWEB)

    1992-09-01

    The US Department of Energy (DOE) has prepared an environmental assessment (EA), DOE/EA-0423, for the construction and operation of a replacement hazardous waste handling facility (HWHF) and decontamination of the existing HWHF at Lawrence Berkeley Laboratory (LBL), Berkeley, California. The proposed facility would replace several older buildings and cargo containers currently being used for waste handling activities and consolidate the LBL`s existing waste handling activities in one location. The nature of the waste handling activities and the waste volume and characteristics would not change as a result of construction of the new facility. Based on the analysis in the EA, DOE has determined that the proposed action would not constitute a major Federal action significantly affecting the quality of the human environment within the meaning of the National Environmental Policy Act (NEPA) of 1969, 42 USC. 4321 et seq. Therefore, an environmental impact statement is not required.

  17. Development of constrained motion control for robot handling of hazardous waste

    International Nuclear Information System (INIS)

    Starr, G.P.

    1993-01-01

    Handling and archiving of hazardous waste is an area where automation and robotics can be of significant benefit, by removing the human operator from the workplace and its associated hazards. For reasons of safety, throughput, and reduced setup time, force-controlled robots are well-suited for hazardous materials handling. The focus of this investigation is the development of advanced force control techniques for commercial industrial robots in the surface sampling of hazardous waste containers. Two particular control strategies are considered, (1) preview control, and (2) adaptive control. Preview control uses a sensor which can ''look ahead'' and thereby reduce the effect of surface irregularity on contact force control. Adaptive control allows the robot controller to compensate for changes in the robot characteristics as it changes position, and likewise improves performance. The resulting control algorithms will be applied to a two-dimensional contour-following task using a PUMA robot at the Robotics Research Laboratory at The University of New Mexico. (author) 9 figs., 13 refs

  18. Radioactive waste handling at the Mochovce NPP, 1998-2008

    International Nuclear Information System (INIS)

    Vasickova, Gabriela

    2009-01-01

    The radioactive waste management system at the Mochovce NPP is described. The system addresses technical aspects as well as administrative provisions related to radioactive waste generated within the controlled area, from the waste generation phase to waste sorting, packaging, storage, recording, measurement, and transportation to the Bohunice waste processing facility or transfer to the Mochovce liquid radioactive waste treatment facility. The article also addresses conditions for release from the controlled area to the environment for radioactive waste which can be exempt from the institutional administrative control system or released to the environment on the basis of a valid permission issued by the relevant regulatory authority

  19. 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

  20. Low-level waste certification plan for the Lawrence Berkeley Laboratory Hazardous Waste Handling Facility. Revision 1

    International Nuclear Information System (INIS)

    1995-01-01

    The purpose of this plan is to describe the organization and methodology for the certification of low-level radioactive waste (LLW) handled in the Hazardous Waste Handling Facility (HWHF) at Lawrence Berkeley Laboratory (LBL). This plan is composed to meet the requirements found in the Westinghouse Hanford Company (WHC) Solid Waste Acceptance Criteria (WAC) and follows the suggested outline provided by WHC in the letter of April 26, 1990, to Dr. R.H. Thomas, Occupational Health Division, LBL. LLW is to be transferred to the WHC Hanford Site Central Waste Complex and Burial Grounds in Hanford, Washington

  1. Low-level waste certification plan for the Lawrence Berkeley Laboratory Hazardous Waste Handling Facility. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-01-10

    The purpose of this plan is to describe the organization and methodology for the certification of low-level radioactive waste (LLW) handled in the Hazardous Waste Handling Facility (HWHF) at Lawrence Berkeley Laboratory (LBL). This plan is composed to meet the requirements found in the Westinghouse Hanford Company (WHC) Solid Waste Acceptance Criteria (WAC) and follows the suggested outline provided by WHC in the letter of April 26, 1990, to Dr. R.H. Thomas, Occupational Health Division, LBL. LLW is to be transferred to the WHC Hanford Site Central Waste Complex and Burial Grounds in Hanford, Washington.

  2. Remote waste handling at the Hot Fuel Examination Facility

    International Nuclear Information System (INIS)

    Vaughn, M.E.

    1982-01-01

    Radioactive solid wastes, some of which are combustible, are generated during disassembly and examination of irradiated fast-reactor fuel and material experiments at the Hot Fuel Examination Facility (HFEF). These wastes are remotely segregated and packaged in doubly contained, high-integrity, clean, retrievable waste packages for shipment to the Radioactive Waste Management Complex (RWMC) at the Idaho National Engineering Laboratory (INEL). This paper describes the equipment and techniques used to perform these operations

  3. Pre-disposal storage, transport and handling of vitrified high level waste

    International Nuclear Information System (INIS)

    Kempe, T.F.; Martin, A.

    1981-05-01

    The objectives of the study were to review non site-specific engineering features of the storage, transport and handling of vitrified high level radioactive waste prior to its transfer into an underground repository, and to identify those features which require validation or development. Section headings are: introduction (historical and technical background); characteristics and arisings of vitrified high level waste; overpacks (additional containment barrier, corrosion resistant); interim storage of HLW; transport of HLW; handling; conclusions and recommendations. (U.K.)

  4. The low-level waste handling challenge at the Feed Materials Production Center

    International Nuclear Information System (INIS)

    Harmon, J.E.; Diehl, D.E.; Gardner, R.L.

    1988-01-01

    The management of low-level wastes from the production of depleted uranium at the Feed Materials Production Center presents an enormous challenge. The recovery of uranium from materials contaminated with depleted uranium is usually not economical. As a result, large volumes of wastes are generated. The Westinghouse Materials Company of Ohio has established an aggressive waste management program. Simple solutions have been applied to problems in the areas of waste handling and waste minimization. The success of this program has been demonstrated by the reduction of low-level waste inventory at the Feed Materials Production Center

  5. The low-level waste handling challenge at the Feed Materials Production Center

    International Nuclear Information System (INIS)

    Harmon, J.E.; Diehl, D.E.; Gardner, R.L.

    1988-02-01

    The management of low-level wastes from the production of depleted uranium at the Feed Materials Production Center presents an enormous challenge. The recovery of uranium from materials contaminated with depleted uranium is usually not economical. As a result, large volumes of wastes are generated. The Westinghouse Materials Company of Ohio has established an aggressive waste management program. Simple solutions have been applied to problems in the areas of waste handling and waste minimization. The success of this program has been demonstrated by the reduction of low-level waste inventory at the Feed Materials Production Center. 8 refs., 4 figs

  6. Decree 2211: Standards to control the generation and handling of dangerous wastes

    International Nuclear Information System (INIS)

    1992-01-01

    This Decree has for object to establish the conditions under which should be carried out the activities of generation and handling of dangerous waste, in order to prevent damages to health and to the atmosphere. It includes: definitions; a list of sources of waste; a list of constituent of dangerous waste; the characteristics of danger; a lists of maximum permissible concentrations in leachates, handling of dangerous waste, criterion for transport, monitoring form, storage areas, treatment and final disposition, storage, elimination, incineration, recycling, reuse and recovery, installation and operation of security backfilling, book of waste record, control of activities, obligations in charge of those who manage dangerous waste, and trans border movements of dangerous waste [es

  7. Considerations for evaluation and selection of solid waste handling apron conveyors

    Energy Technology Data Exchange (ETDEWEB)

    Lisiecki, H.G.

    1976-11-01

    Criteria to be used in evaluating and selecting conveyer equipment for facilities handling solid wastes, such as solid waste resource recovery facilities, are discussed. Types of conveyer pan design and chain mechanisms are described. It is concluded that the conveyer purchaser must be knowledgeable about the equipment available, the specific use of equipment, its performance specifications, and the overall maintenance and operating costs. (LCL)

  8. Development and use of a remote waste handling system for disposal of greater confinement wastes

    International Nuclear Information System (INIS)

    Williams, R.E.

    1985-01-01

    This paper discusses the design and development of a remotely controlled waste handling system (RWHS) for use in radioactive waste disposal operations. A RWHS was developed at the US Department of Energy's (DOE) Nevada Test Site for use in the Greater Confinement Disposal Test (GCDT). The RWHS consists of a remote control console and the following remotely operated features: a crane, a grapple/manipulator module which is suspended by the crane hoist hook, and closed-circuit television cameras. The RWHS was used to safely place high-specific-activity radioactive waste in greater confinement disposal. Between December 15, 1983, and February 23, 1984, five encapsulated sources were open-air transferred from shielded shipping casks and placed 30 m down a 3-m-dia augered shaft using the RWHS. These sources contained approximately 460 kCi of 90 Sr, 21 kCi of 137 Cs, and 390 Ci of 60 Co. Each source was transferred safely and efficiently and operational personnel did not receive any recordable doses. 3 references, 5 figures

  9. Impact of hazardous waste handling legislation on nuclear installations and radioactive waste management in the United States

    International Nuclear Information System (INIS)

    Trosten, L.M.

    1988-01-01

    The United States has enacted complex legislation to help assure proper handling of hazardous waste and the availability of funds to cover the expenditures. There are a number of uncertainties concerning the impact of this legislation, and regulations promulgated by the Environmental Protection Agency and the states, upon nuclear installations and radioactive waste management. This report provides an overview of the U.S. hazardous waste legislation and examines the outlook for its application to the nuclear industry (NEA) [fr

  10. Remote systems and automation in radioactive waste package handling

    International Nuclear Information System (INIS)

    Gneiting, B.C.; Hayward, M.L.

    1987-01-01

    A proof-of-principle test was conducted at the Hanford Engineering Development Laboratory (HEDL) to demonstrate the feasibility of performing cask receiving and unloading operations in a remote and partially automated manner. This development testing showed feasibility of performing critical cask receipt, preparation, and unloading operations from a single control station using remote controls and indirect viewing. Using robotics and remote automation in a cask handling system can result in lower personnel exposure levels and cask turnaround times while maintaining operational flexibility. An automated cask handling system presents a flexible state-of-the-art, cost effective alternative solution to hands-on methods that have been used in the past

  11. Project Execution Plan for the Remote Handled Low-Level Waste Disposal Project

    Energy Technology Data Exchange (ETDEWEB)

    Danny Anderson

    2014-07-01

    As part of ongoing cleanup activities at the Idaho National Laboratory (INL), closure of the Radioactive Waste Management Complex (RWMC) is proceeding under the Comprehensive Environmental Response, Compensation, and Liability Act (42 USC 9601 et seq. 1980). INL-generated radioactive waste has been disposed of at RWMC since 1952. The Subsurface Disposal Area (SDA) at RWMC accepted the bulk of INL’s contact and remote-handled low-level waste (LLW) for disposal. Disposal of contact-handled LLW and remote-handled LLW ion-exchange resins from the Advanced Test Reactor in the open pit of the SDA ceased September 30, 2008. Disposal of remote-handled LLW in concrete disposal vaults at RWMC will continue until the facility is full or until it must be closed in preparation for final remediation of the SDA (approximately at the end of fiscal year FY 2017). The continuing nuclear mission of INL, associated ongoing and planned operations, and Naval spent fuel activities at the Naval Reactors Facility (NRF) require continued capability to appropriately dispose of contact and remote handled LLW. A programmatic analysis of disposal alternatives for contact and remote-handled LLW generated at INL was conducted by the INL contractor in Fiscal Year 2006; subsequent evaluations were completed in Fiscal Year 2007. The result of these analyses was a recommendation to the Department of Energy (DOE) that all contact-handled LLW generated after September 30, 2008, be disposed offsite, and that DOE proceed with a capital project to establish replacement remote-handled LLW disposal capability. An analysis of the alternatives for providing replacement remote-handled LLW disposal capability has been performed to support Critical Decision-1. The highest ranked alternative to provide this required capability has been determined to be the development of a new onsite remote-handled LLW disposal facility to replace the existing remote-handled LLW disposal vaults at the SDA. Several offsite DOE

  12. Audit Report on 'Waste Processing and Recovery Act Acceleration Efforts for Contact-Handled Transuranic Waste at the Hanford Site'

    International Nuclear Information System (INIS)

    2010-01-01

    The Department of Energy's Office of Environmental Management's (EM), Richland Operations Office (Richland), is responsible for disposing of the Hanford Site's (Hanford) transuranic (TRU) waste, including nearly 12,000 cubic meters of radioactive contact-handled TRU wastes. Prior to disposing of this waste at the Department's Waste Isolation Pilot Plant (WIPP), Richland must certify that it meets WIPP's waste acceptance criteria. To be certified, the waste must be characterized, screened for prohibited items, treated (if necessary) and placed into a satisfactory disposal container. In a February 2008 amendment to an existing Record of Decision (Decision), the Department announced its plan to ship up to 8,764 cubic meters of contact-handled TRU waste from Hanford and other waste generator sites to the Advanced Mixed Waste Treatment Project (AMWTP) at Idaho's National Laboratory (INL) for processing and certification prior to disposal at WIPP. The Department decided to maximize the use of the AMWTP's automated waste processing capabilities to compact and, thereby, reduce the volume of contact-handled TRU waste. Compaction reduces the number of shipments and permits WIPP to more efficiently use its limited TRU waste disposal capacity. The Decision noted that the use of AMWTP would avoid the time and expense of establishing a processing capability at other sites. In May 2009, EM allocated $229 million of American Recovery and Reinvestment Act of 2009 (Recovery Act) funds to support Hanford's Solid Waste Program, including Hanford's contact-handled TRU waste. Besides providing jobs, these funds were intended to accelerate cleanup in the short term. We initiated this audit to determine whether the Department was effectively using Recovery Act funds to accelerate processing of Hanford's contact-handled TRU waste. Relying on the availability of Recovery Act funds, the Department changed course and approved an alternative plan that could increase costs by about $25 million

  13. Alternative configurations for the waste-handling building at the Yucca Mountain Repository

    International Nuclear Information System (INIS)

    1990-08-01

    Two alternative configurations of the waste-handling building have been developed for the proposed nuclear waste repository in tuff at Yucca Mountain, Nevada. One configuration is based on criteria and assumptions used in Case 2 (no monitored retrievable storage facility, no consolidation), and the other configuration is based on criteria and assumptions used in Case 5 (consolidation at the monitored retrievable storage facility) of the Monitored Retrievable Storage System Study for the Repository. Desirable waste-handling design concepts have been selected and are included in these configurations. For each configuration, general arrangement drawings, plot plans, block flow diagrams, and timeline diagrams are prepared

  14. Process Knowledge Summary Report for Materials and Fuels Complex Contact-Handled Transuranic Debris Waste

    Energy Technology Data Exchange (ETDEWEB)

    R. P. Grant; P. J. Crane; S. Butler; M. A. Henry

    2010-02-01

    This Process Knowledge Summary Report summarizes the information collected to satisfy the transportation and waste acceptance requirements for the transfer of transuranic (TRU) waste between the Materials and Fuels Complex (MFC) and the Advanced Mixed Waste Treatment Project (AMWTP). The information collected includes documentation that addresses the requirements for AMWTP and the applicable portion of their Resource Conservation and Recovery Act permits for receipt and treatment of TRU debris waste in AMWTP. This report has been prepared for contact-handled TRU debris waste generated by the Idaho National Laboratory at MFC. The TRU debris waste will be shipped to AMWTP for purposes of supercompaction. This Process Knowledge Summary Report includes information regarding, but not limited to, the generation process, the physical form, radiological characteristics, and chemical contaminants of the TRU debris waste, prohibited items, and packaging configuration. This report, along with the referenced supporting documents, will create a defensible and auditable record for waste originating from MFC.

  15. Remote systems and automation in radioactive waste package handling

    International Nuclear Information System (INIS)

    Gneiting, B.C.; Hayward, M.L.

    1987-01-01

    A proof-of-principle test was conducted at the Hanford Engineering Development Laboratory (HEDL) to demonstrate the feasibility of performing cask receiving and unloading operations in a remote and partially automated manner. This development testing showed feasibility of performing critical cask receipt, preparation, and unloading operations from a single control station using remote controls and indirect viewing. Using robotics and remote automation in a cask handling system can result in lower personnel exposure levels and cask turnaround times while maintaining operational flexibility. An automated cask handling system presents a flexible state-of-the-art, cost effective alternative solution to hands-on methods that have been used in the past. 7 refs., 13 figs

  16. The main ecological principles of ensuring safety of man and biosphere in the handling of radioactive wastes

    International Nuclear Information System (INIS)

    Kryshev, I.I.; Sazykina, T.G.

    1999-01-01

    This paper provides an assessment of ecological safety in the handling of radioactive wastes in the territory of Russia. The following problems are considered: the main sources of radioactive wastes and spent nuclear fuel; assessments of collective dose from the enterprises of the nuclear fuel cycle in Russia; and principles and criteria for ensuring ecological safety when handling radioactive wastes

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

    International Nuclear Information System (INIS)

    Callan, J.E.

    1986-01-01

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

  18. Savannah River Certification Plan for newly generated, contact-handled transuranic waste

    International Nuclear Information System (INIS)

    Wierzbicki, K.S.

    1986-01-01

    This Certification Plan document describes the necessary processes and methods for certifying unclassified, newly generated, contact-handled solid transuranic (TRU) waste at the Savannah River Plant and Laboratory (SRP, SRL) to comply with the Waste Isolation Pilot Plant Waste Acceptance Criteria (WIPP-WAC). Section 2 contains the organizational structure as related to waste certification including a summary of functional responsibilities, levels of authority, and lines of communication of the various organizations involved in certification activities. Section 3 describes general plant operations and TRU waste generation. Included is a description of the TRU Waste classification system. Section 4 contains the SR site TRU Waste Quality Assurance Program Plan. Section 5 describes waste container procurement, inspection, and certification prior to being loaded with TRU waste. Certification of waste packages, after package closure in the waste generating areas, is described in Section 6. The packaging and certification of individual waste forms is described in Attachments 1-5. Included in each attachment is a description of controls used to ensure that waste packages meet all applicable waste form compliance requirements for shipment to the WIPP. 3 figs., 3 tabs

  19. Efficient handling of high-level radioactive cell waste in a vitrification facility analytical laboratory

    International Nuclear Information System (INIS)

    Roberts, D.W.; Collins, K.J.

    1998-01-01

    The Savannah River Site''s (SRS) Defense Waste Processing Facility (DWPF) near Aiken, South Carolina, is the world''s largest and the United State''s first high level waste vitrification facility. For the past 1.5 years, DWPF has been vitrifying high level radioactive liquid waste left over from the Cold War. The vitrification process involves the stabilization of high level radioactive liquid waste into borosilicate glass. The glass is contained in stainless steel canisters. DWPF has filled more than 200 canisters 3.05 meters (10 feet) long and 0.61 meters (2 foot) diameter. Since operations began at DWPF in March of 1996, high level radioactive solid waste continues to be generated due to operating the facility''s analytical laboratory. The waste is referred to as cell waste and is routinely removed from the analytical laboratories. Through facility design, engineering controls, and administrative controls, DWPF has established efficient methods of handling the high level waste generated in its laboratory facility. These methods have resulted in the prevention of undue radiation exposure, wasted man-hours, expenses due to waste disposal, and the spread of contamination. This level of efficiency was not reached overnight, but it involved the collaboration of Radiological Control Operations and Laboratory personnel working together to devise methods that best benefited the facility. This paper discusses the methods that have been incorporated at DWPF for the handling of cell waste. The objective of this paper is to provide insight to good radiological and safety practices that were incorporated to handle high level radioactive waste in a laboratory setting

  20. Design and operation of a remotely operated plutonium waste size reduction and material handling process

    International Nuclear Information System (INIS)

    Stewart, J.A. III; Charlesworth, D.L.

    1986-01-01

    Noncombustible 238 Pu and 239 Pu waste is generated as a result of normal operation and decommissioning activity at the Savannah River Plant, and is being retrievably stored there. As part of the long-term plant to process the stored waste and current waste for permanent disposal, a remote size reduction and material handling process is being cold-tested at Savannah River Laboratory. The process consists of a large, low-speed shredder and material handling system, a remote worktable, a bagless transfer system, and a robotically controlled manipulator. Initial testing of the shredder and material handling system and a cycle test of the bagless transfer system has been completed. Fabrication and acceptance testing of the Telerobat, a robotically controlled manipulator has been completed. Testing is scheduled to begin in 3/86. Design features maximizing the ability to remotely maintain the equipment were incorporated. Complete cold-testing of the equipment is scheduled to be completed in 1987

  1. Remote-Handled Low-Level Waste Disposal Project Alternatives Analysis

    Energy Technology Data Exchange (ETDEWEB)

    David Duncan

    2011-04-01

    This report identifies, evaluates, and compares alternatives for meeting the U.S. Department of Energy’s mission need for management of remote-handled low-level waste generated by the Idaho National Laboratory and its tenants. Each alternative identified in the Mission Need Statement for the Remote-Handled Low-Level Waste Treatment Project is described and evaluated for capability to fulfill the mission need. Alternatives that could meet the mission need are further evaluated and compared using criteria of cost, risk, complexity, stakeholder values, and regulatory compliance. The alternative for disposal of remote-handled low-level waste that has the highest confidence of meeting the mission need and represents best value to the government is to build a new disposal facility at the Idaho National Laboratory Site.

  2. Remote-Handled Low-Level Waste Disposal Project Alternatives Analysis

    Energy Technology Data Exchange (ETDEWEB)

    David Duncan

    2011-03-01

    This report identifies, evaluates, and compares alternatives for meeting the U.S. Department of Energy’s mission need for management of remote-handled low-level waste generated by the Idaho National Laboratory and its tenants. Each alternative identified in the Mission Need Statement for the Remote-Handled Low-Level Waste Treatment Project is described and evaluated for capability to fulfill the mission need. Alternatives that could meet the mission need are further evaluated and compared using criteria of cost, risk, complexity, stakeholder values, and regulatory compliance. The alternative for disposal of remote-handled low-level waste that has the highest confidence of meeting the mission need and represents best value to the government is to build a new disposal facility at the Idaho National Laboratory Site.

  3. Remote-Handled Low-Level Waste Disposal Project Alternatives Analysis

    Energy Technology Data Exchange (ETDEWEB)

    David Duncan

    2010-06-01

    This report identifies, evaluates, and compares alternatives for meeting the U.S. Department of Energy’s mission need for management of remote-handled low-level waste generated by the Idaho National Laboratory and its tenants. Each alternative identified in the Mission Need Statement for the Remote-Handled Low-Level Waste Treatment Project is described and evaluated for capability to fulfill the mission need. Alternatives that could meet the mission need are further evaluated and compared using criteria of cost, risk, complexity, stakeholder values, and regulatory compliance. The alternative for disposal of remote-handled low-level waste that has the highest confidence of meeting the mission need and represents best value to the government is to build a new disposal facility at the Idaho National Laboratory Site.

  4. Fuel handling, reprocessing, and waste and related nuclear data aspects

    International Nuclear Information System (INIS)

    Kuesters, H.; Lalovic, M.; Wiese, H.W.

    1979-06-01

    The essential processes in the out-of-pile nuclear fuel cycle are described, i.e. mining and milling of uranium ores, enrichment, fuel fabrication, storage, transportation, reprocessing of irradiated fuel, waste treatment and waste disposal. The aspects of radiation (mainly gammas and neutrons) and of heat production, as well as special safety considerations are outlined with respect to their potential operational impacts and long-term hazards. In this context the importance of nuclear data for the out-of-pile fuel cycle is discussed. Special weight is given to the LWR fuel cycle including recycling; the differences of LMFBR high burn-up fuel with large PuO 2 content are described. The HTR fuel cycle is discussed briefly as well as some alternative fuel cycle concepts. (orig.) [de

  5. High level radioactive waste repositories. Task 3. Review of underground handling and emplacement. 1. Executive summary

    Energy Technology Data Exchange (ETDEWEB)

    1981-01-01

    A review is presented of proposals for transport, handling and emplacement of high-level radioactive waste in an underground repository appropriate to the U.K. context, with particular reference to waste block size and configuration; self-shielded or partially-shielded block; stages of disposal; transport by road/rail to repository site; handling techniques within repository; emplacement in vertical holes or horizontal tunnels; repository access by adit, incline or shaft; conventional and radiological safety; costs; and major areas of uncertainty requiring research or development.

  6. Legal provisions concerning the handling and disposal of radioactive waste in international and national law

    International Nuclear Information System (INIS)

    Bischof, W.

    1980-01-01

    The development and present state of legislation and regulation in the field of handling and disposal of radioactive waste is surveyed. On the basis of the comprehensive collection of all legal sources of atomic energy law, including the radiation protection law of the Institute of Public International Law of the Goettingen University (Germany, F.R.), the report will consider provisions of international organizations (IAEA, OECD-NEA, EURATOM-Basic Norms, ICRP), of international agreements (London, Barcelona, Paris, Helsinki Conventions; civil liability conventions) and of the national law of different countries (USA, UK, France, Germany, F.R. and D.R., Italy, Switzerland, Belgium, the Netherlands, Spain). The following subjects are considered: notion and definition of radioactive waste, license-system for handling, storage and disposal; exemptions; licensing of nuclear installations and waste disposal; obligation to deliver radioactive wastes; centralized interim and final storage installations; penalties. (H.K.)

  7. Transport, handling, and interim storage of intermediate-level transuranic waste at the INEL

    International Nuclear Information System (INIS)

    Metzger, J.C.; Snyder, A.M.

    1977-09-01

    The Idaho National Engineering Laboratory stores transuranic (TRU)-contaminated waste emitting significant amounts of beta-gamma radiation. This material is referred to as intermediate-level TRU waste. The Energy Research and Development Administration requires that this waste be stored retrievably during the interim before a Federal repository becomes operational. Waste form and packaging criteria for the eventual storage of this waste at a Federal repository, i.e., the Waste Isolation Pilot Plant (WIPP), have been tentatively established. The packaging and storage techniques now in use at the Idaho National Engineering Laboratory are compatible with these criteria and also meet the requirement that the waste containers remain in a readily-retrievable, contamination-free condition during the interim storage period. The Intermediate Level Transuranic Storage Facility (ILTSF) provides below-grade storage in steel pipe vaults for intermediate-level TRU waste prior to shipment to the WIPP. Designated waste generating facilities, operated for the Energy Research and Development Administration, use a variety of packaging and transportation methods to deliver this waste to the ILTSF. Transfer of the waste containers to the ILTSF storage vaults is accomplished using handling methods compatible with these waste packaging and transport methods

  8. Engineering solutions of environmental problems in organic waste handling

    Science.gov (United States)

    Briukhanov, A. Y.; Vasilev, E. V.; Shalavina, E. V.; Kucheruk, O. N.

    2017-10-01

    This study shows the urgent need to consider modernization of agricultural production in terms of sustainable development, which takes into account environmental implications of intensive technologies in livestock farming. Some science-based approaches are offered to address related environmental challenges. High-end technologies of organic livestock waste processing were substantiated by the feasibility study and nutrient balance calculation. The technologies were assessed on the basis of best available techniques criteria, including measures such as specific capital and operational costs associated with nutrient conservation and their delivery to the plants.

  9. Methodology in the handling of the waste radioactive material

    International Nuclear Information System (INIS)

    Emeterio H, M.

    2013-10-01

    The methodology in the management of radioactive waste is constituted by an administrative part and seven technical stages: transport, classification, segregation, conditioning, treatment, packages qualification and final disposition (storage). In their diverse stages the management deserves a special attention, due to the increment of the use and application of the nuclear energy and radioactive substances, for such a reason should be managed in such a way that the exposed personnel safety and the public in general is guaranteed, protecting the integrity of the environment. (Author)

  10. TRU [transuranic] waste certification compliance requirements for acceptance of newly generated contact-handled wastes to be shipped to the Waste Isolation Pilot Plant: Revision 2

    International Nuclear Information System (INIS)

    1989-01-01

    Compliance requirements are presented for certifying that unclassified, newly generated (NG), contact-handled (CH) transuranic (TRU) solid wastes from defense programs meet the Waste Isolation Pilot Plant (WIPP) Waste Acceptance Criteria (WAC). Where appropriate, transportation and interim storage requirements are incorporated; however, interim storage sites may have additional requirements consistent with these requirements. All applicable Department of Energy (DOE) orders must continue to be met. The compliance requirements for stored or buried waste are not addressed in this document. The compliance requirements are divided into four sections, primarily determined by the general feature that the requirements address. These sections are General Requirements, Waste Container Requirements, Waste Form Requirements, and Waste Package Requirements. The waste package is the combination of waste container and waste. 10 refs., 1 fig

  11. Benchmarking the Remote-Handled Waste Facility at the West Valley Demonstration Project

    Energy Technology Data Exchange (ETDEWEB)

    O. P. Mendiratta; D. K. Ploetz

    2000-02-29

    ABSTRACT Facility decontamination activities at the West Valley Demonstration Project (WVDP), the site of a former commercial nuclear spent fuel reprocessing facility near Buffalo, New York, have resulted in the removal of radioactive waste. Due to high dose and/or high contamination levels of this waste, it needs to be handled remotely for processing and repackaging into transport/disposal-ready containers. An initial conceptual design for a Remote-Handled Waste Facility (RHWF), completed in June 1998, was estimated to cost $55 million and take 11 years to process the waste. Benchmarking the RHWF with other facilities around the world, completed in November 1998, identified unique facility design features and innovative waste pro-cessing methods. Incorporation of the benchmarking effort has led to a smaller yet fully functional, $31 million facility. To distinguish it from the June 1998 version, the revised design is called the Rescoped Remote-Handled Waste Facility (RRHWF) in this topical report. The conceptual design for the RRHWF was completed in June 1999. A design-build contract was approved by the Department of Energy in September 1999.

  12. Benchmarking the Remote-Handled Waste Facility at the West Valley Demonstration Project

    International Nuclear Information System (INIS)

    Mendiratta, O.P.; Ploetz, D.K.

    2000-01-01

    ABSTRACT Facility decontamination activities at the West Valley Demonstration Project (WVDP), the site of a former commercial nuclear spent fuel reprocessing facility near Buffalo, New York, have resulted in the removal of radioactive waste. Due to high dose and/or high contamination levels of this waste, it needs to be handled remotely for processing and repackaging into transport/disposal-ready containers. An initial conceptual design for a Remote-Handled Waste Facility (RHWF), completed in June 1998, was estimated to cost $55 million and take 11 years to process the waste. Benchmarking the RHWF with other facilities around the world, completed in November 1998, identified unique facility design features and innovative waste processing methods. Incorporation of the benchmarking effort has led to a smaller yet fully functional, $31 million facility. To distinguish it from the June 1998 version, the revised design is called the Rescoped Remote-Handled Waste Facility (RRHWF) in this topical report. The conceptual design for the RRHWF was completed in June 1999. A design-build contract was approved by the Department of Energy in September 1999

  13. Designing shafts for handling high-level radioactive wastes in mined geologic repositories

    International Nuclear Information System (INIS)

    Hambley, D.F.; Morris, J.R.

    1988-01-01

    Waste package conceptual designs developed in the United States by the U.S. Department of Energy's Office of Civilian Radioactive Waste Management are the basis for specifying the dimensions and weights of the waste package and transfer cask combinations to be hoisted in the waste handling shafts in mined geologic repositories for high-level radioactive waste. The hoist, conveyance, counterweight, and hoist ropes are then sized. Also taken into consideration are overwind and underwind arrestors and safety features required by the U.S. Nuclear Regulatory Commission. Other design features such as braking systems, chairing system design, and hoisting speed are considered in specifying waste hoisting system parameters for example repository sites

  14. Development of devices for handling with BN-350 radioactive waste

    International Nuclear Information System (INIS)

    Iksanov, A.G.; Pustobaev, S.N.; Shirobokov, Yu.P.; Pugachyev, G.P.; Baldov, A.N.; Tikhomirov, L.N.; Tkachenko, V.V.; Tazhibayeva, I.L.; Klepikov, A.Kh.; Romanenko, O.G.; Kenzhin, E.A.; Yakovlev, V.V.; Khametov, S.; Kalinkin, V.L.; Skvortsov, A.I.; Dmitriev, S.A.; Arustamov, A.E.; Zelenski, D.I.; Serebrennikov, Yu.A.

    2010-01-01

    The package of activity performed proves the correctness of the concept accepted by the Government of the Republic of Kazakhstan on the BN-350 decommissioning (three successive steps above) targeted at minimization of cost, exposure and amount of radioactive waste. Decommissioning of the high power fast breeder reactor plant is carried out for the first time and therefore the normative documents and design decisions elaborated, accepted technologies and estimation of capital expenditure and maintenance costs may enrich the database and serve as orientation for decommissioning of similar units. According to the concept accepted the BN-350 decommissioning is the process of top level of complexity that is characterized with the requirement of concurrent execution of a large scope of work by means of international teams from Kazakhstan, Russia, USA, EC, etc. Such approach needs the creation of modern effective organization schemes of interfaces and management of the Projects and will be further used in other complicated Projects

  15. Economic Floating Waste Detectionfor Surface Cleaning Robots

    Directory of Open Access Journals (Sweden)

    Sumroengrit Jakkrit

    2017-01-01

    Full Text Available Removing waste out of water surface is a routine task and can be operated by using autonomous surface cleaning robots. This paper presents amethodoflaser-based floating waste detection for surface robot guidance when waste positions are unknown beforehand. Basing on concept of refraction and reflection of laser ray, the proposed laser-based technique is proven to be applicable on floating waste detection. The economic waste detector is constructed and mounted on the robot. Five DOF equations of motion are formulated for calculation of waste position incorporating distance measured by the laser and also the robot motion caused by external wind force as well as water surface tension. Experiments were conducted on a pond with calm water and results show that the presented economic waste detection successfully identify and locate position of plastic bottles floating on water surface within the range of 5 meters.

  16. Preliminary risk analysis applied to the handling of health-care waste

    Directory of Open Access Journals (Sweden)

    Carvalho S.M.L.

    2002-01-01

    Full Text Available Between 75% and 90% of the waste produced by health-care providers no risk or is "general" health-care waste, comparable to domestic waste. The remaining 10-25% of health-care waste is regarded as hazardous due to one or more of the following characteristics: it may contain infectious agents, sharps, toxic or hazardous chemicals or it may be radioactive. Infectious health-care waste, particularly sharps, has been responsible for most of the accidents reported in the literature. In this work the preliminary risks analysis (PRA technique was used to evaluate practices in the handling of infectious health-care waste. Currently the PRA technique is being used to identify and to evaluate the potential for hazard of the activities, products, and services from facilities and industries. The system studied was a health-care establishment which has handling practices for infectious waste. Thirty-six procedures related to segregation, containment, internal collection, and storage operation were analyzed. The severity of the consequences of the failure (risk that can occur from careless management of infectious health-care waste was classified into four categories: negligible, marginal, critical, and catastrophic. The results obtained in this study showed that events with critics consequences, about 80%, may occur during the implementation of the containment operation, suggesting the need to prioritize this operation. As a result of the methodology applied in this work, a flowchart the risk series was also obtained. In the flowchart the events that can occur as a consequence of a improper handling of infectious health-care waste, which can cause critical risks such as injuries from sharps and contamination (infection from pathogenic microorganisms, are shown.

  17. Robotics and remote handling concepts for disposal of high-level nuclear waste

    International Nuclear Information System (INIS)

    McAffee, Douglas; Raczka, Norman; Schwartztrauber, Keith

    1997-01-01

    This paper summarizes preliminary remote handling and robotic concepts being developed as part of the US Department of Energy's (DOE) Yucca Mountain Project. The DOE is currently evaluating the Yucca Mountain Nevada site for suitability as a possible underground geologic repository for the disposal of high level nuclear waste. The current advanced conceptual design calls for the disposal of more than 12,000 high level nuclear waste packages within a 225 km underground network of tunnels and emplacement drifts. Many of the waste packages may weigh as much as 66 tonnes and measure 1.8 m in diameter and 5.6 m long. The waste packages will emit significant levels of radiation and heat. Therefore, remote handling is a cornerstone of the repository design and operating concepts. This paper discusses potential applications areas for robotics and remote handling technologies within the subsurface repository. It also summarizes the findings of a preliminary technology survey which reviewed available robotic and remote handling technologies developed within the nuclear, mining, rail and industrial robotics and automation industries, and at national laboratories, universities, and related research institutions and government agencies

  18. Safety of handling, storing and transportation of spent nuclear fuel and vitrified high-level wastes

    International Nuclear Information System (INIS)

    Ericsson, A.M.

    1977-11-01

    The safety of handling and transportation of spent fuel and vitrified high-level waste has been studied. Only the operations which are performed in Sweden are included. That is: - Transportation of spent fuel from the reactors to an independant spent fuel storage installation (ISFSI). - Temporary storage of spent fuel in the ISFSI. - Transportation of the spent fuel from the ISFSI to a foreign reprocessing plant. - Transportation of vitrified high-level waste to an interim storage facility. - Interim storage of vitrified high-level waste. - Handling of the vitrified high-level waste in a repository for ultimate disposal. For each stage in the handling sequence above the following items are given: - A brief technical description. - A description of precautionary measures considered in the design. - An analysis of the discharges of radioactive materials to the environment in normal operation. - An analysis of the discharges of radioactive materials due to postulated accidents. The dose to the public has been roughly and conservatively estimated for both normal and accident conditions. The expected rate of occurence are given for the accidents. The results show that above described handling sequence gives only a minor risk contribution to the public

  19. Handling and final disposal of nuclear waste. Programme for research development and other measures

    International Nuclear Information System (INIS)

    1989-09-01

    The report is divided into two parts. Part 1 presents the premises for waste management in Sweden and the waste types that are produced in Sweden. A brief description is then provided of the measures required for the handling and disposal of the various waste forms. An account of measures for decommissioning of nuclear power plants is also included. Part 2 describes the research program for 1990-1995, which includes plans for siting, repository design; studies of rock properties and chemistry, biosphere, technological barriers. Activities within two large projects, the Stripa laboratory and Natural analogues are also described. 240 refs. 40 figs

  20. Influence of Handling Practices on Material Recovery from Residential Solid Waste

    Directory of Open Access Journals (Sweden)

    Jairo F. Pereira

    2010-07-01

    Full Text Available Material recovery from municipal solid waste (MSW is becoming widely adopted in several developing countries. Residential solid waste is one of the most important components of MSW and the handling practices of the MSW by the generators have a major impact on the quality and quantity of the materials for recovery. This article analyzes the generation and composition of residential solid waste and the handling practices by users in three municipalities in Colombia that have a solid waste management plant (SWMP. The findings show that, although there are significant amounts of useful materials, their handling of the materials as “garbage”, the low recognition of recovery work, and the inadequate storage and source management practices, affect material recovery and the operation of SWMPs. These results may be taken as a reference for this type of municipality, because the solid waste management system and the type of operation of the SWMPs analyzed is similar to all of the SWMPs in the country as well as in other countries in the region.

  1. Remote-Handled Low-Level Waste Disposal Project Code of Record

    Energy Technology Data Exchange (ETDEWEB)

    S.L. Austad, P.E.; L.E. Guillen, P.E.; C. W. McKnight, P.E.; D. S. Ferguson, P.E.

    2012-06-01

    The Remote-Handled Low-Level Waste (LLW) Disposal Project addresses an anticipated shortfall in remote-handled LLW disposal capability following cessation of operations at the existing facility, which will continue until it is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). Development of a new onsite disposal facility will provide necessary remote-handled LLW disposal capability and will ensure continuity of operations that generate remote-handled LLW. This report documents the Code of Record for design of a new LLW disposal capability. The report is owned by the Design Authority, who can authorize revisions and exceptions. This report will be retained for the lifetime of the facility.

  2. Remote-Handled Low-Level Waste Disposal Project Code of Record

    Energy Technology Data Exchange (ETDEWEB)

    S.L. Austad, P.E.; L.E. Guillen, P.E.; C. W. McKnight, P.E.; D. S. Ferguson, P.E.

    2014-06-01

    The Remote-Handled Low-Level Waste (LLW) Disposal Project addresses an anticipated shortfall in remote-handled LLW disposal capability following cessation of operations at the existing facility, which will continue until it is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). Development of a new onsite disposal facility will provide necessary remote-handled LLW disposal capability and will ensure continuity of operations that generate remote-handled LLW. This report documents the Code of Record for design of a new LLW disposal capability. The report is owned by the Design Authority, who can authorize revisions and exceptions. This report will be retained for the lifetime of the facility.

  3. Remote-Handled Low-Level Waste Disposal Project Code of Record

    Energy Technology Data Exchange (ETDEWEB)

    Austad, S. L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Guillen, L. E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); McKnight, C. W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ferguson, D. S. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-04-01

    The Remote-Handled Low-Level Waste (LLW) Disposal Project addresses an anticipated shortfall in remote-handled LLW disposal capability following cessation of operations at the existing facility, which will continue until it is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). Development of a new onsite disposal facility will provide necessary remote-handled LLW disposal capability and will ensure continuity of operations that generate remote-handled LLW. This report documents the Code of Record for design of a new LLW disposal capability. The report is owned by the Design Authority, who can authorize revisions and exceptions. This report will be retained for the lifetime of the facility.

  4. Conceptual Design Report for Remote-Handled Low-Level Waste Disposal Facility

    Energy Technology Data Exchange (ETDEWEB)

    Lisa Harvego; David Duncan; Joan Connolly; Margaret Hinman; Charles Marcinkiewicz; Gary Mecham

    2010-10-01

    This conceptual design report addresses development of replacement remote-handled low-level waste disposal capability for the Idaho National Laboratory. Current disposal capability at the Radioactive Waste Management Complex is planned until the facility is full or until it must be closed in preparation for final remediation (approximately at the end of Fiscal Year 2017). This conceptual design report includes key project assumptions; design options considered in development of the proposed onsite disposal facility (the highest ranked alternative for providing continued uninterrupted remote-handled low level waste disposal capability); process and facility descriptions; safety and environmental requirements that would apply to the proposed facility; and the proposed cost and schedule for funding, design, construction, and operation of the proposed onsite disposal facility.

  5. Conceptual Design Report for the Remote-Handled Low-Level Waste Disposal Project

    Energy Technology Data Exchange (ETDEWEB)

    David Duncan

    2011-05-01

    This conceptual design report addresses development of replacement remote-handled low-level waste disposal capability for the Idaho National Laboratory. Current disposal capability at the Radioactive Waste Management Complex is planned until the facility is full or until it must be closed in preparation for final remediation (approximately at the end of Fiscal Year 2017). This conceptual design report includes key project assumptions; design options considered in development of the proposed onsite disposal facility (the highest ranked alternative for providing continued uninterrupted remote-handled low level waste disposal capability); process and facility descriptions; safety and environmental requirements that would apply to the proposed facility; and the proposed cost and schedule for funding, design, construction, and operation of the proposed onsite disposal facility.

  6. Conceptual Design Report for the Remote-Handled Low-Level Waste Disposal Project

    Energy Technology Data Exchange (ETDEWEB)

    Lisa Harvego; David Duncan; Joan Connolly; Margaret Hinman; Charles Marcinkiewicz; Gary Mecham

    2011-03-01

    This conceptual design report addresses development of replacement remote-handled low-level waste disposal capability for the Idaho National Laboratory. Current disposal capability at the Radioactive Waste Management Complex is planned until the facility is full or until it must be closed in preparation for final remediation (approximately at the end of Fiscal Year 2017). This conceptual design report includes key project assumptions; design options considered in development of the proposed onsite disposal facility (the highest ranked alternative for providing continued uninterrupted remote-handled low level waste disposal capability); process and facility descriptions; safety and environmental requirements that would apply to the proposed facility; and the proposed cost and schedule for funding, design, construction, and operation of the proposed onsite disposal facility.

  7. Handling, assessment, transport and disposal of tritiated waste materials at JET

    International Nuclear Information System (INIS)

    Newbert, G.; Haigh, A.; Atkins, G.

    1995-01-01

    All types of JET radioactive wastes are received for disposal at the Waste Handling Facility (WHF) which features a waste sorting and sampling station, a glove box, a compactor, and packaging and transfer systems. The WHF is operated as a contamination control area with monitored tritium discharges. Two main types of tritium monitors used are liquid scintillation counters and ionization chambers, and samples of various components and materials have now been assessed for tritium. The results so far indicate a widespread of tritium levels from 2Bq/g for cold gas transfer lines to 200kBq/g for in-vessel tiles. General soft housekeeping waste is assessed by a sniffing technique which has a limit of detection corresponding to 120Bq/g. Investigation of improved methods of tritium measurement and of component detritiation was made to facilitate future waste disposal. 8 refs., 6 figs., 2 tabs

  8. Vault lining for 340 waste handling facility, 300 area

    International Nuclear Information System (INIS)

    Hollenbeck, R.G.

    1997-01-01

    Coating systems by Protection Enterprises, Ameron, Carboline, and Steelcote were evaluated. Each manufacturer has a coating system that is acceptable for use in the 340 Vault (see Appendix A). The choice of which system to use will be made after in-place adhesion tests are complete. The Protection Enterprises coating has the greatest potential for acceptable adhesion with minimal surface preparation. Total project cost for engineering and construction is $1,220,000 including 50% for contingency (see Appendix B). If the existing vault coverblock access hatch can satisfy entry requirements, $95,000 can be saved from the removal of coverblocks and the erection and disassembly of the greenhouse

  9. Application of advanced remote systems technology to future waste handling facilities

    International Nuclear Information System (INIS)

    Kring, C.T.; Meacham, S.A.

    1987-01-01

    The Consolidated Fuel Reprocessing Program (CFRP) at Oak Ridge National Laboratory (ORNL) has been advancing the technology of remote handling and remote maintenance for in-cell systems planned for future nuclear fuel reprocessing plants. Much of the experience and technology developed over the past decade in this endeavor is directly applicable to the proposed in-cell systems being considered for the facilities of the Federal Waste Management System (FWMS). The application of teleoperated, force-reflecting servomanipulators with television viewing could be a major step forward in waste handling facility design. Primary emphasis in the current program is the operation of a prototype remote handling and maintenance system, the advanced servomanipulator (ASM), which specifically addresses the requirements of fuel reprocessing and waste handling with emphasis on force reflection, remote maintainability, reliability, radiation tolerance, and corrosion resistance. Concurrent with the evolution of dexterous manipulators, concepts have also been developed that provide guidance for standardization of the design of the remotely operated and maintained equipment, the interface between the maintenance tools and the equipment, and the interface between the in-cell components and the facility

  10. Remote Handled TRU Waste Status and Activities and Challenges at the Hanford Site

    International Nuclear Information System (INIS)

    MCKENNEY, D.E.

    2000-01-01

    A significant portion of the Department of Energy's forecast volume of remote-handled (RH) transuranic (TRU) waste will originate from the Hanford Site. The forecasted Hanford RH-TRU waste volume of over 2000 cubic meters may constitute over one-third of the forecast inventory of RH-TRU destined for disposal at the Waste Isolation Pilot Plant (WIPP). To date, the Hanford TRU waste program has focused on the retrieval, treatment and certification of the contact-handled transuranic (CH-TRU) wastes. This near-term focus on CH-TRU is consistent with the National TRU Program plans and capabilities. The first shipment of CH-TRU waste from Hanford to the WIPP is scheduled early in Calendar Year 2000. Shipments of RH-TRU from Hanford to the WIPP are scheduled to begin in Fiscal Year 2006 per the National TRU Waste Management Plan. This schedule has been incorporated into milestones within the Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement). These Tri-Party milestones (designated the ''M-91'' series of milestones) relate to development of project management plans, completion of design efforts, construction and contracting schedules, and initiation of process operations. The milestone allows for modification of an existing facility, construction of a new facility, and/or commercial contracting to provide the capabilities for processing and certification of RH-TRU wastes for disposal at the WIPP. The development of a Project Management Plan (PMP) for TRU waste is the first significant step in the development of a program for disposal of Hanford's RH-TRU waste. This PMP will address the path forward for disposition of waste streams that cannot be prepared for disposal in the Hanford Waste Receiving and Processing facility (a contact-handled, small container facility) or other Site facilities. The PMP development effort has been initiated, and the PMP will be provided to the regulators for their approval by June 30, 2000. This plan will detail the

  11. Demonstration of a remotely operated TRU waste size-reduction and material handling process

    International Nuclear Information System (INIS)

    Stewart, J.A. III; Schuler, T.F.; Ward, C.R.

    1986-01-01

    Noncombustible Pu-238 and Pu-239 waste is generated as a result of normal operation and decommissioning activity at the Savannah River Plant and is being retrievably stored at the site. As part of the long-term plan to process the stored waste and current waste for permanent disposal, a remote size-reduction and material handling process is being tested at Savannah River Laboratory to provide design support for the plant TRU Waste Facility scheduled to be completed in 1993. The process consists of a large, low-speed shredder and material handling system, a remote worktable, a bagless transfer system, and a robotically controlled manipulator, or Telerobot. Initial testing of the shredder and material handling system and a cycle test of the bagless transfer system were completed. Initial Telerobot run-in and system evaluation was completed. User software was evaluated and modified to support complete menu-driven operation. Telerobot prototype size-reduction tooling was designed and successfully tested. Complete nonradioactive testing of the equipment is scheduled to be completed in 1987

  12. Rustler Formation in the waste handling and exhaust shafts, Waste Isolation Pilot Plant (WIPP) site, southeastern New Mexico

    International Nuclear Information System (INIS)

    Holt, R.M.; Powers, D.W.

    1987-01-01

    The Permian Rustler Formation was recently examined in detail in two shafts at the WIPP site: the waste handling shaft (waste shaft) and the exhaust shaft. Fresh exposures of the Rustler in the shafts exhibited abundant primary sedimentary structures. The abundance of primary sedimentary structures observed in the shafts is unequaled in previously described sections. Data are reported here in their stratigraphic context as an initial basis for evaluation of depositional environments of the Rustler and reevaluating the role of dissolution in the formation of the Rustler. 10 refs

  13. Legal provisions concerning the handling and disposal of radioactive waste in international and national law

    International Nuclear Information System (INIS)

    Bischof, W.

    1980-01-01

    A short survey is given on the situation of international legislation concerning radioactive waste handling and disposal. There are special rules on the disposal of nuclear waste in a number of conventions (Geneva 1958, London 1972, Helsinki 1974, Paris 1974, Barcellone 1976) on the protection of the marine environment and of the high sea against pollutions. In 1974 and 1978, the International Atomic Energy Agency made further recommendations concerning radioactive wastes referred to in the London Convention. In 1977, the Organisation for Economic Cooperation and Development also set up within its Nuclear Energy Agency (NEA) a multilateral consultation and surveillance mechanism for the sea-dumping of radioactive waste. The NEA has since published recommendations on the sea-dumping of radioactive waste. In 1975, it was agreed to abide by the Antarctic Treaty of 1959 not to dispose any nuclear waste on the Antarctic Region. There is at present no absolute prohibition of radioactive waste disposal in outer space but the Member States of the United Nations are responsible for such activities. As regards national legislation, the legal provisions for 13 different countries on radioactive waste disposal are listed. (UK)

  14. Generation, on-site storage; handling and processing of industrial waste of Tehran

    International Nuclear Information System (INIS)

    Abduli, M.A.

    1997-01-01

    This paper describes out the present status of generation, on-site handling, processing and storage of industrial waste in Tehran. In this investigation, 67 large scale factories of different industrial groups were randomly selected. Above cited functional elements of these factories were surveyed. In this investigation a close contact with each factory was required, thus a questionnaire was prepared and distributed among these factories. The relationship between daily weight of the industrial waste (Y) and number of employer of each factory (x) is found to be Y=547.4 + 0.58 x. The relationship between daily volume of industrial waste (V), and daily weight of waste generated in each factory (Y) can be described by V=1.56 + 0.00078 Y. About 68% of the factories have their own interim storage site and the rest of the factories do not possess any on-site storage facility

  15. French surface disposal experience. The disposal of large waste

    International Nuclear Information System (INIS)

    Dutzer, Michel; Lecoq, Pascal; Duret, Franck; Mandoki, Robert

    2006-01-01

    More than 90 percent of the volume of radioactive waste that are generated in France can be managed in surface disposal facilities. Two facilities are presently operated by ANDRA: the Centre de l'Aube disposal facility that is dedicated to low and intermediate short lived waste and the Morvilliers facility for very low level waste. The Centre de l'Aube facility was designed at the end of the years 1980 to replace the Centre de la Manche facility that ended operation in 1994. In order to achieve as low external exposure as possible for workers it was decided to use remote handling systems as much as possible. Therefore it was necessary to standardize the types of waste containers. But taking into account the fact that these waste were conditioned in existing facilities, it was not possible to change a major part of existing packages. As a consequence, 6 mobile roofs were constructed to handle 12 different types of waste packages in the disposal vaults. The scope of Centre de l'Aube was mainly to dispose operational waste. However some packages, as 5 or 10 m 3 metallic boxes, could be used for larger waste generated by decommissioning activities. The corresponding flow was supposed to be small. After the first years of operations, it appeared interesting to develop special procedures to dispose specific large waste in order to avoid external exposure costly cutting works in the generating facilities. A 40 m 3 box and a large remote handling device were disposed in vaults that were currently used for other types of packages. Such a technique could not be used for the disposal of vessel heads that were replaced in 55 pressurised water power reactors. The duration of disposal and conditioning operation was not compatible with the flow of standard packages that were delivered in the vaults. Therefore a specific type of vault was designed, including handling and conditioning equipment. The first pressure vessel head was delivered on the 29 of July 2004, 6 heads have been

  16. Preliminary seismic design cost-benefit assessment of the tuff repository waste-handling facilities

    International Nuclear Information System (INIS)

    Subramanian, C.V.; Abrahamson, N.; Hadjian, A.H.

    1989-02-01

    This report presents a preliminary assessment of the costs and benefits associated with changes in the seismic design basis of waste-handling facilities. The objectives of the study are to understand the capability of the current seismic design of the waste-handling facilities to mitigate seismic hazards, evaluate how different design levels and design measures might be used toward mitigating seismic hazards, assess the costs and benefits of alternative seismic design levels, and develop recommendations for possible modifications to the seismic design basis. This preliminary assessment is based primarily on expert judgment solicited in an interdisciplinary workshop environment. The estimated costs for individual attributes and the assumptions underlying these cost estimates (seismic hazard levels, fragilities, radioactive-release scenarios, etc.) are subject to large uncertainties, which are generally identified but not treated explicitly in this preliminary analysis. The major conclusions of the report do not appear to be very sensitive to these uncertainties. 41 refs., 51 figs., 35 tabs

  17. Environmental Assessment for the Independent Waste Handling Facility, 211-F at the Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-08-01

    Currently, liquid Low Activity Waste (LAW) and liquid High Activity Waste (HAW) are generated from various process operational facilities/processes throughout the Savannah River Site (SRS) as depicted on Figure 2-1. Prior to storage in the F-Area tank farm, these wastes are neutralized and concentrated to minimize their volume. The Waste Handling Facility (211-3F) at Building 211-F Complex (see Figure 2-2) is the only existing facility onsite equipped to receive acidic HAW for neutralization and volume reduction processing. Currently, Building 221-F Canyon (see Figure 2-2) houses the neutralization and evaporation facilities for HAW volume reduction and provides support services such as electric power and plant, process, and instrument air, waste transfer capabilities, etc., for 21 1-F operations. The future plan is to deactivate the 221-F building. DOE`s purpose is to be able to process the LAW/HAW that will continue to be generated on site. DOE needs to establish an alternative liquid waste receipt and treatment capability to support site facilities with a continuing mission. The desire is for Building 211-F to provide the receipt and neutralization functions for LAW and HAW independent of 221-F Canyon. The neutralization capability is required to be part of the Nuclear Materials Stabilization Programs (NMSP) facilities since the liquid waste generated by the various site facilities is acidic. Tn order for Waste Management to receive the waste streams, the solutions must be neutralized to meet Waste Management`s acceptance criteria. The Waste Management system is caustic in nature to prevent corrosion and the subsequent potential failure of tanks and associated piping and hardware.

  18. A systematic critical review of epidemiological studies on public health concerns of municipal solid waste handling.

    Science.gov (United States)

    Ncube, France; Ncube, Esper Jacobeth; Voyi, Kuku

    2017-03-01

    The ultimate aim of this review was to summarise the epidemiological evidence on the association between municipal solid waste management operations and health risks to populations residing near landfills and incinerators, waste workers and recyclers. To accomplish this, the sub-aims of this review article were to (1) examine the health risks posed by municipal solid waste management activities, (2) determine the strengths and gaps of available literature on health risks from municipal waste management operations and (3) suggest possible research needs for future studies. The article reviewed epidemiological literature on public health concerns of municipal solid waste handling published in the period 1995-2014. The PubMed and MEDLINE computerised literature searches were employed to identify the relevant papers using the keywords solid waste, waste management, health risks, recycling, landfills and incinerators. Additionally, all references of potential papers were examined to determine more articles that met the inclusion criteria. A total of 379 papers were identified, but after intensive screening only 72 met the inclusion criteria and were reviewed. Of these studies, 33 were on adverse health effects in communities living near waste dumpsites or incinerators, 24 on municipal solid waste workers and 15 on informal waste recyclers. Reviewed studies were unable to demonstrate a causal or non-causal relationship due to various limitations. In light of the above findings, our review concludes that overall epidemiological evidence in reviewed articles is inadequate mainly due to methodological limitations and future research needs to develop tools capable of demonstrating causal or non-causal relationships between specific waste management operations and adverse health endpoints.

  19. Mission Need Statement for the Idaho National Laboratory Remote-Handled Low-Level Waste Disposal Project

    International Nuclear Information System (INIS)

    Harvego, Lisa

    2009-01-01

    The Idaho National Laboratory proposes to establish replacement remote-handled low-level waste disposal capability to meet Nuclear Energy and Naval Reactors mission-critical, remote-handled low-level waste disposal needs beyond planned cessation of existing disposal capability at the end of Fiscal Year 2015. Remote-handled low-level waste is generated from nuclear programs conducted at the Idaho National Laboratory, including spent nuclear fuel handling and operations at the Naval Reactors Facility and operations at the Advanced Test Reactor. Remote-handled low-level waste also will be generated by new programs and from segregation and treatment (as necessary) of remote-handled scrap and waste currently stored in the Radioactive Scrap and Waste Facility at the Materials and Fuels Complex. Replacement disposal capability must be in place by Fiscal Year 2016 to support uninterrupted Idaho operations. This mission need statement provides the basis for the laboratory's recommendation to the Department of Energy to proceed with establishing the replacement remote-handled low-level waste disposal capability, project assumptions and constraints, and preliminary cost and schedule information for developing the proposed capability. Without continued remote-handled low-level waste disposal capability, Department of Energy missions at the Idaho National Laboratory would be jeopardized, including operations at the Naval Reactors Facility that are critical to effective execution of the Naval Nuclear Propulsion Program and national security. Remote-handled low-level waste disposal capability is also critical to the Department of Energy's ability to meet obligations with the State of Idaho

  20. Mission Need Statement for the Idaho National Laboratory Remote-Handled Low-Level Waste Disposal Project

    Energy Technology Data Exchange (ETDEWEB)

    Lisa Harvego

    2009-06-01

    The Idaho National Laboratory proposes to establish replacement remote-handled low-level waste disposal capability to meet Nuclear Energy and Naval Reactors mission-critical, remote-handled low-level waste disposal needs beyond planned cessation of existing disposal capability at the end of Fiscal Year 2015. Remote-handled low-level waste is generated from nuclear programs conducted at the Idaho National Laboratory, including spent nuclear fuel handling and operations at the Naval Reactors Facility and operations at the Advanced Test Reactor. Remote-handled low-level waste also will be generated by new programs and from segregation and treatment (as necessary) of remote-handled scrap and waste currently stored in the Radioactive Scrap and Waste Facility at the Materials and Fuels Complex. Replacement disposal capability must be in place by Fiscal Year 2016 to support uninterrupted Idaho operations. This mission need statement provides the basis for the laboratory’s recommendation to the Department of Energy to proceed with establishing the replacement remote-handled low-level waste disposal capability, project assumptions and constraints, and preliminary cost and schedule information for developing the proposed capability. Without continued remote-handled low-level waste disposal capability, Department of Energy missions at the Idaho National Laboratory would be jeopardized, including operations at the Naval Reactors Facility that are critical to effective execution of the Naval Nuclear Propulsion Program and national security. Remote-handled low-level waste disposal capability is also critical to the Department of Energy’s ability to meet obligations with the State of Idaho.

  1. Graphical models for simulation and control of robotic systems for waste handling

    International Nuclear Information System (INIS)

    Drotning, W.D.; Bennett, P.C.

    1992-01-01

    This paper discusses detailed geometric models which have been used within a graphical simulation environment to study transportation cask facility design and to perform design and analyses of robotic systems for handling of nuclear waste. The models form the basis for a robot control environment which provides safety, flexibility, and reliability for operations which span the spectrum from autonomous control to tasks requiring direct human intervention

  2. Treatment of plutonium-contaminated solid waste: a review of handling systems

    International Nuclear Information System (INIS)

    Meredith, B.E.; Hardy, A.R.

    1985-02-01

    Handling techniques are reviewed to identify those suitable for adaptation for use in transporting large items of redundant plutonium contaminated plant and equipment to a remotely operated size reduction facility, moving them into the facility, presenting them to size reduction equipment and loading the processed waste into drums. It is concluded that an integrated system based on a combination of slatted conveyors, roller tables, air transporters and manipulators, merits further consideration. An appropriate experimental programme is outlined. (author)

  3. A passive-active neutron device for assaying remote-handled transuranic waste

    International Nuclear Information System (INIS)

    Estep, R.J.; Coop, K.L.; Deane, T.M.; Lujan, J.E.

    1990-01-01

    A combined passive-active neutron assay device was constructed for assaying remote-handled transuranic waste. A study of matrix and source position effects in active assays showed that a knowledge of the source position alone is not sufficient to correct for position-related errors in highly moderating or absorbing matrices. An alternate function for the active assay of solid fuel pellets was derived, although the efficacy of this approach remains to be established

  4. A comparison of the consequences of different waste handling systems in two Danish communities

    DEFF Research Database (Denmark)

    Grunert, Suzanne C.; Thøgersen, John

    1995-01-01

    a system based solely on non-economic incentives. The main objective was to compare citizen`s beliefs and attitudes towards waste handling systems and their consequence for motivations to co-operate. Th groups of hypotheses concerning the beliefs-attitude relationship, differences in attitudes between...... cities, and the use of economic incentives were tested. Whereas beliefs influenced attitudes in the expected direction, the consequences of economi incentives for differences in attitudes were less clear....

  5. DOE assay methods used for characterization of contact-handled transuranic waste

    Energy Technology Data Exchange (ETDEWEB)

    Schultz, F.J. (Oak Ridge National Lab., TN (United States)); Caldwell, J.T. (Pajarito Scientific Corp., Los Alamos, NM (United States))

    1991-08-01

    US Department of Energy methods used for characterization of contact-handled transuranic (CH-TRU) waste prior to shipment to the Waste Isolation Pilot Plant (WIPP) are described and listed by contractor site. The methods described are part of the certification process. All CH-TRU waste must be assayed for determination of fissile material content and decay heat values prior to shipment and prior to storage on-site. Both nondestructive assay (NDA) and destructive assay methods are discussed, and new NDA developments such as passive-action neutron (PAN) crate counter improvements and neutron imaging are detailed. Specifically addressed are assay method physics; applicability to CH-TRU wastes; calibration standards and implementation; operator training requirements and practices; assay procedures; assay precision, bias, and limit of detection; and assay limitation. While PAN is a new technique and does not yet have established American Society for Testing and Materials. American National Standards Institute, or Nuclear Regulatory Commission guidelines or methods describing proper calibration procedures, equipment setup, etc., comparisons of PAN data with the more established assay methods (e.g., segmented gamma scanning) have demonstrated its reliability and accuracy. Assay methods employed by DOE have been shown to reliable and accurate in determining fissile, radionuclide, alpha-curie content, and decay heat values of CH-TRU wastes. These parameters are therefore used to characterize packaged waste for use in certification programs such as that used in shipment of CH-TRU waste to the WIPP. 36 refs., 10 figs., 7 tabs.

  6. Assessment of work-related accidents associated with waste handling in Belo Horizonte (Brazil).

    Science.gov (United States)

    Mol, Marcos Pg; Pereira, Amanda F; Greco, Dirceu B; Cairncross, Sandy; Heller, Leo

    2017-10-01

    As more urban solid waste is generated, managing it becomes ever more challenging and the potential impacts on the environment and human health also become greater. Handling waste - including collection, treatment and final disposal - entails risks of work accidents. This article assesses the perception of waste management workers regarding work-related accidents in domestic and health service contexts in Belo Horizonte, Brazil. These perceptions are compared with national data from the Ministry of Social Security on accidents involving workers in solid waste management. A high proportion of accidents involves cuts and puncture injuries; 53.9% among workers exposed to domestic waste and 75% among those exposed to health service waste. Muscular lesions and fractures accounted for 25.7% and 12.5% of accidents, respectively. Data from the Ministry of Social Security diverge from the local survey results, presumably owing to under-reporting, which is frequent in this sector. Greater commitment is needed from managers and supervisory entities to ensure that effective measures are taken to protect workers' health and quality of life. Moreover, workers should defend their right to demand an accurate registry of accidents to complement monitoring performed by health professionals trained in risk identification. This would contribute to the improved recovery of injured workers and would require managers in waste management to prepare effective preventive action.

  7. DOE assay methods used for characterization of contact-handled transuranic waste

    International Nuclear Information System (INIS)

    Schultz, F.J.; Caldwell, J.T.

    1991-08-01

    US Department of Energy methods used for characterization of contact-handled transuranic (CH-TRU) waste prior to shipment to the Waste Isolation Pilot Plant (WIPP) are described and listed by contractor site. The methods described are part of the certification process. All CH-TRU waste must be assayed for determination of fissile material content and decay heat values prior to shipment and prior to storage on-site. Both nondestructive assay (NDA) and destructive assay methods are discussed, and new NDA developments such as passive-action neutron (PAN) crate counter improvements and neutron imaging are detailed. Specifically addressed are assay method physics; applicability to CH-TRU wastes; calibration standards and implementation; operator training requirements and practices; assay procedures; assay precision, bias, and limit of detection; and assay limitation. While PAN is a new technique and does not yet have established American Society for Testing and Materials. American National Standards Institute, or Nuclear Regulatory Commission guidelines or methods describing proper calibration procedures, equipment setup, etc., comparisons of PAN data with the more established assay methods (e.g., segmented gamma scanning) have demonstrated its reliability and accuracy. Assay methods employed by DOE have been shown to reliable and accurate in determining fissile, radionuclide, alpha-curie content, and decay heat values of CH-TRU wastes. These parameters are therefore used to characterize packaged waste for use in certification programs such as that used in shipment of CH-TRU waste to the WIPP. 36 refs., 10 figs., 7 tabs

  8. The presence and partitioning behavior of flame retardants in waste, leachate, and air particles from Norwegian waste-handling facilities.

    Science.gov (United States)

    Morin, Nicolas A O; Andersson, Patrik L; Hale, Sarah E; Arp, Hans Peter H

    2017-12-01

    Flame retardants in commercial products eventually make their way into the waste stream. Herein the presence of flame retardants in Norwegian landfills, incineration facilities and recycling sorting/defragmenting facilities is investigated. These facilities handled waste electrical and electronic equipment (WEEE), vehicles, digestate, glass, combustibles, bottom ash and fly ash. The flame retardants considered included polybrominated diphenyl ethers (∑BDE-10) as well as dechlorane plus, polybrominated biphenyls, hexabromobenzene, pentabromotoluene and pentabromoethylbenzene (collectively referred to as ∑FR-7). Plastic, WEEE and vehicles contained the largest amount of flame retardants (∑BDE-10: 45,000-210,000μg/kg; ∑FR-7: 300-13,000μg/kg). It was hypothesized leachate and air concentrations from facilities that sort/defragment WEEE and vehicles would be the highest. This was supported for total air phase concentrations (∑BDE-10: 9000-195,000pg/m 3 WEEE/vehicle facilities, 80-900pg/m 3 in incineration/sorting and landfill sites), but not for water leachate concentrations (e.g., ∑BDE-10: 15-3500ng/L in WEEE/Vehicle facilities and 1-250ng/L in landfill sites). Landfill leachate exhibited similar concentrations as WEEE/vehicle sorting and defragmenting facility leachate. To better account for concentrations in leachates at the different facilities, waste-water partitioning coefficients, K waste were measured (for the first time to our knowledge for flame retardants). WEEE and plastic waste had elevated K waste compared to other wastes, likely because flame retardants are directly added to these materials. The results of this study have implications for the development of strategies to reduce exposure and environmental emissions of flame retardants in waste and recycled products through improved waste management practices. Copyright © 2017. Published by Elsevier B.V.

  9. The crane handling system for 500 litre drums of cemented radioactive waste

    International Nuclear Information System (INIS)

    Staples, A.T.

    1991-01-01

    As part of the AEA Technology strategy for dealing with radioactive wastes new waste treatment facilities are being built at the Winfrith Technology Centre (WTC), Dorset. One of the facilities at WTC is the Treated Radwaste Store (TRS) which is designed to store sealed 500 litre capacity drums of treated waste for an interim period until the national disposal facility is operational. Within the TRS two cranes have been incorporated, one spanning the entire width and travelling the length of the Store. The second operates within the area designated for drum handling during inspection work. The development of the design of these cranes and their associated control systems, to meet the complex requirements of operations whilst also satisfying the reliability and safety criteria, is discussed within the paper. (author)

  10. Handling e-waste in developed and developing countries: Initiatives, practices, and consequences

    Energy Technology Data Exchange (ETDEWEB)

    Sthiannopkao, Suthipong, E-mail: suthisuthi@gmail.com [Department of Environmental Engineering, College of Engineering, Dong-A University, 37 Nakdong-Daero 550 beon-gil Saha-gu, Busan (Korea, Republic of); Wong, Ming Hung [Croucher Institute for Environmental Sciences, Hong Kong Baptist University, Hong Kong (China)

    2013-10-01

    Discarded electronic goods contain a range of toxic materials requiring special handling. Developed countries have conventions, directives, and laws to regulate their disposal, most based on extended producer responsibility. Manufacturers take back items collected by retailers and local governments for safe destruction or recovery of materials. Compliance, however, is difficult to assure, and frequently runs against economic incentives. The expense of proper disposal leads to the shipment of large amounts of e-waste to China, India, Pakistan, Nigeria, and other developing countries. Shipment is often through middlemen, and under tariff classifications that make quantities difficult to assess. There, despite the intents of national regulations and hazardous waste laws, most e-waste is treated as general refuse, or crudely processed, often by burning or acid baths, with recovery of only a few materials of value. As dioxins, furans, and heavy metals are released, harm to the environment, workers, and area residents is inevitable. The faster growth of e-waste generated in the developing than in the developed world presages continued expansion of a pervasive and inexpensive informal processing sector, efficient in its own way, but inherently hazard-ridden. - Highlights: ► Much e-waste, expensive to process safely, illegally goes to developing countries. ► E-waste processing in developing countries pollutes with heavy metals and dioxins. ► Well-conceived developing world waste regulations lack enforceability. ► Crude e-waste processing cannot recover several rare materials. ► The amount of e-waste unsafely processed will continue to grow.

  11. Handling e-waste in developed and developing countries: Initiatives, practices, and consequences

    International Nuclear Information System (INIS)

    Sthiannopkao, Suthipong; Wong, Ming Hung

    2013-01-01

    Discarded electronic goods contain a range of toxic materials requiring special handling. Developed countries have conventions, directives, and laws to regulate their disposal, most based on extended producer responsibility. Manufacturers take back items collected by retailers and local governments for safe destruction or recovery of materials. Compliance, however, is difficult to assure, and frequently runs against economic incentives. The expense of proper disposal leads to the shipment of large amounts of e-waste to China, India, Pakistan, Nigeria, and other developing countries. Shipment is often through middlemen, and under tariff classifications that make quantities difficult to assess. There, despite the intents of national regulations and hazardous waste laws, most e-waste is treated as general refuse, or crudely processed, often by burning or acid baths, with recovery of only a few materials of value. As dioxins, furans, and heavy metals are released, harm to the environment, workers, and area residents is inevitable. The faster growth of e-waste generated in the developing than in the developed world presages continued expansion of a pervasive and inexpensive informal processing sector, efficient in its own way, but inherently hazard-ridden. - Highlights: ► Much e-waste, expensive to process safely, illegally goes to developing countries. ► E-waste processing in developing countries pollutes with heavy metals and dioxins. ► Well-conceived developing world waste regulations lack enforceability. ► Crude e-waste processing cannot recover several rare materials. ► The amount of e-waste unsafely processed will continue to grow

  12. Conditioning and handling of tritiated wastes at Canadian nuclear power facilities

    International Nuclear Information System (INIS)

    Krochmalnek, L.S.; Krasznai, J.P.; Carney, M.

    1987-04-01

    Ontario Hydro operates a 10,000 MW capacity nuclear power system utilizing the CANDU pressurized heavy water reactor design. The use of D 2 O as moderator and coolant results in the production of about 2400 Ci of tritium per MWe-yr. As a result, there is significant Canadian experience in the treatment, handling, transport and storage of tritiated wastes. Ontario Hydro operates its own reactor waste storage site which includes systems for volume reduction, immobilization and packaging of wastes. In addition, a facility to remove tritium from heavy water is presently being commissioned at the Darlington nuclear site. This facility will generate tritiated liquid and solid waste that will have to be properly conditioned prior to storage or disposal. The nature of these various wastes and the processes/packaging required to meet storage/disposal criteria are judged to have relevance to investigations in fusion facility waste arisings. Experience to date, planned operational procedures and ongoing R and D in this area are described

  13. Handling of spent nuclear fuel and final storage of nitrified high level reprocessing waste

    International Nuclear Information System (INIS)

    The following stages of handling and transport of the fuel on its way to final storage are dealt with in the report. 1) The spent nuclear fuel is stored at the power station or in the central fuel storage facility awaiting reprocessing. 2) The fuel is reprocessed, i.e. uranium, plutonium and waste are separated from each other. Reprocessing does not take place in Sweden. The highlevel waste is vitrified and can be sent back to Sweden in the 1990s. 3) Vitrified waste is stored for about 30 years awaiting deposition in the final repository. 4) The waste is encapsulated in highly durable materials to prevent groundwater from coming into contact with the waste glass while the radioactivity of the waste is still high. 5) The canisters are emplaced in a final repository which is built at a depth of 500 m in rock of low permeability. 6) All tunnels and shafts are filled with a mixture of clay and sand of low permeability. A detailed analysis of possible harmful effects resulting from normal acitivties and from conceivable accidents is presented in a special section. (author)

  14. Handling of micro objects: investigation of mechanical gripper functional surfaces

    DEFF Research Database (Denmark)

    Gegeckaite, Asta; Hansen, Hans Nørgaard; De Chiffre, Leonardo

    2007-01-01

    between the micro object and the gripper do not allow simple picking and releasing of the object. This effect can be overcome by modifying the functional surface of the gripper. The functional surface of the gripper was modified by different machining techniques. The results of this investigation...

  15. Siting Study for the Remote-Handled Low-Level Waste Disposal Project

    Energy Technology Data Exchange (ETDEWEB)

    Lisa Harvego; Joan Connolly; Lance Peterson; Brennon Orr; Bob Starr

    2010-10-01

    The U.S. Department of Energy has identified a mission need for continued disposal capacity for remote-handled low-level waste (LLW) generated at the Idaho National Laboratory (INL). An alternatives analysis that was conducted to evaluate strategies to achieve this mission need identified two broad options for disposal of INL generated remote-handled LLW: (1) offsite disposal and (2) onsite disposal. The purpose of this study is to identify candidate sites or locations within INL boundaries for the alternative of an onsite remote handled LLW disposal facility and recommend the highest-ranked locations for consideration in the National Environmental Policy Act process. The study implements an evaluation based on consideration of five key elements: (1) regulations, (2) key assumptions, (3) conceptual design, (4) facility performance, and (5) previous INL siting study criteria, and uses a five-step process to identify, screen, evaluate, score, and rank 34 separate sites located across INL. The result of the evaluation is identification of two recommended alternative locations for siting an onsite remote-handled LLW disposal facility. The two alternative locations that best meet the evaluation criteria are (1) near the Advanced Test Reactor Complex and (2) west of the Idaho Comprehensive Environmental Response, Compensation, and Liability Act Disposal Facility.

  16. Highlights of the American Nuclear Society topical meeting on the treatment and handling of radioactive wastes

    International Nuclear Information System (INIS)

    Blasewitz, A.G.; Lerch, R.E.; Richardson, G.L.

    1983-01-01

    The American Nuclear Society Topical Meeting on the Treatment and Handling of Radioactive Wastes was held in Richland, Washington, from 19-22 April 1982. The object of the meeting was to provide a thorough assessment of the status of technology. The response to the meeting was excellent: 123 papers were presented. There were 505 registrations; 83 were from outside the USA, representing 13 countries. The large and diverse attendance provided a broad technological view and perspective. The following major points emerged from the conference: (1) In an extensive world-wide effort, techniques are being developed to cover all phases of radioactive waste management. (2) A broad and deep technological base has been developed. (3) Many adequate processes are ready for actual application while others are ready for demonstration of applicability. These demonstrations are important to further public acceptance of nuclear energy. (4) At the present level of maturity, systematic analyses should be performed to determine actual requirements for the treatment and handling of radioactive wastes. These analyses can be used to focus our research and development, and demonstration activities to achieve treatment and conditioning systems which are both appropriate and cost-effective. (author)

  17. Responsible handling of the radioactive waste at the Universidad de Costa Rica

    International Nuclear Information System (INIS)

    Mora Rodriguez, Patricia; Varela, Alfonso

    2006-01-01

    The Radiation Safety Program (RSP) of the Universidad de Costa Rica established in 1990, handles the radioactive waste generated at the University. A centralized storage waste room is used by the Centro de Investigacion en Ciencias Atomicas, Nucleares y Moleculares, Instituto de Investigacion en Salud, Centro de Investigacion en Biologia Celular y Molecular, and the Centro de Investigacion en Contaminacion Ambiental. The RSP has pre-storage procedures, internal controls, protocols for storage, withdrawal of sources and discharges to the environment, according to national and international legislation. The main radionuclides in liquid and solid wastes are P32, I125, S35 y C14; which after a storage period will be disposed of as exempted materials. The waste room also permanently stores sources with the following radionuclides Cs137, U238, Th232, Sr90, Ra226, Cd109, Cf252 and Am241. It has 96 permanent sources and 52 that will be disposed of. The RSP allows the University to have a centralized facility for the safe management of all radioactive waste generated locally. (Author)

  18. The presence and partitioning behavior of flame retardants in waste, leachate, and air particles from Norwegian waste-handling facilities

    Institute of Scientific and Technical Information of China (English)

    Nicolas A.O.Morin; Patrik L.Andersson; Sarah E.Hale; Hans Peter H.Arp

    2017-01-01

    Flame retardants in commercial products eventually make their way into the waste stream.Herein the presence of flame retardants in Norwegian landfills,incineration facilities and recycling sorting/defragmenting facilities is investigated.These facilities handled waste electrical and electronic equipment (WEEE),vehicles,digestate,glass,combustibles,bottom ash and fly ash.The flame retardants considered included polybrominated diphenyl ethers (∑BDE-10) as well as dechlorane plus,polybrominated biphenyls,hexabromobenzene,pentabromotoluene and pentabromoethylbenzene (collectively referred to as ∑FR-7).Plastic,WEEE and vehicles contained the largest amount of flame retardants (∑BDE-10:45,000-210,000 μg/kg;∑FR-7:300-13,000 μg/kg).It was hypothesized leachate and air concentrations from facilities that sort/defragment WEEE and vehicles would be the highest.This was supported for total air phase concenttations (∑BDE-10:9000-195,000 pg/m3 WEEE/vehicle facilities,80-900 pg/m3 in incineration/sorting and landfill sites),but not for water leachate concentrations (e.g.,ΣBDE-10:15-3500 ng/L in WEEE/Vehicle facilities and 1-250 ng/L in landfill sites).Landfill leachate exhibited similar concentrations as WEEE/vehicle sorting and defragmenting facility leachate.To better account for concentrations in leachates at the different facilities,waste-water partitioning coefficients,Kwaste were measured (for the first time to our knowledge for flame retardants).WEEE and plastic waste had elevated Kwaste compared to other wastes,likely because flame retardants are directly added to these materials.The results of this study have implications for the development of strategies to reduce exposure and environmental emissions of flame retardants in waste and recycled products through improved waste management practices.

  19. Preparation and handling of surfaces for superconducting radio frequency cavities

    International Nuclear Information System (INIS)

    Bloess, D.

    1988-01-01

    Fortunately, surface treatment for s.c. cavities knows only one simple rule. If one observes this rule strictly one will be successful, if not, one will fail! The rule is CLEANLINESS. This means: clean material (high purity niobium without inclusions), clean (analytical grade) polishing chemicals and solvents, ultraclean (semiconductor grade) rinsing water, ultraclean (class 100) assembly environment. In general, if one applies the same working practice as the semiconductor industry, one will produce surfaces that are less clean than silicon wafers, due to the shape of the cavity (an inner surface is much more difficult to clean than a flat wafer); due to its size and due to the material (niobium is hydrophilic which makes the water with all the dirt in it stick to the surface). 9 references

  20. Diagnose and Redesign of the handling and treatment processes of the solid waste in the Hospital Mexico

    International Nuclear Information System (INIS)

    Campos Arrieta, G.; Navarro Blanco, D

    1999-01-01

    In the Hospital Mexico a program for the handling of the solid waste was implemented. The program consists on placing recipients, in all the corridors, for each type of waste (recyclable, toxic, dangerous, kitchens). However, this measure doesn't eliminate the risk that the waste represents for the community and the environment. The handling of the solid waste includes the selection or classification, the gathering, the transportation, and the temporary storage. While the treatment consists on the application of procedures that reduce the polluting properties of the waste. The planning of the topic is: To diagnose and to redesign of the handling processes and internal treatment of the hospital solid waste (HSW) in the Hospital Mexico. The contribution of the Industrial Engineering is given in the thematic of redesign of processes; the complementary areas are engineering of the human factor, environmental impact and normalization. The current problem that undergoes the Hospital was defined as follows: The Hospital Mexico cannot assure that the handling and current treatment of the solid waste diminish the risk that they represent to the health of the hospital community and the deterioration of the environment. This problem contains the independent variables such as the handling and current treatment of the solid waste, and the dependent variables such as the risk to the health of the community and deterioration of the environment. Based on the problem, the following hypothesis is established: The current conditions of handling and the lack of internal treatment of the solid waste in the Hospital Mexico, causes that the waste is a risk for the health of the hospital community and the deterioration of the environment. The project was structured in three denominated stages: Diagnose, Design and Validation, which respond to different general and specific objectives. In the stage of diagnose, to determine that the waste generated in the centers of health contain

  1. Management of waste cladding hulls. Part II. An assessment of zirconium pyrophoricity and recommendations for handling waste hulls

    International Nuclear Information System (INIS)

    Kullen, B.J.; Levitz, N.M.; Steindler, M.J.

    1977-11-01

    This report reviews experience and research related to the pyrophoricity of zirconium and zirconium alloys. The results of recent investigations of the behavior of Zircaloy and some observations of industrial handling and treatment of Zircaloy tubing and scrap are also discussed. A model for the management of waste Zircaloy cladding hulls from light water reactor fuel reprocessing is offered, based on an evaluation of the reviewed information. It is concluded that waste Zircaloy cladding hulls do not constitute a pyrophoric hazard if, following the model flow sheet, finely divided metal is oxidized during the management procedure. Steps alternative to the model are described which yield zirconium in deactivated form and also accomplish varying degrees of transuranic decontamination. Information collected into appendixes is (1) a collation of zirconium pyrophoricity data from the literature, (2) calculated radioactivity contents in Zircaloy cladding hulls from spent LWR fuels, and (3) results of a laboratory study on volatilization of zirconium from Zircaloy using HCl or Cl 2

  2. Preliminary Project Execution Plan for the Remote-Handled Low-Level Waste Disposal Project

    International Nuclear Information System (INIS)

    Duncan, David

    2011-01-01

    This preliminary project execution plan (PEP) defines U.S. Department of Energy (DOE) project objectives, roles and responsibilities of project participants, project organization, and controls to effectively manage acquisition of capital funds for construction of a proposed remote-handled low-level waste (LLW) disposal facility at the Idaho National Laboratory (INL). The plan addresses the policies, requirements, and critical decision (CD) responsibilities identified in DOE Order 413.3B, 'Program and Project Management for the Acquisition of Capital Assets.' This plan is intended to be a 'living document' that will be periodically updated as the project progresses through the CD process to construction and turnover for operation.

  3. Determination of Waste Isolation Pilot Plant (WIPP) management and institutional requirements documents for contact-handled (CH) critical systems

    International Nuclear Information System (INIS)

    1990-01-01

    This document lists the critical requirements documents applicable to the receipt of contact-handled waste at the Waste Isolation Pilot Plant. It also describes the processes used to determine the applicability of each document. This analysis is based on the applicable documents that were in effect in the February 1988 time frame. 2 refs

  4. Management of radioactive waste in nuclear power: handling of irradiated graphite from water-cooled graphite reactors

    International Nuclear Information System (INIS)

    Anfimov, S.S.

    2001-01-01

    In this paper an radioactive waste processing of graphite from graphite moderated nuclear reactors at its decommissioning is discussed. Methods of processing of irradiated graphite are presented. It can be concluded that advanced methods for graphite radioactive waste handling are available nowadays. Implementation of these methods will allow to enhance environmental safety of nuclear power that will benefit its progress in the future

  5. Report of the remote-handled transuranic waste mock retrieval demonstration

    International Nuclear Information System (INIS)

    1987-05-01

    This report documents the results of the mock, onsite retrieval demonstration that was conducted on May 19 and 20, 1987, for representatives of the New Mexico Environmental Evaluation Group (EEG). Demonstration of the retrievability of remote-handled transuranic (RH TRU) waste is part of a milestone included in the Agreement for Consultation and Cooperation between the state of New Mexico and the United States Department of Energy. Retrieval equipment design documents and a retrievability demonstration plan for RH TRU waste were previously transmitted to the EEG. This report documents the results of the demonstration by evaluating the demonstration against the acceptance criteria that were established in the Demonstration Plan. 1 fig., 2 tabs

  6. A pilot survey of the U.S. medical waste industry to determine training needs for safely handling highly infectious waste.

    Science.gov (United States)

    Le, Aurora B; Hoboy, Selin; Germain, Anne; Miller, Hal; Thompson, Richard; Herstein, Jocelyn J; Jelden, Katelyn C; Beam, Elizabeth L; Gibbs, Shawn G; Lowe, John J

    2018-02-01

    The recent Ebola outbreak led to the development of Ebola virus disease (EVD) best practices in clinical settings. However, after the care of EVD patients, proper medical waste management and disposal was identified as a crucial component to containing the virus. Category A waste-contaminated with EVD and other highly infectious pathogens-is strictly regulated by governmental agencies, and led to only several facilities willing to accept the waste. A pilot survey was administered to determine if U.S. medical waste facilities are prepared to handle or transport category A waste, and to determine waste workers' current extent of training to handle highly infectious waste. Sixty-eight percent of survey respondents indicated they had not determined if their facility would accept category A waste. Of those that had acquired a special permit, 67% had yet to modify their permit since the EVD outbreak. This pilot survey underscores gaps in the medical waste industry to handle and respond to category A waste. Furthermore, this study affirms reports a limited number of processing facilities are capable or willing to accept category A waste. Developing the proper management of infectious disease materials is essential to close the gaps identified so that states and governmental entities can act accordingly based on the regulations and guidance developed, and to ensure public safety. Copyright © 2018 Association for Professionals in Infection Control and Epidemiology, Inc. Published by Elsevier Inc. All rights reserved.

  7. Westinghouse Hanford Company plan for certifying newly generated contact-handled transuranic waste for emplacement in the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Lipinski, R.M.; Sheehan, J.S.

    1992-07-01

    Westinghouse Hanford Company (Westinghouse Hanford) currently manages an interim storage site for Westinghouse Hanford and non-Westinghouse Hanford-generated transuranic (TRU) waste and operates TRU waste generating facilities within the Hanford Site in Washington State. Approval has been received from the Waste Acceptance Criteria Certification Committee (WACCC) and Westinghouse Hanford TRU waste generating facilities to certify newly generated contact-handled TRU (CH-TRU) solid waste to meet the Waste Acceptance Criteria (WAC). This document describes the plan for certifying newly generated CH-TRU solid waste to meet the WAC requirements for storage at the Waste Isolation Pilot Plant (WIPP) site. Attached to this document are facility-specific certification plans for the Westinghouse Hanford TRU waste generators that have received WACCC approval. The certification plans describe operations that generate CH-TRU solid waste and the specific procedures by which these wastes will be certified and segregated from uncertified wastes at the generating facilities. All newly generated CH-TRU solid waste is being transferred to the Transuranic Storage and Assay Facility (TRUSAF) and/or a controlled storage facility. These facilities will store the waste until the certified TRU waste can be sent to the WIPP site and the non-certified TRU waste can be sent to the Waste Receiving and Processing Facility. All non-certifiable TRU waste will be segregated and clearly identified

  8. The application of advanced remote systems technology to future waste handling facilities: Waste Systems Data and Development Program

    International Nuclear Information System (INIS)

    Kring, C.T.; Herndon, J.N.; Meacham, S.A.

    1987-01-01

    The Consolidated Fuel Reprocessing Program (CFRP) at the Oak Ridge National Laboratory (ORNL) has been advancing the technology in remote handling and remote maintenance of in-cell systems planned for future US nuclear fuel reprocessing plants. Much of the experience and technology developed over the past decade in this endeavor are directly applicable to the in-cell systems being considered for the facilities of the Federal Waste Management System (FWMS). The ORNL developments are based on the application of teleoperated force-reflecting servomanipulators controlled by an operator completely removed from the hazardous environment. These developments address the nonrepetitive nature of remote maintenance in the unstructured environments encountered in a waste handling facility. Employing technological advancements in dexterous manipulators, as well as basic design guidelines that have been developed for remotely maintained equipment and processes, can increase operation and maintenance system capabilities, thereby allowing the attainment of two FWMS major objectives: decreasing plant personnel radiation exposure and increasing plant availability by decreasing the mean-time-to-repair in-cell maintenance and process equipment. 5 refs., 7 figs

  9. B cell remote-handled waste shipment cask alternatives study; TOPICAL

    International Nuclear Information System (INIS)

    RIDDELLE, J.G.

    1999-01-01

    The decommissioning of the 324 Facility B Cell includes the onsite transport of grouted remote-handled radioactive waste from the 324 Facility to the 200 Areas for disposal. The grouted waste has been transported in the leased ATG Nuclear Services 3-82B Radioactive Waste Shipping Cask (3-82B cask). Because the 3-82B cask is a U.S. Nuclear Regulatory Commission (NRC)-certified Type B shipping cask, the lease cost is high, and the cask operations in the onsite environment may not be optimal. An alternatives study has been performed to develop cost and schedule information on alternative waste transportation systems to assist in determining which system should be used in the future. Five alternatives were identified for evaluation. These included continued lease of the 3-82B cask, fabrication of a new 3-82B cask, development and fabrication of an onsite cask, modification of the existing U.S. Department of Energy-owned cask (OH-142), and the lease of a different commercially available cask. Each alternative was compared to acceptance criteria for use in the B Cell as an initial screening. Only continued leasing of the 3-82B cask, fabrication of a new 3-82B cask, and the development and fabrication of an onsite cask were found to meet all of the B Cell acceptance criteria

  10. Handling e-waste in developed and developing countries: initiatives, practices, and consequences.

    Science.gov (United States)

    Sthiannopkao, Suthipong; Wong, Ming Hung

    2013-10-01

    Discarded electronic goods contain a range of toxic materials requiring special handling. Developed countries have conventions, directives, and laws to regulate their disposal, most based on extended producer responsibility. Manufacturers take back items collected by retailers and local governments for safe destruction or recovery of materials. Compliance, however, is difficult to assure, and frequently runs against economic incentives. The expense of proper disposal leads to the shipment of large amounts of e-waste to China, India, Pakistan, Nigeria, and other developing countries. Shipment is often through middlemen, and under tariff classifications that make quantities difficult to assess. There, despite the intents of national regulations and hazardous waste laws, most e-waste is treated as general refuse, or crudely processed, often by burning or acid baths, with recovery of only a few materials of value. As dioxins, furans, and heavy metals are released, harm to the environment, workers, and area residents is inevitable. The faster growth of e-waste generated in the developing than in the developed world presages continued expansion of a pervasive and inexpensive informal processing sector, efficient in its own way, but inherently hazard-ridden. Copyright © 2012 Elsevier B.V. All rights reserved.

  11. Radioactive waste handling and disposal at King Faisal Specialist Hospital and Research Centre.

    Science.gov (United States)

    Al-Haj, Abdalla N; Lobriguito, Aida M; Al Anazi, Ibrahim

    2012-08-01

    King Faisal Specialist Hospital & Research Centre (KFSHRC) is the largest specialized medical center in Saudi Arabia. It performs highly specialized diagnostic imaging procedures with the use of various radionuclides required by sophisticated dual imaging systems. As a leading institution in cancer research, KFSHRC uses both long-lived and short-lived radionuclides. KFSHRC established the first cyclotron facility in the Middle East, which solved the in-house high demand for radionuclides and the difficulty in importing them. As both user and producer of high standard radiopharmaceuticals, KFSHRC generates large volumes of low and high level radioactive wastes. An old and small radioactive facility that was used for storage of radioactive waste was replaced with a bigger warehouse provided with facilities that will reduce radiation exposure of the staff, members of the public, and of the environment in the framework of "as low as reasonably achievable." The experiences and the effectiveness of the radiation protection program on handling and storage of radioactive wastes are presented.

  12. How problems of storing waste nuclear fuel are handled in some countries

    International Nuclear Information System (INIS)

    Langhe, R.

    1983-01-01

    This report is a survey of the situation in a number of European countries, in the United States and the Soviet Union as well. In all democratic countries, the nuclear power issue is controversial. Everywhere it has met with opposition and criticism, even in countries where nuclear power is officially promoted. Which of the elements comprised in the nuclear power issue is regarded as most controversial varies from country to country. In some countries, final storage and handling of waste nuclear fuel are referred to this category, in others nuclear power plant safety is claimed to be of greater importance. In the last few months, some public opinion has been coupling the peaceful use of nuclear power with nuclear weapons, thereby deeming the greatest danger to be the risk of unwanted distribution of nuclear weapons. Technical difficulties as well as public opinion have indefinitely adjourned the final solution of the disposal of waste nuclear fuel. This problem is of such magnitude that a final solution is urgently needed. Apart from opinions, the existence of waste nuclear power fuel emitting dangerous radiation for over 40 generations to come, makes it a moral obligation to find a way to spare future generations that heritage. (author)

  13. Remote handling of canisters containing nuclear waste in glass at the Savannah River Plant

    International Nuclear Information System (INIS)

    Callan, J.E.

    1986-01-01

    The Defense Waste Processing Facility is being constructed at the Savannah River Plant at a cost of $870 million to immobilize the defense high-level radioactive waste. This radioactive waste is being added to borosilicate glass for later disposal in a federal repository. The borosilicate glass is poured into stainless steel canisters for storage. These canisters must be handled remotely because of their high radioactivity, up to 5000 R/h. After the glass has been poured into the canister which will be temporarily sealed, it is transferred to a decontamination cell and decontaminated. The canister is then transferred to the weld cell where a permanent cap is welded into place. The canisters must then be transported from the processing building to a storage vault on the plant until the federal repository is available. A shielded canister transporter (SCT) has been designed and constructed for this purpose. The design of the SCT vehicle allows the safe transport of a highly radioactive canister containing borosilicate glass weighing 2300 kg with a radiation level up to 5000 R/h from one building to another. The design provides shielding for the operator in the cab of the vehicle to be below 0.5 rem/h. The SCT may also be used to load the final shipping cask when the federal repository is ready to receive the canisters

  14. The surface disposal concept for VLL waste

    International Nuclear Information System (INIS)

    2011-01-01

    Disposal facilities for very-low-level (VLL) waste have been designed to accommodate both residues originating from the decommissioning of nuclear facilities and used components. Those residues have very low specific-activity levels that lie below a few hundreds of becquerels per gram (Bq/g). As for the average activity found in any disposal facility, it never exceeds more than a few tens of becquerels per gram. In that case, waste disposal involves no special processing or conditioning, except for handling requirements or volume-gain purposes. The main barrier against radionuclide dispersion is provided by the geological formation being used for waste disposal. Basic disposal concept The design and construction provisions allow for the optimal operation of the disposal facility without any risk of altering the required safety level. They also ensure a satisfactory containment level for several centuries at the end of the operating lifetime. Hence, the natural materials in their original context constitute a particular advantage for the safety demonstration over the long term. With due account of the nature of VLL waste, their containment envelope (drums, big bags, etc.) has no role in confining radioactivity, but rather in facilitating handling and disposal operations, while protecting operators. Approximately 30% of all waste received at the CSTFA undergo a specific treatment before disposal. Low-density residues (plastics, thermal-insulation materials, etc.) are first compacted by a baling press, then strapped and wrapped in clear plastic-sheet. Another bundle press is used to reduce the volume of scrap metal. Some waste, such as the polluted waters generated on site or the sludges sent by producers, are processed in the solidification and stabilisation unit. Disposal cells are excavated progressively, as needed, directly in the clay formation down to a depth of 8 m and are operated in sequence. Cell design has evolved to maximize the disposal volume, and now

  15. The surface disposal concept for VLL waste

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-07-01

    Disposal facilities for very-low-level (VLL) waste have been designed to accommodate both residues originating from the decommissioning of nuclear facilities and used components. Those residues have very low specific-activity levels that lie below a few hundreds of becquerels per gram (Bq/g). As for the average activity found in any disposal facility, it never exceeds more than a few tens of becquerels per gram. In that case, waste disposal involves no special processing or conditioning, except for handling requirements or volume-gain purposes. The main barrier against radionuclide dispersion is provided by the geological formation being used for waste disposal. Basic disposal concept The design and construction provisions allow for the optimal operation of the disposal facility without any risk of altering the required safety level. They also ensure a satisfactory containment level for several centuries at the end of the operating lifetime. Hence, the natural materials in their original context constitute a particular advantage for the safety demonstration over the long term. With due account of the nature of VLL waste, their containment envelope (drums, big bags, etc.) has no role in confining radioactivity, but rather in facilitating handling and disposal operations, while protecting operators. Approximately 30% of all waste received at the CSTFA undergo a specific treatment before disposal. Low-density residues (plastics, thermal-insulation materials, etc.) are first compacted by a baling press, then strapped and wrapped in clear plastic-sheet. Another bundle press is used to reduce the volume of scrap metal. Some waste, such as the polluted waters generated on site or the sludges sent by producers, are processed in the solidification and stabilisation unit. Disposal cells are excavated progressively, as needed, directly in the clay formation down to a depth of 8 m and are operated in sequence. Cell design has evolved to maximize the disposal volume, and now

  16. Systematic handling of requirements and conditions (in compliance with waste acceptance requirements for a radioactive waste disposal facility)

    International Nuclear Information System (INIS)

    Keyser, Peter; Helander, Anita

    2012-01-01

    This Abstract and presentation will demonstrate the need for a structured requirement management and draw upon experiences and development from SKB requirements data base and methodology, in addition to international guidelines and software tools. The presentation will include a discussion on how requirement management can be applied for the decommissioning area. The key issue in the decommissioning of nuclear facilities is the progressive removal of hazards, by stepwise decontamination and dismantling activities that have to be carried out safely and within the boundaries of an approved safety case. For decommissioning there exists at least two safety cases, one for the pre-disposal activities and one for the disposal facility, and a need for a systematic handling of requirements and conditions to safely manage the radioactive waste in the long term. The decommissioning safety case is a collection of arguments and evidence to demonstrate the safety of a decommissioning project. It also includes analyzing and updating the decommissioning safety case in accordance with the waste acceptance criteria's and the expected output, i.e. waste packages. It is a continuous process to confirm that all requirements have been met. On the other hand there is the safety case for a radioactive waste disposal facility, which may include the following processes and requirements: i) Integrating relevant scientific (and other) information in a structured, traceable and transparent way and, thereby, developing and demonstrating an understanding of the potential behavior and performance of the disposal system; ii) Identifying uncertainties in the behavior and performance of the disposal system, describing the possible significance of the uncertainties, and identifying approaches for the management of significant uncertainties; iii) Demonstrating long-term safety and providing reasonable assurance that the disposal facility will perform in a manner that protects human health and the

  17. Feature-based handling of surface faults in compact disc players

    DEFF Research Database (Denmark)

    Odgaard, Peter Fogh; Stoustrup, Jakob; Andersen, Palle

    2006-01-01

    In this paper a novel method called feature-based control is presented. The method is designed to improve compact disc players’ handling of surface faults on the discs. The method is based on a fault-tolerant control scheme, which uses extracted features of the surface faults to remove those from...... the detector signals used for control during the occurrence of surface faults. The extracted features are coefficients of Karhunen–Loève approximations of the surface faults. The performance of the feature-based control scheme controlling compact disc players playing discs with surface faults has been...... validated experimentally. The proposed scheme reduces the control errors due to the surface faults, and in some cases where the standard fault handling scheme fails, our scheme keeps the CD-player playing....

  18. Influence on grip of knife handle surface characteristics and wearing protective gloves.

    Science.gov (United States)

    Claudon, Laurent

    2006-11-01

    Ten subjects were asked to apply maximum torques on knife handles with either their bare hand or their hand wearing a Kevlar fibre protective glove. Four knife handles (2 roughnesses, 2 hardnesses) were tested. Surface electromyograms of 6 upper limb and shoulder muscles were recorded and subject opinions on both knife handle hardness and friction in the hand were also assessed. The results revealed the significant influence of wearing gloves (pgloves greatly increased the torque independently of the other two parameters. Under the bare hand condition, a 90 degrees ShA slightly rough handle provided the greatest torque. Subject opinion agreed with the observed effects on recorded torque values except for the hardness factor, for which a preference for the 70 degrees ShA value over the 90 degrees ShA value emerged.

  19. Preliminary Project Execution Plan for the Remote-Handled Low-Level Waste Disposal Project

    Energy Technology Data Exchange (ETDEWEB)

    David Duncan

    2011-05-01

    This preliminary project execution plan (PEP) defines U.S. Department of Energy (DOE) project objectives, roles and responsibilities of project participants, project organization, and controls to effectively manage acquisition of capital funds for construction of a proposed remote-handled low-level waste (LLW) disposal facility at the Idaho National Laboratory (INL). The plan addresses the policies, requirements, and critical decision (CD) responsibilities identified in DOE Order 413.3B, 'Program and Project Management for the Acquisition of Capital Assets.' This plan is intended to be a 'living document' that will be periodically updated as the project progresses through the CD process to construction and turnover for operation.

  20. The Handling of Liquid Waste at the Research Station of Studsvik, Sweden

    Energy Technology Data Exchange (ETDEWEB)

    Lindhe, Soeren; Linder, Per

    1965-03-15

    The following quantities of radioactive waste are allowed to be released into a strait between the islands of Stora Bergoe and Studsviksholme: Total {alpha}-activity 0,2 curie/month. Total {beta}-activity 36 curie/month of which cerium, yttrium, rare earth 15 curie/month, strontium 2,4 curie/month. Before the release the radioactive waste has to be collected and controlled. Quantities approaching or exceeding the disposal limits have to be removed and concentrated by evaporation. The liquid waste is classified in several categories depending upon the level of activity: high active and medium active waste, low active waste, process water, sanitary water, surface water and reactor cooling water. The technical dimensioning of each category was based upon expected specific production figures (volume/man - month and activity/ man - month). These figures are based upon information obtained from Harwell. Actual production figures obtained during 1963 and the first half of 1964 are shown and compared with the expected ones. As a conclusion is stated that the actual production follows the predictions fairly well.

  1. The Handling of Liquid Waste at the Research Station of Studsvik, Sweden

    International Nuclear Information System (INIS)

    Lindhe, Soeren; Linder, Per

    1965-03-01

    The following quantities of radioactive waste are allowed to be released into a strait between the islands of Stora Bergoe and Studsviksholme: Total α-activity 0,2 curie/month. Total β-activity 36 curie/month of which cerium, yttrium, rare earth 15 curie/month, strontium 2,4 curie/month. Before the release the radioactive waste has to be collected and controlled. Quantities approaching or exceeding the disposal limits have to be removed and concentrated by evaporation. The liquid waste is classified in several categories depending upon the level of activity: high active and medium active waste, low active waste, process water, sanitary water, surface water and reactor cooling water. The technical dimensioning of each category was based upon expected specific production figures (volume/man - month and activity/ man - month). These figures are based upon information obtained from Harwell. Actual production figures obtained during 1963 and the first half of 1964 are shown and compared with the expected ones. As a conclusion is stated that the actual production follows the predictions fairly well

  2. Plan of environmental administration for the handling of ordinary and specific accustomed to waste of Los Chiles Hospital (Alajuela)

    International Nuclear Information System (INIS)

    Parini Corella, P.

    1999-01-01

    The area of study of the present project is the treatment of waste, likewise, the objective of this project was to design a plan of Environmental Administration for the handling of the ordinary and specific accustomed to waste generated in the Los Chiles Hospital. For all this, firstly one carries out an diagnose. In the first stage of diagnose, it was necessary to elaborate an initial tool that allowed to know the position of the Institution in environmental matter. To develop this they take into account different elements of the norm ISO-14000, specifically of the norms ISO-14001 and ISO-14004, the environmental legislation of our country, aspects of the strategic planning, elements of occupational security and some existent politicians at Managerial level of the CCSS related with the administration in the handling of hospital waste. With regard to this finish, one carries out a study on the situation of the Hospital, since this information constitutes the base for the elaboration of the Institutional Program, for the acquisition of inputs, the assignment of resources and for the establishment of the Program of Control of Monitoreo the diagnoses sandal five points: Generation and composition of the waste handling,resources, knowledge and attitudes,mechanism of Control. As for the evaluation of the different stages that you/they constitute the administration of the manipulation of the hospital accustomed to waste, three stages could be identified in the Los Chiles Hospital, that is: generation and deposit, gathering, transporting and final decomposition. The first one is since a complex stage it depends on several such factors as: the activity type that is carried out when the waste, the place is generated where is taken to end happiness activity, the type or nature of the waste and different people that can give origin to these waste. The second stage, the handling of the accustomed to waste, involves exclusively personal of toilet and some infirmary assistants

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

    International Nuclear Information System (INIS)

    Wu, C.L.; Lee, J.; Lu, D.L.; Jardine, L.J.

    1991-12-01

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

  4. Decommissioning high-level waste surface facilities

    International Nuclear Information System (INIS)

    1978-04-01

    The protective storage, entombment and dismantlement options of decommissioning a High-Level Waste Surface Facility (HLWSF) was investigated. A reference conceptual design for the facility was developed based on the designs of similar facilities. State-of-the-art decommissioning technologies were identified. Program plans and cost estimates for decommissioning the reference conceptual designs were developed. Good engineering design concepts were on the basis of this work identified

  5. SRTC criticality technical review: Nuclear Criticality Safety Evaluation 93-18 Uranium Solidification Facility's Waste Handling Facility

    International Nuclear Information System (INIS)

    Rathbun, R.

    1993-01-01

    Separate review of NMP-NCS-930058, open-quotes Nuclear Criticality Safety Evaluation 93-18 Uranium Solidification Facility's Waste Handling Facility (U), August 17, 1993,close quotes was requested of SRTC Applied Physics Group. The NCSE is a criticality assessment to determine waste container uranium limits in the Uranium Solidification Facility's Waste Handling Facility. The NCSE under review concludes that the NDA room remains in a critically safe configuration for all normal and single credible abnormal conditions. The ability to make this conclusion is highly dependent on array limitation and inclusion of physical barriers between 2x2x1 arrays of boxes containing materials contaminated with uranium. After a thorough review of the NCSE and independent calculations, this reviewer agrees with that conclusion

  6. Validation of the plan of handling of hospital accustomed to waste of the Costa Rica Box of the Public Health

    International Nuclear Information System (INIS)

    Salazar Monge, F. J.

    1999-01-01

    The handling that is given at the moment to the hospital solid waste in Costa Rica is considered inadequate, due to the lack of the personnel's of cleaning training, segregation, vehicles of transport and recipients, signaling, team of personal protection, schedule and gathering route and recycle. This situation represents a serious danger for the population's health and the since environment is exposed infectious agents toxic substances and even radioactive products that are generated in the residuals of the centers of health. In this work they intend improvements to the system of handling of solid waste of the hospital Calderon Guardia, and you evaluates each one of the points mentioned previously. A revision was made about the properties, the quality and the quantity of the produced solid waste and an I diagnose of the current situation in this center of health. The proposed improvements are based on the regional program of hospital solid waste, agreement ALA91/33 (this it was elaborated in 1997 by a cooperation initiative between the European Union and the Governments from Costa Rica, El Salvador, Guatemala, Honduras, Nicaragua and Panama) and the Norms for the Handling of dangerous Solid Waste in Establishments of Health (elaborated by the Costa Rican Box of the Public Health) their advantages and disadvantages are also exposed and he/she is carried out a comparative analysis among the handling plan proposed in these documents and the one that is used at the moment in the Calderon Guardia hospital. Some of the detected problems are: it lacks of team of personal protection, the appropriate recipients are not used, storage of waste inside the hospital, the vehicles used in the internal transport are not appropriate, bad organization of the storing center, among others [es

  7. Electrostatic chuck consisting of polymeric electrostatic inductive fibers for handling of objects with rough surfaces

    International Nuclear Information System (INIS)

    Dhelika, Radon; Sawai, Kenji; Saito, Shigeki; Takahashi, Kunio; Takarada, Wataru; Kikutani, Takeshi

    2013-01-01

    An electrostatic chuck (ESC) is a type of reversible dry adhesive which clamps objects by means of electrostatic force. Currently an ESC is used only for objects having flat surfaces because the attractive force is reduced for rough surfaces. An ESC that can handle objects with rough surfaces will expand its applications to MEMS (micro electro mechanical system) or optical parts handling. An ESC consisting of compliant electrostatic inductive fibers which conform to the profile of the surface has been proposed for such use. This paper aims at furthering previous research by observing the attractive force/pressure generated, both theoretically and experimentally, through step-by-step fabrication and analysis. Additionally, how the proposed fiber ESC behaves toward rough surfaces is also observed. The attractive force/pressure of the fiber ESC is theoretically investigated using a robust mechano-electrostatic model. Subsequently, a prototype of the fiber ESC consisting of ten fibers arranged at an angle is employed to experimentally observe its attractive force/pressure for objects with rough surfaces. The attractive force of the surface which is modeled as a sinusoidal wave with various amplitudes is observed, through which the feasibility of a fiber ESC is justified. (paper)

  8. Cultural Resource Investigations for the Remote Handled Low Level Waste Facility at the Idaho National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Brenda R. Pace; Hollie Gilbert; Julie Braun Williams; Clayton Marler; Dino Lowrey; Cameron Brizzee

    2010-06-01

    The U. S. Department of Energy, Idaho Operations Office is considering options for construction of a facility for disposal of Idaho National Laboratory (INL) generated remote-handled low-level waste. Initial screening has resulted in the identification of two recommended alternative locations for this new facility: one near the Advanced Test Reactor (ATR) Complex and one near the Idaho Comprehensive Environmental Response, Compensation, and Liability Act Disposal Facility (ICDF). In April and May of 2010, the INL Cultural Resource Management Office conducted archival searches, intensive archaeological field surveys, and initial coordination with the Shoshone-Bannock Tribes to identify cultural resources that may be adversely affected by new construction within either one of these candidate locations. This investigation showed that construction within the location near the ATR Complex may impact one historic homestead and several historic canals and ditches that are potentially eligible for nomination to the National Register of Historic Places. No resources judged to be of National Register significance were identified in the candidate location near the ICDF. Generalized tribal concerns regarding protection of natural resources were also documented in both locations. This report outlines recommendations for protective measures to help ensure that the impacts of construction on the identified resources are not adverse.

  9. Analysis of Welding Joint on Handling High Level Waste-Glass Canister

    International Nuclear Information System (INIS)

    Herlan Martono; Aisyah; Wati

    2007-01-01

    The analysis of welding joint of stainless steel austenitic AISI 304 for canister material has been studied. At the handling of waste-glass canister from melter below to interim storage, there is a step of welding of canister lid. Welding quality must be kept in a good condition, in order there is no gas out pass welding pores and canister be able to lift by crane. Two part of stainless steel plate in dimension (200 x 125 x 3) mm was jointed by welding. Welding was conducted by TIG machine with protection gas is argon. Electric current were conducted for welding were 70, 80, 90, 100, 110, 120, 130, and 140 A. Welded plates were cut with dimension according to JIS 3121 standard for tensile strength test. Hardness test in welding zone, HAZ, and plate were conducted by Vickers. Analysis of microstructure by optic microscope. The increasing of electric current at the welding, increasing of tensile strength of welding yields. The best quality welding yields using electric current was 110 A. At the welding with electric current more than 110 A, the electric current influence towards plate quality, so that decreasing of stainless steel plate quality and breaking at the plate. Tensile strength of stainless steel plate welding yields in requirement conditions according to application in canister transportation is 0.24 kg/mm 2 . (author)

  10. Surface layer effects on waste glass corrosion

    International Nuclear Information System (INIS)

    Feng, X.

    1993-01-01

    Water contact subjects waste glass to chemical attack that results in the formation of surface alteration layers. Two principal hypotheses have been advanced concerning the effect of surface alteration layers on continued glass corrosion: (1) they act as a mass transport barrier and (2) they influence the chemical affinity of the glass reaction. In general, transport barrier effects have been found to be less important than affinity effects in the corrosion of most high-level nuclear waste glasses. However, they can be important under some circumstances, for example, in a very alkaline solution, in leachants containing Mg ions, or under conditions where the matrix dissolution rate is very low. The latter suggests that physical barrier effect may affect the long-term glass dissolution rate. Surface layers influence glass reaction affinity through the effects of the altered glass and secondary phases on the solution chemistry. The reaction affinity may be controlled by various precipitates and crystalline phases, amorphous silica phases, gel layer, or all the components of the glass. The surface alteration layers influence radionuclide release mainly through colloid formation, crystalline phase incorporation, and gel layer retention. This paper reviews current understanding and uncertainties

  11. Liquid waste handling facilities for a conceptual LWR spent fuel reprocessing complex

    International Nuclear Information System (INIS)

    Witt, D.C.; Bradley, R.F.

    1978-01-01

    The waste evaporator systems and the methods for evaporating the liquid wastes of various radioactivity levels are discussed. After the liquid wastes are evaporated and nitric acid is recovered the high-level liquid waste is incorporated into borosilicate glass and the intermediate-level liquid waste into concrete for final disposal

  12. Waste water treatment in surface mines

    Energy Technology Data Exchange (ETDEWEB)

    Navasardyants, M A; Esipov, V Z; Ryzhkov, Yu A

    1981-01-01

    This paper evaluates problems associated with waste water from coal surface mines of the Kemerovougol' association in the Kuzbass. Waste water treatment in the Kuzbass is of major importance as the region is supplied with water from only one river, the Tom river. Water influx to Kemerovougol' surface mines in a year amounts to 136 million m/sup 3/. The water is used during technological processes, for fire fighting, and spraying to prevent dusting; the rest, about 82.1 million m/sup 3/, is discharged into surface waters. Of this amount, 25.1 million m/sup 3/ is heavily polluted water, 46.6 million m3 are polluted but within limits, and 10.4 million m/sup 3/ are characterized as relatively clean. Waste water is polluted with: suspended matters, oils and oil products, nitrates, nitrides and chlorides. Suspended matter content sometimes reaches 4,000 and 5,000 mg/l, and oil product content in water amounts to 2.17 mg/l. Water treatment in surface mines is two-staged: sumps and sedimentation tanks are used. Water with suspended matter content of 50 to 100 mg/l in winter and summer, and 200 to 250 mg/l in spring and autumn is reduced in sumps to 25 to 30 mg/l in summer and winter and to 40 to 50 mg/l in autumn and spring. During the first stage water treatment efficiency ranges from 50 to 80%. During the second stage water is collected in sedimentation tanks. It is noted that so-called secondary pollution is one of the causes of the relatively high level of suspended matter in discharged water. Water discharged from sedimentation tanks carries clay and loam particles from the bottom and walls of water tanks and channels.

  13. The impact of uncontrolled waste disposal on surface water quality ...

    African Journals Online (AJOL)

    The main threat to the surface water quality in Addis Ababa is environmental pollution derived from domestic and industrial activities. Due to the inadequacy of controlled waste management strategies and waste treatment plants, people are forced to discharge wastes both on open surface and within water bodies.

  14. A Review and Analysis of European Industrial Experience in Handling LWR Spent Fuel and Vitrified High-Level Waste

    Energy Technology Data Exchange (ETDEWEB)

    Blomeke, J.O.

    2001-07-10

    The industrial facilities that have been built or are under construction in France, the United Kingdom, Sweden, and West Germany to handle light-water reactor (LWR) spent fuel and canisters of vitrified high-level waste before ultimate disposal are described and illustrated with drawings and photographs. Published information on the operating performance of these facilities is also given. This information was assembled for consideration in planning and design of similar equipment and facilities needed for the Federal Waste Management System in the United States.

  15. FRIDA: A model for the generation and handling of solid waste in Denmark

    DEFF Research Database (Denmark)

    Larsen, Helge V.; Møller Andersen, Frits

    2012-01-01

    Since 1994, Danish waste treatment plants have been obliged to report to the Danish EPA the annual amounts of waste treated. Applying these data, we analyse the development, link amounts of waste to economic and demographic variables, and present a model for the generation and treatment of waste...... in Denmark. Using the model and official projections of the economic development, a baseline projection for the generation and treatment of waste is presented. © 2012 Elsevier B.V. All rights reserved....

  16. Thermal response modeling of a contact-handled transuranic waste shipping container system to a fire

    International Nuclear Information System (INIS)

    Suchsland, K.E.; Kwong, K.C.; Fretter, E.F.; Boyd, R.D.; Auerbach, I.; Yoshimura, H.R.

    1980-01-01

    A one-dimensional thermal model has been developed to predict the response of a transuranic (TRU) waste shipping container accidentally exposed to a fire environment. The basic wall structure of the container consists of polyurethane foam (64 kg/m 3 ) sandwiched between two steel plates. The foam thermal model, based on high temperature experimental data, is developed for the case in which the virgin foam is in a nonoxidizing environment. The experimental results indicate that foam decomposition is highly heat rate dependent. At low quasi-steady heating rates, the foam changes to a bubbling black viscous liquid. At very high heating rates, pyrolysis gases are formed as the foam decomposes and a 20% (by weight) residual char remains. This porous char acts as a radiation shield which can significantly reduce thermal transport. In the case of a TRU shipping container wall, this char will slow the thermal penetration rate and drastically reduce the heat load to the container contents. When the front surface of the wall was subjected to 1333 0 K, numerical computations predict that after approximately 1800 s the foam temperature rise at a depth of 10.2 cm was less than 200 K (uncharred). After approximately 3600 s the foam temperature rise at a depth of 20.4 cm was 23 0 K. Typical waste contents temperature rise was predicted to be less than 56 0 K after 3600 s of heating

  17. Guide to the safe handling of radioactive wastes at nuclear power plants

    International Nuclear Information System (INIS)

    1980-01-01

    This guide discusses the responsibilities of the regulatory authorities, the design considerations of waste management systems and the source and characteristics of waste. Present techniques for treating, conditioning, storing and disposing of gaseous, liquid and solid wastes on and from the site are summarized, and a consensus of good practice in waste management based on current knowledge and experience is given. The guide also contains brief chapters on transport of wastes, monitoring systems, safety analyses and a review of future trends in waste management

  18. Status of microwave process development for RH-TRU [remote-handled transuranic] wastes at 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. The microwave fields are uniform in one dimension to reduce the formation of hot spots on the microwaved wasteform. 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., 1 fig., 1 tab

  19. Investigating cross-contamination by yeast strains from dental solid waste to waste-handling workers by DNA sequencing.

    Science.gov (United States)

    Vieira, Cristina Dutra; Tagliaferri, Thaysa Leite; de Carvalho, Maria Auxiliadora Roque; de Resende-Stoianoff, Maria Aparecida; Holanda, Rodrigo Assuncao; de Magalhães, Thais Furtado Ferreira; Magalhães, Paula Prazeres; Dos Santos, Simone Gonçalves; de Macêdo Farias, Luiz

    2018-04-01

    Trying to widen the discussion on the risks associated with dental waste, this study proposed to investigate and genetically compare yeast isolates recovered from dental solid waste and waste workers. Three samples were collected from workers' hands, nasal mucosa, and professional clothing (days 0, 30, and 180), and two from dental waste (days 0 and 180). Slide culture, microscopy, antifungal drug susceptibility, intersimple sequence repeat analysis, and amplification and sequencing of internal transcribed spacer regions were performed. Yeast strains were recovered from all waste workers' sites, including professional clothes, and from waste. Antifungal susceptibility testing demonstrated that some yeast recovered from employees and waste exhibited nonsusceptible profiles. The dendrogram demonstrated the presence of three major clusters based on similarity matrix and UPGMA grouping method. Two branches displayed 100% similarity: three strains of Candida guilliermondii isolated from different employees, working in opposite work shifts, and from diverse sites grouped in one part of branch 1 and cluster 3 that included two samples of Candida albicans recovered from waste and the hand of one waste worker. The results suggested the possibility of cross-contamination from dental waste to waste workers and reinforce the need of training programs focused on better waste management routines. © 2017 The Authors. MicrobiologyOpen published by John Wiley & Sons Ltd.

  20. 77 FR 11112 - Proposed Approval of the Central Characterization Project's Remote-Handled Transuranic Waste...

    Science.gov (United States)

    2012-02-24

    ... debris waste from the FB-Line at SRS. This waste was generated by glovebox operations, decontamination... summary category group solids (S3000) or soils and gravel (S4000) is characterized for WIPP disposal; and...

  1. Too hot to handle. Social and policy issues in the management of radioactive wastes

    International Nuclear Information System (INIS)

    Walker, C.A.; Gould, L.C.; Woodhouse, E.J.

    1983-01-01

    Information about the management of radioactive wastes is provided in this book. Specifically, the book attempts to supply information to further the understanding of the history of radioactive waste management in this country and the role of nuclear energy in the future of the US; the science and technology of the processes that produce radioactive wastes and of the methods proposed for managing them; the biological effects of radiation; the public attitudes about nuclear power; the nature of risks resulting from technological developments and ways of managing them; and the political institutions and processes that govern radioactive waste management. The authors have attempted to present an objective view of nuclear waste management taking a stand neither for nor against nuclear power but placing special emphasis on radioactive waste management rather than nuclear power, because they feel that the latter aspect of the subject has received much more extensive coverage elsewhere. The contents of the book are divided into 7 chapters entitled: The Radioactive Waste Management Problem, Science and Technology of the Sources and Management of Radioactive Wastes, Nuclear Waste Management and Risks to Human Health, Public Attitudes toward Radioactive Wastes, How Safe Is Safe Enough; Determinants of Perceived and Acceptable Risk, The Politics of Nuclear Waste Management, and Value Issues in Radioactive Waste Management

  2. Infectious Risk Assessment of Unsafe Handling Practices and Management of Clinical Solid Waste

    Science.gov (United States)

    Hossain, Md. Sohrab; Rahman, Nik Norulaini Nik Ab; Balakrishnan, Venugopal; Puvanesuaran, Vignesh R.; Sarker, Md. Zaidul Islam; Kadir, Mohd Omar Ab

    2013-01-01

    The present study was undertaken to determine the bacterial agents present in various clinical solid wastes, general waste and clinical sharp waste. The waste was collected from different wards/units in a healthcare facility in Penang Island, Malaysia. The presence of bacterial agents in clinical and general waste was determined using the conventional bacteria identification methods. Several pathogenic bacteria including opportunistic bacterial agent such as Pseudomonas aeruginosa, Salmonella spp., Klebsiella pneumoniae, Serratia marcescens, Acinetobacter baumannii, Staphylococcus aureus, Staphylococcus epidermidis, Enterococcus faecalis, Streptococcus pyogenes were detected in clinical solid wastes. The presence of specific pathogenic bacterial strains in clinical sharp waste was determined using 16s rDNA analysis. In this study, several nosocomial pathogenic bacteria strains of Escherichia coli, Klebsiella pneumoniae, Proteus mirabilis, Lysinibacillus sphaericus, Serratia marcescens, and Staphylococcus aureus were detected in clinical sharp waste. The present study suggests that waste generated from healthcare facilities should be sterilized at the point of generation in order to eliminate nosocomial infections from the general waste or either of the clinical wastes. PMID:23435587

  3. Infectious Risk Assessment of Unsafe Handling Practices and Management of Clinical Solid Waste

    Directory of Open Access Journals (Sweden)

    Md. Zaidul Islam Sarker

    2013-01-01

    Full Text Available The present study was undertaken to determine the bacterial agents present in various clinical solid wastes, general waste and clinical sharp waste. The waste was collected from different wards/units in a healthcare facility in Penang Island, Malaysia. The presence of bacterial agents in clinical and general waste was determined using the conventional bacteria identification methods. Several pathogenic bacteria including opportunistic bacterial agent such as Pseudomonas aeruginosa, Salmonella spp., Klebsiella pneumoniae, Serratia marcescens, Acinetobacter baumannii, Staphylococcus aureus, Staphylococcus epidermidis, Enterococcus faecalis, Streptococcus pyogenes were detected in clinical solid wastes. The presence of specific pathogenic bacterial strains in clinical sharp waste was determined using 16s rDNA analysis. In this study, several nosocomial pathogenic bacteria strains of Escherichia coli, Klebsiella pneumoniae, Proteus mirabilis, Lysinibacillus sphaericus, Serratia marcescens, and Staphylococcus aureus were detected in clinical sharp waste. The present study suggests that waste generated from healthcare facilities should be sterilized at the point of generation in order to eliminate nosocomial infections from the general waste or either of the clinical wastes.

  4. Identifying potential environmental impacts of waste handling strategies in textile industry.

    Science.gov (United States)

    Yacout, Dalia M M; Hassouna, M S

    2016-08-01

    Waste management is a successful instrument to minimize generated waste and improve environmental conditions. In spite of the large share of developing countries in the textile industry, limited information is available concerning the waste management strategies implemented for textiles on those countries and their environmental impacts. In the current study, two waste management approaches for hazardous solid waste treatment of acrylic fibers (landfill and incineration) were investigated. The main research questions were: What are the different impacts of each waste management strategy? Which waste management strategy is more ecofriendly? Life cycle assessment was employed in order to model the environmental impacts of each waste streaming approach separately then compare them together. Results revealed that incineration was the more ecofriendly approach. Highest impacts of both approaches were on ecotoxicity and carcinogenic potentials due to release of metals from pigment wastes. Landfill had an impact of 46.8 % on human health as compared to 28 % by incineration. Incineration impact on ecosystem quality was higher than landfill impact (68.4 and 51.3 %, respectively). As for resources category, incineration had a higher impact than landfill (3.5 and 2.0 %, respectively). Those impacts could be mitigated if state-of-the-art landfill or incinerator were used and could be reduced by applying waste to energy approaches for both management systems In conclusion, shifting waste treatment from landfill to incineration would decrease the overall environmental impacts and allow energy recovery. The potential of waste to energy approach by incineration with heat recovery could be considered in further studies. Future research is needed in order to assess the implementation of waste management systems and the preferable waste management strategies in the textile industry on developing countries.

  5. Robust and efficient handling of yield surface discontinuities in elasto-plastic finite element calculations

    DEFF Research Database (Denmark)

    Clausen, Johan Christian; Damkilde, Lars; Andersen, Lars Vabbersgaard

    2015-01-01

    Purpose – The purpose of this paper is to present several methods on how to deal with yield surface discontinuities. The explicit formulations, first presented by Koiter (1953), result in multisingular constitutive matrices which can cause numerical problems in elasto-plastic finite element...... documented in the literature all present “easy” calculation examples, e.g. low friction angles and few elements. The amendments presented in this paper result in robust elasto-plastic computations, making the solution of “hard” problems possible without introducing approximations in the yield surfaces...... calculations. These problems, however, are not documented in previous literature. In this paper an amendment to the Koiter formulation of the constitutive matrices for stress points located on discontinuities is proposed. Design/methodology/approach – First, a review of existing methods of handling yield...

  6. State fund of decommissioning of nuclear installations and handling of spent nuclear fuels and nuclear wastes (Slovak Republic)

    International Nuclear Information System (INIS)

    Kozma, Milos

    2006-01-01

    State Fund for Decommissioning of Nuclear Installations and Handling of Spent Nuclear Fuels and Nuclear Wastes was established by the Act 254/1994 of the National Council of the Slovak Republic as a special-purpose fund which concentrates financial resources intended for decommissioning of nuclear installations and for handling of spent nuclear fuels and radioactive wastes. The Act was amended in 2000, 2001 and 2002. The Fund is legal entity and independent from operator of nuclear installations Slovak Power Facilities Inc. The Fund is headed by Director, who is appointed and recalled by Minister of Economy of the Slovak Republic. Sources of the Fund are generated from: a) contributions by nuclear installation operators; b) penalties imposed by Nuclear Regulatory Authority of the Slovak Republic upon natural persons and legal entities pursuant to separate regulation; c) bank credits; d) interest on Fund deposits in banks; e) grants from State Budget; f) other sources as provided by special regulation. Fund resources may be used for the following purposes: a) decommissioning of nuclear installations; b) handling of spent nuclear fuels and radioactive wastes after the termination of nuclear installation operation; c) handling of radioactive wastes whose originator is not known, including occasionally seized radioactive wastes and radioactive materials stemming from criminal activities whose originator is not known, as confirmed by Police Corps investigator or Ministry of Health of the Slovak Republic; d) purchase of land for the establishment of nuclear fuel and nuclear waste repositories; e) research and development in the areas of decommissioning of nuclear installations and handling of nuclear fuels and radioactive wastes after the termination of the operation of nuclear installations; f) selection of localities, geological survey, preparation, design, construction, commissioning, operation and closure of repositories of spent nuclear fuels and radioactive wastes

  7. Effect of solid waste landfill on underground and surface water ...

    African Journals Online (AJOL)

    Effect of solid waste landfill on underground and surface water quality at ring road, Ibadan, Nigeria. ... parameters showed increased concentrations over those from control sites. ... Keywords: Landfill, groundwater, surface-water, pollution.

  8. Fatigue crack growth from handling surface anomalies in a nickel based superalloy at high temperature

    Directory of Open Access Journals (Sweden)

    Gourdin Stéphane

    2014-01-01

    Full Text Available Aircraft engine manufacturers have to demonstrate that handling surface anomalies in sensitive areas of discs are not critical for in-service life of a component. Currently, the models used consider anomalies as long cracks propagating from the first cycle, which introduces a certain degree of conservatism when calculating the fatigue life of surface flaws. Preliminary studies have shown that the first stages of crack propagation from surface anomalies are responsible for the conservative results. Thus, the aim of the study is to characterize the crack propagation from typical surface anomalies and to establish a new crack growth model, which can account for the micro-propagation stage. To separate the effects of the geometry of the anomalies and the residual stress state after introduction of the surface flaws, two V-type anomalies are studied: scratches and dents. Different studies have shown that the residual stresses beneath the anomalies seem to control the fatigue life of samples exhibiting scratches and dents. In order to monitor the crack micro-propagation, a direct current potential drop technique, coupled with heat tints is used during fatigue tests at elevated temperature. Thermal treatments releasing the residual stresses are also used to decouple the effect of crack morphology and residual stresses.

  9. Handling of spent nuclear fuel and final storage of vitrified high level reprocessing waste

    International Nuclear Information System (INIS)

    1978-01-01

    The report gives a general summary of the Swedish KBS-project on management and disposal of vitrified reprocessed waste. Its final aim is to demostrate that the means of processing and managing power reactor waste in an absolutely safe way, as stipulated in the Swedish so called Conditions Act, already exist. Chapters on Storage facility for spent fuel, Intermidiate storage of reprocessed waste, Geology, Final repository, Transportation, Protection, and Siting. (L.E.)

  10. Inspection and verification of waste packages for near surface disposal

    International Nuclear Information System (INIS)

    2000-01-01

    Extensive experience has been gained with various disposal options for low and intermediate level waste at or near surface disposal facilities. Near surface disposal is based on proven and well demonstrated technologies. To ensure the safety of near surface disposal facilities when available technologies are applied, it is necessary to control and assure the quality of the repository system's performance, which includes waste packages, engineered features and natural barriers, as well as siting, design, construction, operation, closure and institutional controls. Recognizing the importance of repository performance, the IAEA is producing a set of technical publications on quality assurance and quality control (QA/QC) for waste disposal to provide Member States with technical guidance and current information. These publications cover issues on the application of QA/QC programmes to waste disposal, long term record management, and specific QA/QC aspects of waste packaging, repository design and R and D. Waste package QA/QC is especially important because the package is the primary barrier to radionuclide release from a disposal facility. Waste packaging also involves interface issues between the waste generator and the disposal facility operator. Waste should be packaged by generators to meet waste acceptance requirements set for a repository or disposal system. However, it is essential that the disposal facility operator ensure that waste packages conform with disposal facility acceptance requirements. Demonstration of conformance with disposal facility acceptance requirements can be achieved through the systematic inspection and verification of waste packages at both the waste generator's site and at the disposal facility, based on a waste package QA/QC programme established by the waste generator and approved by the disposal operator. However, strategies, approaches and the scope of inspection and verification will be somewhat different from country to country

  11. Synthesis long life storage studies surface storage of vitrified wastes

    International Nuclear Information System (INIS)

    Beziat, A.; Breton, E.; Ranc, G.; Gaillard, J.P.; Lagrave, H.; Hollender, F.; Jourdain, F.; Piault, E.; Garnier, J.; Lamare, V.; Duret, B.; Helie, M.; Ferry, C.; Mijuin, D.; Gagnier, E.

    2004-01-01

    This document is realized in the framework of the axis 3 of the law of 1991 on the radioactive wastes management. It justifies the choices concerning long time surface storage installation of vitrified wastes, called high activity wastes. The long time of the installation would reach 300 years at the maximum. These wastes represent 1 % at the maximum, of radioactive wastes in France but 95 % of the whole radioactivity. Three main objectives were followed: provide a permanent containment of radionuclides; give the possibility of wastes containers retrieval at all the time; minimize the maintenance and the control. The results allow to conclude that the long time surface storage of high activity wastes is feasible. (A.L.B.)

  12. Westinghouse Hanford Company plan for certifying newly generated contact -- handled transuranic waste. Revision 1

    International Nuclear Information System (INIS)

    Lipinski, R.M.; Backlund, E.G.

    1995-09-01

    All transuranic (TRU) waste generators are required by US Department of Energy (DOE) Order 5820.2A to package their TRU waste in order to comply wit the Waste Isolation Pilot Plant (WIPP) -- Waste Acceptance Criteria (WAC) or keep non-certifiable containers segregated. The Westinghouse Hanford Company (WHC) Transuranic Waste Certification Plan was developed to ensure that TRU newly generated waste at WHC meets the DOE Order 5820.2A and the WHC-WAC which includes the State of Washington Department of Ecology -- Washington Administrative Code (DOE-WAC). The metho used at WHC to package TRU waste are described in sufficient detail to meet the regulations. This document is organized to provide a brief overview of waste generation operations at WHC. The methods used to implement this plan are discussed briefly along with the responsibilities and authorities of applicable organizations. This plan describes how WHC complies with all applicable regulations and requirements set forth in the latest approved revision of WHC-EP-0063-4

  13. Ceramics for Molten Materials Containment, Transfer and Handling on the Lunar Surface

    Science.gov (United States)

    Standish, Evan; Stefanescu, Doru M.; Curreri, Peter A.

    2009-01-01

    As part of a project on Molten Materials Transfer and Handling on the Lunar Surface, molten materials containment samples of various ceramics were tested to determine their performance in contact with a melt of lunar regolith simulant. The test temperature was 1600 C with contact times ranging from 0 to 12 hours. Regolith simulant was pressed into cylinders with the approximate dimensions of 1.25 dia x 1.25cm height and then melted on ceramic substrates. The regolith-ceramic interface was examined after processing to determine the melt/ceramic interaction. It was found that the molten regolith wetted all oxide ceramics tested extremely well which resulted in chemical reaction between the materials in each case. Alumina substrates were identified which withstood contact at the operating temperature of a molten regolith electrolysis cell (1600 C) for eight hours with little interaction or deformation. This represents an improvement over alumina grades currently in use and will provide a lifetime adequate for electrolysis experiments lasting 24 hours or more. Two types of non-oxide ceramics were also tested. It was found that they interacted to a limited degree with the melt resulting in little corrosion. These ceramics, Sic and BN, were not wetted as well as the oxides by the melt, and so remain possible materials for molten regolith handling. Tests wing longer holding periods and larger volumes of regolith are necessary to determine the ultimate performance of the tested ceramics.

  14. Review of Solid Waste Management Practice, Handling and Planning in the Construction Industry

    Directory of Open Access Journals (Sweden)

    Fiza Mohd Noh

    2017-11-01

    Full Text Available The building and construction industry is a major contributor to the source of national economy. However, inappropriate construction waste management lead to various problems such as illegal dumping along the roadsides, demolition waste and disposal of construction at landfills that Malaysia is facing serious shortage of landfill space and recently the issue has become more serious throughout the country, which these have caused major government sources and environmental issue. Solid waste management is one of the environmental issues which always been a concerned to most governments. In urban areas, 46% of the population in the statistics that shows the world population has reached six billion. In 1997, generation of the municipal solid waste was about 0.49 billion tons around the globe with an estimated annual growth rate of 3.2-4.5% in develop nations and 2-3% in developing countries. The characteristics of solid waste generated were changed due to the rapid urbanization and industrialization.

  15. Handling and storage of high-level liquid wastes from reprocessing of spent fuel

    International Nuclear Information System (INIS)

    Finsterwalder, L.

    1982-01-01

    The high level liquid wastes arise from the reprocessing of irradiated nuclear fuels, which are dissolved in aqueous acid solution, and the plutonium and unburned uranium removed in the chemical separation plant. The remaining solution, containing more than 99% of the dissolved fission products, together with impurities from cladding materials, corrosion products, traces of unseparated plutonium and uranium and most of the transuranic elements, constitutes the high-level waste. At present, these liquid wastes are usually concentrated by evaporation and stored as an aqueous nitric acid solution in high-integrity stainless-steel tanks. There is now world-wide agreement that, for the long term, these liquid wastes should be converted to solid form and much work is in progress to develop techniques for the solidification of these wastes. This paper considers the design requirements for such facilities and the experience gained during nearly 30 years of operation. (orig./RW)

  16. The Remote Handled Immobilization Low-Activity Waste Disposal Facility Environmental Permits and Approval Plan

    International Nuclear Information System (INIS)

    DEFFENBAUGH, M.L.

    2000-01-01

    The purpose of this document is to revise Document HNF-SD-ENV-EE-003, ''Permitting Plan for the Immobilized Low-Activity Waste Project, which was submitted on September 4, 1997. That plan accounted for the interim storage and disposal of Immobilized-Low Activity Waste at the existing Grout Treatment Facility Vaults (Project W-465) and within a newly constructed facility (Project W-520). Project W-520 was to have contained a combination of concrete vaults and trenches. This document supersedes that plan because of two subsequent items: (1) A disposal authorization that was received on October 25, 1999, in a U. S. Department of Energy-Headquarters, memorandum, ''Disposal Authorization Statement for the Department of Energy Hanford site Low-Level Waste Disposal facilities'' and (2) ''Breakthrough Initiative Immobilized Low-Activity Waste (ILAW) Disposal Alternative,'' August 1999, from Lucas Incorporated, Richland, Washington. The direction within the U. S. Department of Energy-Headquarters memorandum was given as follows: ''The DOE Radioactive Waste Management Order requires that a Disposal authorization statement be obtained prior to construction of new low-level waste disposal facility. Field elements with the existing low-level waste disposal facilities shall obtain a disposal authorization statement in accordance with the schedule in the complex-wide Low-Level Waste Management Program Plan. The disposal authorization statement shall be issued based on a review of the facility's performance assessment and composite analysis or appropriate CERCLA documentation. The disposal authorization shall specify the limits and conditions on construction, design, operations, and closure of the low-level waste facility based on these reviews. A disposal authorization statement is a part of the required radioactive waste management basis for a disposal facility. Failure to obtain a disposal authorization statement or record of decision shall result in shutdown of an operational

  17. Management of radioactive waste in nuclear power: handling of irradiated graphite from water-cooled graphite reactors

    International Nuclear Information System (INIS)

    Anfimov, S.S.

    2000-01-01

    As a result of decommissioning of water-cooled graphite-moderated reactors, a large amount of rad-waste in the form of graphite stack fragments is generated (on average 1500-2000 tons per reactor). That is why it is essentially important, although complex from the technical point of view, to develop advanced technologies based on up-to-date remotely-controlled systems for unmanned dismantling of the graphite stack containing highly-active long-lived radionuclides and for conditioning of irradiated graphite (IG) for the purposes of transportation and subsequent long term and ecologically safe storage either on NPP sites or in special-purpose geological repositories. The main characteristics critical for radiation and nuclear hazards of the graphite stack are as follows: the graphite stack is contaminated with nuclear fuel that has gotten there as a result of the accidents; the graphite mass is 992 tons, total activity -6?104 Ci (at the time of unit shutdown); the fuel mass in the reactor stack amounts to 100-140 kg, as estimated by IPPE and RDIPE, respectively; γ-radiation dose rate in the stack cells varies from 4 to 4300 R/h, with the prevailing values being in the range from 50 to 100 R/h. In this paper the traditional methods of rad-waste handling as bituminization technology, cementing technology are discussed. In terms of IG handling technology two lines were identified: long-term storage of conditioned IG and IG disposal by means of incineration. The specific cost of graphite immobilization in a radiation-resistant polymeric matrix amounts to -2600 USD per 1 t of graphite, whereas the specific cost of immobilization in slag-stone containers with an inorganic binder (cement) is -1400 USD per 1 t of graphite. On the other hand, volume of conditioned IG rad-waste subject for disposal, if obtained by means of the first technology, is 2-2.5 times less than the volume of rad-waste generated by means of the second technology. It can be concluded from the above that

  18. Potential of Briquetting as a Waste-Management Option for Handling Market-Generated Vegetable Waste in Port Harcourt, Nigeria

    Directory of Open Access Journals (Sweden)

    Olugbemiro M. Akande

    2018-03-01

    Full Text Available The conversion of biomass to high-density briquettes is a potential solution to solid waste problems as well as to a high dependence on fuel wood in developing countries. In this study, the potential of converting vegetable waste to briquettes using waste paper as a binder was investigated. A sample size of 30 respondents was interviewed using a self-administered questionnaire at the D-line fruit and vegetable market in Port Harcourt, Nigeria. Carrot and cabbage leaves were selected for briquetting based on their availability and heating value. This waste was sun-dried, pulverized, torrefied and fermented. Briquettes were produced with a manual briquette press after the processed vegetable waste was mixed with waste paper in four paper:waste ratios, i.e., 10:90, 15:85, 20:80 and 25:75. The moisture content, densities and cooking efficiency of the briquettes were determined using the oven-drying method, the water-displacement method, and the water-boiling test, respectively. There was no observed trend in moisture content values of the briquettes, which varied significantly between 3.0% and 8.5%. There was no significant variation in the densities, which ranged from 0.79 g/cm3 to 0.96 g/cm3 for all the briquette types. A degree of compaction above 300% was achieved for all the briquette types. Water-boiling test results revealed that 10:90 paper:sun-dried cabbage briquettes had the highest ignitability of 0.32 min. Torrefied carrot briquettes with 25% paper had the least boiling time and the highest burning rates of 9.21 min and 4.89 g/min, respectively. It was concluded that cabbage and carrot waste can best be converted into good-quality briquettes after torrefaction.

  19. Proposal for Managing Eco-efficient Operations Plant Dedicated to Waste Handling at Costa Rican Institute of Electricity

    Directory of Open Access Journals (Sweden)

    Annie Chinchilla

    2015-06-01

    Full Text Available In the present study, different eco-efficient specifications were established considered by Ingeniería y Construcciónor IC (Engineering and Construction, a business of the Costa Rican Institute of Electricity (ICE, in Spanish, at the time of developing an operational plant devoted to the handling of waste, in order to make rational use of resources and generate the lowest environmental impact. Initially a general diagnosis was conducted to learn about the current process of waste management in IC, as well as the identification and assessment of its aspects and environmental impacts. An ecoefficiency proposal program was subsequently prepared to be implemented once the ordinary, special and hazardous waste plant is operating. As part of this investigation, eco-efficient measures and technologies were also identified; this can be adopted by IC or any organization to improve its waste management. Finally, it is necessary that the Eco-efficient Management Program (PGE, in Spanish is organized, planned and systematized over time; in addition, the need to have an Ecoefficiency Management Committee arises, which will allow to implement it and measure it through a series of indicators.

  20. Handling of spent nuclear fuel and final storage of vitrified high level reprocessing waste

    International Nuclear Information System (INIS)

    1978-01-01

    A summary of the planning of transportation and plant design in the Swedish KBS project on management and disposal reprocessed radioactive waste. It describes a transportation system, a central storage facility for used fuel elements, a plant for intermediate storage and encapsulation and a final repository for the vitrified waste. Accounts are given for the reprocessing and vitrification. The safety of the entire system is discussed

  1. Engineering studies: high-level radioactive waste repositories task 3 - review of underground handling and emplacement. 1. Executive summary

    Energy Technology Data Exchange (ETDEWEB)

    1981-02-01

    The report reviews proposals for transport, handling and emplacement of high-level radioactive waste in an underground repository with particular reference to: waste block size and configuration; self-shielded or partially-shielded block; stages of disposal; transport by road/rail to repository site; handling techniques within repository; emplacement in vertical holes or horizontal tunnels; repository access by adit, incline or shaft; conventional and radiological safety; costs; and major areas of uncertainty requiring research or development. In carrying out this programme due attention was given to work already carried out both in the U.K. and overseas and where appropriate comparisons with this work have been made to substantiate and explain the observations made in this report. The examination and use of this previous work however has been confined to those proposals which were considered capable of meeting the basic design criterion for a U.K. based repository, that the maximum temperature achieved by the host rock should not exceed 100/sup 0/C.

  2. Development of the remote-handled transuranic waste radioassay data quality objectives. An evaluation of RH-TRU waste inventories, characteristics, radioassay methods and capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Meeks, A.M.; Chapman, J.A.

    1997-09-01

    The Waste Isolation Pilot Plant will accept remote-handled transuranic waste as early as October of 2001. Several tasks must be accomplished to meet this schedule, one of which is the development of Data Quality Objectives (DQOs) and corresponding Quality Assurance Objectives (QAOs) for the assay of radioisotopes in RH-TRU waste. Oak Ridge National Laboratory (ORNL) was assigned the task of providing to the DOE QAO, information necessary to aide in the development of DQOs for the radioassay of RH-TRU waste. Consistent with the DQO process, information needed and presented in this report includes: identification of RH-TRU generator site radionuclide data that may have potential significance to the performance of the WIPP repository or transportation requirements; evaluation of existing methods to measure the identified isotopic and quantitative radionuclide data; evaluation of existing data as a function of site waste streams using documented site information on fuel burnup, radioisotope processing and reprocessing, special research and development activities, measurement collection efforts, and acceptable knowledge; and the current status of technologies and capabilities at site facilities for the identification and assay of radionuclides in RH-TRU waste streams. This report is intended to provide guidance in developing the RH-TRU waste radioassay DQOs, first by establishing a baseline from which to work, second, by identifying needs to fill in the gaps between what is known and achievable today and that which will be required before DQOs can be formulated, and third, by recommending measures that should be taken to assure that the DQOs in fact balance risk and cost with an achievable degree of certainty.

  3. Development of the remote-handled transuranic waste radioassay data quality objectives. An evaluation of RH-TRU waste inventories, characteristics, radioassay methods and capabilities

    International Nuclear Information System (INIS)

    Meeks, A.M.; Chapman, J.A.

    1997-09-01

    The Waste Isolation Pilot Plant will accept remote-handled transuranic waste as early as October of 2001. Several tasks must be accomplished to meet this schedule, one of which is the development of Data Quality Objectives (DQOs) and corresponding Quality Assurance Objectives (QAOs) for the assay of radioisotopes in RH-TRU waste. Oak Ridge National Laboratory (ORNL) was assigned the task of providing to the DOE QAO, information necessary to aide in the development of DQOs for the radioassay of RH-TRU waste. Consistent with the DQO process, information needed and presented in this report includes: identification of RH-TRU generator site radionuclide data that may have potential significance to the performance of the WIPP repository or transportation requirements; evaluation of existing methods to measure the identified isotopic and quantitative radionuclide data; evaluation of existing data as a function of site waste streams using documented site information on fuel burnup, radioisotope processing and reprocessing, special research and development activities, measurement collection efforts, and acceptable knowledge; and the current status of technologies and capabilities at site facilities for the identification and assay of radionuclides in RH-TRU waste streams. This report is intended to provide guidance in developing the RH-TRU waste radioassay DQOs, first by establishing a baseline from which to work, second, by identifying needs to fill in the gaps between what is known and achievable today and that which will be required before DQOs can be formulated, and third, by recommending measures that should be taken to assure that the DQOs in fact balance risk and cost with an achievable degree of certainty

  4. The Remote Handled Immobilization Low Activity Waste Disposal Facility Environmental Permits & Approval Plan

    Energy Technology Data Exchange (ETDEWEB)

    DEFFENBAUGH, M.L.

    2000-08-01

    The purpose of this document is to revise Document HNF-SD-ENV-EE-003, ''Permitting Plan for the Immobilized Low-Activity Waste Project, which was submitted on September 4, 1997. That plan accounted for the interim storage and disposal of Immobilized-Low Activity Waste at the existing Grout Treatment Facility Vaults (Project W-465) and within a newly constructed facility (Project W-520). Project W-520 was to have contained a combination of concrete vaults and trenches. This document supersedes that plan because of two subsequent items: (1) A disposal authorization that was received on October 25, 1999, in a U. S. Department of Energy-Headquarters, memorandum, ''Disposal Authorization Statement for the Department of Energy Hanford site Low-Level Waste Disposal facilities'' and (2) ''Breakthrough Initiative Immobilized Low-Activity Waste (ILAW) Disposal Alternative,'' August 1999, from Lucas Incorporated, Richland, Washington. The direction within the U. S. Department of Energy-Headquarters memorandum was given as follows: ''The DOE Radioactive Waste Management Order requires that a Disposal authorization statement be obtained prior to construction of new low-level waste disposal facility. Field elements with the existing low-level waste disposal facilities shall obtain a disposal authorization statement in accordance with the schedule in the complex-wide Low-Level Waste Management Program Plan. The disposal authorization statement shall be issued based on a review of the facility's performance assessment and composite analysis or appropriate CERCLA documentation. The disposal authorization shall specify the limits and conditions on construction, design, operations, and closure of the low-level waste facility based on these reviews. A disposal authorization statement is a part of the required radioactive waste management basis for a disposal facility. Failure to obtain a disposal authorization statement

  5. Emergy Evaluation of the Urban Solid Waste Handling in Liaoning Province, China

    Directory of Open Access Journals (Sweden)

    Lixiao Zhang

    2013-10-01

    Full Text Available Waste management is a distinct practice aimed at reducing its effects on health and the environment and increasing energy and material recovery. The urban waste management industry has been slow to adopt new technologies, such as sanitary landfills and incineration, which enable better treatment results. The aim of a thorough ecological-economic evaluation of different treatment technologies is to extract the maximum practical benefits from investments and to ensure the minimum environmental impacts of wastes. This paper compares four garbage treatment systems, including sanitary landfills systems, fluidized bed incineration system, grate type incineration system and the current landfills system in Liaoning Province, China. By considering the economic and environmental impacts of waste treatment and disposal, impact of emissions, and contribution of wastes input, this paper constructed an emergy-based urban solid waste model for evaluating the sustainability of the holistic systems. The results in Liaoning indicate that the human health losses caused by the harmful air emissions are ranked in this order: fluidized bed incineration > grate type incineration > current landfills > sanitary landfills, while the ecosystem losses are ranked: grate type incineration > fluidized bed incineration > sanitary landfills > current landfills. The electricity yield ratios are ranked: grate type incineration > fluidized bed incineration > sanitary landfills > current landfills. Taken together this suggests that in considering the incineration option, decision makers must weigh the benefits of incineration against the significant operating costs, potential environmental impacts, and technical difficulties of operating. Emergy analysis of the urban solid treatment systems can provide a set of useful tools which can be used to compare the comprehensive performances of different waste treatment processes for decision-making and optimizing the whole process.

  6. Safety evaluation report of the Waste Isolation Pilot Plant safety analysis report: Contact-handled transuranic waste disposal operations

    International Nuclear Information System (INIS)

    1997-02-01

    DOE 5480.23, Nuclear Safety Analysis Reports, requires that the US Department of Energy conduct an independent, defensible, review in order to approve a Safety Analysis Report (SAR). That review and the SAR approval basis is documented in this formal Safety Evaluation Report (SER). This SER documents the DOE's review of the Waste Isolation Pilot Plant SAR and provides the Carlsbad Area Office Manager, the WIPP SAR approval authority, with the basis for approving the safety document. It concludes that the safety basis documented in the WIPP SAR is comprehensive, correct, and commensurate with hazards associated with planned waste disposal operations

  7. Electrostatic application of antimicrobial sprays to sanitize food handling and processing surfaces for enhanced food safety

    Energy Technology Data Exchange (ETDEWEB)

    Lyons, Shawn M; Harrison, Mark A [Food Science and Technology Department, University of Georgia, Athens, GA, 30602-2610 (United States); Law, S Edward, E-mail: edlaw@engr.uga.edu [Biological and Agricultural Engineering Department, Applied Electrostatics Laboratory www.ael.engr.uga.edu, University of Georgia, Athens, GA, 30602-4435 (United States)

    2011-06-23

    Human illnesses and deaths caused by foodborne pathogens (e.g., Salmonella enterica, Listeria monocytogenes, Escherichia coli O157:H7, etc.) are of increasing concern globally in maintaining safe food supplies. At various stages of the food production, processing and supply chain antimicrobial agents are required to sanitize contact surfaces. Additionally, during outbreaks of contagious pathogenic microorganisms (e.g., H1N1 influenza), public health requires timely decontamination of extensive surfaces within public schools, mass transit systems, etc. Prior publications verify effectiveness of air-assisted, induction-charged (AAIC) electrostatic spraying of various chemical and biological agents to protect on-farm production of food crops...typically doubling droplet deposition efficiency with concomitant increases in biological control efficacy. Within a biosafety facility this present work evaluated the AAIC electrostatic-spraying process for application of antimicrobial liquids onto various pathogen-inoculated food processing and handling surfaces as a food safety intervention strategy. Fluoroanalysis of AAIC electrostatic sprays (-7.2 mC/kg charge-to-mass ratio) showed significantly greater (p<0.05) mass of tracer active ingredient (A.I.) deposited onto target surfaces at various orientations as compared both to a similar uncharged spray nozzle (0 mC/kg) and to a conventional hydraulic-atomizing nozzle. Per unit mass of A.I. dispensed toward targets, for example, A.I. mass deposited by AAIC electrostatic sprays onto difficult to coat backsides was 6.1-times greater than for similar uncharged sprays and 29.0-times greater than for conventional hydraulic-nozzle sprays. Even at the 56% reduction in peracetic acid sanitizer A.I. dispensed by AAIC electrostatic spray applications, they achieved equal or greater CFU population reductions of Salmonella on most target orientations and materials as compared to uncharged sprays and conventional full-rate hydraulic

  8. Electrostatic application of antimicrobial sprays to sanitize food handling and processing surfaces for enhanced food safety

    International Nuclear Information System (INIS)

    Lyons, Shawn M; Harrison, Mark A; Law, S Edward

    2011-01-01

    Human illnesses and deaths caused by foodborne pathogens (e.g., Salmonella enterica, Listeria monocytogenes, Escherichia coli O157:H7, etc.) are of increasing concern globally in maintaining safe food supplies. At various stages of the food production, processing and supply chain antimicrobial agents are required to sanitize contact surfaces. Additionally, during outbreaks of contagious pathogenic microorganisms (e.g., H1N1 influenza), public health requires timely decontamination of extensive surfaces within public schools, mass transit systems, etc. Prior publications verify effectiveness of air-assisted, induction-charged (AAIC) electrostatic spraying of various chemical and biological agents to protect on-farm production of food crops...typically doubling droplet deposition efficiency with concomitant increases in biological control efficacy. Within a biosafety facility this present work evaluated the AAIC electrostatic-spraying process for application of antimicrobial liquids onto various pathogen-inoculated food processing and handling surfaces as a food safety intervention strategy. Fluoroanalysis of AAIC electrostatic sprays (-7.2 mC/kg charge-to-mass ratio) showed significantly greater (p<0.05) mass of tracer active ingredient (A.I.) deposited onto target surfaces at various orientations as compared both to a similar uncharged spray nozzle (0 mC/kg) and to a conventional hydraulic-atomizing nozzle. Per unit mass of A.I. dispensed toward targets, for example, A.I. mass deposited by AAIC electrostatic sprays onto difficult to coat backsides was 6.1-times greater than for similar uncharged sprays and 29.0-times greater than for conventional hydraulic-nozzle sprays. Even at the 56% reduction in peracetic acid sanitizer A.I. dispensed by AAIC electrostatic spray applications, they achieved equal or greater CFU population reductions of Salmonella on most target orientations and materials as compared to uncharged sprays and conventional full-rate hydraulic

  9. INNOVATIVE TECHNIQUES AND TECHNOLOGY APPLICATION IN MANAGEMENT OF REMOTE HANDLED AND LARGE SIZED MIXED WASTE FORMS

    International Nuclear Information System (INIS)

    BLACKFORD LT

    2008-01-01

    CH2M HILL Hanford Group, Inc. (CH2M HILL) plays a critical role in Hanford Site cleanup for the U. S. Department of Energy, Office of River Protection (ORP). CH2M HILL is responsible for the management of 177 tanks containing 53 million gallons of highly radioactive wastes generated from weapons production activities from 1943 through 1990. In that time, 149 single-shell tanks, ranging in capacity from 50,000 gallons to 500,000 gallons, and 28 double-shell tanks with a capacity of 1 million gallons each, were constructed and filled with toxic liquid wastes and sludges. The cleanup mission includes removing these radioactive waste solids from the single-shell tanks to double-shell tanks for staging as feed to the Waste Treatment Plant (WTP) on the Hanford Site for vitrification of the wastes and disposal on the Hanford Site and Yucca Mountain repository. Concentrated efforts in retrieving residual solid and sludges from the single-shell tanks began in 2003; the first tank retrieved was C-106 in the 200 East Area of the site. The process for retrieval requires installation of modified sluicing systems, vacuum systems, and pumping systems into existing tank risers. Inherent with this process is the removal of existing pumps, thermo-couples, and agitating and monitoring equipment from the tank to be retrieved. Historically, these types of equipment have been extremely difficult to manage from the aspect of radiological dose, size, and weight of the equipment, as well as their attendant operating and support systems such as electrical distribution and control panels, filter systems, and mobile retrieval systems. Significant effort and expense were required to manage this new waste stream and resulted in several events over time that were both determined to be unsafe for workers and potentially unsound for protection of the environment. Over the last four years, processes and systems have been developed that reduce worker exposures to these hazards, eliminate violations

  10. Nukem Nuclear GmbH activity in CIS countries in the sphere of radioactive wastes and nuclear fuel handling

    International Nuclear Information System (INIS)

    Vaihard, A.

    1997-01-01

    NUKEM was founded in 1960 as one of the first nuclear companies in the German Federal Republic. With this work, Nukem developed not only processes for producing fuels and fuel elements, but also the plant and equipment necessary for this production. NUKEM engineers further planned and built the total infrastructure for operation of these manufacturing plants, including the supply and waste plants, the nuclear ventilation technology, the laboratory and the remote handling manipulators. The scope of activities extends from the design to the manufacture and construction of turnkey plants. The points of emphasis are plants and processes for the Treatment of radioactive wastes, storage and container technology, the decommissioning of nuclear plants, the planning and building of nuclear laboratories, the design of fuel elements and safety and monitoring technology. NUKEM Nuclear Technology is an independent division within the plant construction of the NUKEM Group. Additionally, five further subsidiary and partner companies have a spectrum of nuclear technology activities. Altogether, Nukem Nuclear Technology counts around 300 highly qualified engineers, scientists and technicians. Numerous Designs and patents underline the strength of innovative output in this area. The engineering service offered by NUKEM includes the whole spectrum of process and technology as well as construction and start-up as general engineer or general contractor: Basic engineering; Detail engineering; Procurement; Personnel Training; Start-up. Engineering and safety for nuclear technology: Process and plant planing; Media supply and disposal; Building and architecture; Electrical, measurement and control technology; Safety and accident analysis; Licensing procedures. Treatment of Radioactive Wastes: - Volume reduction of soil and liquid wastes: vaporizer plants; evaporator plants; incineration plants; pyrolysis plants; compactors. - Chemical/physical processes for residue treatment: boric acid

  11. Evaluation of doses during the handling and transport of radioactive wastes containers

    International Nuclear Information System (INIS)

    Kubik, I.; Kusovska, Z.; Hanusik, V.; Mrskova, A.; Kapisovsky, V.

    2000-01-01

    Radioactive waste products from the nuclear power plants (NPPs) must be isolated from contact with people for very long period of time. Low and intermediate-level waste will be disposed of in Slovakia in specially licensed Regional disposal facility which is located near the NPP Mochovce site. Radioactive waste accumulated in the Jaslovsk. Bohunice site, during the decommissioning process of the NPP A-1 and arising from the NPP V-1 and NPP V-2 operation, will be processed and shipped in standard concrete containers to the Mochovce Regional disposal facility. The treatment centre was build at the NPP Jaslovsk? Bohunice site which is in the trial operation now. It is supposed that radioactive waste containers will be transported by train from the treatment centre Jaslovsk? Bohunice to the site of Radioactive Waste Repository at Mochovce and by truck in the area of repository. To estimate the occupational radiation exposure during the transport the calculations of dose rates from the containers are necessary. The national regulations allow low level of radiation to emanate from the casks and containers. The maximum permissible volume radioactivity of wastes inside the container is limited in such a way that irradiation level should not exceed 2 mGy/h for the contact irradiation level and 0,1 mGy/h at 2-meter distance. MicroShield code was used to analyse shielding and assessing exposure from gamma radiation of containers to people. A radioactive source was conservatively modelled by homogenous mixture of radionuclides with concrete. Standard rectangular volume source and shield geometry is used in model calculations. The activities of the personnel during the transport and storage of containers are analysed and results of the evaluation of external dose rates and effective doses are described. (author)

  12. Neutron and gamma-ray nondestructive examination of contact-handled transuranic waste at the ORNL TRU Waste Drum Assay Facility

    International Nuclear Information System (INIS)

    Schultz, F.J.; Coffey, D.E.; Norris, L.B.; Haff, K.W.

    1985-03-01

    A nondestructive assay system, which includes the Neutron Assay System (NAS) and the Segmented Gamma Scanner (SGS), for the quantification of contact-handled (<200 mrem/h total radiation dose rate at contact with container) transuranic elements (CH-TRU) in bulk solid waste contained in 208-L and 114-L drums has been in operation at the Oak Ridge National Laboratory since April 1982. The NAS has been developed and demonstrated by Los Alamos National Laboratory (LANL) and the Oak Ridge National Laboratory (ORNL) for use by most US Department of Energy Defense Plant (DOE-DP) sites. More research and development is required, however, before the NAS can provide complete assay results for other than routine defense waste. To date, 525 ORNL waste drums have been assayed, with varying degrees of success. The isotopic complexity of the ORNL waste creates a correspondingly complex assay problem. The NAS and SGS assay data are presented and discussed. Neutron matrix effects, the destructive examination facility, and enriched uranium fuel-element assays are also discussed

  13. The ministry of environment: advanced measures to handle wastes and combat pollution

    International Nuclear Information System (INIS)

    Chartouni, Talal

    1997-01-01

    Lebanon is facing environmental problems that stem from wartime chaos, negligence and lack of governmental control. Since the end of the war, environmental protection and getting rid of pollution factors and hazardous waste introduced to Lebanon during the war have become a main concern to the government and the people. The author as an adviser to the environment minister, presents the environment ministry's plan to fight pollution and remove hazardous waste and reveals the steps already executed and the ones to be taken in the future

  14. Assessment of Geochemical Environment for the Proposed INL Remote-Handled Low-Level Waste Disposal Facility

    Energy Technology Data Exchange (ETDEWEB)

    D. Craig Cooper

    2011-11-01

    Conservative sorption parameters have been estimated for the proposed Idaho National Laboratory Remote-Handled Low-Level Waste Disposal Facility. This analysis considers the influence of soils, concrete, and steel components on water chemistry and the influence of water chemistry on the relative partitioning of radionuclides over the life of the facility. A set of estimated conservative distribution coefficients for the primary media encountered by transported radionuclides has been recommended. These media include the vault system, concrete-sand-gravel mix, alluvium, and sedimentary interbeds. This analysis was prepared to support the performance assessment required by U.S. Department of Energy Order 435.1, 'Radioactive Waste Management.' The estimated distribution coefficients are provided to support release and transport calculations of radionuclides from the waste form through the vadose zone. A range of sorption parameters are provided for each key transport media, with recommended values being conservative. The range of uncertainty has been bounded through an assessment of most-likely-minimum and most-likely-maximum distribution coefficient values. The range allows for adequate assessment of mean facility performance while providing the basis for uncertainty analysis.

  15. A compost bin for handling privy wastes: its fabrication and use

    Science.gov (United States)

    R.E. Leonard; S.C. Fay

    1978-01-01

    A 24-ft3 (6.8-m3) fiberglass bin was constructed and tested for its effectiveness in composting privy wastes. A mixture of ground hardwood bark and raw sewage was used for composting. Temperatures in excess of 60°C for 36 hours were produced in the bin by aerobic, thermophilic composting. This temperature is...

  16. Handling and storage of high-level radioactive liquid wastes requiring cooling

    International Nuclear Information System (INIS)

    1979-01-01

    The technology of high-level liquid wastes storage and experience in this field gained over the past 25 years are reviewed in this report. It considers the design requirements for storage facilities, describes the systems currently in use, together with essential accessories such as the transfer and off-gas cleaning systems, and examines the safety and environmental factors

  17. Preliminary siting activities for new waste handling facilities at the Idaho National Engineering Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, D.D.; Hoskinson, R.L.; Kingsford, C.O.; Ball, L.W.

    1994-09-01

    The Idaho Waste Processing Facility, the Mixed and Low-Level Waste Treatment Facility, and the Mixed and Low-Level Waste Disposal Facility are new waste treatment, storage, and disposal facilities that have been proposed at the Idaho National Engineering Laboratory (INEL). A prime consideration in planning for such facilities is the selection of a site. Since spring of 1992, waste management personnel at the INEL have been involved in activities directed to this end. These activities have resulted in the (a) identification of generic siting criteria, considered applicable to either treatment or disposal facilities for the purpose of preliminary site evaluations and comparisons, (b) selection of six candidate locations for siting,and (c) site-specific characterization of candidate sites relative to selected siting criteria. This report describes the information gathered in the above three categories for the six candidate sites. However, a single, preferred site has not yet been identified. Such a determination requires an overall, composite ranking of the candidate sites, which accounts for the fact that the sites under consideration have different advantages and disadvantages, that no single site is superior to all the others in all the siting criteria, and that the criteria should be assigned different weighing factors depending on whether a site is to host a treatment or a disposal facility. Stakeholder input should now be solicited to help guide the final selection. This input will include (a) siting issues not already identified in the siting, work to date, and (b) relative importances of the individual siting criteria. Final site selection will not be completed until stakeholder input (from the State of Idaho, regulatory agencies, the public, etc.) in the above areas has been obtained and a strategy has been developed to make a composite ranking of all candidate sites that accounts for all the siting criteria.

  18. Preliminary siting activities for new waste handling facilities at the Idaho National Engineering Laboratory

    International Nuclear Information System (INIS)

    Taylor, D.D.; Hoskinson, R.L.; Kingsford, C.O.; Ball, L.W.

    1994-09-01

    The Idaho Waste Processing Facility, the Mixed and Low-Level Waste Treatment Facility, and the Mixed and Low-Level Waste Disposal Facility are new waste treatment, storage, and disposal facilities that have been proposed at the Idaho National Engineering Laboratory (INEL). A prime consideration in planning for such facilities is the selection of a site. Since spring of 1992, waste management personnel at the INEL have been involved in activities directed to this end. These activities have resulted in the (a) identification of generic siting criteria, considered applicable to either treatment or disposal facilities for the purpose of preliminary site evaluations and comparisons, (b) selection of six candidate locations for siting,and (c) site-specific characterization of candidate sites relative to selected siting criteria. This report describes the information gathered in the above three categories for the six candidate sites. However, a single, preferred site has not yet been identified. Such a determination requires an overall, composite ranking of the candidate sites, which accounts for the fact that the sites under consideration have different advantages and disadvantages, that no single site is superior to all the others in all the siting criteria, and that the criteria should be assigned different weighing factors depending on whether a site is to host a treatment or a disposal facility. Stakeholder input should now be solicited to help guide the final selection. This input will include (a) siting issues not already identified in the siting, work to date, and (b) relative importances of the individual siting criteria. Final site selection will not be completed until stakeholder input (from the State of Idaho, regulatory agencies, the public, etc.) in the above areas has been obtained and a strategy has been developed to make a composite ranking of all candidate sites that accounts for all the siting criteria

  19. Thermal impact of waste emplacement and surface cooling associated with geologic disposal of nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    Wang, J.S.Y.; Mangold, D.C.; Spencer, R.K.; Tsang, C.F.

    1982-08-01

    The thermal effects associated with the emplacement of aged radioactive wastes in a geologic repository were studied, with emphasis on the following subjects: the waste characteristics, repository structure, and rock properties controlling the thermally induced effects; the current knowledge of the thermal, thermomechanical, and thermohydrologic impacts, determined mainly on the basis of previous studies that assume 10-year-old wastes; the thermal criteria used to determine the repository waste loading densities; and the technical advantages and disadvantages of surface cooling of the wastes prior to disposal as a means of mitigating the thermal impacts. The waste loading densities determined by repository designs for 10-year-old wastes are extended to older wastes using the near-field thermomechanical criteria based on room stability considerations. Also discussed are the effects of long surface cooling periods determined on the basis of far-field thermomechanical and thermohydrologic considerations. The extension of the surface cooling period from 10 years to longer periods can lower the near-field thermal impact but have only modest long-term effects for spent fuel. More significant long-term effects can be achieved by surface cooling of reprocessed high-level waste.

  20. Thermal impact of waste emplacement and surface cooling associated with geologic disposal of nuclear waste

    International Nuclear Information System (INIS)

    Wang, J.S.Y.; Mangold, D.C.; Spencer, R.K.; Tsang, C.F.

    1982-08-01

    The thermal effects associated with the emplacement of aged radioactive wastes in a geologic repository were studied, with emphasis on the following subjects: the waste characteristics, repository structure, and rock properties controlling the thermally induced effects; the current knowledge of the thermal, thermomechanical, and thermohydrologic impacts, determined mainly on the basis of previous studies that assume 10-year-old wastes; the thermal criteria used to determine the repository waste loading densities; and the technical advantages and disadvantages of surface cooling of the wastes prior to disposal as a means of mitigating the thermal impacts. The waste loading densities determined by repository designs for 10-year-old wastes are extended to older wastes using the near-field thermomechanical criteria based on room stability considerations. Also discussed are the effects of long surface cooling periods determined on the basis of far-field thermomechanical and thermohydrologic considerations. The extension of the surface cooling period from 10 years to longer periods can lower the near-field thermal impact but have only modest long-term effects for spent fuel. More significant long-term effects can be achieved by surface cooling of reprocessed high-level waste

  1. Human factors programs for high-level radioactive waste handling systems

    International Nuclear Information System (INIS)

    Pond, D.J.

    1992-01-01

    Human Factors is the discipline concerned with the acquisition of knowledge about human capabilities and limitations, and the application of such knowledge to the design of systems. This paper discusses the range of human factors issues relevant to high-level radioactive waste (HLRW) management systems and, based on examples form other organizations, presents mechanisms through which to assure application of such expertise in the safe, efficient, and effective management and disposal of high-level waste. Additionally, specific attention is directed toward consideration of who might be classified as a human factors specialist, why human factors expertise is critical to the success of the HLRW management system, and determining when human factors specialists should become involved in the design and development process

  2. Human factors programs for high-level radioactive waste handling systems

    International Nuclear Information System (INIS)

    Pond, D.J.

    1992-04-01

    Human Factors is the discipline concerned with the acquisition of knowledge about human capabilities and limitations, and the application of such knowledge to the design of systems. This paper discusses the range of human factors issues relevant to high-level radioactive waste (HLRW) management systems and, based on examples from other organizations, presents mechanisms through which to assure application of such expertise in the safe, efficient, and effective management and disposal of high-level waste. Additionally, specific attention is directed toward consideration of who might be classified as a human factors specialist, why human factors expertise is critical to the success of the HLRW management system, and determining when human factors specialists should become involved in the design and development process

  3. Tolerancing requirements for remote handling at the Hanford Waste Vitrification Plant

    International Nuclear Information System (INIS)

    Van Katwijk, C.; Keenan, R.M.; Bullis, R.E.

    1993-01-01

    The Hanford Waste Vitrification Plant (HWVP) is being designed by Fluor Daniel, Inc. with Waste Chem Corporation as Fluor Daniel, Inc.'s major subcontractor specializing in vitrification and remote system technologies. United Engineers and Constructors (UE ampersand C)/Catalytic (UCAT) will construct the plant. Westinghouse Hanford Company is the Project Integration manager and Business manager, and as the plant operator it provides technical direction to the Architect/ Engineer team (A/E) and constructor on behalf of the US Department of Energy - Richland Field Office. The A/E has developed, in cooperation with UE ampersand C, Westinghouse Hanford Company, and the US Department of Energy, a new and innovative approach to installations of the many remote nozzles and electrical connectors that must be installed to demanding tolerances. This paper summarizes the key elements of the HWVP approach

  4. Science for safety in nuclear waste handling; Aspects scientifiques de la surete des dechets radioactifs

    Energy Technology Data Exchange (ETDEWEB)

    Bonin, B. [CEA Saclay, Dir. de l' Energie Nucleaire (DEN), 91 - Gif sur Yvette (France)

    2005-05-01

    A facility for disposing of nuclear waste has the objective of protecting mankind. It is shown how nuclear safety is taken into account in the definition of the facilities, particularly through the use of the multi-barriers concept. Elements on safety assessment are provided, with emphasis on the scenario (normal or altered approach; examples of research programs required by these assessment tasks, particularly by means of underground laboratories are given. (author)

  5. Waste handling and storage in the decontamination pilot projects of JAEA for environments of Fukushima

    Energy Technology Data Exchange (ETDEWEB)

    Nakayama, S.; Kawase, K.; Iijima, K.; Kato, M. [Fukushima Environmental Safety Center, Headquarters of Fukushima Partnership Operations, Japan Atomic Energy Agency, Fukushima (Japan)

    2013-07-01

    After the Fukushima Daiichi nuclear accident, Japan Atomic Energy Agency (JAEA) was chosen by the national government to conduct decontamination pilot projects at selected sites in Fukushima prefecture. Despite tight boundary conditions in terms of timescale and resources, the projects served their primary purpose to develop a knowledge base to support more effective planning and implementation of stepwise regional remediation of the evacuated zone. A range of established, modified and newly developed techniques were tested under realistic field conditions and their performance characteristics were determined. The results of the project can be summarized in terms of site characterization, cleanup and waste management. A range of options were investigated to reduce the volumes of waste produced and to ensure that decontamination water could be cleaned to the extent that it could be discharged to normal drainage. Resultant solid wastes were packaged in standard flexible containers, labelled and stored at the remediation site (temporary storage until central interim storage becomes available). The designs of such temporary storage facilities were tailored to available sites, but all designs included measures to ensure mechanical stability (e.g., filling void spaces between containers with sand, graded cover with soil) and prevent releases to groundwater (impermeable base and cap, gravity flow drainage including radiation monitors and catch tanks). Storage site monitoring was also needed to check that storage structures would not be perturbed by external events that could include typhoons, heavy snowfalls, freeze/thaw cycles and earthquakes. (authors)

  6. Surface analysis: its uses and abuses in waste form evaluation

    International Nuclear Information System (INIS)

    McVay, G.L.; Pederson, L.R.

    1981-01-01

    Surface and near-surface analytical techniques are significant aids in understanding waste form-aqueous solution interactions. They can be beneficially employed to evaluate reaction layers on waste forms, to assess surface treatments prior to and after leaching, and to identify interactions with waste forms. Surface analyses are best used in conjunction with other types of analyses, such as solution analyses, in order to obtain a better overall understanding of reaction processes. In spite of all the benefits to be gained by using surface analyses, misinterpretations can result if care is not taken to properly obtain and analyze the data. In particular, the density variations through a reaction layer must be accounted for in both sputtering and data analysis techniques

  7. Application of the air/water cushion technology for handling of heavy waste packages in Sweden and France

    International Nuclear Information System (INIS)

    Bosgiraud, Jean-Michel; Seidler, Wolf K.; Londe, Louis; Thurner, Erik; Pettersson, Stig

    2008-01-01

    The disposal of certain types of radioactive waste canisters in a deep repository involves handling and emplacement of very heavy loads. The weight of these particular canisters can be in the order of 20 to 50 metric tons. They generally have to be handled underground in openings that are not much larger than the canisters themselves as it is time consuming and expensive to excavate and backfill large openings in a repository. This therefore calls for the development of special technology that can meet the requirements for safe operation in an industrial scale in restrained operating spaces. Air/water cushion lifting systems are used world wide in the industry for moving heavy loads. However, until now the technology needed for emplacing heavy cylindrical radioactive waste packages in bored drifts (with narrow annular gaps) has not been developed or demonstrated previously. This paper describes the related R and D work carried out by ANDRA (for air cushion technology) and by SKB and Posiva (for water cushion technology) respectively, mainly within the framework of the European Commission (EC) funded Integrated Project called ESDRED (6th European Framework Programme). The background for both the air and the water cushion applications is presented. The specific characteristics of the two different emplacement concepts are also elaborated. The various phases of the Test Programmes (including the Prototype phases) are detailed and illustrated for the two lifting media. Conclusions are drawn for each system developed and evaluated. Finally, based on the R and D experience, improvements deemed necessary for an industrial application are listed. The tests performed so far have shown that the emplacement equipment developed is operating efficiently. However further tests are required to verify the availability and the reliability of the equipment over longer periods of time and to identify the modifications that would be needed for an industrial application in a nuclear

  8. Potential problems from shipment of high-curie content contact-handled transuranic (CH-TRU) waste to WIPP

    International Nuclear Information System (INIS)

    Neill, R.H.; Channell, J.K.

    1983-08-01

    There are about 1000 drums of contact-handled transuranic (CH-TRU) wastes containing more than 100 Ci/drum of Pu-238 that are stored at the Savannah River Plant and at the Los Alamos National Laboratory. Studies performed at DOE laboratories have shown that large quantities of gases are generated in stored drums containing 100 Ci of 238 Pu. Concentrations of hydrogen gas in the void space of the drums are often found to be high enough to be explosive. None of the analyses in the DOE WIPP Final Environmental Impact Statement, Safety Analysis Report, and Preliminary Transportation Analysis have considered the possibility that the generation of hydrogen gas by radiolysis may create an explosive or flammable hazard that could increase the frequency and severity of accidental releases of radionuclides during transportation or handling. These high 238 Pu concentration containers would also increase the estimated doses received by individuals and populations from transportation, WIPP site operations, and human intrusion scenarios even if the possibility of gas-enhanced releases is ignored. The WIPP Project Office has evaluated this effect on WIPP site operations and is suggesting a maximum limit of 140 239 Pu equivalent curies (P-Ci) per drum so that postulated accidental off-site doses will not be larger than those listed in the FEIS. The TRUPACT container, which is being designed for the transportation of CH-TRU wastes to WIPP, does not appear to meet the Nuclear Regulatory Commission regulations requiring double containment for the transportation of plutonium in quantities >20 Ci. A 20 alpha Ci/shipment limit would require about 200,000 shipments for the 4 million curies of alpha emitters slated for WIPP

  9. Awareness, practice of safety measures and the handling of medical wastes at a tertiary hospital in Nigeria.

    Science.gov (United States)

    Samuel, S O; Kayode, O O; Musa, O I

    2010-12-01

    The study is prompted by the significant public health impact of continuing rise in the emerging and re-emerging infectious diseases.and to determine the awareness and practice of safety measures in the handling of medical wastes among health workers in a teaching hospital. MATERIALS, SUBJECTS AND METHODS: Stratified sampling technique was used to choose 325 respondents from different professional groups and cadres of health workers and these included medical doctors, nurses/midwives, laboratory workers, ward attendants, porters, cleaners and laundry workers at the University of Ilorin teaching hospital between January and June 2008. Simple random sampling method by balloting was used to select subjects in each group. Data was collected using structured, self administered questionnaires which considered all the variables under study. Data collected were analyzed using Epi-Info computer software program. Three hundred and twenty five (325) questionnaires were administered, out of which 320 were returned giving a response rate of 98.5%. Respondents are nurses 128 (40.0%), doctors 107 (33.4%) and pharmacists 10 (3.1%). Years of work experience ranged from 3 to 27 years with respondents who had working experience between 11 to 15 years constituting over one quarter, 88 (27.5%) while those below 5 years were 8 (2.5'%). Two hundred and ninety eight (93.0%) respondents knew about hospital wastes while 193 (60.3%) only knew about general wastes. Majority of the health workers have appreciable knowledge of collection, minimization and personal risks associated with hospital wastes 299 (93.4%), 302 (94.4%) and 311 (97.2%) respectively. The most common routine safety practice is putting on protective clothing. This study revealed a high level of awareness of hospital wastes among health workers; however, the practice of standard safety measures was low. It is recommended that hospital wastes disposal and management policy be formulated and appropriate committee constituted to

  10. Exhibit of ADS transmutation system to-Handle MA contained in Highly Radioactive Waste

    International Nuclear Information System (INIS)

    Marsodi; Lasman, A.N.; Nishihara, K.; Marsongkohadi; Su'ud, Z.

    2002-01-01

    This ADS transmutation system consists of a high intensity proton beam accelerator, spallation target, and sub-critical reactor core. The general approach was conducted using N-15 fuel to choose a strategy for destroying or minimizing the dangerously radioactive waste using a fast neutron spectrum. The fuel of this system was put surrounding the target with the some composition, i.e. the composition of MOX from PWR reactor spent-fuel with 5 year cooling time. Basic characteristics of this system have been conducted based on analysis of neutronics calculation results using ATRAS codes system

  11. Historical summary of the fuel and waste handling and disposition activities of the TMI-2 Information and Examination Program (1980-1988)

    International Nuclear Information System (INIS)

    Reno, H.W.; Schmitt, R.C.

    1988-10-01

    This report is a historical summary of the major activities conducted by the TMI-2 Information and Examination Program in managing fuel and special radioactive wastes resulting from the accident at the Unit 2 reactor of the Three Mile Island Nuclear Power Station (TMI-2). The activities often required the development and use of advanced handling, processing, and/or disposal technologies for those wastes

  12. Efforts to Handle Waste through Science, Environment, Technology and Society (SETS)

    Science.gov (United States)

    Rahmawati, D.; Rahman, T.; Amprasto, A.

    2017-09-01

    This research to identify the attempt to deal with the waste through a learning SETS to facilitate troubleshooting and environmentally conscious high school students. The research method is weak experiment, with the design of the study “The One-group pretest-Posttest Design”. The population used in this study is an entire senior high school class in Ciamis Regency of Indonesia many as 10 classes totaling 360 students. The sample used in this study were 1 class. Data collected through pretest and posttest to increase problem-solving skills and environmental awareness of students. Instruments used in this research is to test the ability to solve the problem on the concept of Pollution and Environmental Protection, in the form of essays by 15 matter, the attitude scale questionnaire of 28 statements. The analysis N-gain average showed that the SETS problem-solving skills and environmental awareness of students in the medium category. In addition, students’ creativity in finding out pretty good waste management by creating products that are aesthetically valuable and economic appropriately.

  13. Survey and evaluation of handling and disposing of solid low-level nuclear fuel cycle wastes

    International Nuclear Information System (INIS)

    Mullarkey, T.B.; Jentz, T.L.; Connelly, J.M.; Kane, J.P.

    1976-10-01

    The report identifies the types and quantities of low-level solid radwaste for each portion of the nuclear fuel cycle, based on operating experiences at existing sites and design information for future installations. These facts are used to evaluate reference 1000 MWe reactor plants in terms of solid radwaste generation. The effect of waste volumes on disposal methods and land usage has also been determined, based on projections of nuclear power growth through the year 2000. The relative advantages of volume reduction alternatives are included. Major conclusions are drawn concerning available land burial space, light water reactors and fuel fabrication and reprocessing facilities. Study was conducted under the direction of an industry task force and the National Environmental Studies Project, a technical program of the Atomic Industrial Forum. Data was obtained from questionnaires sent to 8 fuel fabrication facilities, 39 reactor sites and 6 commercial waste disposal sites. Additional data were gathered from interviews with architect engineering firms, site visits, contacts with regulatory agencies and published literature

  14. Legal aspects of handling and disposal of nuclear waste - an Indian perspective

    International Nuclear Information System (INIS)

    Sree Sudha, P.

    2014-01-01

    India's rise as a global power has made it an extremely lucrative market, especially in the field of nuclear energy. Nuclear energy is often painted as a 'clean- energy option, and therefore a solution to climate change. Splitting the atom doesn't produce greenhouse gases, but the nuclear fuel cycle is far from clean: it produces radioactive waste that pollutes the environment for generations. As the most populous democracy in the world, India's energy needs far exceed its current capacity and to achieve this, the Government of India intends to draw twenty-five percent of its energy from nuclear power by the year 2050. This plan includes 20,000 MW of installed capacity from nuclear energy by 2020, and 63,000 MW by 2032. There are currently twenty operational nuclear power reactors in India, across six states. They contribute less than three per cent of the country's total energy generation, yet radioactively pollute at every stage of the nuclear fuel cycle: from mining and milling to reprocessing or disposal. There is no long-term radioactive waste disposal policy in India. India is one of the few countries in the world that is expanding its nuclear power sector at an enormous rate. Seven more nuclear reactors of 4800 MW installed capacity are under construction. At least thirty-six new nuclear reactors are planned or proposed. A critical subset of any country's nuclear safety approach is its radioactive waste management, in particular management of High Level Waste. By recognizing the facts that nuclear safety and waste management are of utmost importance for success of the nuclear energy program, India ratified the Convention on Nuclear Safety (CNS) in 2005 and has recently submitted its second National Report for review. The CNS essentially seeks to commit Parties to maintain a high level of safety by setting international benchmarks based on the IAEA fundamental principles of safety, which cover design, construction, operation, the

  15. Preliminary radiological analysis of the transportation of remote-handled transuranic waste within the state of New Mexico

    International Nuclear Information System (INIS)

    Daer, G.; Harvill, J.

    1985-07-01

    This analysis assesses the potential radiological impacts on the citizens of New Mexico from the transport of RH-TRU waste to the WIPP by rail or by truck. Assuming exclusive use of the truck transport mode, the combined annual exposure to the public from accident-free shipment of waste is estimated to be 11.5 person-rem/year. It is estimated that a theoretical member of the public receiving maximum exposure to the combined truck shipments of RH-TRU waste to the WIPP would receive an annual whole body dose equivalent of 0.00072 rem. Such an exposure is insignificant in comparison to the average annual whole body dose equivalent to an individual living in the Colorado Plateau area of between 0.075 and 0.140 rem from naturally occurring radiation. The highest average annual dose commitment to any organ from potential accidents along all New Mexico truck routes to the WIPP is projected as 0.012 person-rem/year to bone surfaces. Assuming sole use of the rail transport mode, the combined annual exposure to the public from accident-free shipment of waste is estimated to be 1.3 person-rem/year. A theoretical member of the public receiving combined maximum exposure to rail shipments of RH-TRU waste to the WIPP would receive an annual whole body dose equivalent of 0.000014 rem. The highest average annual dose commitment to any organ from potential accidents along the New Mexico rail routes to the WIPP is projected as 0.0004 person-rem/year to bone surfaces

  16. Thermal impact of waste emplacement and surface cooling associated with geologic disposal of nuclear waste

    International Nuclear Information System (INIS)

    Wang, J.S.Y.; Mangold, D.C.; Spencer, R.K.; Tsang, C.F.

    1982-01-01

    The age of nuclear waste - the length of time between its removal from the reactor cores and its emplacement in a repository - is a significant factor in determining the thermal loading of a repository. The surface cooling period as well as the density and sequence of waste emplacement affects both the near-field repository structure and the far-field geologic environment. To investigate these issues, a comprehensive review was made of the available literature pertaining to thermal effects and thermal properties of mined geologic repositories. This included a careful evaluation of the effects of different surface cooling periods of the wastes, which is important for understanding the optimal thermal loading of a repository. The results led to a clearer understanding of the importance of surface cooling in evaluating the overall thermal effects of a radioactive waste repository. The principal findings from these investigations are summarized in this paper

  17. Implications of long-term surface or near-surface storage of intermediate and low-level wastes in the UK

    International Nuclear Information System (INIS)

    Murray, N.; Vande Putte, D.; Ware, R.J.

    1986-02-01

    Various options for 200 year-long storage of all Low- and Intermediate-Level wastes generated to the year 2030 are considered. On-site storage and centralised storage have been examined and compared. The feasibility of storing some of the wastes in underground facilities that are convertible to repositories has been demonstrated, but it is shown that centralised, surface storage of wastes would be more economical. There appears to be little merit in storing Intermediate Level wastes in separate facilities that could be converted to repositories. Storage is shown to be more expensive than direct disposal, except if future costs are discounted by more than about 10%. With carefully designed stores and remote handling, the collective dose to operators could be limited to about 20-40 man Sv over the whole period of storage. (author)

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

    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. GSAK approach is described, while test results are presented in Part II. (author)

  19. 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)

  20. Reactor component chemical decontamination-developments in waste handling and disposal

    International Nuclear Information System (INIS)

    Papesch, R.; Atwood, K.L.

    1989-01-01

    Because of restrictive limits on man-rem exposure in European nuclear plants, a company has developed and applied a number of chemical decontamination techniques for components that must be periodically maintained. These techniques are particularly effective for components that can be placed in a decontamination bath for dose reduction prior to performing maintenance. The cleaning technique has the ability to achieve decontamination factors of at least 20 and in some cases much greater. For components with before cleaning dose rates of between 1 to as high as 80 R/hr, significant man-rem reductions are achieved when hundreds of manhours may be required to complete required component maintenance. Transferring this solvent technology to the U.S. required a program to develop solidification formulas to allow the solvent wastes to be disposed of in accordance with regulations and in a cost effective manner. This paper demonstrates in chemical decontaminations with small liquid volume systems that concentrated decontamination solvents can be employed to achieve high decontamination factors

  1. Safety Analysis of 'Older/Aged' Handling and Transportation Equipment for Heavy Loads, Radioactive Waste and Materials in Accordance with German Nuclear Standards KTA 3902, 3903 and 3905

    International Nuclear Information System (INIS)

    Macias, P.; Prucker, E.; Stang, W.

    2006-01-01

    The purpose of this paper is to present a general safety analysis of important handling and transportation processes and their related equipment ('load chains' consisting of cranes, load-bearing equipment and load-attaching points). This project was arranged by the responsible Bavarian ministry for environment, health and consumer protection (StMUGV) in agreement with the power plant operators of all Bavarian nuclear power plants to work out potential safety improvements. The range of the equipment (e.g. reactor building, crane, refuelling machine, load-bearing equipment and load-attaching points) covers the handling and transportation of fuel elements (e. g. with fuel flasks), heavy loads (e.g. reactor pressure vessel closure head, shielding slabs) and radioactive materials and waste (e.g. waste flasks, control elements, fuel channels, structure elements). The handling equipment was subjected to a general safety analysis taking into account the ageing of the equipment and the progress of standards. Compliance with the current valid requirements of the state of science and technology as required by German Atomic Act and particularly of the nuclear safety KTA-standards (3902, 3903 and 3905) was examined. The higher protection aims 'safe handling and transportation of heavy loads and safe handling of radioactive materials and waste' of the whole analysis are to avoid a criticality accident, the release of radioactivity and inadmissible effects on important technical equipment and buildings. The scope of the analysis was to check whether these protection aims were fulfilled for all important technical handling and transportation processes. In particularly the design and manufacturing of the components and the regulations of the handling itself were examined. (authors)

  2. 324 Building Compliance Project: Selection and evaluation of alternatives for the removal of solid remote-handled mixed wastes from the 324 Building

    International Nuclear Information System (INIS)

    Ross, W.A.; Bierschbach, M.C.; Dukelow, J.S. Jr.

    1995-06-01

    Six alternatives for the interim storage of remote-handled mixed wastes from the 324 Building on the Hanford Site have been identified and evaluated. The alternatives focus on the interim storage facility and include use of existing facilities in the 200 Area, the construction of new facilities, and the vitrification of the wastes within the 324 Building to remove the majority of the wastes from under RCRA regulations. The six alternatives are summarized in Table S.1, which identifies the primary facilities to be utilized, the anticipated schedule for removal of the wastes, the costs of the transfer from 324 Building to the interim storage facility (including any capital costs), and an initial risk comparison of the alternatives. A recently negotiated Tri-Party Agreement (TPA) change requires the last of the mixed wastes to be removed by May 1999. The ability to use an existing facility reduces the costs since it eliminates the need for new capital construction. The basic regulatory approvals for the storage of mixed wastes are in place for the PUREX facility, but the Form HI permit will need some minor modifications since the 324 Building wastes have some additional characteristic waste codes and the current permit limits storage of wastes to those from the facility itself. Regulatory reviews have indicated that it will be best to use the tunnels to store the wastes. The PUREX alternatives will only provide storage for about 65% of the wastes. This results from the current schedule of the B-Cell Clean Out Project, which projects that dispersible debris will continue to be collected in small quantities until the year 2000. The remaining fraction of the wastes will then be stored in another facility. Central Waste Complex (CWC) is currently proposed for that residual waste storage; however, other options may also be available

  3. Handling, treatment, conditioning and storage of biological radioactive wastes. Technical manual for the management of low and intermediate level wastes generated at small nuclear research centres and by radioisotope users in medicine, research and industry

    International Nuclear Information System (INIS)

    1994-12-01

    Biological materials that contain radioactive isotopes have many important applications. During the production and use of these materials, waste will inevitably arise which must be managed with particular care due to their potential biological as well as radiological hazards. This report deals with wastes that arise outside the nuclear fuel cycle and is directed primarily to countries without nuclear power programmes. It is intended to provide guidance to Member States in the handling, treatment and conditioning of biological radioactive materials. The objective of radioactive waste management is to handle, pretreat, treat, condition, store, transport and dispose of radioactive waste in a manner that protects human health and the environment without imposing undue burdens on future generations. 31 refs, 15 figs, 3 tabs

  4. Handling, treatment, conditioning and storage of biological radioactive wastes. Technical manual for the management of low and intermediate level wastes generated at small nuclear research centres and by radioisotope users in medicine, research and industry

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-12-01

    Biological materials that contain radioactive isotopes have many important applications. During the production and use of these materials, waste will inevitably arise which must be managed with particular care due to their potential biological as well as radiological hazards. This report deals with wastes that arise outside the nuclear fuel cycle and is directed primarily to countries without nuclear power programmes. It is intended to provide guidance to Member States in the handling, treatment and conditioning of biological radioactive materials. The objective of radioactive waste management is to handle, pretreat, treat, condition, store, transport and dispose of radioactive waste in a manner that protects human health and the environment without imposing undue burdens on future generations. 31 refs, 15 figs, 3 tabs.

  5. Sample handling in surface sensitive chemical and biological sensing: a practical review of basic fluidics and analyte transport.

    Science.gov (United States)

    Orgovan, Norbert; Patko, Daniel; Hos, Csaba; Kurunczi, Sándor; Szabó, Bálint; Ramsden, Jeremy J; Horvath, Robert

    2014-09-01

    This paper gives an overview of the advantages and associated caveats of the most common sample handling methods in surface-sensitive chemical and biological sensing. We summarize the basic theoretical and practical considerations one faces when designing and assembling the fluidic part of the sensor devices. The influence of analyte size, the use of closed and flow-through cuvettes, the importance of flow rate, tubing length and diameter, bubble traps, pressure-driven pumping, cuvette dead volumes, and sample injection systems are all discussed. Typical application areas of particular arrangements are also highlighted, such as the monitoring of cellular adhesion, biomolecule adsorption-desorption and ligand-receptor affinity binding. Our work is a practical review in the sense that for every sample handling arrangement considered we present our own experimental data and critically review our experience with the given arrangement. In the experimental part we focus on sample handling in optical waveguide lightmode spectroscopy (OWLS) measurements, but the present study is equally applicable for other biosensing technologies in which an analyte in solution is captured at a surface and its presence is monitored. Explicit attention is given to features that are expected to play an increasingly decisive role in determining the reliability of (bio)chemical sensing measurements, such as analyte transport to the sensor surface; the distorting influence of dead volumes in the fluidic system; and the appropriate sample handling of cell suspensions (e.g. their quasi-simultaneous deposition). At the appropriate places, biological aspects closely related to fluidics (e.g. cellular mechanotransduction, competitive adsorption, blood flow in veins) are also discussed, particularly with regard to their models used in biosensing. Copyright © 2014 Elsevier B.V. All rights reserved.

  6. Hydrogeological aspects of radioactive waste disposal into surface formations

    International Nuclear Information System (INIS)

    Dlouhy, Z.

    1980-01-01

    Safe disposal is discussed of radioactive wastes in geological surface formations and basic criteria are shown of the radiological protection of population from possible effects of radioactive materials diffused in the environment. Main principles are listed governing the selection of suitable localities with respect to possible interactions of the locality and the storage site with the environment. (author)

  7. A review and analysis of European industrial experience in handling LWR [light water reactor] spent fuel and vitrified high-level waste

    International Nuclear Information System (INIS)

    Blomeke, J.O.

    1988-06-01

    The industrial facilities that have been built or are under construction in France, the United Kingdom, Sweden, and West Germany to handle light-water reactor (LWR) spent fuel and canisters of vitrified high-level waste before ultimate disposal are described and illustrated with drawings and photographs. Published information on the operating performances of these facilities is also given. This information was assembled for consideration in planning and design of similar equipment and facilities needed for the Federal Waste Management System in the United States. 79 refs., 71 figs., 10 tabs

  8. Remote-handled transuranic system assessment appendices. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-11-01

    Volume 2 of this report contains six appendices to the report: Inventory and generation of remote-handled transuranic waste; Remote-handled transuranic waste site storage; Characterization of remote-handled transuranic waste; RH-TRU waste treatment alternatives system analysis; Packaging and transportation study; and Remote-handled transuranic waste disposal alternatives.

  9. Remote-handled transuranic system assessment appendices. Volume 2

    International Nuclear Information System (INIS)

    1995-11-01

    Volume 2 of this report contains six appendices to the report: Inventory and generation of remote-handled transuranic waste; Remote-handled transuranic waste site storage; Characterization of remote-handled transuranic waste; RH-TRU waste treatment alternatives system analysis; Packaging and transportation study; and Remote-handled transuranic waste disposal alternatives

  10. Evaluation of handling and reuse approaches for the waste generated from MEA-based CO2 capture with the consideration of regulations in the UAE.

    Science.gov (United States)

    Nurrokhmah, Laila; Mezher, Toufic; Abu-Zahra, Mohammad R M

    2013-01-01

    A waste slip-stream is generated from the reclaiming process of monoethanolamine (MEA) based Post-Combustion Capture (PCC). It mainly consists of MEA itself, ammonium, heat-stable salts (HSS), carbamate polymers, and water. In this study, the waste quantity and nature are characterized for Fluor's Econamine FGSM coal-fired CO2 capture base case. Waste management options, including reuse, recycling, treatment, and disposal, are investigated due to the need for a more environmentally sound handling. Regulations, economic potential, and associated costs are also evaluated. The technical, economic, and regulation assessment suggests waste reuse for NOx scrubbing. Moreover, a high thermal condition is deemed as an effective technique for waste destruction, leading to considerations of waste recycling into a coal burner or incineration. As a means of treatment, three secondary-biological processes covering Complete-Mix Activated Sludge (CMAS), oxidation ditch, and trickling filter are designed to meet the wastewater standards in the United Arab Emirates (UAE). From the economic point of view, the value of waste as a NOx scrubbing agent is 6,561,600-7,348,992 USD/year. The secondary-biological treatment cost is 0.017-0.02 USD/ton of CO2, while the cost of an on-site incinerator is 0.031 USD/ton of CO2 captured. In conclusion, secondary biological treatment is found to be the most economical option.

  11. Control of Eolian soil erosion from waste site surface barriers

    International Nuclear Information System (INIS)

    Ligotke, M.W.

    1994-11-01

    Physical models were tested in a wind tunnel to determine optimum surface-ravel admixtures for protecting silt-loam soil from erosion by, wind and saltating, sand stresses. The tests were performed to support the development of a natural-material surface barrier for and waste sites. Plans call for a 2-m deep silt-loam soil reservoir to retain infiltrating water from rainfall and snowmelt. The objective of the study was to develop a gravel admixture that would produce an erosion-resistant surface layer during, periods of extended dry climatic stress. Thus, tests were performed using simulated surfaces representing dry, unvegetated conditions present just after construction, after a wildfire, or during an extended drought. Surfaces were prepared using silt-loam soil mixed with various grades of sand and Travel. Wind-induced surface shear stresses were controlled over the test surfaces, as were saltating, sand mass flow rates and intensities. Tests were performed at wind speeds that approximated and exceeded local 100-year peak gust intensities. Surface armors produced by pea gravel admixtures were shown to provide the best protection from wind and saltating sand stresses. Compared with unprotected silt-loam surfaces, armored surfaces reduced erosion rates by more than 96%. Based in part on wind tunnel results, a pea gravel admixture of 15% will be added to the top 1 in of soil in a prototype barrier under construction in 1994. Field tests are planned at the prototype site to provide data for comparison with wind tunnel results

  12. Maximum surface level and temperature histories for Hanford waste tanks

    International Nuclear Information System (INIS)

    Flanagan, B.D.; Ha, N.D.; Huisingh, J.S.

    1994-01-01

    Radioactive defense waste resulting from the chemical processing of spent nuclear fuel has been accumulating at the Hanford Site since 1944. This waste is stored in underground waste-storage tanks. The Hanford Site Tank Farm Facilities Interim Safety Basis (ISB) provides a ready reference to the safety envelope for applicable tank farm facilities and installations. During preparation of the ISB, tank structural integrity concerns were identified as a key element in defining the safety envelope. These concerns, along with several deficiencies in the technical bases associated with the structural integrity issues and the corresponding operational limits/controls specified for conduct of normal tank farm operations are documented in the ISB. Consequently, a plan was initiated to upgrade the safety envelope technical bases by conducting Accelerated Safety Analyses-Phase 1 (ASA-Phase 1) sensitivity studies and additional structural evaluations. The purpose of this report is to facilitate the ASA-Phase 1 studies and future analyses of the single-shell tanks (SSTs) and double-shell tanks (DSTs) by compiling a quantitative summary of some of the past operating conditions the tanks have experienced during their existence. This report documents the available summaries of recorded maximum surface levels and maximum waste temperatures and references other sources for more specific data

  13. Handling sticky Resin by Stingless Bees: Adhesive Properties of Surface Structures

    Directory of Open Access Journals (Sweden)

    MARKUS GASTAUER

    2013-09-01

    Full Text Available Many Stingless Bees (Hymenoptera: Meliponini like Tetragonisca angustula collect resin to defend their nests against intruders like ants or Robber Bees. Small portions of resin are attached to intruders bodies and extremities causing their immobilization. It has been observed that resin is removed easily from the bee's mandible but adheres strongly to the intruder's cuticle. We tested the hypothesis that resin sticks lesser to the mandibles of Stingless Bees than to the surface of intruders due to special surface structures or adhesive properties of these structures. The surface structures of the mandible of T. angustula and the trochanter of Camponotus sericeiventris were studied by scanning electron microscopy. To measure adhesion properties, selected surfaces were fixed on a fine glass pin and withdrawn from a glass tip covered with resin. The deformation of the glass pin indicates adhesion forces operating between the resin and the selective surface. The absolute value of the forces is computed from the glass pin's stiffness. It has been shown that resin sticks more to the smooth mandible of the bee than to the structured trochanter of the ant. A new hypothesis to be tested says that the bees might lubricate their mandibles with nectar or honey to reduce the resin's adhesion temporarily.

  14. Radiological Characterization Methodology for INEEL-Stored Remote-Handled Transuranic (RH TRU) Waste from Argonne National Laboratory-East

    International Nuclear Information System (INIS)

    Kuan, P.; Bhatt, R.N.

    2003-01-01

    An Acceptable Knowledge (AK)-based radiological characterization methodology is being developed for RH TRU waste generated from ANL-E hot cell operations performed on fuel elements irradiated in the EBR-II reactor. The methodology relies on AK for composition of the fresh fuel elements, their irradiation history, and the waste generation and collection processes. Radiological characterization of the waste involves the estimates of the quantities of significant fission products and transuranic isotopes in the waste. Methods based on reactor and physics principles are used to achieve these estimates. Because of the availability of AK and the robustness of the calculation methods, the AK-based characterization methodology offers a superior alternative to traditional waste assay techniques. Using the methodology, it is shown that the radiological parameters of a test batch of ANL-E waste is well within the proposed WIPP Waste Acceptance Criteria limits

  15. Report on how to handle the recommendations in expert reports and statements of views on the demonstration of the nuclear waste disposal programme. Technical report 08-02

    International Nuclear Information System (INIS)

    2008-10-01

    In the process of evaluating the Opalinus Clay project demonstrating the feasibility of disposing of spent fuel, vitrified high-level waste and long-lived intermediate-level waste in Switzerland, the authorities and their experts made numerous recommendations regarding future procedures and activities to be implemented in the disposal programme. Nagra analysed these reviews and expert opinions and took the recommendations into consideration when preparing its future work programme. In 2006, the Federal Council decided that the feasibility of disposing of these waste categories had been demonstrated successfully, but called on the waste producers to prepare a report that systematically addresses the open questions and recommendations and shows how these will be dealt with in a timely and technically appropriate manner. In the present report, Nagra fulfils this requirement on behalf of the waste producers. The report sets out the recommendations made by the authorities and explains how they will be handled. In many cases, the work required has already begun; in other cases plans are already in place. To facilitate the discussion in the report, the recommendations and the responses of Nagra in each case are divided into topical areas. The main part of the report provides a summary discussion, which also addresses time-related aspects of implementing the recommendations. An appendix to the report provides a detailed overview, in the form of a table, of all the recommendations and the associated responses of Nagra. (author)

  16. Ways of solving the problems of radiation safety and environmental protection in handling radioactive waste at atomic power stations in the USSR

    International Nuclear Information System (INIS)

    Gusev, D.I.; Belitskij, A.S.; Turkin, A.D.; Kozlov, V.M.

    1977-01-01

    Requirements of the State Sanitary Supervision on radiation safety of the personnel and population and on protection of the environment in handling radioactive wastes from nuclear power stations in the USSR are regulated by the Standards of Radiation Safety, the Main Sanitary Rules for Handling Radioactive Materials and by the Sanitary Rules for Designing Nuclear Power Stations. The regulations contained in these documents are obligatory for all the establishments at the stages of design, building and operation of nuclear power stations. The main requirement for handling radioactive wastes from nuclear power stations in the USSR is to dispose of them near the place of their production. In nuclear power station siting and designing the special territory is provided for liquid and solid radioactive wastes storage taking into account the whole period of nuclear power station operation. These storage sites are located within the controlled area. They are built as required, usually for five years. The report contains hygienic and hydrological requirements to the radiation waste burial sites and data on the accepted system of controlling leak-proof qualities of the disposal cavities and radioactivity of the ground water in this region. The results of long-term studies on radionuclide leaching from the bituminic blocks are given and it is shown that the bituminizing method used for solidification of intermediate activity wastes is very promising. In the USSR much attention is given to the problem of sanitary protection of the cooling ponds at nuclear power stations. No limits to the national-economic use of these ponds outside the nuclear power station site are established. Therefore in determining the requirements to the discharge of effluents into the cooling ponds of nuclear power stations the possibility of radionuclide transfer to the population through the aquaeous and terrestrial biological chains is taken into account. The possibility of human diet contamination

  17. Handling and treatment of low-level radioactive wastes from gaseous diffusion plants in the United States of America

    International Nuclear Information System (INIS)

    Wing, J.F.; Behrend, J.E.

    1984-01-01

    Gaseous diffusion plants in the United States of America currently generate very small quantities of low-level radioactive wastes. These wastes consist primarily of airborne effluent solid trapping media and liquid scrubber solutions, liquid effluent treatment sludges, waste oils and solvents, scrap metals and conventional combustible wastes such as floor sweepings, cleaning rags and shoe covers. In addition to waste emanating from current operations, large quantities of scrap metal generated during the Cascade Improvement Program are stored above ground at each of the diffusion plants. The radionuclides of primary concern are uranium and 99 Tc. Current radioactive waste treatment consists of uranium dissolution in weak acids followed by chemical precipitation and/or solvent extraction for uranium recovery. Current disposal operations consist of above ground storage of scrap metals, shallow land burial of inorganic solids and incineration of combustible wastes. With increased emphasis on reducing the potential for off-site radiological dose, several new treatment and disposal options are being studied and new projects are being planned. One project of particular interest involves the installation of a high temperature incinerator to thermally degrade hazardous organic wastes contaminated with low-level radioactive wastes. Other technologies being studied include fixation of uranium-bearing sludges in concrete before burial, decontamination of scrap metals by smelting and use of specially engineered centralized burial grounds. (author)

  18. Low-waste electrochemical decontamination of stainless-steel surface

    International Nuclear Information System (INIS)

    Babain, V.A.; Smirnov, I.V.; Shadrin, A.Yu.; Firsin, N.G.; Zakharchuk, G.A.; Pavlov, A.B.; Shilov, V.V.

    2002-01-01

    An electrochemical decontamination method using a formic acid-based recycling electrolyte was proposed to remove firmly fixed contaminants from stainless-steel surfaces. The following provisions make for minimisation of the amounts of waste: (i) use of specially designed electrodes with vacuum removal of spent electrolyte; (ii) inter-cycle removal of radionuclides from the electrolyte by using an inorganic sorbent; (iii) periodic regeneration of the spent electrolyte. the dissolved metals (Fe, Cr, Ni) being transformed into acidic phosphates; (iv) solidification of residues arising from the regeneration of the electrolyte and spent sorbent into iron-phosphate ceramics. The technology and equipment developed were used for decontamination of a plutonium glove-box. The level of surface contamination was reduced 100-fold in two decontamination cycles. The depth of metal skimming was 1.5 μ for the ceiling and walls and 4.5 μ for the table top. Each square meter of stainless-steel surface provides about 100 g of solid radioactive waste in the form of iron-phosphate ceramic blocks

  19. Vertical Drop Of 21-PWR Waste Package On Unyielding Surface

    International Nuclear Information System (INIS)

    S. Mastilovic; A. Scheider; S.M. Bennett

    2001-01-01

    The objective of this calculation is to determine the structural response of a 21-PWR (pressurized-water reactor) Waste Package (WP) subjected to the 2-m vertical drop on an unyielding surface at three different temperatures. The scope of this calculation is limited to reporting the calculation results in terms of stress intensities in two different WP components. The information provided by the sketches (Attachment I) is that of the potential design of the type of WP considered in this calculation, and all obtained results are valid for that design only

  20. Design and testing of a unique active Compton-suppressed LaBr3(Ce) detector system for improved sensitivity assays of TRU in remote-handled TRU wastes

    Energy Technology Data Exchange (ETDEWEB)

    J. K. Hartwell; M. E. McIlwain; J. A. Kulisek

    2007-10-01

    The US Department of Energy’s transuranic (TRU) waste inventory includes about 4,500 m3 of remote-handled TRU (RH-TRU) wastes composed of a variety of containerized waste forms having a contact surface dose rate that exceeds 2 mSv/hr (200 mrem/hr) containing waste materials with a total TRU concentration greater than 3700 Bq/g (100 nCi/g). As part of a research project to investigate the use of active Compton-suppressed room-temperature gamma-ray detectors for direct non-destructive quantification of the TRU content of these RH-TRU wastes, we have designed and purchased a unique detector system using a LaBr3(Ce) primary detector and a NaI(Tl) suppression mantle. The LaBr3(Ce) primary detector is a cylindrical unit ~25 mm in diameter by 76 mm long viewed by a 38 mm diameter photomultiplier. The NaI(Tl) suppression mantle (secondary detector) is 175 mm by 175 mm with a center well that accommodates the primary detector. An important feature of this arrangement is the lack of any “can” between the primary and secondary detectors. These primary and secondary detectors are optically isolated by a thin layer (.003") of aluminized kapton, but the hermetic seal and thus the aluminum can surrounds the outer boundary of the detector system envelope. The hermetic seal at the primary detector PMT is at the PMT wall. This arrangement virtually eliminates the “dead” material between the primary and secondary detectors, a feature that preliminary modeling indicated would substantially improve the Compton suppression capability of this device. This paper presents both the expected performance of this unit determined from modeling with MCNPX, and the performance measured in our laboratory with radioactive sources.

  1. Metabolism of tritium uptake due to handling of metal surfaces exposed to tritiated hydrogen gas

    International Nuclear Information System (INIS)

    Johnson, J.R.; Peterman, B.F.

    1987-08-01

    Hairless rats were exposed to tritium by rubbing HT contaminated stainless steel planchets on them. The pattern of tritium excretion in the urine (n=4), shows the OBT (organically bound tritium) retention curve to be approximated by the sum of 2 exponential curves, one with a half-life of 0.4 days and another with a half-life of 1.4 days. The retention of HTO fit a single exponential curve with a half-life of 3.1 days. Exposed skin, unexposed skin, liver, muscle and blood (n=6) were assayed for HBO, and free HTO. Highest activity was found in the exposed skin, other organs with high activity are the unexposed skin and liver. Examination of the exposed skin showed HTO to be concentrated in the uppermost layers. The distribution of OBT was similar but was incorporated at a faster rate. The basal layer is exposed to a tritium concentration between 70-90% of that of the surface. The two macromolecule fractions with the highest amount of radioactivity were lipid and insoluble protein (mainly collagen)

  2. Development of safety-relevant components for the transport and handling of final storage casks for waste from decommissioning

    International Nuclear Information System (INIS)

    Bruening, D.; Geiser, H.; Kloeckner, F.; Rittscher, D.; Schlesinger, H.J.

    1992-10-01

    The aim of the study was the development, construction and testing of a transportation system that is able to transport cylindrical waste containers as well as containers from the deliverer to the 'KONRAD' final repository. A transport palette has been developed that can carry two cylindrical waste containers with type B requirement or classification II. An Open-All-Container for the transport of palettes and 'KONRAD' containers has been developed. A storage of cylindrical waste containers and containers in the final repository is possible with the newly developed transportation system. Safety specifications of the transportation system have been passed successfully. (orig.). 30 refs., 8 tabs., 74 figs [de

  3. Ecological solution of the problem of handling liquid radioactive wastes - Lr (by the example of Flue SSC RF RIA R)

    International Nuclear Information System (INIS)

    Polyakov, V.I.; Bukvich, B.A.

    2006-01-01

    A sharp reduction of nuclear waste amounts is possible if their elements are considered as source material of atomic complexes - SMAC. The prospect of their possible salvaging will require technological changes and ensuring safety of storage of the material till the need arises. Long experience in deep liquid radioactive waste disposal and accounting, calculations, and motivations demonstrate that a corresponding choice of geological formations makes it possible to abandon liquid radioactive waste solidification and ensure their isolation from environment when the most rigid radiation safety requirements are fulfilled. (author)

  4. From energy resource to riddance problem. The issue of nuclear waste handling in the public dialogue in Sweden, 1950-2002

    International Nuclear Information System (INIS)

    Anshelm, Jonas

    2006-10-01

    Which risks are associated with the handling of high level radioactive wastes? Where should they be kept? Who is responsible for their safe keeping? How should a repository for safe final disposal be designed? Is there, at all, a safe solution for all future time? How could we possibly know that? These questions and many more have been given much attention in the public debate in Sweden, ever since the plans for a Swedish nuclear power program were approved by the parliament in the 1950s. If the questions largely have remained the same, the answers have varied a lot. Representatives for both the nuclear industry and the environmental movement have changed their attitudes and claims for knowing the truth as the technological, political, economical, scientific and cultural circumstances change. This report examines the changes in value base and what was held for truth regarding the plans for a Swedish repository for high-level radioactive waste. E.g. in the 1950s the waste was regarded as an energy resource for the future breeder reactors - in contrast to the conflict-ridden debates of the 1970s when the possibility to manage the waste by any means was questioned. The opposing views on how to select a site for the repository and the diverging opinions on risks, responsibilities, knowledge, technologies, science and nature during the 1980s and 1990s are also analyzed

  5. Derivation of Waste Acceptance Criteria for Low and Intermediate Level Waste in Surface Disposal Facility

    International Nuclear Information System (INIS)

    Gagner, L.; Voinis, S.

    2000-01-01

    In France, low- and intermediate-level radioactive wastes are disposed in a near-surface facility, at Centre de l'Aube disposal facility. This facility, which was commissioned in 1992, has a disposal capacity of one million cubic meters, and will be operated up to about 2050. It took over the job from Centre de la Manche, which was commissioned in 1969 and shut down in 1994, after having received about 520,000 cubic meters of wastes. The Centre de l'Aube disposal facility is designed to receive a many types of waste produced by nuclear power plants, reprocessing, decommissioning, as well as by the industry, hospitals and armed forces. The limitation of radioactive transfer to man and the limitation of personnel exposure in all situations considered plausible require limiting the total activity of the waste disposed in the facility as well as the activity of each package. The paper presents how ANDRA has derived the activity-related acceptance criteria, based on the safety analysis. In the French methodology, activity is considered as end-point for deriving the concentration limits per package, whereas it is the starting point for deriving the total activity limits. For the concentration limits (called here LMA) the approach consists of five steps: the determination of radionuclides important for safety with regards to operational and long-term safety, the use of relevant safety scenarios as a tool to derive quantitative limits, the setting of dose constraint per situation associated with scenarios, the setting of contribution factor per radionuclide, and the calculation of concentration activity limits. An exhaustive survey has been performed and has shown that the totality of waste packages which should be delivered by waste generators are acceptable in terms of activity limits in the Centre de l'Aube. Examples of concentration activity limits derived from this methodology are presented. Furthermore those limits have been accepted by the French regulatory body and

  6. Report on the handling of safety information concerning flammable gases and ferrocyanide at the Hanford waste tanks

    International Nuclear Information System (INIS)

    1990-07-01

    This report discusses concerns safety issues, and management at Hanford Tank Farm. Concerns center on the issue of flammable gas generation which could ignite, and on possible exothermic reactions of ferrocyanide compounds which were added to single shell tanks in the 1950's. It is believed that information concerning these issues has been mis-handled and the problems poorly managed

  7. Report on the handling of safety information concerning flammable gases and ferrocyanide at the Hanford waste tanks

    Energy Technology Data Exchange (ETDEWEB)

    1990-07-01

    This report discusses concerns safety issues, and management at Hanford Tank Farm. Concerns center on the issue of flammable gas generation which could ignite, and on possible exothermic reactions of ferrocyanide compounds which were added to single shell tanks in the 1950's. It is believed that information concerning these issues has been mis-handled and the problems poorly managed. (CBS)

  8. Quantities and characteristics of the contact-handled low-level mixed waste streams for the DOE complex

    International Nuclear Information System (INIS)

    Huebner, T.L.; Wilson, J.M.; Ruhter, A.H.; Bonney, S.J.

    1994-08-01

    This report supports the Integrated Thermal Treatment System (ITTS) Study initiated by the Department of Energy (DOE) Office of Technology Development (EM-50), which is a system engineering assessment of a variety of mixed waste treatment process. The DOE generates and stores large quantities of mixed wastes that are contaminated with both chemically hazardous and radioactive species. The treatment of these mixed wastes requires meeting the standards established by the Environmental Protection Agency for the specific hazardous contaminants regulated under the Resource Conservation and Recovery Act while also providing adequate control of the radionuclides. The thrust of the study is to develop preconceptual designs and life-cycle cost estimates for integrated thermal treatment systems ranging from conventional incinerators, such as rotary kiln and controlled air systems, to more innovative but not yet established technologies, such as molten salt and molten metal waste destruction systems. Prior to this engineering activity, the physical and chemical characteristics of the DOE low-level mixed waste streams to be treated must be defined or estimated. This report describes efforts to estimate the DOE waste stream characteristics

  9. Cocaine and metabolites in waste and surface water across Belgium

    Energy Technology Data Exchange (ETDEWEB)

    Nuijs, Alexander L.N. van [Toxicological Centre, Department of Pharmaceutical Sciences, University of Antwerp (Ukraine), Universiteitsplein 1, 2610 Antwerp (Belgium)], E-mail: alexander.vannuijs@ua.ac.be; Pecceu, Bert [Laboratory for Ecophysiology, Biochemistry and Toxicology, Department of Biology, University of Antwerp (Ukraine), Groenenborgerlaan 171, 2020 Antwerp (Belgium); Theunis, Laetitia; Dubois, Nathalie; Charlier, Corinne [Laboratory of Clinical, Forensic and Environmental Toxicology, University of Liege, (ULg), CHU Sart-Tilman, 4000 Liege (Belgium); Jorens, Philippe G. [Department of Clinical Pharmacology/Clinical Toxicology, University of Antwerp (Ukraine), University Hospital of Antwerp, Universiteitsplein 1, 2610 Antwerp (Belgium); Bervoets, Lieven; Blust, Ronny [Laboratory for Ecophysiology, Biochemistry and Toxicology, Department of Biology, University of Antwerp (Ukraine), Groenenborgerlaan 171, 2020 Antwerp (Belgium); Neels, Hugo [Toxicological Centre, Department of Pharmaceutical Sciences, University of Antwerp (Ukraine), Universiteitsplein 1, 2610 Antwerp (Belgium); Laboratory of Toxicology, ZNA Stuivenberg, Lange Beeldekensstraat 267, 2060 Antwerp (Belgium); Covaci, Adrian [Toxicological Centre, Department of Pharmaceutical Sciences, University of Antwerp (Ukraine), Universiteitsplein 1, 2610 Antwerp (Belgium); Laboratory for Ecophysiology, Biochemistry and Toxicology, Department of Biology, University of Antwerp (Ukraine), Groenenborgerlaan 171, 2020 Antwerp (Belgium)

    2009-01-15

    Cocaine abuse, a growing social problem, is currently estimated from population surveys, consumer interviews and crime statistics. A new approach based on the analysis of cocaine (COC) and metabolites, benzoylecgonine (BE) and ecgonine methyl ester (EME), in water samples was applied to 28 rivers and 37 waste water treatment plants in Belgium using solid-phase extraction and liquid chromatography coupled to tandem mass spectrometry. While EME was undetectable, COC and BE were detectable with concentrations ranging from <1 to 753 ng/L and <1 to 2258 ng/L, respectively. BE concentrations were employed to calculate the local amount of abused cocaine. The highest values (up to 1.8 g/day cocaine per 1000 inhabitants) were found in large cities and during weekends. The estimation of cocaine abuse through water analysis can be executed on regular basis without cooperation of patients. It also gives clear geographical information, while prevention campaigns can easily be implemented and evaluated. - Cocaine consumption can be evaluated through analysis of waste and surface water.

  10. Cocaine and metabolites in waste and surface water across Belgium

    International Nuclear Information System (INIS)

    Nuijs, Alexander L.N. van; Pecceu, Bert; Theunis, Laetitia; Dubois, Nathalie; Charlier, Corinne; Jorens, Philippe G.; Bervoets, Lieven; Blust, Ronny; Neels, Hugo; Covaci, Adrian

    2009-01-01

    Cocaine abuse, a growing social problem, is currently estimated from population surveys, consumer interviews and crime statistics. A new approach based on the analysis of cocaine (COC) and metabolites, benzoylecgonine (BE) and ecgonine methyl ester (EME), in water samples was applied to 28 rivers and 37 waste water treatment plants in Belgium using solid-phase extraction and liquid chromatography coupled to tandem mass spectrometry. While EME was undetectable, COC and BE were detectable with concentrations ranging from <1 to 753 ng/L and <1 to 2258 ng/L, respectively. BE concentrations were employed to calculate the local amount of abused cocaine. The highest values (up to 1.8 g/day cocaine per 1000 inhabitants) were found in large cities and during weekends. The estimation of cocaine abuse through water analysis can be executed on regular basis without cooperation of patients. It also gives clear geographical information, while prevention campaigns can easily be implemented and evaluated. - Cocaine consumption can be evaluated through analysis of waste and surface water

  11. Enlargement of the Baldone near-surface radioactive waste repository

    International Nuclear Information System (INIS)

    Dreimanis, A.

    2007-01-01

    A unified analysis of the enlargement of the Baldone near-surface radioactive waste (RW) repository RADONS considers the interplay of the existing engineering, safety and infrastructure premises, with the foreseen newly socio-technical features. This enlargement consists in construction of two additional RW disposal vaults and in building a long-term storage facility for spent sealed sources at the RADONS territory. Our approach is based on consecutive analysis of following basic elements: - the origin of enlargement - the RADONS safety analysis and a set of optimal socio-technical solutions of Salaspils research reactor decommissioning waste management; - the enlargement - a keystone of the national RW management concept, including the long-term approach; - the enlargement concept - the result of international co-operation and obligations; - arrangement optimization of new disposal and storage space; - environmental impact assessment for the repository enlargement - the update of socio-technical studies. The study of the public opinion revealed: negative attitude to repository enlargement is caused mainly due to missing information on radiation level and on the RADONS previous operations. These results indicate: basic measures to improve the public attitude to repository enlargement: the safety upgrade, public education and compensation mechanisms. A detailed stakeholders engagement and public education plan is elaborated. (author)

  12. Surface water management at a mixed waste remediation site

    International Nuclear Information System (INIS)

    Schlotzhauer, D.S.; Warbritton, K.R.

    1991-01-01

    The Weldon Spring Remedial Action Project (WSSRAP) deals with chemical and radiological contaminants. MK-Ferguson Company is managing the project under contract with the US Department of Energy. Remedial activities include demolishing buildings, constructing material storage and staging areas, excavating and consolidating waste materials, and treating and disposing of the materials in a land disposal facility. Due to the excavation and construction required during remediation, a well-planned surface water management system is essential. Planning involves characterization of source areas and surface water transport mechanisms and identification of applicable regulations. System components include: erosion control sediment control, flow attenuation, and management of contaminated water. Combinations of these components may be utilized during actual construction and remediation to obtain optimum control. Monitoring is performed during implementation in order to assess the effectiveness of control measures. This management scheme provides for comprehensive management of surface water at this site by providing control and/or treatment to appropriate standards. Although some treatment methodologies for contaminated water are specific to site contaminants, this comprehensive program provides a management approach which is applicable to many remedial projects in order to minimize contaminant release and meet Clean Water Act requirements

  13. Near-surface groundwater responses to injection of geothermal wastes

    Energy Technology Data Exchange (ETDEWEB)

    Arnold, S.C.

    1984-06-01

    Experiences with injecting geothermal fluids have identified technical problems associated with geothermal waste disposal. This report assesses the feasibility of injection as an alternative for geothermal wastewater disposal and analyzes hydrologic controls governing the upward migration of injected fluids. Injection experiences at several geothermal developments are presented, including: Raft River, Salton Sea, East Mesa, Otake and Hatchobaru in Japan, and Ahuachapan in El Salvador. Hydrogeologic and design/operational factors affecting the success of an injection program are identified. Hydrogeologic factors include subsidence, near-surface effects of injected fluids, and seismicity. Design/operational factors include hydrodynamic breakthrough, condition of the injection system and reservoir maintenance. Existing and potential effects of production/injection on these factors are assessed.

  14. Preliminary radiological analysis of the transportation of contact-handled transuranic waste within the state of New Mexico. Revision 1

    International Nuclear Information System (INIS)

    Tappen, J.; Fredrickson, C.; Daer, G.

    1985-06-01

    This analysis assesses the potential radiological impacts on the citizens of New Mexico from the transport of CH-TRU waste to WIPP by rail or by truck. Assuming exclusive use of the truck transport mode, the combined annual exposure to the public from accident-free shipment of waste is estimated to be 3.3 person-rem/year. It is estimated that a theoretical member of the public receiving maximum exposure to the combined truck shipments of CH-TRU waste to WIPP would receive an annual whole body dose equivalent of 0.000016 rem. Such an exposure is insignificant in comparison to the average annual whole body dose equivalent to an individual living in the Colorado Plateau area of between 0.075 and 0.140 rem from natural occurring radiation. The combined annual radiological risk to the public living along the new Mexico truck routes to WIPP from potential accidents is projected as 0.031 person-rem/year. Assuming exclusive use of the rail transport mode, the combined annual exposure to the public from accident-free shipment of waste is estimated to be 1.2 person-rem/year. A theoretical member of the public receiving combined maximum exposure to rail shipments of CH-TRU waste to WIPP would receive an annual whole body dose equivalent of 0.000012 rem. The combined annual radiological risk to the public living along the New Mexico rail routes to WIPP from potential accidents is projected as 0.0022 person-rem/year. An estimate of the radiological impacts in a year of maximum waste receipt can be made by multiplying the above results for rail or truck by 2. This estimate is based upon the WIPP design waste throughput rate of 500,000 ft 3 per year. An estimate of the radiological impacts of CH-TRU waste transport to WIPP over the facility life can be made by multiplying the above results by 25

  15. Viscosity of ashes from energy production and municipal solid waste handling: A comparative study between two different experimental setups

    DEFF Research Database (Denmark)

    Arvelakis, Stelios; Frandsen, Flemming; Folkedahl, B.

    2008-01-01

    This paper discusses the viscosity characteristics of ash fractions produced from the co-combustion of coal and biomass in a pilot-scale pulverized fuel (PF) boiler and from the incineration of municipal solid waste (MSW) in a Danish incinerator that were determined using the high...

  16. Waste Volume Reduction Using Surface Characterization and Decontamination By Laser Ablation

    International Nuclear Information System (INIS)

    Pellin, Michael J.; Savina, Michael R.; Reed, Claude B.; Zhiyue, Xu; Yong, Wang

    2000-01-01

    Cs, in and out of model waste forms and engineered barriers[9-21]. Few studies exist on the chemical speciation of the contaminants themselves in concrete [22-25]. For example, the extent to which various contaminants react with the cement and various aggregate particles is currently unknown, as is the role of the high pH of the cement pore water on ion partitioning and chemical speciation. DOE has designated understanding the chemical nature of the contaminants as important in the rational design of characterization, decontamination, and waste handling strategies[26, 27]. We have investigated laser ablation as a means of concrete surface removal[28-31]. Lasers are attractive since the power can be delivered remotely via articulated mirrors or fiber optic cables and the ablation head can be manipulated by robots, thus avoiding exposing workers and the laser system to the radiation field. In addition, lasers can be instrumented with spectrometers or effluent sampling devices to provide for on-line analysis. In contrast to mechanical scabbling systems, laser beams can penetrate cracks or follow very rough or irregularly shaped surfaces. Finally, a laser ablation system produces the smallest possible waste stream since no cleaning agents such as detergents or grit (from grit blasting systems) are mixed with the effluent

  17. A Two-Phase Cooling Loop for Fission Surface Power Waste Heat Transport, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Current lunar-based Fission Surface Power (FSP) Systems that will support sustained surface outposts consist of a nuclear reactor with power converters, whose waste...

  18. Optimizing supercritical carbon dioxide in the inactivation of bacteria in clinical solid waste by using response surface methodology

    Energy Technology Data Exchange (ETDEWEB)

    Hossain, Md. Sohrab [Department of Environmental Technology, School of Industrial Technology, Universiti Sains Malaysia, 11800 Penang (Malaysia); Nik Ab Rahman, Nik Norulaini [School of Distance Education, Universiti Sains Malaysia, 11800 Penang (Malaysia); Balakrishnan, Venugopal [Institute for Research in Molecular Medicine, Universiti Sains Malaysia, 11800 Penang (Malaysia); Alkarkhi, Abbas F.M. [Department of Environmental Technology, School of Industrial Technology, Universiti Sains Malaysia, 11800 Penang (Malaysia); Ahmad Rajion, Zainul [School of Dental Science, Universiti Sains Malaysia, 16150 Kubang Kerian, Kelantan (Malaysia); Ab Kadir, Mohd Omar, E-mail: akmomar@usm.my [Department of Environmental Technology, School of Industrial Technology, Universiti Sains Malaysia, 11800 Penang (Malaysia)

    2015-04-15

    Highlights: • Supercritical carbon dioxide sterilization of clinical solid waste. • Inactivation of bacteria in clinical solid waste using supercritical carbon dioxide. • Reduction of the hazardous exposure of clinical solid waste. • Optimization of the supercritical carbon dioxide experimental conditions. - Abstract: Clinical solid waste (CSW) poses a challenge to health care facilities because of the presence of pathogenic microorganisms, leading to concerns in the effective sterilization of the CSW for safe handling and elimination of infectious disease transmission. In the present study, supercritical carbon dioxide (SC-CO{sub 2}) was applied to inactivate gram-positive Staphylococcus aureus, Enterococcus faecalis, Bacillus subtilis, and gram-negative Escherichia coli in CSW. The effects of SC-CO{sub 2} sterilization parameters such as pressure, temperature, and time were investigated and optimized by response surface methodology (RSM). Results showed that the data were adequately fitted into the second-order polynomial model. The linear quadratic terms and interaction between pressure and temperature had significant effects on the inactivation of S. aureus, E. coli, E. faecalis, and B. subtilis in CSW. Optimum conditions for the complete inactivation of bacteria within the experimental range of the studied variables were 20 MPa, 60 °C, and 60 min. The SC-CO{sub 2}-treated bacterial cells, observed under a scanning electron microscope, showed morphological changes, including cell breakage and dislodged cell walls, which could have caused the inactivation. This espouses the inference that SC-CO{sub 2} exerts strong inactivating effects on the bacteria present in CSW, and has the potential to be used in CSW management for the safe handling and recycling-reuse of CSW materials.

  19. Optimizing supercritical carbon dioxide in the inactivation of bacteria in clinical solid waste by using response surface methodology

    International Nuclear Information System (INIS)

    Hossain, Md. Sohrab; Nik Ab Rahman, Nik Norulaini; Balakrishnan, Venugopal; Alkarkhi, Abbas F.M.; Ahmad Rajion, Zainul; Ab Kadir, Mohd Omar

    2015-01-01

    Highlights: • Supercritical carbon dioxide sterilization of clinical solid waste. • Inactivation of bacteria in clinical solid waste using supercritical carbon dioxide. • Reduction of the hazardous exposure of clinical solid waste. • Optimization of the supercritical carbon dioxide experimental conditions. - Abstract: Clinical solid waste (CSW) poses a challenge to health care facilities because of the presence of pathogenic microorganisms, leading to concerns in the effective sterilization of the CSW for safe handling and elimination of infectious disease transmission. In the present study, supercritical carbon dioxide (SC-CO 2 ) was applied to inactivate gram-positive Staphylococcus aureus, Enterococcus faecalis, Bacillus subtilis, and gram-negative Escherichia coli in CSW. The effects of SC-CO 2 sterilization parameters such as pressure, temperature, and time were investigated and optimized by response surface methodology (RSM). Results showed that the data were adequately fitted into the second-order polynomial model. The linear quadratic terms and interaction between pressure and temperature had significant effects on the inactivation of S. aureus, E. coli, E. faecalis, and B. subtilis in CSW. Optimum conditions for the complete inactivation of bacteria within the experimental range of the studied variables were 20 MPa, 60 °C, and 60 min. The SC-CO 2 -treated bacterial cells, observed under a scanning electron microscope, showed morphological changes, including cell breakage and dislodged cell walls, which could have caused the inactivation. This espouses the inference that SC-CO 2 exerts strong inactivating effects on the bacteria present in CSW, and has the potential to be used in CSW management for the safe handling and recycling-reuse of CSW materials

  20. Engineering considerations for the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Scully, L.W.

    1978-01-01

    The WIPP, located at Los Medanos in New Mexico, is to be used for DOE transuranic and high-level defense wastes. On the surface, there are contact-handled and remote-handled waste facilities. Package size, delivery rates, shipping, shielding and thermal considerations, underground transport and emplacement, retrievability, ventilation, and hoist conveyence safety are discussed

  1. 40 CFR Appendix Vii to Part 268 - LDR Effective Dates of Surface Disposed Prohibited Hazardous Wastes

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 26 2010-07-01 2010-07-01 false LDR Effective Dates of Surface Disposed Prohibited Hazardous Wastes VII Appendix VII to Part 268 Protection of Environment ENVIRONMENTAL... VII to Part 268—LDR Effective Dates of Surface Disposed Prohibited Hazardous Wastes Table 1—Effective...

  2. Summary of Conceptual Models and Data Needs to Support the INL Remote-Handled Low-Level Waste Disposal Facility Performance Assessment and Composite Analysis

    International Nuclear Information System (INIS)

    Sondrup, A. Jeff; Schafter, Annette L.; Rood, Arthur S.

    2010-01-01

    An overview of the technical approach and data required to support development of the performance assessment, and composite analysis are presented for the remote handled low-level waste disposal facility on-site alternative being considered at Idaho National Laboratory. Previous analyses and available data that meet requirements are identified and discussed. Outstanding data and analysis needs are also identified and summarized. The on-site disposal facility is being evaluated in anticipation of the closure of the Radioactive Waste Management Complex at the INL. An assessment of facility performance and of the composite performance are required to meet the Department of Energy's Low-Level Waste requirements (DOE Order 435.1, 2001) which stipulate that operation and closure of the disposal facility will be managed in a manner that is protective of worker and public health and safety, and the environment. The corresponding established procedures to ensure these protections are contained in DOE Manual 435.1-1, Radioactive Waste Management Manual (DOE M 435.1-1 2001). Requirements include assessment of (1) all-exposure pathways, (2) air pathway, (3) radon, and (4) groundwater pathway doses. Doses are computed from radionuclide concentrations in the environment. The performance assessment and composite analysis are being prepared to assess compliance with performance objectives and to establish limits on concentrations and inventories of radionuclides at the facility and to support specification of design, construction, operation and closure requirements. Technical objectives of the PA and CA are primarily accomplished through the development of an establish inventory, and through the use of predictive environmental transport models implementing an overarching conceptual framework. This document reviews the conceptual model, inherent assumptions, and data required to implement the conceptual model in a numerical framework. Available site-specific data and data sources

  3. Handling wood shavings

    Energy Technology Data Exchange (ETDEWEB)

    1974-09-18

    Details of bulk handling equipment suitable for collection and compressing wood waste from commercial joinery works are discussed. The Redler Bin Discharger ensures free flow of chips from storage silo discharge prior to compression into briquettes for use as fuel or processing into chipboard.

  4. Research, development and optimization of real time radioscopic characterization of remote handled waste and intermediate level waste, using X-ray imaging at MeV energies

    International Nuclear Information System (INIS)

    Halliwell, Stephen

    2007-01-01

    Available in abstract form only. Full text of publication follows: Real time radioscopy (RTR) using X-ray energies of up to 450 keV, is used extensively in the characterization of nuclear waste. The majority of LLW and some ILW in drums and boxes can be penetrated, for successful imaging, by X-rays with energies of up to 450 keV. However, the shielding of many waste packages, and the range of higher density waste matrices, require X-rays at MeV energies, for X-ray imaging to achieve the performance criteria. A broad imaging performance is required to enable the identification of a range of prohibited items, including the ability to see a moving liquid meniscus which indicates the presence of free liquid, in a high density or a waste matrix with substantial containment shielding. Enhanced, high energy X-ray imaging technology to meet the future characterization demands of the nuclear industry required the design and build of a high energy facility, and the implementation of a program of research and development. The initial phase of development has confirmed that digital images meeting the required performance criteria can be made using high energy X-rays. The evaluation of real time imaging and the optimization of imaging with high energy X-rays is currently in progress. (author)

  5. Methodology in the handling of the waste radioactive material; Metodologia en el manejo del material radiactivo de desecho

    Energy Technology Data Exchange (ETDEWEB)

    Emeterio H, M., E-mail: miguel.emeterio@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    The methodology in the management of radioactive waste is constituted by an administrative part and seven technical stages: transport, classification, segregation, conditioning, treatment, packages qualification and final disposition (storage). In their diverse stages the management deserves a special attention, due to the increment of the use and application of the nuclear energy and radioactive substances, for such a reason should be managed in such a way that the exposed personnel safety and the public in general is guaranteed, protecting the integrity of the environment. (Author)

  6. Method and equipment to prepare aqueous radioactive waste for non-polluting and safe handling, transport, and final storage

    Energy Technology Data Exchange (ETDEWEB)

    Drobmik, S; Hild, W; Kaufmann, F; Koschorke, H

    1977-09-15

    A system is proposed which allows the joint execution of the process steps of denitration, drying and calcination necessary for the treatment of radioactive waste and the subsequent injection of superheated water vapor into the reaction chamber so as to also clean the offgases produced and discharge them from the facility through a filter system without polluting the environment. Several design variants are described which mainly relate to the arrangement of the spray nozzle and its environment. The vitrified radioactive materials are then put into repositories.

  7. Quarter-scale modeling of room convergence effects on CH [contact-handled] TRU drum waste emplacements using WIPP [Waste Isolation Pilot Plant] reference design geometries

    International Nuclear Information System (INIS)

    VandeKraats, J.

    1987-11-01

    This study investigates the effect of horizontal room convergence on CH waste packages emplaced in the WIPP Reference Design geometry (rooms 13 feet high by 33 feet wide, with minus 3/8 inch screened backfill emplaced over and around the waste packages) as a function of time. Based on two tests, predictions were made with regard to full-scale 6-packs emplaced in the Reference Design geometry. These are that load will be transmitted completely through the stack within the first five years after waste emplacement and all drums in all 6-packs will be affected; that virtually all drums will show some deformation eight years after emplacement; that some drums may breach before the eighth year after emplacement has elapsed; and that based on criteria developed during testing, it is predicted that 1% of the drums emplaced will be breached after 8 years and, after 15 years, approximately 12% of the drums are predicted to be breached. 8 refs., 41 figs., 3 tabs

  8. Development of the scientific concept of the phosphate methods for actinide-containing waste handling (pyrochemical fuel reprocessing)

    International Nuclear Information System (INIS)

    Orlova, A.I.; Orlova, V.A.; Skiba, O.V.; Bychkov, A.V.; Volkov, Yu.F.; Lukinykh, A.N.; Tomilin, S.V.; Lizin, A.A.

    2008-01-01

    Full text of publication follows: The crystallochemical phosphate concept in question is developed successfully in the new pyro-electrochemical reprocessing technology of irradiated fuel in molten chlorides of alkaline elements at one of the leading scientific nuclear centers - Research Institute of Atomic Reactors. Irradiated fuel is dissolved in molten chlorides of alkaline elements by mean of treating by chlorine. Then uranium and plutonium dioxides are removed electrochemically. The melt, when used many times, is contaminated by the residual actinide and contains fission products and the so called 'process' elements. This melt is unacceptable for future use. Phosphate methods can be applied for the solution of the following tasks: a) reprocessing (purification) of molten chloride salt solvents; b) conversion of the spent chloride melts to the insoluble stable crystalline product for safe storage and disposal. Within the framework of task 'a' phosphate methods may be realized by the several ways: 1) phosphate concentrating of impurities and their extraction from molten chlorides into solid phase by mean of chemical precipitation, co-precipitation, ion exchange and other chemical interactions, 2) conversion of precipitated waste phosphates to stable crystalline phosphate powders or ceramics for safe storage and disposal. (authors)

  9. KS 20322007 Near-Surface Disposal Radioactive Waste - Code Of Practice

    International Nuclear Information System (INIS)

    Omondi, C.

    2017-01-01

    To provide a basis for the near-surface disposal of solid radioactive waste to ensures that there is no unacceptable risk to humans, other biota or the environment. Near-Surface Disposal is the disposal of radioactive waste in below or above the natural ground surface, within app. 30 m. The code deals with management aspects associated with radioactive waste disposal only, and is not intended to cover issues related to the production and use of radionuclides. The objective of waste disposal is to isolate radioactive waste in order to ensure that there is no unacceptable health risk to humans and no long-term unacceptable effect to the environment. Radiation protection annual effective dose for exposure of members of the public should not exceed 1 mSv/year and occupational exposure of 20 mSv/year

  10. Design surface covers: an approach to long-term waste site stabilization

    International Nuclear Information System (INIS)

    Beedlow, P.A.; Cadwell, L.L.; McShane, M.C.

    1983-02-01

    The wide range of existing environmental conditions, potential contaminants and available cover materials at waste disposal sites necessitates site-specific designing of surface covers for effective long-term erosion resistance. This paper presents a systematic approach to designing surface covers for hazardous waste repositories that can be tailored to conditions at any site. The approach consists of three phases: (1) an assessment, during which the degree of required surface protection (erosion potential) is determined; (2) a preliminary design that integrates surface cover design with the need to minimize transport of contaminants; and (3) a final design, where the cost and effectiveness of the surface cover are determined. 1 figure

  11. Research on near-surface disposal of very low level radioactive waste

    International Nuclear Information System (INIS)

    Wang Shaowei; Yue Huiguo; Hou Jie; Chen Haiying; Zuo Rui; Wang Jinsheng

    2012-01-01

    Radioactive waste disposal is one of the most sensitive environmental problems to control and solve. As the arriving of decommissioning of early period nuclear facilities in China, large amounts of very low level radioactive waste will be produced inevitably. The domestic and abroad definitions about very low level radioactive waste and its disposal were introduced, and then siting principles of near-surface disposal of very low level radioactive waste were discussed. The near- surface disposal siting methods of very low level radioactive waste were analyzed from natural and geographical conditions assessment, geological conditions analysis, hydrogeological conditions analysis, geological hazard assessment and radioactive background investigation; the near-surface disposal sites'natural barriers of very low level radioactive waste were analyzed from the crustal structure and physico-chemical characteristics, the dynamics characteristics of groundwater, the radionuclide adsorption characteristics of natural barriers and so on; the near-surface disposal sites' engineered barriers of very low level radioactive waste were analyzed from the repository design, the repository barrier materials selection and so on. Finally, the improving direction of very low level radioactive waste disposal was proposed. (authors)

  12. Handling and Emplacement Options for Deep Borehole Disposal Conceptual Design.

    Energy Technology Data Exchange (ETDEWEB)

    Cochran, John R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hardin, Ernest [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-07-01

    This report presents conceptual design information for a system to handle and emplace packages containing radioactive waste, in boreholes 16,400 ft deep or possibly deeper. Its intended use is for a design selection study that compares the costs and risks associated with two emplacement methods: drill-string and wireline emplacement. The deep borehole disposal (DBD) concept calls for siting a borehole (or array of boreholes) that penetrate crystalline basement rock to a depth below surface of about 16,400 ft (5 km). Waste packages would be emplaced in the lower 6,560 ft (2 km) of the borehole, with sealing of appropriate portions of the upper 9,840 ft (3 km). A deep borehole field test (DBFT) is planned to test and refine the DBD concept. The DBFT is a scientific and engineering experiment, conducted at full-scale, in-situ, without radioactive waste. Waste handling operations are conceptualized to begin with the onsite receipt of a purpose-built Type B shipping cask, that contains a waste package. Emplacement operations begin when the cask is upended over the borehole, locked to a receiving flange or collar. The scope of emplacement includes activities to lower waste packages to total depth, and to retrieve them back to the surface when necessary for any reason. This report describes three concepts for the handling and emplacement of the waste packages: 1) a concept proposed by Woodward-Clyde Consultants in 1983; 2) an updated version of the 1983 concept developed for the DBFT; and 3) a new concept in which individual waste packages would be lowered to depth using a wireline. The systems described here could be adapted to different waste forms, but for design of waste packaging, handling, and emplacement systems the reference waste forms are DOE-owned high- level waste including Cs/Sr capsules and bulk granular HLW from fuel processing. Handling and Emplacement Options for Deep Borehole Disposal Conceptual Design July 23, 2015 iv ACKNOWLEDGEMENTS This report has

  13. Alpha waste management at the Valduc Research Center

    International Nuclear Information System (INIS)

    Jouan, A.; Cartier, R.; Durec, J.P.; Flament, T.

    1995-01-01

    Operation of the reprocessing facilities at the Valduc Research Center of the French Atomic Energy Commission (CEA) generates waste with a variety of characteristics. The waste compatible with surface storage requirements is transferred to the French Radioactive Waste Management Agency (ANDRA); rest is reprocessed under a program which enables storage in compliance with the requirements of permits issued by safety Authorities. The waste reprocessing program provides for the construction of an incinerator capable of handling nearly all of the combustible waste generated by the Center and vitrification facility for treating liquid waste generated by the plutonium handling plant. (authors)

  14. 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

  15. National Enforcement Initiative: Preventing Animal Waste from Contaminating Surface and Ground Water

    Science.gov (United States)

    This page describes EPA's goal in preventing animal waste from contaminating surface and ground Water. It is an EPA National Enforcement Initiative. Both enforcement cases, and a map of enforcement actions are provided.

  16. Remote handling machines

    International Nuclear Information System (INIS)

    Sato, Shinri

    1985-01-01

    In nuclear power facilities, the management of radioactive wastes is made with its technology plus the automatic techniques. Under the radiation field, the maintenance or aid of such systems is important. To cope with this situation, MF-2 system, MF-3 system and a manipulator system as remote handling machines are described. MF-2 system consists of an MF-2 carrier truck, a control unit and a command trailer. It is capable of handling heavy-weight objects. The system is not by hydraulic but by electrical means. MF-3 system consists of a four-crawler truck and a manipulator. The truck is versatile in its posture by means of the four independent crawlers. The manipulator system is bilateral in operation, so that the delicate handling is made possible. (Mori, K.)

  17. Pre-treatment of bituminized NPP wastes for disposal in near-surface repository

    Energy Technology Data Exchange (ETDEWEB)

    Vieira, Vanessa Mota; Tello, Clédola Cássia Oliveira de, E-mail: vanessamotavieira@gmail.com, E-mail: tellocc@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2017-07-01

    The implementation of the national repository is an important technical requirement, and a legal requirement for the entry into operation of the nuclear power plant Angra 3. The Brazilian repository is being planned to be a near-surface one. In Brazil the low and intermediate level radioactive wastes are immobilized using cement and bitumen for Angra 1 and Angra 2 NPP, respectively. The main problems due to the disposal of bituminized wastes in repositories are swelling of the waste products and their degradation in the long term. To accommodate the swelling of the bituminized wastes, the drums are filled up to 70 - 90% of their volume, which reduces the structural the repository stability and the disposal availability. Countries, which use bitumen in the solidification of NPP's radioactive waste and have near-surface repositories, need to immobilize this bituminized waste within other drums containing cement pastes or mortars to disposal them. This study aims to find solutions for the storage in surface repository of bituminized radioactive waste products, making them compatible with the acceptance criteria of cemented waste products. It was also performed a modeling with the results obtained in the leaching test using the ALT program and defined the transport model of the cesium leachate element and it was verified that in the early times the leaching was governed by the diffusion model and later by the partition model. The results obtained in this study can be used in the evaluation of performance of repositories. (author)

  18. Recent advances in yeast cell-surface display technologies for waste biorefineries.

    Science.gov (United States)

    Liu, Zhuo; Ho, Shih-Hsin; Hasunuma, Tomohisa; Chang, Jo-Shu; Ren, Nan-Qi; Kondo, Akihiko

    2016-09-01

    Waste biorefinery aims to maximize the output of value-added products from various artificial/agricultural wastes by using integrated bioprocesses. To make waste biorefinery economically feasible, it is thus necessary to develop a low-cost, environment-friendly technique to perform simultaneous biodegradation and bioconversion of waste materials. Cell-surface display engineering is a novel, cost-effective technique that can auto-immobilize proteins on the cell exterior of microorganisms, and has been applied for use with waste biofinery. Through tethering different enzymes (e.g., cellulase, lipase, and protease) or metal-binding peptides on cell surfaces, various yeast strains can effectively produce biofuels and biochemicals from sugar/protein-rich waste materials, catalyze waste oils into biodiesels, or retrieve heavy metals from wastewater. This review critically summarizes recent applications of yeast cell-surface display on various types of waste biorefineries, highlighting its potential and future challenges with regard to commercializing this technology. Copyright © 2016 Elsevier Ltd. All rights reserved.

  19. Pre-treatment of bituminized NPP wastes for disposal in near-surface repository

    International Nuclear Information System (INIS)

    Vieira, Vanessa Mota; Tello, Clédola Cássia Oliveira de

    2017-01-01

    The implementation of the national repository is an important technical requirement, and a legal requirement for the entry into operation of the nuclear power plant Angra 3. The Brazilian repository is being planned to be a near-surface one. In Brazil the low and intermediate level radioactive wastes are immobilized using cement and bitumen for Angra 1 and Angra 2 NPP, respectively. The main problems due to the disposal of bituminized wastes in repositories are swelling of the waste products and their degradation in the long term. To accommodate the swelling of the bituminized wastes, the drums are filled up to 70 - 90% of their volume, which reduces the structural the repository stability and the disposal availability. Countries, which use bitumen in the solidification of NPP's radioactive waste and have near-surface repositories, need to immobilize this bituminized waste within other drums containing cement pastes or mortars to disposal them. This study aims to find solutions for the storage in surface repository of bituminized radioactive waste products, making them compatible with the acceptance criteria of cemented waste products. It was also performed a modeling with the results obtained in the leaching test using the ALT program and defined the transport model of the cesium leachate element and it was verified that in the early times the leaching was governed by the diffusion model and later by the partition model. The results obtained in this study can be used in the evaluation of performance of repositories. (author)

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  1. Retrievable surface storage: interim storage of solidified high-level waste

    International Nuclear Information System (INIS)

    LaRiviere, J.R.; Nelson, D.C.

    1976-01-01

    Studies have been conducted on retrievable-surface-storage concepts for the interim storage of solidified high-level wastes. These studies have been reviewed by the Panel on Engineered Storage, convened by the Committee on Radioactive Waste Management of the National Research Council-National Academy of Sciences. The Panel has concluded that ''retrievable surface storage is an acceptable interim stage in a comprehensive system for managing high-level radioactive wastes.'' The scaled storage cask concept, which was recommended by the Panel on Engineered Storage, consists of placing a canister of waste inside a carbon-steel cask, which in turn is placed inside a thick concrete cylinder. The waste is cooled by natural convection air flow through an annulus between the cask and the inner wall of the concrete cylinder. The complete assembly is placed above ground in an outdoor storage area

  2. Near-surface storage facilities for vitrified high-level wastes

    International Nuclear Information System (INIS)

    Kondrat'ev, A.N.; Kulichenko, V.V.; Kryukov, I.I.; Krylova, N.V.; Paramoshkin, V.I.; Strakhov, M.V.

    1980-01-01

    Concurrently with the development of methods for solidifying liquid radioactive wastes, reliable and safe methods for the storage and disposal of solidified wastes are being devised in the USSR and other countries. One of the main factors affecting the choice of storage conditions for solidified wastes originating from the vitrification of high-level liquid wastes from fuel reprocessing plants is the problem of removing the heat produced by radioactive decay. In order to prevent the temperature of solidified wastes from exceeding the maximum permissible level for the material concerned, it is necessary to limit either the capacity of waste containers or the specific heat release of the wastes themselves. In order that disposal of high-level wastes in geological formations should be reliable and economic, solidified wastes undergo interim storage in near-surface storage facilities with engineered cooling systems. The paper demonstrates the relative influences of specific heat release, of the maximum permissible storage temperature for vitrified wastes and of the methods chosen for cooling wastes in order for the dimensions of waste containers to be reduced to the extent required. The effect of concentrating wastes to a given level in the vitrification process on the cost of storage in different types of storage facility is also examined. Calculations were performed for the amount of vitrified wastes produced by a reprocessing plant with a capacity of five tonnes of uranium per 24 hours. Fuel elements from reactors of the water-cooled, water-moderated type are sent for reprocessing after having been held for about two years. The dimensions of the storage facility are calculated on the assumption that it will take five years to fill

  3. Handling and Transport Problems

    Energy Technology Data Exchange (ETDEWEB)

    Pomarola, J. [Head of Technical Section, Atomic Energy Commission, Saclay (France); Savouyaud, J. [Head of Electro-Mechanical Sub-Division, Atomic Energy Commission, Saclay (France)

    1960-07-01

    Arrangements for special or dangerous transport operations by road arising out of the activities of the Atomic Energy Commission are made by the Works and Installations Division which acts in concert with the Monitoring and Protection Division (MPD) whenever radioactive substances or appliances are involved. In view of the risk of irradiation and contamination entailed in handling and transporting radioactive substances, including waste, a specialized transport and storage team has been formed as a complement to the emergency and decontamination teams.

  4. Operable Unit 3-13, Group 3, Other Surface Soils Remediation Sets 4-6 (Phase II) Waste Management Plan

    International Nuclear Information System (INIS)

    G. L. Schwendiman

    2006-01-01

    This Waste Management Plan describes waste management and waste minimization activities for Group 3, Other Surface Soils Remediation Sets 4-6 (Phase II) at the Idaho Nuclear Technology and Engineering Center located within the Idaho National Laboratory. The waste management activities described in this plan support the selected response action presented in the Final Record of Decision for Idaho Nuclear Technology and Engineering Center, Operable Unit 3-13. This plan identifies the waste streams that will be generated during implementation of the remedial action and presents plans for waste minimization, waste management strategies, and waste disposition

  5. Impact of Water Recovery from Wastes on the Lunar Surface Mission Water Balance

    Science.gov (United States)

    Fisher, John W.; Hogan, John Andrew; Wignarajah, Kanapathipi; Pace, Gregory S.

    2010-01-01

    Future extended lunar surface missions will require extensive recovery of resources to reduce mission costs and enable self-sufficiency. Water is of particular importance due to its potential use for human consumption and hygiene, general cleaning, clothes washing, radiation shielding, cooling for extravehicular activity suits, and oxygen and hydrogen production. Various water sources are inherently present or are generated in lunar surface missions, and subject to recovery. They include: initial water stores, water contained in food, human and other solid wastes, wastewaters and associated brines, ISRU water, and scavenging from residual propellant in landers. This paper presents the results of an analysis of the contribution of water recovery from life support wastes on the overall water balance for lunar surface missions. Water in human wastes, metabolic activity and survival needs are well characterized and dependable figures are available. A detailed life support waste model was developed that summarizes the composition of life support wastes and their water content. Waste processing technologies were reviewed for their potential to recover that water. The recoverable water in waste is a significant contribution to the overall water balance. The value of this contribution is discussed in the context of the other major sources and loses of water. Combined with other analyses these results provide guidance for research and technology development and down-selection.

  6. Application of advanced handling techniques to transportation cask design

    International Nuclear Information System (INIS)

    Bennett, P.C.

    1992-01-01

    Sandia National Laboratories supports the US Department of Energy (DOE) Office of Civilian Radioactive Waste Management (OCRWM) applying technology to the safe transport of nuclear waste. Part of that development effort includes investigation of advanced handling technologies for automation of cask operations at nuclear waste receiving facilities. Although low radiation levels are expected near transport cask surfaces, cumulative occupational exposure at a receiving facility can be significant. Remote automated cask handling has the potential to reduce both the occupational exposure and the time necessary to process a cask. Thus, automated handling is consistent with DOE efforts to reduce the lifecycle costs of the waste disposal system and to maintain public and occupational radiological risks as low as reasonably achievable. This paper describes the development of advanced handling laboratory mock-ups and demonstrations for spent fuel casks. Utilizing the control enhancements described below, demonstrations have been carried out including cask location and identification, contact and non-contact surveys, impact limiter removal, tiedown release, uprighting, swing-free movement, gas sampling, and lid removal operations. Manually controlled movement around a cask under off-normal conditions has also been demonstrated

  7. Handling, conditioning and storage of spent sealed radioactive sources. Technical manual for the management of low and intermediate level wastes generated at small nuclear research centres and by radioisotope users in medicine, research and industry

    International Nuclear Information System (INIS)

    2000-05-01

    This report is intended to provide reference material, guidance and know-how on handling, conditioning and storage of spent sealed radioactive sources (SRS) to both users of SRS and operators of waste management facilities. The scope of this report covers all types of SRS except those exempted from regulatory control. The report contains in some detail technical procedures for the conditioning of spent SRS, describes the means required to assure the quality of the resulting package and discusses the measures to prepare waste packages with a certain flexibility to accommodate possible future disposal requirements

  8. Surface sealing systems for dumps and old waste sites. Oberflaechenabdichtungssysteme fuer Deponien und Altlasten

    Energy Technology Data Exchange (ETDEWEB)

    Egloffstein, T [ed.; ICP Ingenieurgesellschaft Prof. Czurda und Partner mbH, Karlsruhe (Germany); Burkhardt, G [ed.; ICP Ingenieurgesellschaft Prof. Czurda und Partner mbH, Karlsruhe (Germany)

    1994-01-01

    The primary task of surface lining is to prevent rainwater from seeping into the wastes deposited below the surface. Its function is therefore to seal the surface. Above the sealing layer there is a drainage system; if large amount of gas are generated (domestic wastes) the gas is channeled through a gas drainage layer. The top layer of the system serves as recultivation horizon for plants. The functions of the liner are: - sealing against precipitating water; - channeling off gas or - seepage water; - plant site. (orig./EF)

  9. The surface disposal concept for LIL/SL waste

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-07-01

    Most low-level and intermediate-level short-lived (LIL/SL) waste result from the nuclear-power industry. Their specific activity level is sufficiently high to justify a protective conditioning and to ensure proper confinement until that level has decreased to harmless levels for human beings and the environment (a few centuries considering the half lives of the radionuclides contained in LIL/SL waste). The disposal concept for such residues relies on a multi-barrier protective system, each barrier being designed to fulfil different or redundant functions in order to delay or mitigate radionuclide transfers first into the environment and onwards to human beings. The originality of the concept pertains to its flexibility, since: it is adaptable to various geological environments and its overall performance may be guaranteed by modulating that of the engineered barriers, and it is suitable for the disposal of different types and sizes of waste packages, as long as their characteristics are consistent with acceptance criteria, which are de facto specific to each case. To provide its wide-ranging competences in the field of waste management and disposal, ANDRA offers multiple solutions, from consultancy and documents reviewing, to technology transfer and turnkey projects. The safety of the disposal facility is guaranteed by the combination of the package, the concrete structures, the filling materials between packages and the watertight clay cap that will be installed at the end of the operating lifetime of the facility. That layout also takes all natural risks into account. Lastly, all disposal structures are built away from any potential flood zones and from the highest possible level of the groundwater table. Concrete and metal packages are disposed of in slightly different structures. Once a structure is full, concrete packages are immobilised with gravel, whereas metal packages are blocked in place by pouring concrete between them. Once a disposal structure is

  10. The surface disposal concept for LIL/SL waste

    International Nuclear Information System (INIS)

    2011-01-01

    Most low-level and intermediate-level short-lived (LIL/SL) waste result from the nuclear-power industry. Their specific activity level is sufficiently high to justify a protective conditioning and to ensure proper confinement until that level has decreased to harmless levels for human beings and the environment (a few centuries considering the half lives of the radionuclides contained in LIL/SL waste). The disposal concept for such residues relies on a multi-barrier protective system, each barrier being designed to fulfil different or redundant functions in order to delay or mitigate radionuclide transfers first into the environment and onwards to human beings. The originality of the concept pertains to its flexibility, since: it is adaptable to various geological environments and its overall performance may be guaranteed by modulating that of the engineered barriers, and it is suitable for the disposal of different types and sizes of waste packages, as long as their characteristics are consistent with acceptance criteria, which are de facto specific to each case. To provide its wide-ranging competences in the field of waste management and disposal, ANDRA offers multiple solutions, from consultancy and documents reviewing, to technology transfer and turnkey projects. The safety of the disposal facility is guaranteed by the combination of the package, the concrete structures, the filling materials between packages and the watertight clay cap that will be installed at the end of the operating lifetime of the facility. That layout also takes all natural risks into account. Lastly, all disposal structures are built away from any potential flood zones and from the highest possible level of the groundwater table. Concrete and metal packages are disposed of in slightly different structures. Once a structure is full, concrete packages are immobilised with gravel, whereas metal packages are blocked in place by pouring concrete between them. Once a disposal structure is

  11. Evaluation of Groundwater Impacts to Support the National Environmental Policy Act Environmental Assessment for the INL Remote-Handled Low-Level Waste Disposal Project

    Energy Technology Data Exchange (ETDEWEB)

    Annette Schafer, Arthur S. Rood, A. Jeffrey Sondrup

    2011-12-23

    Groundwater impacts have been analyzed for the proposed remote-handled low-level waste disposal facility. The analysis was prepared to support the National Environmental Policy Act environmental assessment for the top two ranked sites for the proposed disposal facility. A four-phase screening and analysis approach was documented and applied. Phase I screening was site independent and applied a radionuclide half-life cut-off of 1 year. Phase II screening applied the National Council on Radiation Protection analysis approach and was site independent. Phase III screening used a simplified transport model and site-specific geologic and hydrologic parameters. Phase III neglected the infiltration-reducing engineered cover, the sorption influence of the vault system, dispersion in the vadose zone, vertical dispersion in the aquifer, and the release of radionuclides from specific waste forms. These conservatisms were relaxed in the Phase IV analysis which used a different model with more realistic parameters and assumptions. Phase I screening eliminated 143 of the 246 radionuclides in the inventory from further consideration because each had a half-life less than 1 year. An additional 13 were removed because there was no ingestion dose coefficient available. Of the 90 radionuclides carried forward from Phase I, 57 radionuclides had simulated Phase II screening doses exceeding 0.4 mrem/year. Phase III and IV screening compared the maximum predicted radionuclide concentration in the aquifer to maximum contaminant levels. Of the 57 radionuclides carried forward from Phase II, six radionuclides were identified in Phase III as having simulated future aquifer concentrations exceeding maximum contaminant limits. An additional seven radionuclides had simulated Phase III groundwater concentrations exceeding 1/100th of their respective maximum contaminant levels and were also retained for Phase IV analysis. The Phase IV analysis predicted that none of the thirteen remaining

  12. Multiple containment for LSA [low specific activity] and SCO [surface contaminated objects] wastes

    International Nuclear Information System (INIS)

    Burgess, M.H.

    1993-09-01

    Radioactive wastes are generally transported in the form of Low Specific Activity (LSA) materials or Surface Contaminated Objects (SCO). This report proposes that a method of acknowledging the beneficial effects of multiple containment for such wastes should be written into the 1996 Edition of the IAEA Transport Regulations. Experience used to assess risks from on-site movements of radioactive material in the UK can be applied to develop safety arguments justifying the alleviation of off-site transport risks. (UK)

  13. Safe handling of tritium

    International Nuclear Information System (INIS)

    1991-01-01

    The main objective of this publication is to provide practical guidance and recommendations on operational radiation protection aspects related to the safe handling of tritium in laboratories, industrial-scale nuclear facilities such as heavy-water reactors, tritium removal plants and fission fuel reprocessing plants, and facilities for manufacturing commercial tritium-containing devices and radiochemicals. The requirements of nuclear fusion reactors are not addressed specifically, since there is as yet no tritium handling experience with them. However, much of the material covered is expected to be relevant to them as well. Annex III briefly addresses problems in the comparatively small-scale use of tritium at universities, medical research centres and similar establishments. However, the main subject of this publication is the handling of larger quantities of tritium. Operational aspects include designing for tritium safety, safe handling practice, the selection of tritium-compatible materials and equipment, exposure assessment, monitoring, contamination control and the design and use of personal protective equipment. This publication does not address the technologies involved in tritium control and cleanup of effluents, tritium removal, or immobilization and disposal of tritium wastes, nor does it address the environmental behaviour of tritium. Refs, figs and tabs

  14. Pollution distribution of heavy metals in surface soil at an informal electronic-waste recycling site.

    Science.gov (United States)

    Fujimori, Takashi; Takigami, Hidetaka

    2014-02-01

    We studied distribution of heavy metals [lead (Pb), copper (Cu) and zinc (Zn)] in surface soil at an electronic-waste (e-waste) recycling workshop near Metro Manila in the Philippines to evaluate the pollution size (spot size, small area or the entire workshop), as well as to assess heavy metal transport into the surrounding soil environment. On-site length-of-stride-scale (~70 cm) measurements were performed at each surface soil point using field-portable X-ray fluorescence (FP-XRF). The surface soil at the e-waste recycling workshop was polluted with Cu, Zn and Pb, which were distributed discretely in surface soil. The site was divided into five areas based on the distance from an entrance gate (y-axis) of the e-waste recycling workshop. The three heavy metals showed similar concentration gradients in the y-axis direction. Zn, Pb and Cu concentrations were estimated to decrease to half of their maximum concentrations at ~3, 7 and 7 m from the pollution spot, respectively, inside the informal e-waste recycling workshop. Distance from an entrance may play an important role in heavy metal transport at the soil surface. Using on-site FP-XRF, we evaluated the metal ratio to characterise pollution features of the solid surface. Variability analysis of heavy metals revealed vanishing surficial autocorrelation over metre ranges. Also, the possibility of concentration prediction at unmeasured points using geostatistical kriging was evaluated, and heavy metals had a relative "small" pollution scales and remained inside the original workshop compared with toxic organohalogen compounds. Thus, exposure to heavy metals may directly influence the health of e-waste workers at the original site rather than the surrounding habitat and environmental media.

  15. Near-surface facilities for disposal radioactive waste from non-nuclear application

    International Nuclear Information System (INIS)

    Barinov, A.

    2000-01-01

    The design features of the near-surface facilities of 'Radon', an estimation of the possible emergency situations, and the scenarios of their progress are given. The possible safety enhancing during operation of near-surface facilities, so called 'Historical facilities', and newly developed ones are described. The Moscow SIA 'Radon' experience in use of mobile module plants for liquid radioactive waste purification and principal technological scheme of the plant are presented. Upgrading of the technological scheme for treatment and conditioning of radioactive waste for new-developed facilities is shown. The main activities related to management of spent ionizing sources are mentioned

  16. Scoping survey of perceived concerns, issues, and problems for near-surface disposal of FUSRAP waste

    International Nuclear Information System (INIS)

    Robinson, J.E.; Gilbert, T.L.

    1982-12-01

    This report is a scoping summary of concerns, issues, and perceived problems for near-surface disposal of radioactive waste, based on a survey of the current literature. Near-surface disposal means land burial in or within 15 to 20 m of the earth's surface. It includes shallow land burial (burial in trenches, typically about 6 m deep with a 2-m cap and cover) and some intermediate-depth land burial (e.g., trenches and cap similar to shallow land burial, but placed below 10 to 15 m of clean soil). Proposed solutions to anticipated problems also are discussed. The purpose of the report is to provide a better basis for identifying and evaluating the environmental impacts and related factors that must be analyzed and compared in assessing candidate near-surface disposal sites for FUSRAP waste. FUSRAP wastes are of diverse types, and their classification for regulatory purposes is not yet fixed. Most of it may be characterized as low-activity bulk solid waste, and is similar to mill tailings, but with somewhat lower average specific activity. It may also qualify as Class A segregated waste under the proposed 10 CFR 61 rules, but the parent radionuclides of concern in FUSRAP (primarily U-238 and Th-232) have longer half-lives than do the radionuclides of concern in most low-level waste. Most of the references reviewed deal with low-level waste or mill tailings, since there is as yet very little literature in the public domain on FUSRAP per se

  17. Assessment of Environmental Factors of Geology on Waste and Engineering Barriers for Waste Storage Near Surface

    International Nuclear Information System (INIS)

    Arimuladi SP

    2007-01-01

    Geological environment factors include features and processes occurring within that spatial and temporal (post-closure) domain whose principal effect is to determine the evolution of the physical, chemical, biological and human conditions of the domain that are relevant to estimating the release and migration of radionuclide and consequent exposure to man. Hardness of radioactive waste and engineer barrier can be decrease by environmental factors. Disposal system domain geological environment factors is a category in the International FEP list and is divided into sub-categories. There are 13 sub-factors of geological environment, 12 sub-factors influence hardness of radioactive waste and engineer barrier, thermal processes and conditions in geosphere can be excluded. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-12-31

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

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

    International Nuclear Information System (INIS)

    1994-01-01

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

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

    International Nuclear Information System (INIS)

    1994-01-01

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

  1. Hanford Site annual dangerous waste report: Volume 3, Part 2, Waste Management Facility report, dangerous waste

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1944-12-31

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

  2. Activation/waste management

    International Nuclear Information System (INIS)

    Maninger, C.

    1984-10-01

    The selection of materials and the design of the blankets for fusion reactors have significant effects upon the radioactivity generated by neutron activation in the materials. This section considers some aspects of materials selection with respect to waste management. The activation of the materials is key to remote handling requirements for waste, to processing and disposal methods for waste, and to accident severity in waste management operations. In order to realize the desirable evnironmental potentials of fusion power systems, there are at least three major goals for waste management. These are: (a) near-surface burial; (b) disposal on-site of the fusion reactor; (c) acceptable radiation doses at least cost during and after waste management operations

  3. Application of a modified electrochemical system for surface decontamination of radioactive metal waste

    International Nuclear Information System (INIS)

    Lee, J.H.; Lim, Y.K.; Yang, H.Y.; Shin, S.W.; Song, M.J.

    2003-01-01

    Conventional and modified electrolytic decontamination experiments were performed in a solution of sodium sulfate for the decontamination of carbon steel as the simulated metal wastes which are generated in large amounts from nuclear power plants. The effect of reaction time, current density and concentration of electrolytes in the modified electrolytic decontamination system were examined to remove the surface contamination of the simulated radioactive metal wastes. As for the results of this research, the modified electrochemical decontamination process can decontaminate more effectively than the conventional decontamination process by applying different anode material which causes higher induced electro-motive forces. When 0.5 M sodium sulfate, 0.4 A/cm 2 current density and 30 minutes reaction time were applied in the modified process, a 16 μm thickness change that is expected to remove most surface contamination in radioactive metal wastes was achieved on carbon steel which is the main material of radioactive metal waste in nuclear power plants. The decontamination efficiency of metal waste showed similar results with the small and large lab-scale modified electrochemical system. The application of this modified electrolytic decontamination system is expected to play a considerable role for decontamination of radioactive metal waste in nuclear power plants in the near future. (author)

  4. Utilization of surface-treated rubber particles from waste tires

    Energy Technology Data Exchange (ETDEWEB)

    Smith, F.G. [Argonne National Lab., IL (United States). Energy Systems Div.]|[Environmental Technologies Alternatives, Inc., Lima, OH (United States)

    1994-12-01

    During a 12-month program, the author successfully demonstrated commercial applications for surface-treated rubber particles in two major markets: footwear (shoe soles and components) and urethane-foam carpet underlay (padding). In these markets, he has clearly demonstrated the ease of using R-4080 and R-4030 surface-treated rubber particles in existing manufacturing plants and processes and have shown that the material meets or exceeds existing standards for performance, quality, and cost-effectiveness. To produce R-4080 and R-4030, vulcanized rubber, whole-tire material is finely ground to particles of nominal 80 and mesh size respectively. Surface treatment is achieved by reacting these rubber particles with chlorine gas. In this report, the author describes the actual test and evaluations of the participant companies, and identifies other potential end uses.

  5. Issues and Recommendations Arising from the Idaho National Laboratory Remote-Handled Low-Level Waste Disposal Facility Composite Analysis - 13374

    Energy Technology Data Exchange (ETDEWEB)

    Rood, Arthur S.; Schafer, Annette L.; Sondrup, A. Jeff [Idaho National Laboratory, Battelle Energy Alliance, P.O. Box 1625, Idaho Falls, ID 83401-2107 (United States)

    2013-07-01

    Development of the composite analysis (CA) for the Idaho National Laboratory's (INLs) proposed remote-handled (RH) low-level waste (LLW) disposal facility has underscored the importance of consistency between analyses conducted for site-specific performance assessments (PAs) for LLW disposal facilities, sites regulated by the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) [1], and residual decontamination and decommissioning (D and D) inventories. Consistency is difficult to achieve because: 1) different legacy sources and compliance time-periods were deemed important for each of the sites evaluated at INL (e.g., 100 years for CERCLA regulated facilities vs. 1,000 years for LLW disposal facilities regulated under U.S. Department of Energy (DOE) Order 435.1 [2]); 2) fate and transport assumptions, parameters, and models have evolved through time at the INL including the use of screening-level parameters vs. site-specific values; and 3) evaluation objectives for the various CERCLA sites were inconsistent with those relevant to either the PA or CA including the assessment of risk rather than effective dose. The proposed single site-wide CA approach would provide needed consistency, allowing ready incorporation of new information and/or facilities in addition to being cost effective in terms of preparation of CAs and review by the DOE. A single site-wide CA would include a central database of all existing INL sources, including those from currently operating LLW facilities, D and D activities, and those from the sites evaluated under CERCLA. The framework presented for the INL RH-LLW disposal facility allows for development of a single CA encompassing air and groundwater impacts. For groundwater impacts, a site-wide MODFLOW/MT3D-MS model was used to develop unit-response functions for all potential sources providing responses for a grid of receptors. Convolution and superposition of the response functions are used to compute

  6. Surface-type repository for low and intermediate level radioactive waste in the Republic of Croatia

    International Nuclear Information System (INIS)

    Kucar-Dragicevic, S.; Zarkovic, V.; Subasic, D.

    1995-01-01

    The low-level intermediate-level (LL/IL) radioactive waste repository siting and construction project is one of the activities related to establishing the rad waste management system in the Republic of Croatia. The repository project design is one in an array of project activities which also include the site selection procedure and public attitude issues. The prepared design documentation gives technical, safety and financial background relevant for making a final decision on the waste disposal type, and it includes the technological, mechanical, civil and financial documentation on the preliminary/basic design level. During the last few years, the preliminary design has been prepared and safety assessment conducted for the tunnel-type LL/IL rad waste repository. As the surface-type repository is one of alternatives for final disposal the design documentation for that repository type was prepared during 1994. (author)

  7. Contents and Sample Arguments of a Safety Case for Near Surface Disposal of Radioactive Waste

    International Nuclear Information System (INIS)

    2017-06-01

    This publication arises from the results of two projects to assist Member States in understanding and developing safety cases for near-surface radioactive waste disposal facilities. The objective of the publication is to give detailed information on the contents of safety cases for radioactive waste disposal and the types of arguments that may be included. It is written for technical experts preparing a safety case, and decision makers in the regulatory body and government. The publication outlines the key uses and aspects of the safety case, its evolution in parallel with that of the disposal facility, the key decision steps in the development of the waste disposal facility, the components of the safety case, their place in the Matrix of Arguments for a Safety Case (the MASC matrix), and a detailed description of the development of sample arguments that might be included in a safety case for each of two hypothetical radioactive waste disposal facilities.

  8. Hot Laboratories and Remote Handling

    International Nuclear Information System (INIS)

    Bart, G.; Blanc, J.Y.; Duwe, R.

    2003-01-01

    The European Working Group on ' Hot Laboratories and Remote Handling' is firmly established as the major contact forum for the nuclear R and D facilities at the European scale. The yearly plenary meetings intend to: - Exchange experience on analytical methods, their implementation in hot cells, the methodologies used and their application in nuclear research; - Share experience on common infrastructure exploitation matters such as remote handling techniques, safety features, QA-certification, waste handling; - Promote normalization and co-operation, e.g., by looking at mutual complementarities; - Prospect present and future demands from the nuclear industry and to draw strategic conclusions regarding further needs. The 41. plenary meeting was held in CEA Saclay from September 22 to 24, 2003 in the premises and with the technical support of the INSTN (National Institute for Nuclear Science and Technology). The Nuclear Energy Division of CEA sponsored it. The Saclay meeting was divided in three topical oral sessions covering: - Post irradiation examination: new analysis methods and methodologies, small specimen technology, programmes and results; - Hot laboratory infrastructure: decommissioning, refurbishment, waste, safety, nuclear transports; - Prospective research on materials for future applications: innovative fuels (Generation IV, HTR, transmutation, ADS), spallation source materials, and candidate materials for fusion reactor. A poster session was opened to transport companies and laboratory suppliers. The meeting addressed in three sessions the following items: Session 1 - Post Irradiation Examinations. Out of 12 papers (including 1 poster) 7 dealt with surface and solid state micro analysis, another one with an equally complex wet chemical instrumental analytical technique, while the other four papers (including the poster) presented new concepts for digital x-ray image analysis; Session 2 - Hot laboratory infrastructure (including waste theme) which was

  9. A methodology for evaluating alternative sites for a near-surface radioactive waste repository

    International Nuclear Information System (INIS)

    Watson, S.R.; Brownlow, S.A.

    1986-02-01

    This report addresses the issue of constructing an evaluation procedure for a near-surface radioactive waste repository. It builds on earlier work of the authors, and describes a basis for a practicable methodology for assessing the relative merits of different sites. (author)

  10. Sampling and sample handling procedures for priority pollutants in surface coal mining wastewaters. [Detailed list to be analyzed for

    Energy Technology Data Exchange (ETDEWEB)

    Hayden, R. S.; Johnson, D. O.; Henricks, J. D.

    1979-03-01

    The report describes the procedures used by Argonne National Laboratory to sample surface coal mine effluents in order to obtain field and laboratory data on 110 organic compounds or classes of compounds and 14 metals and minerals that are known as priority pollutants, plus 5-day biochemical oxygen demand (BOD/sub 5/), total organic carbon (TOC), chemical oxygen demand (COD), total dissolved solids (TDS), and total suspended solids (TSS). Included are directions for preparation of sampling containers and equipment, methods of sampling and sample preservation, and field and laboratory protocols, including chain-of-custody procedures. Actual analytical procedures are not described, but their sources are referenced.

  11. Use of Multiple Innovative Technologies for Retrieval and Handling of Low-Level Radioactive Tank Wastes at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Noble-Dial, J.; Riner, G.; Robinson, S.; Lewis, B.; Bolling, D.; Ganapathi, G.; Harper, M.; Billingsley, K.; Burks, B.

    2002-01-01

    The U.S. Department of Energy (DOE) successfully implemented an integrated tank waste management plan at Oak Ridge National Laboratory (ORNL) (1), which resulted in the cleanup, removal, or stabilization of 37 inactive underground storage tanks (USTs) since 1998, and the reduction of risk to human health and the environment. The integrated plan helped accelerate the development and deployment of innovative technologies for the retrieval of radioactive sludge and liquid waste from inactive USTs. It also accelerated the pretreatment of the retrieved waste and newly generated waste from ORNL research and development activities to provide for volume and contamination reduction of the liquid waste. The integrated plan included: retrieval of radioactive sludge, contaminated material, and other debris from USTs at ORNL using a variety of robotic and remotely operated equipment; waste conditioning and transfer of retrieved waste to pretreatment facilities and interim, double contained storage tanks; the development and deployment of technologies for pretreating newly generated and retrieved waste transferred to interim storage tanks; waste treatment and packaging for final off-site disposal; stabilization of the inactive USTs that did not meet the regulatory requirements of the Federal Facilities Agreement between the DOE, the Environmental Protection Agency (EPA), and the Tennessee Department of Environment and Conservation (TDEC); and the continued monitoring of the active USTs that remain in long-term service. This paper summarizes the successful waste retrieval and tank stabilization operations conducted during two ORNL tank remediation projects (The Gunite Tanks Remediation Project and the Old Hydrofracture Facility Tanks Remediation Project), the sludge retrieval operations from the active Bethel Valley Evaporator Service Tanks, and pretreatment operations conducted for the tank waste. This paper also provides the status of ongoing activities conducted in preparation

  12. Waste-surface mapping of the Fernald K-65 silos using a structured light measurement system

    International Nuclear Information System (INIS)

    Burks, B.L.; DePiero, F.W.; Dinkins, M.A.; Rowe, J.C.; Selleck, C.B.; Jacoboski, D.L.

    1992-10-01

    A remotely operated surface-mapping measurement system was developed by the Robotics ampersand Process Systems Division at Oak Ridge National Laboratory for use in the K-65 waste-storage silos at Fernald, Ohio. The mapping system used three infrared line-generating laser diodes as illumination sources and three high-resolution, low-lux, calibrated, black-and-white, charge-coupled-device video cameras as receivers. These components were combined to form structured light source range and direction sensors with six different possible emitter-receiver pairs. A technology demonstration and predeployment tests were performed at Fernald using the empty Silo 4 into which was placed rectangular objects of known dimensions. These objects were scanned by the structured light sources to demonstrate functionality and verify that the system was giving sufficiently accurate range data in three dimensions. The structured light sources were deployed in Silos 1 and 2 to scan the waste surfaces. The resulting data were merged to create three-dimensional maps of those surfaces. A bentonite clay cap was placed over the waste surfaces and surface maps were obtained. The change in surface height before and after bentonite addition was utilized as a measure of clay cap thickness

  13. Waste management, waste resource facilities and waste conversion processes

    International Nuclear Information System (INIS)

    Demirbas, Ayhan

    2011-01-01

    In this study, waste management concept, waste management system, biomass and bio-waste resources, waste classification, and waste management methods have been reviewed. Waste management is the collection, transport, processing, recycling or disposal, and monitoring of waste materials. A typical waste management system comprises collection, transportation, pre-treatment, processing, and final abatement of residues. The waste management system consists of the whole set of activities related to handling, treating, disposing or recycling the waste materials. General classification of wastes is difficult. Some of the most common sources of wastes are as follows: domestic wastes, commercial wastes, ashes, animal wastes, biomedical wastes, construction wastes, industrial solid wastes, sewer, biodegradable wastes, non-biodegradable wastes, and hazardous wastes.

  14. Preparing, Loading and Shipping Irradiated Metals in Canisters Classified as Remote-Handled (RH) Low-Level Waste (LLW) From Oak Ridge National Laboratory (ORNL) to the Nevada Test Site (NTS)

    International Nuclear Information System (INIS)

    McClelland, B.C.; Moore, T.D.

    2006-01-01

    Irradiated metals, classified as remote-handled low-level waste generated at the Oak Ridge National Laboratory (ORNL) in Oak Ridge, Tennessee, were containerised in various sized canisters for long-term storage. The legacy waste canisters were placed in below-grade wells located at the 7827 Facility until a pathway for final disposal at the Nevada Test Site (NTS) could be identified and approved. Once the pathway was approved, WESKEM, LLC was selected by Bechtel Jacobs Company, LLC to prepare, load, and ship these canisters from ORNL to the NTS. This paper details some of the technical challenges encountered during the retrieval process and solutions implemented to ensure the waste was safely and efficiently over-packed and shipped for final disposal. The technical challenges detailed in this paper include: 1) how to best perform canister/lanyard pre-lift inspections since some canisters had not been moved in ∼10 years, so deterioration was a concern; 2) replacing or removing damaged canister lanyards; 3) correcting a mis-cut waste canister lanyard resulting in a shielded overpack lid not seating properly; 4) retrieving a stuck canister; and 5) developing a path forward after an overstrained lanyard failed causing a well shield plug to fall and come in contact with a waste canister. Several of these methods can serve as positive lessons learned for other projects encountering similar situations. (authors)

  15. A unified framework for developing effective hygiene procedures for hands, environmental surfaces and laundry in healthcare, domestic, food handling and other settings

    Science.gov (United States)

    Bloomfield, Sally F.; Carling, Philip C.; Exner, Martin

    2017-01-01

    Hygiene procedures for hands, surfaces and fabrics are central to preventing spread of infection in settings including healthcare, food production, catering, agriculture, public settings, and home and everyday life. They are used in situations including hand hygiene, clinical procedures, decontamination of environmental surfaces, respiratory hygiene, food handling, laundry hygiene, toilet hygiene and so on. Although the principles are common to all, approaches currently used in different settings are inconsistent. A concern is the use of inconsistent terminology which is misleading, especially to people we need to communicate with such as the public or cleaning professionals. This paper reviews the data on current approaches, alongside new insights to developing hygiene procedures. Using this data, we propose a more scientifically-grounded framework for developing procedures that maximize protection against infection, based on consistent principles and terminology, and applicable across all settings. A key feature is use of test models which assess the state of surfaces after treatment rather than product performance alone. This allows procedures that rely on removal of microbes to be compared with those employing chemical or thermal inactivation. This makes it possible to ensure that a consistent “safety target level” is achieved regardless of the type of procedure used, and allows us deliver maximum health benefit whilst ensuring prudent usage of antimicrobial agents, detergents, water and energy. PMID:28670508

  16. Cask system design guidance for robotic handling

    International Nuclear Information System (INIS)

    Griesmeyer, J.M.; Drotning, W.D.; Morimoto, A.K.; Bennett, P.C.

    1990-10-01

    Remote automated cask handling has the potential to reduce both the occupational exposure and the time required to process a nuclear waste transport cask at a handling facility. The ongoing Advanced Handling Technologies Project (AHTP) at Sandia National Laboratories is described. AHTP was initiated to explore the use of advanced robotic systems to perform cask handling operations at handling facilities for radioactive waste, and to provide guidance to cask designers regarding the impact of robotic handling on cask design. The proof-of-concept robotic systems developed in AHTP are intended to extrapolate from currently available commercial systems to the systems that will be available by the time that a repository would be open for operation. The project investigates those cask handling operations that would be performed at a nuclear waste repository facility during cask receiving and handling. The ongoing AHTP indicates that design guidance, rather than design specification, is appropriate, since the requirements for robotic handling do not place severe restrictions on cask design but rather focus on attention to detail and design for limited dexterity. The cask system design features that facilitate robotic handling operations are discussed, and results obtained from AHTP design and operation experience are summarized. The application of these design considerations is illustrated by discussion of the robot systems and their operation on cask feature mock-ups used in the AHTP project. 11 refs., 11 figs

  17. Weathering behavior of mine tailings and waste rock: A surface investigation

    International Nuclear Information System (INIS)

    Domvile, S.J.; Li, M.G.; Sollner, D.D.; Nesbitt, W.

    1994-01-01

    A study focusing on the ion movement in the near surface of sulfide minerals was conducted to better understand the weathering mechanisms of mine waste materials. Tailings and waste rock samples from Canadian mines were subjected to controlled weathering studies using various chemical leachants. Leachates were analyzed for various parameters, and petrographic analyses were conducted on the solid residues. Laboratory oxidation studies of pure pyrrhotite and arsenopyrite were carried out using the surface techniques X-ray photoelectron spectroscopy (XPS) and auger electron spectroscopy (AES). The data derived from the weathering study and the surface techniques were correlated to determine mechanisms of oxidation. Several results were observed during the project: ferric iron constitutes one third of the iron present in pyrrhotite, sulfide oxidation is initiated when rock is blasted, sulfide sulfur is oxidized to di- and poly-sulfides prior to forming sulfates, and significantly more sulfate is produced upon exposure to aqueous environments than to air alone

  18. Treatment and utilization of waste waters of surface mines in Ukraine

    Energy Technology Data Exchange (ETDEWEB)

    Khmel' , N S

    1981-01-01

    Waste water of brown coal surface mines in the Dnieper basin is characterized. The water's pH value is 7, alkalinity ranges from 5.1 to 5.9 mg equivalent/1, it has no odor, a low mineralization level ranging from 1000 to 1100 mg/l. Concentration of mechanical impurities (suspended matter) ranges from 90 to 900 mg/l, and its maximum level can reach 5000 mg/l. An improved design of tanks in which waste water from surface mines is treated, and mechanical impurities settle, is proposed. Conventional design of a water sedimentation tank consists of a long ditch in which suspended matter settles, and a rectangular water reservoir at its end. In the improved version the long ditch is enlarged in some places to create additional tanks and to reduce velocity of flowing waste water. This improvement increases the amount of suspended matter which settles in the ditch and in its enlarged zones. When water reaches the rectangular sedimentation tank at the end of the system its suspended matter content is reduced to 40-45 mg/l. Formulae used to calculate dimensions of water treatment system, gradient of the ditch and size of sedimentation tank are presented. Methods of discharging treated waste water to surface water, rivers and stagnant waters, are evaluated. (In Russian)

  19. Synthesis of long live storage studies surface storage of MA-VL wastes

    International Nuclear Information System (INIS)

    Hollender, F.; Jourdain, F.; Piault, E.; Blanchet, Y.; Avakian, G.; Goger, F.; Caillaud, J.; Devictor, N.; Bary, B.; Moitier, C.; Breton, E.; Ranc, G.; Gaillard, J.P.; Lagrave, H.

    2004-01-01

    This document is realized in the framework of the axis 3 of the law of 1991 on the radioactive wastes management. It presents a long time surface storage installation of medium activity long life wastes. The long time of the installation would reach 300 years at the maximum. The feasibility is demonstrated and the design choices are presented and justified. The specific points of the long time storage installation, which are different from a classical industrial storage installation, are also discussed. (A.L.B.)

  20. Ergonomic material-handling device

    Science.gov (United States)

    Barsnick, Lance E.; Zalk, David M.; Perry, Catherine M.; Biggs, Terry; Tageson, Robert E.

    2004-08-24

    A hand-held ergonomic material-handling device capable of moving heavy objects, such as large waste containers and other large objects requiring mechanical assistance. The ergonomic material-handling device can be used with neutral postures of the back, shoulders, wrists and knees, thereby reducing potential injury to the user. The device involves two key features: 1) gives the user the ability to adjust the height of the handles of the device to ergonomically fit the needs of the user's back, wrists and shoulders; and 2) has a rounded handlebar shape, as well as the size and configuration of the handles which keep the user's wrists in a neutral posture during manipulation of the device.

  1. Handling of disused radioactive materials in Ecuador

    International Nuclear Information System (INIS)

    Benitez, Manuel

    1999-10-01

    This paper describes the handling of disused radioactive sources. It also shows graphic information of medical and industrial equipment containing radioactive sources. This information was prepared as part of a training course on radioactive wastes. (The author)

  2. Vertical Drop of the Naval SNF Long Waste Package On Unyielding Surface

    International Nuclear Information System (INIS)

    S. Mastilovic

    2006-01-01

    The purpose of this calculation is to determine the structural response of a Naval SNF (Spent Nuclear Fuel) Long Waste Package (WP) subjected to 2 m-vertical drop on unyielding surface (US). The scope of this document is limited to reporting the calculation results in terms of maximum stress intensities. This calculation is associated with the waste package design; calculation is performed by the Waste Package Design group. AP-3.12Q, Revision 0, ICN 0, Calculations, is used to perform the calculation and develop the document. The finite element calculation is performed by using the commercially available ANSYS Version (V) 5.4 finite element code. The result of this calculation is provided in terms of maximum stress intensities

  3. Consideration of Criteria for a Conceptual Near Surface Radioactive Waste disposal Facility in Kenya

    Energy Technology Data Exchange (ETDEWEB)

    Nderitu, Stanley Werugia; Kim, Changlak [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2014-05-15

    The purpose of the criteria is to limit the consequences of events which could lead to radiation exposures. This study will present an approach for establishing radiological waste acceptance criteria using a safety assessment methodology and illustrate some of its application in establishing limits on the total activity and the activity concentrations of radioactive waste to be disposed in a conceptual near surface disposal facility in Kenya. The approach will make use of accepted methods and computational schemes currently used in assessing the safety of near surface disposal facilities. The study will mainly focus on post-closure periods. The study will employ some specific inadvertent human intrusion scenarios in the development of example concentration ranges for the disposal of near-surface wastes. The overall goal of the example calculations is to illustrate the application of the scenarios in a performance assessment to assure that people in the future cannot receive a dose greater than an established limit. The specific performance assessments will use modified scenarios and data to establish acceptable disposal concentrations for specific disposal sites and conditions. Safety and environmental impacts assessments is required in the post-closure phase to support particular decisions in development, operation, and closure of a near surface repository.

  4. Radioactive wastes

    International Nuclear Information System (INIS)

    Grass, F.

    1982-01-01

    Following a definition of the term 'radioactive waste', including a discussion of possible criteria allowing a delimitation of low-level radioactive against inactive wastes, present techniques of handling high-level, intermediate-level and low-level wastes are described. The factors relevant for the establishment of definitive disposals for high-level wastes are discussed in some detail. Finally, the waste management organization currently operative in Austria is described. (G.G.)

  5. Nuclear waste repository in basalt: a design description

    International Nuclear Information System (INIS)

    Ritchie, J.S.; Schmidt, B.

    1982-01-01

    The conceptual design of a nuclear waste repository in basalt is described. Nuclear waste packages are placed in holes drilled into the floor of tunnels at a depth of 3700 ft. About 100 miles of tunnels are required to receive 35,000 packages. Five shafts bring waste packages, ventilation air, excavated rock, personnel, material, and services to and from the subsurface. The most important surface facility is the waste handling building, located over the waste handling shaft, where waste is received and packaged for storage. Two independent ventilation systems are provided to avoid potential contamination of spaces that do not contain nuclear waste. Because of the high temperatures at depth, an elaborate air chilling system is provided. Because the waste packages deliver a considerable amount of heat energy to the rock mass, particular attention is paid to heat transfer and thermal stress studies. 3 references, 31 figures, 3 tables

  6. Management of Radioactive Wastes

    International Nuclear Information System (INIS)

    Tchokosa, P.

    2010-01-01

    Management of Radioactive Wastes is to protect workers and the public from the radiological risk associated with radioactive waste for the present and future. It application of the principles to the management of waste generated in a radioisotope uses in the industry. Any material that contains or is contaminated with radionuclides at concentrations or radioactivity levels greater than ‘exempt quantities’ established by the competent regulatory authorities and for which no further use is foreseen or intended. Origin of the Radioactive Waste includes Uranium and Thorium mining and milling, nuclear fuel cycle operations, Operation of Nuclear power station, Decontamination and decommissioning of nuclear facilities and Institutional uses of isotopes. There are types of radioactive waste: Low-level Waste (LLW) and High-level Waste. The Management Options for Radioactive Waste Depends on Form, Activity, Concentration and half-lives of the radioactive waste, Storage and disposal methods will vary according to the following; the radionuclides present, and their concentration, and radio toxicity. The contamination results basically from: Contact between radioactive materials and any surface especially during handling. And it may occur in the solid, liquid or gas state. Decontamination is any process that will either reduce or completely remove the amount of radionuclides from a contaminated surface

  7. Considerations in the development of near surface repositories for radioactive waste

    International Nuclear Information System (INIS)

    2003-01-01

    The report presents an integrated, stepwise approach for the development (includes pre-operational, operational and post-closure phases) of near surface disposal facilities for low and intermediate level radioactive waste. It has been developed in light of the considerable experience that has accumulated on the development of such disposal systems and is consistent with the current international requirements, principles, standards and guidance for the disposal of radioactive waste. It is considered that the systematic application of the various steps of the approach can contribute to the successful development of a repository programme. The approach is designed to be generic, integrating the various technical and nontechnical factors, and flexible enough to be suitable for use in the various Member States, ranging from countries that have nuclear power plants to countries that have small inventories of radioactive waste from nuclear applications. It is anticipated that this report will be particularly useful and of direct relevance to Member States that are currently developing, or have plans to develop, disposal facilities for low and intermediate level radioactive waste in the near future. The report is intended to respond to the disposal needs of the various Member States, ranging from countries with nuclear power plants to countries having small inventories of radioactive waste from nuclear applications. It was developed with the help of consultants and through an Advisory Group Meeting held in November 2001

  8. Waste Tyres as Heat Sink to Reduce the Driveway Surface Temperatures in Malaysia

    Directory of Open Access Journals (Sweden)

    Aniza Abdul Aziz

    2013-12-01

    Full Text Available The development of roads and driveways are on the rise as automobiles are now a necessity to all. This excessive development with its requirements increased the urban heat temperature and the generation of waste tyres. Waste tyre management has therefore been taken seriously by developed countries and since the European directive to ban used tyre products and whole tire disposal from landfill in 2003 and 2006 respectively, many researchers have looked for alternative ways to use the waste tyre. In Malaysia, The Smart and Cool Home Developer attempted to develop an eco-house by utilising waste tyre as the foundation for the driveway and claimed that the buried tyres act as a heat sink for the concrete and reduce the surface temperature of the driveway. Hence investigations were conducted on two sample houses to investigate this phenomenon. Findings from this pilot study show that waste tyres do act as a heat sink to the concrete driveways which affect the ambient temperature and relative humidity of the immediate surroundings.

  9. Evaluation of bubbler/diaphragm techniques to measure surface level in the waste storage tanks

    International Nuclear Information System (INIS)

    Peters, T.J.; Hickman, B.J.; Colson, J.B.

    1993-10-01

    This report describes the results of tests conducted at the Pacific Northwest Laboratory (PNL) to determine if a bubbler technique can be used to determine the surface level in the waste tanks. Two techniques were evaluated. The first technique is a standard bubbler system in which a tube is placed below the surface of the liquid to be measured and air pressure inside a tube is increased until bubbles begin to become emitted from the tube. The air pressure then is a function of the pressure at the bottom of the tube. The second technique involves a system similar to the standard bubbler technique, but instead of bubbles being released into the material to be gauged, air pressure is increased against a diaphragm until enough pressure is applied to overcome the pressure of the liquid at the given depth, at which time the air then flows in a return loop back to a vent. The advantage of the diaphragm system is that it is a sealed system; thus no air is released into the waste tank materials, and it is not possible for the waste tank materials to get into the air flow. Based on the results of the tests conducted in this program, it appears that the bubbler and diaphragm systems that were tested could not be used for accurate measurements of the level in the waste tanks. Both exhibited deposits of simulated waste tank material at the end of the devices which affected the ability of the gauge to accurately determine changes in the surface level even though the measured value of the level was inaccurate. Further investigations into the cause of this inaccuracy may be warranted. Alternate diaphragm materials may improve the performance of this gauge

  10. From energy resource to riddance problem. The issue of nuclear waste handling in the public dialogue in Sweden, 1950-2002; Fraan energiresurs till kvittblivningsproblem. Fraagan om kaernavfallets hantering i det offentliga samtalet i Sverige, 1950-2002

    Energy Technology Data Exchange (ETDEWEB)

    Anshelm, Jonas [Linkoeping Univ. (Sweden)

    2006-10-15

    Which risks are associated with the handling of high level radioactive wastes? Where should they be kept? Who is responsible for their safe keeping? How should a repository for safe final disposal be designed? Is there, at all, a safe solution for all future time? How could we possibly know that? These questions and many more have been given much attention in the public debate in Sweden, ever since the plans for a Swedish nuclear power program were approved by the parliament in the 1950s. If the questions largely have remained the same, the answers have varied a lot. Representatives for both the nuclear industry and the environmental movement have changed their attitudes and claims for knowing the truth as the technological, political, economical, scientific and cultural circumstances change. This report examines the changes in value base and what was held for truth regarding the plans for a Swedish repository for high-level radioactive waste. E.g. in the 1950s the waste was regarded as an energy resource for the future breeder reactors - in contrast to the conflict-ridden debates of the 1970s when the possibility to manage the waste by any means was questioned. The opposing views on how to select a site for the repository and the diverging opinions on risks, responsibilities, knowledge, technologies, science and nature during the 1980s and 1990s are also analyzed.

  11. Technical considerations in the design of near surface disposal facilities for radioactive waste

    International Nuclear Information System (INIS)

    2001-11-01

    Good design is an important step towards ensuring operational as well as long term safety of low and intermediate level waste (LILW) disposal. The IAEA has produced this report with the objective of outlining the most important technical considerations in the design of near surface disposal facilities and to provide some examples of the design process in different countries. This guidance has been developed in light of experience gained from the design of existing near surface disposal facilities in a range of Member States. In particular the report provide information on design objective, design requirements, and design phases. The report focuses on: near surface disposal facilities accepting solidified LILW; disposal facilities on or just below the ground surface, where the final protective covering is of the order of a few metres thick; and disposal facilities several tens of metres below the ground surface (including rock cavern type facilities)

  12. Durability of simulated waste glass: effects of pressure and formation of surface layers

    International Nuclear Information System (INIS)

    Wicks, G.G.; Mosley, W.C.; Whitkop, P.G.; Saturday, K.A.

    1981-01-01

    The leaching behavior of simulated Savannah River Plant (SRP) waste glass was studied at elevated pressures and anticipated storage temperatures. An integrated approach, which combined leachate solution analyses with both bulk and surface studies, was used to study the corrosion process. Compositions of leachates were evaluated by colorimetry and atomic absorption. Used in the bulk and surface analyses were optical microscopy, scanning electron microscopy, x-ray energy spectroscopy, wide-angle x-ray, diffraction, electron microprobe analysis, infrared reflectance spectroscopy, electron spectroscopy for chemical analysis, and Auger electron spectroscopy. Results from this study show that there is no significant adverse effect of pressure, up to 1500 psi and 90 0 C, on the chemical durability of simulated SPR waste glass leached for one month in deionized water. In addition, the leached glass surface layer was characterized by an adsorbed film rich in minor constituents from the glass. This film remained on the glass surface even after leaching in relatively alkaline solutions at elevated pressures at 90 0 C for one month. The sample surface area to volume of leachant ratios (SA/V) was 10:1 cm -1 and 1:10 cm -1 . The corrosion mechanisms and surface and subsurface layers produced will be discussed along with the potential importance of these results to repository storage

  13. Evaluation of disposal possibilities and potential in the Konrad iron-ore mine based on experiments for the handling and isolation of radioactive wastes

    International Nuclear Information System (INIS)

    Brewitz, W.

    1984-01-01

    The Gesellschaft fuer Strahlen und Umwelftorschung (GSF) has established, at the conclusion of feasibility studies, that the suitability of the Konrad mine for the disposal of low-level waste and decommisioning waste has been proved, as well as the radiological safety of the disposal operations. On 31 August 1982, the Physikalisch-Technische Bundesanstalt (PTB) submitted an application for licensing to the State Authority of Lower Saxony

  14. Objectives and limitations of scientific studies with reference to the Swedish R ampersand D programme 1992 for handling and final disposal of nuclear waste

    International Nuclear Information System (INIS)

    Sjoeblom, R.; Dverstorp, B.; Wingefors, S.

    1994-01-01

    The Swedish Nuclear Power Inspectorate (SKI) has recently concluded its evaluation of the Swedish programme for the development of a system for the management of nuclear waste. The programme was compiled and issued by the Swedish Nuclear Fuel and Waste Management Company (SKB). In this process of programme formulation and review, considerable attention has been paid to the question of how scientific studies should be directed and performed in order to provide the support needed in the programme

  15. Use of commercial robotics in radioactive waste shipping and receiving

    International Nuclear Information System (INIS)

    Berger, J.D.

    1985-01-01

    Radioactive waste shipping and receiving facilities presently planned for commercial and defense nuclear waste will handle waste packages at frequencies far in excess of those in common practice today. Unacceptable personnel exposure to ionizing radiation is projected to occur if current limits for radiation levels at the cask surface and current handling methods are used. To reduce these exposure levels, alternate handling methods are being developed and demonstrated. The production nature of cask receiving operations suggests commercial robotics be incorporated into a remote handling system to reduce predicted worker exposure to acceptable levels, while maintaining or increasing throughput. The first phase of cask handling system development culminated in a proof-of-principle test demonstrating the feasibility of performing cask receiving and unloading operations in a remote and partially automated manner. 6 refs., 12 figs

  16. Geologic waste disposal and a model for the surface movement of radionuclides

    International Nuclear Information System (INIS)

    Helton, J.; Iman, R.; Brown, J.; Schreurs, S.

    1979-01-01

    A model for the surface movement of radionuclides is presented. This model, which is referred to as the Pathways Model, was constructed in a NRC project to develop a methodology to assess the risk associated with the goelogic disposal of high-level radioactive waste. The methodology development involves work in two major areas: (a) models for physical processes, and (b) statistical techniques for the use and assessment of these models. The presentation of the Pathways Model involves topics from both areas

  17. Source tracking swine fecal waste in surface water proximal to swine concentrated animal feeding operations.

    Science.gov (United States)

    Heaney, Christopher D; Myers, Kevin; Wing, Steve; Hall, Devon; Baron, Dothula; Stewart, Jill R

    2015-04-01

    Swine farming has gone through many changes in the last few decades, resulting in operations with a high animal density known as confined animal feeding operations (CAFOs). These operations produce a large quantity of fecal waste whose environmental impacts are not well understood. The purpose of this study was to investigate microbial water quality in surface waters proximal to swine CAFOs including microbial source tracking of fecal microbes specific to swine. For one year, surface water samples at up- and downstream sites proximal to swine CAFO lagoon waste land application sites were tested for fecal indicator bacteria (fecal coliforms, Escherichia coli and Enterococcus) and candidate swine-specific microbial source-tracking (MST) markers (Bacteroidales Pig-1-Bac, Pig-2-Bac, and Pig-Bac-2, and methanogen P23-2). Testing of 187 samples showed high fecal indicator bacteria concentrations at both up- and downstream sites. Overall, 40%, 23%, and 61% of samples exceeded state and federal recreational water quality guidelines for fecal coliforms, E. coli, and Enterococcus, respectively. Pig-1-Bac and Pig-2-Bac showed the highest specificity to swine fecal wastes and were 2.47 (95% confidence interval [CI]=1.03, 5.94) and 2.30 times (95% CI=0.90, 5.88) as prevalent proximal down- than proximal upstream of swine CAFOs, respectively. Pig-1-Bac and Pig-2-Bac were also 2.87 (95% CI=1.21, 6.80) and 3.36 (95% CI=1.34, 8.41) times as prevalent when 48 hour antecedent rainfall was greater than versus less than the mean, respectively. Results suggest diffuse and overall poor sanitary quality of surface waters where swine CAFO density is high. Pig-1-Bac and Pig-2-Bac are useful for tracking off-site conveyance of swine fecal wastes into surface waters proximal to and downstream of swine CAFOs and during rain events. Copyright © 2014 Elsevier B.V. All rights reserved.

  18. Site characterization field manual for near surface geologic disposal of low-level radioactive waste

    International Nuclear Information System (INIS)

    McCray, J.G.; Nowatzki, E.A.

    1985-01-01

    This field manual has been developed to aid states and regions to do a detailed characterization of a proposed near-surface low-level waste disposal site. The field manual is directed at planners, staff personnel and experts in one discipline to acquaint them with the requirements of other disciplines involved in site characterization. While it can provide a good review, it is not designed to tell experts how to do their job within their own discipline

  19. Environmental analysis of the system for the households' wastewater and organic waste. Synthesis of handling systems studied in the research program 'Organic waste as a plant nutrient resource'; Miljoesystemanalys av hushaallens avlopp och organiska avfall. Syntes av hanteringssystem undersoekta inom FoU-programmet 'Organiskt avfall som vaextnaeringsresurs'

    Energy Technology Data Exchange (ETDEWEB)

    Kaerrman, E. [Chalmers Inst. of Technology, Goeteborg (Sweden). Sanitary Engineering; Joensson, Haakan; Sonesson, Ulf [Swedish Univ. of Agricultural Sciences, Uppsala (Sweden). Agricultural Engineering; Gruvberger, C.; Dalemo, M. [Swedish Inst. of Agricultural Engineering, Uppsala (Sweden); Stenstroem, T.A. [Swedish Inst. for Infections Disease Control, Stockholm (Sweden)

    2001-11-01

    Knowledge derived within the RandD programme 'Organic waste as a plant nutrient resource' was in this study used to compare four systems for handling wastewater and organic waste. In the conventional system, the sewage is treated in a sewage plant (SP) with mechanical, biological (incl. 70% nitrogen reduction) and chemical treatment. The sludge is anaerobically digested, dewatered and stored for 20 days. In the energy forest irrigation system the SP has no special nitrogen or phosphorus reduction and the effluent water is stored during winter and used for irrigation during summer. In the liquid compost system the wastewater from the vacuum toilets is co-treated with organic household waste in a reactor at 55 deg C for 7 days. In the other systems, the organic household waste is treated in a central windrow compost. In the urine separation system, the source-separated urine is hygienised by separate storage for 6 months. In the liquid compost and urine separation systems the grey water or the grey water plus faecal wastewater respectively is treated in a SP with phosphorus reduction but without nitrogen reduction. The residual products of the systems fertilise arable land 10 km away. The mass flows, environmental effects, energy turnover and resource utilisation were simulated using ORWARE (ORganic WAste REsearch model), assuming that the wastewater and organic household waste from 20,000 people were treated by the systems. Table values were used for composition of wastewater etc. Industrial wastewater and surface water were excluded from the study. The risk of disease transmission was discussed from the flows and faecal concentrations of the wastewater and from estimated intake of Salmonella, EHEC and Cryptosporidium by some animals under assumed conditions. Throughout the study, it was assumed that the systems function well, without disruptions, and that they are used as intended. The compared environmental effects and resource usage were prioritised in

  20. Direct Measurement of Surface Dissolution Rates in Potential Nuclear Waste Forms: The Example of Pyrochlore.

    Science.gov (United States)

    Fischer, Cornelius; Finkeldei, Sarah; Brandt, Felix; Bosbach, Dirk; Luttge, Andreas

    2015-08-19

    The long-term stability of ceramic materials that are considered as potential nuclear waste forms is governed by heterogeneous surface reactivity. Thus, instead of a mean rate, the identification of one or more dominant contributors to the overall dissolution rate is the key to predict the stability of waste forms quantitatively. Direct surface measurements by vertical scanning interferometry (VSI) and their analysis via material flux maps and resulting dissolution rate spectra provide data about dominant rate contributors and their variability over time. Using pyrochlore (Nd2Zr2O7) pellet dissolution under acidic conditions as an example, we demonstrate the identification and quantification of dissolution rate contributors, based on VSI data and rate spectrum analysis. Heterogeneous surface alteration of pyrochlore varies by a factor of about 5 and additional material loss by chemo-mechanical grain pull-out within the uppermost grain layer. We identified four different rate contributors that are responsible for the observed dissolution rate range of single grains. Our new concept offers the opportunity to increase our mechanistic understanding and to predict quantitatively the alteration of ceramic waste forms.

  1. Acidic Microenvironments in Waste Rock Characterized by Neutral Drainage: Bacteria–Mineral Interactions at Sulfide Surfaces

    Directory of Open Access Journals (Sweden)

    John W. Dockrey

    2014-03-01

    Full Text Available Microbial populations and microbe-mineral interactions were examined in waste rock characterized by neutral rock drainage (NRD. Samples of three primary sulfide-bearing waste rock types (i.e., marble-hornfels, intrusive, exoskarn were collected from field-scale experiments at the Antamina Cu–Zn–Mo mine, Peru. Microbial communities within all samples were dominated by neutrophilic thiosulfate oxidizing bacteria. However, acidophilic iron and sulfur oxidizers were present within intrusive waste rock characterized by bulk circumneutral pH drainage. The extensive development of microbially colonized porous Fe(III (oxyhydroxide and Fe(III (oxyhydroxysulfate precipitates was observed at sulfide-mineral surfaces during examination by field emission-scanning electron microscopy-energy dispersive X-ray spectroscopy (FE-SEM-EDS. Linear combination fitting of bulk extended X-ray absorption fine structure (EXAFS spectra for these precipitates indicated they were composed of schwertmannite [Fe8O8(OH6–4.5(SO41–1.75], lepidocrocite [γ-FeO(OH] and K-jarosite [KFe3(OH6(SO42]. The presence of schwertmannite and K-jarosite is indicative of the development of localized acidic microenvironments at sulfide-mineral surfaces. Extensive bacterial colonization of this porous layer and pitting of underlying sulfide-mineral surfaces suggests that acidic microenvironments can play an important role in sulfide-mineral oxidation under bulk circumneutral pH conditions. These findings have important implications for water quality management in NRD settings.

  2. An overview of technical requirements on durable concrete production for near surface disposal facilities for radioactive wastes

    International Nuclear Information System (INIS)

    Tolentino, Evandro; Tello, Cledola Cassia Oliveira de

    2013-01-01

    Radioactive waste can be generated by a wide range of activities varying from activities in hospitals to nuclear power plants, to mines and mineral processing facilities. General public have devoted nowadays considerable attention to the subject of radioactive waste management due to heightened awareness of environmental protection. The preferred strategy for the management of all radioactive waste is to contain it and to isolate it from the accessible biosphere. The Federal Government of Brazil has announced the construction for the year of 2014 and operation for the year of 2016 of a near surface disposal facility for low and intermediate level radioactive waste. The objective of this paper is to provide an overview of technical requirements related to production of durable concrete to be used in near surface disposal facilities for radioactive waste concrete structures. These requirements have been considered by researchers dealing with ongoing designing effort of the Brazilian near surface disposal facility. (author)

  3. How should the household waste be handled? Evaluation of different treatment methods; Hur skall hushaallsavfallet tas om hand? Utvaerdering av olika behandlingsmetoder

    Energy Technology Data Exchange (ETDEWEB)

    Sundqvist, J.O.; Carlsson Reich, M.; Granath, J. [The Swedish Environmental Research Inst., Goeteborg (Sweden); Baky, Andras [Swedish Inst. of Agriculatural and Environmental Engineering, Uppsala (Sweden); Eriksson, Ola [Royal Inst. of Tech., Stockholm (Sweden). Div. of Industrial Ecology

    2002-02-01

    Energy turnover, and environmental and economic consequences of different municipal solid waste management systems have been studied in a systems analysis. Different combinations of incineration, recycling of separated plastic and cardboard packages and biological treatment (anaerobic digestion and composting) of easy degradable organic waste, were studied and also compared to landfilling. A simulation model (ORWARE), based on LCA methodology, was used. The following parameters were used for evaluating the different waste management options: consumption of energy resources, emissions of greenhouse gases, emissions of acidifying substances, emissions of eutrophicating substances, emissions of photo oxidant formers, heavy metal flows, financial economy and welfare economy. The analysis was based on a model municipality. In a sensitivity analysis different technical, geographic and demographic parameters have been varied, making the result to cover several different types of municipalities and regions. The conclusions from the study are as follows. The most obvious conclusion is that landfilling should be avoided. Wastes that can be incinerated (combusted), material recycled, anaerobically digested or composted should not be landfilled. This is valid even if landfill gas is extracted and utilised, and the leachate is collected and treated. This is due to that the resources in the waste are inefficiently utilised when landfilled, making it necessary to produce materials, fuels and nutrients from virgin resources. It is impossible to draw unambiguous conclusions of which of the other treatment options that is is most preferable. There are advantages and disadvantages with all options. In a systems perspective there are small differences between incineration and aerobic digestion of easy degradable organic material, and between incineration and material recycling of e.g. plastics and cardboard. Material recycling, anaerobic digestion and incineration should not be seen

  4. Evaluation of pressure transducers to measure surface level in the waste storage tanks

    International Nuclear Information System (INIS)

    Peters, T.J.; Colson, J.B.

    1994-06-01

    This report describes the results of tests conducted at the Pacific Northwest Laboratory (PNL) to determine if pressure transducers can be used to measure the surface level in the waste tanks. A survey was first conducted to evaluate which, if any, commercially available pressure transducers were available that could meet the requirements for use in the waste tanks. More than 35 companies were contacted to determine if they manufactured a pressure transducer that could be used in the 101-SY waste tank. The three basic requirements for a pressure transducer for this application were that they were radiation-hardened, could withstand a caustic environment, and were certified to be intrinsically safe. No manufacturer was able to meet all three of these requirements with a commercially available product. Seven companies were able to meet the requirements for being radiation-hardened and being able to withstand the caustic environment. However, only two of the nine companies were willing to supply a pressure transducer for laboratory testing. The two pressure transducers that were tested in this program were the VEGA D36-38 from HiTech Technologies, Inc., and the KP-1911-A from Kaman Instrumentation Corporation. Pressure transducers operate on the principle that the pressure at the location of a sensor increases directly with the depth of the liquid above it. A liquid is required in order for these devices to operate. For these tests, water was first used to determine the ideal operation of the devices, then the devices were placed in a 101-SY waste tank simulant. The simulant had a specific gravity of 1.96 and had the consistency similar to the convective layer in the 101-SY waste tank. In order to determine the surface level with pressure transducers, the density of the material needs to be known

  5. Validation of the plan of handling of hospital accustomed to waste of the Costa Rica Box of the Public Health; Validacion del plan de manejo de desechos solidos hospitalarios de la Caja Costarricense del Seguro Social

    Energy Technology Data Exchange (ETDEWEB)

    Salazar Monge, F J

    1999-07-01

    The handling that is given at the moment to the hospital solid waste in Costa Rica is considered inadequate, due to the lack of the personnel's of cleaning training, segregation, vehicles of transport and recipients, signaling, team of personal protection, schedule and gathering route and recycle. This situation represents a serious danger for the population's health and the since environment is exposed infectious agents toxic substances and even radioactive products that are generated in the residuals of the centers of health. In this work they intend improvements to the system of handling of solid waste of the hospital Calderon Guardia, and you evaluates each one of the points mentioned previously. A revision was made about the properties, the quality and the quantity of the produced solid waste and an I diagnose of the current situation in this center of health. The proposed improvements are based on the regional program of hospital solid waste, agreement ALA91/33 (this it was elaborated in 1997 by a cooperation initiative between the European Union and the Governments from Costa Rica, El Salvador, Guatemala, Honduras, Nicaragua and Panama) and the Norms for the Handling of dangerous Solid Waste in Establishments of Health (elaborated by the Costa Rican Box of the Public Health) their advantages and disadvantages are also exposed and he/she is carried out a comparative analysis among the handling plan proposed in these documents and the one that is used at the moment in the Calderon Guardia hospital. Some of the detected problems are: it lacks of team of personal protection, the appropriate recipients are not used, storage of waste inside the hospital, the vehicles used in the internal transport are not appropriate, bad organization of the storing center, among others. [Spanish] El manejo que se da actualmente a los desechos solidos hospitalarios en Costa Rica se considera inadecuado, debido a la falta de capacitacion del personal de limpieza, segregacion

  6. Waste Handeling Building Conceptual Study

    Energy Technology Data Exchange (ETDEWEB)

    G.W. Rowe

    2000-11-06

    The objective of the ''Waste Handling Building Conceptual Study'' is to develop proposed design requirements for the repository Waste Handling System in sufficient detail to allow the surface facility design to proceed to the License Application effort if the proposed requirements are approved by DOE. Proposed requirements were developed to further refine waste handling facility performance characteristics and design constraints with an emphasis on supporting modular construction, minimizing fuel inventory, and optimizing facility maintainability and dry handling operations. To meet this objective, this study attempts to provide an alternative design to the Site Recommendation design that is flexible, simple, reliable, and can be constructed in phases. The design concept will be input to the ''Modular Design/Construction and Operation Options Report'', which will address the overall program objectives and direction, including options and issues associated with transportation, the subsurface facility, and Total System Life Cycle Cost. This study (herein) is limited to the Waste Handling System and associated fuel staging system.

  7. Assessment of management alternatives for LWR wastes. Volume 8. Cost and radiological impact associated with near-surface disposal of reactor waste (Spanish concept)

    International Nuclear Information System (INIS)

    Alamo Berna, S.; Sanchez Delgado, N.

    1993-01-01

    This report deals with the determination of the cost and the radiological impact associated with a near-surface disposal site (Spanish concept) for low and medium-level radioactive waste generated during operation of a 20 GWe nuclear park composed of LWRs for 30 years. This study is part of an overall theoretical exercise aimed at evaluating a selection of management routes for LWR waste based on economical and radiological criteria

  8. Assessment of management alternatives for LWR wastes. Volume 7. Cost and radiological impact associated with near-surface disposal of reactor waste (French concept)

    International Nuclear Information System (INIS)

    Malherbe, J.

    1993-01-01

    This report deals with the determination of the cost and the radiological impact associated with a near-surface disposal site (French concept) for low and medium-level radioactive waste generated during operation of a 20 GWe nuclear park composed of LWRs for 30 years. This study is part of an overall theoretical exercise aimed at evaluating a selection of management routes for LWR waste based on economical and radiological criteria

  9. Bacterial community dynamics in surface flow constructed wetlands for the treatment of swine waste.

    Science.gov (United States)

    Ibekwe, A M; Ma, J; Murinda, Shelton; Reddy, G B

    2016-02-15

    Constructed wetlands are generally used for the removal of waste from contaminated water. In the swine production system, wastes are traditionally flushed into an anaerobic lagoon which is then sprayed on agricultural fields. However, continuous spraying of lagoon wastewater on fields can lead to high N and P accumulations in soil or lead to runoff which may contaminate surface or ground water with pathogens and nutrients. In this study, continuous marsh constructed wetland was used for the removal of contaminants from swine waste. Using pyrosequencing, we assessed bacterial composition within the wetland using principal coordinate analysis (PCoA) which showed that bacterial composition from manure influent and lagoon water were significantly different (P=0.001) from the storage pond to the final effluent. Canonical correspondence analysis (CCA) showed that different bacterial populations were significantly impacted by ammonium--NH4 (P=0.035), phosphate--PO4(3-) (P=0.010), chemical oxygen demand--COD (P=0.0165), total solids--TS (P=0.030), and dissolved solids--DS (P=0.030) removal, with 54% of the removal rate explained by NH4+PO4(3-) according to a partial CCA. Our results showed that different bacterial groups were responsible for the composition of different wetland nutrients and decomposition process. This may be the major reason why most wetlands are very efficient in waste decomposition. Published by Elsevier B.V.

  10. Charging generators for waste management costs

    International Nuclear Information System (INIS)

    Berry, J.B.; Homan, F.J.

    1988-01-01

    The Department of Energy (DOE) has recognized the need for waste management that incorporates improved waste-handling techniques and more stringent regulatory requirements to prevent future liabilities such as Superfund sites. DOE-Oak Ridge Operations (DOE-ORO) has recognized that an effective waste management program focuses on control at the source and that the burden for responsible waste management can be placed on generators by charging for waste management costs. The principle of including the waste management costs in the total cost of the product, even when the product is research and development, is being implemented at Oak Ridge National Laboratory (ORNL). This paper summarizes a plan to charge waste generators, the administrative structure of the plan, a comparison between the rate structure and changes in waste disposal operations, and issues that have surfaced as the plan is implemented

  11. Removal of selenium species from waters using various surface-modified natural particles and waste materials

    Energy Technology Data Exchange (ETDEWEB)

    Yigit, Nevzat O.; Tozum, Seda [Department of Environmental Engineering, Suleyman Demirel University, Isparta (Turkey)

    2012-07-15

    Waste red mud and natural pumice/volcanic slag particles were surface modified and their selenium adsorption from waters was investigated. Acid activation/heat treatment of original red mud (ORM) particles significantly increased their micropore and external surface area and cumulative volume of pores. Iron oxide coating of pumice/slags and acid activation of ORM decreased their pH{sub pzc} values and increased surface acidity. Selenite/selenate adsorption on iron oxide surfaces and acid activated red mud (AARM) was very fast with approximately first-order adsorption kinetics. Iron oxide coating of pumice/slag and acid activation of ORM particles significantly enhanced their selenite and selenate uptakes. Maximum Se adsorption capacities as high as 6.3 (mg Se/g adsorbent) were obtained by AARM. The extent of selenate uptakes by the surface modified particles was generally lower than those of selenite. Due to competition among Se species and other background water matrix for iron oxide adsorption sites, reduced selenite/selenate uptakes were found in natural water compared to single solute tests. Higher Se uptakes by iron oxide surfaces were found at pH 7.5 compared to pH 8.9, due to increased electrostatic repulsion among iron oxides and Se species at higher pH. The most effective adsorbents among the tested 17 different particles for Se uptake were AARM and iron oxide coated pumice. Se concentrations less than drinking water standards (5-10 {mu}g/L) can be achieved by these particles. These low-cost, natural, or recyclable waste particles appear to be promising adsorbents for Se removal after their surface modification. (Copyright copyright 2012 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  12. Brief summary of slag handling options reviewed for the slagging pyrolysis incinerator in the transuranic waste treatment facility (TWIF) at the INEL

    International Nuclear Information System (INIS)

    Darnell, G.R.

    1980-06-01

    This report summarizes the technical problems associated with molten transuranic waste slag as it flows from the incinerator shaft (gasifier) of the slagging pyrolysis incinerator. It addresses essential gasifier seals, slag casting and pouring technology, and transportation and packaging problems. Areas requiring further study and testing are identified

  13. Chemical decontamination method for radioactive metal waste

    International Nuclear Information System (INIS)

    Onuma, Tsutomu; Akimoto, Hidetoshi

    1991-01-01

    The invention relates to a decontamination method for radioactive metal waste products derived from equipment that handles radioactive materials whose surfaces have been contaminated; in particular it concerns a decontamination method that reduces the amount of radioactive waste by decontaminating radioactive waste substances to a level of radioactivity in line with normal waste products. In order to apply chemical decontamination to metal waste products whose surfaces are divided into carbon steel waste and stainless steel waste; the carbon steel waste is treated using only a primary process in which the waste is immersed in a sulfuric acid solution, while the stainless steel waste must be treated with both the primary process and then electrolytically reduces it for a specific length of time and a secondary process that uses a solution of sulfuric acid mixed with oxidizing metal salts. The method used to categorize metal waste into carbon steel waste and stainless steel waste involves determining the presence, or absence, of magnetism. Voltage is applied for a fixed duration; once that has stopped, electrolytic reduction repeats the operative cycle of applying, then stopping voltage until the potential of the radioactive metal waste is retained in the active region. 1 fig. 2 tabs

  14. Confinement facilities for handling plutonium

    International Nuclear Information System (INIS)

    Maraman, W.J.; McNeese, W.D.; Stafford, R.G.

    1975-01-01

    Plutonium handling on a multigram scale began in 1944. Early criteria, equipment, and techniques for confining contamination have been superseded by more stringent criteria and vastly improved equipment and techniques for in-process contamination control, effluent air cleaning and treatment of liquid wastes. This paper describes the evolution of equipment and practices to minimize exposure of workers and escape of contamination into work areas and into the environment. Early and current contamination controls are compared. (author)

  15. A review of construction techniques available for surface and underground radioactive waste repositories

    International Nuclear Information System (INIS)

    Godfrey, D.G.; Davies, I.L.; MacKenzie, R.D.

    1985-01-01

    In terms of engineering requirements the construction of surface or indeed underground radioactive waste repositories is not unduly difficult. The civil engineering techniques likely to be required have generally been carried out previously, albeit not in the context of radioactive waste repositories in this country. The emphasis will have to be very much on the quality of construction. This paper emphasises the need for quality construction and describes the techniques likely to be used in the construction of repositories. Reference is made to the materials likely to be used in the construction of repositories and also to the need for being able to convince the designers, regulating authorities and the general public that the materials used will indeed last for the required time. Brief reference is made at the end of the paper to the civil engineering parameters requiring consideration in the location of repository siting. (author)

  16. An Investigation of Technologies for Hazardous Sludge Reduction at AFLC (Air Force Logistics Command) Industrial Waste Treatment Plants. Volume 1. Sodium Borohydride Treatment and Sludge Handling Technologies.

    Science.gov (United States)

    1983-12-01

    Fisher Cupric sulfate-CuSO 4 . 5H20, Certified ACS Fisher Sodium Bicarbonate-NaHCO3, Certified ACS Fisher NaOH-Certified ACS Electrolytic Pellets , Fisher...The dryer (D-1), burner , and air handling system are part of a package unit including a 4-foot diameter by 24 foot long free-standing rotary dryer, a...blower with a rated capacity of 6,200 scfm of air at 500C, a burner capable of heating that volume of air to 125*C and a cyclonic dust separator to

  17. Study of waterproof capabilities of the engineered barrier containing bentonite in near surface radioactive waste repositories

    International Nuclear Information System (INIS)

    Luu Cao Nguyen; Nguyen Ba Tien; Doan Thi Thu Hien; Nguyen Van Chinh; Vuong Huu Anh

    2017-01-01

    In Vietnam, the study of nuclear fuel cycle is in first steps, such as the exploitation and uranium processing. These processes generated large amounts of radioactive waste over-timing. The naturally occurring radioactive material and technologically enhanced radioactive material (NORM/TENORM) waste, which would be large, needs to be managed and disposed reasonably by effective methods. These wastes were used to be disposal in the near surface. It was therefore very important to study the model of radioactive waste repository, where bentonite waterproofing layer would be applied for the engineered barrier. The aim of this study was to obtain the preliminary parameters for low-level radioactive waste disposal site being suitable with the conditions of Vietnam. The investigation of the ratio between soil and bentonite was taken part. The experiments with some layers of waterproofing material with the ratio of soil and bentonite as 75/25, 50/50 and 25/75 were carried out to test the moving of uranium nuclide through these waterproofing material layers. Analyzing the uranium content in each layer (0.1 cm) of pressed soil - bentonite mixture (as a block) to determine the uranium nuclide adsorption from solution into the materials in the different ratios at the different times: 1, 2 and 3 months was carried out. The results showed that the calculated average rate of uranium nuclide migration into the soil - bentonite layer was 5.4x10 -10 , 5.4x10 -10 and 3.85x10 -10 m/s corresponding to the waterproofing layer thickness (for 300 years) 4.86 m, 4.86 m and 3.63 m respectively, which was due on the ratio of soil and bentonite 75/25, 50/50, 25/75 to keep the safety for the repository. (author)

  18. A literature review of surface alteration layer effects on waste glass behavior

    International Nuclear Information System (INIS)

    Feng, X.; Cunnane, J.C.; Bates, J.K.

    1993-01-01

    When in contact with an aqueous solution, nuclear waste glass is subject to a chemical attack that results in progressive alteration. During tills alteration, constituent elements of the glass pass into the solution; elements initially in solution diffuse into, or are adsorbed onto, the solid; and new phases appear. This results in the formation of surface layers on the reacted glass. The glass corrosion and radionuclide release can be better understood by investigating these surface layer effects. In the past decade, there have been numerous studies regarding the effects of surface layers on glass reactions. This paper presents a systematic analysis and summary of the past knowledge regarding the effects of surface layers on glass-water interaction. This paper describes the major formation mechanisms of surface layers; reviews the role of surface layers in controlling mass transport and glass reaction affinity (through crystalline phases, an amorphous silica, a gel layer, or all the components in the glass); and discusses how the surface layers contribute to the retention of radionuclides during glass dissolution

  19. A technique for the geometric modeling of underground surfaces: Nevada Nuclear Waste Storage Investigations Project

    International Nuclear Information System (INIS)

    Williams, R.L.

    1988-03-01

    There is a need within the Nevada Nuclear Waste Storage Investigation (NNWSI) project to develop three-dimensional surface definitions for the subterranean stratigraphies at Yucca Mountain, Nevada. The nature of the data samples available to the project require an interpolation technique that can perform well with sparse and irregularly spaced data. Following an evaluation of the relevant existing methods, a new technique, Multi-Kernel Modulation (MKM), is presented. MKM interpolates sparse and irregularly spaced data by modulating a polynomial trend surface with a linear summation of regular surfaces (kernels). A perspective discussion of MKM, Kriging, and Multiquadric Analysis reveals that MKM has the advantage of simplicity and efficiency when used with sparse samples. An example of the use of MKM to model a complex topography is presented. 24 refs., 6 figs., 2 tabs

  20. Materials design considerations and selection for a large rad waste incinerator

    International Nuclear Information System (INIS)

    Vormelker, P.R.; Jenkins, C.F.; Burns, H.H.

    1997-01-01

    A new incinerator has been built to process self-generated, low level radioactive wastes at the Department of Energy's Savannah River Site. Wastes include protective clothing and other solid materials used during the handling of radioactive materials, and liquid chemical wastes resulting from chemical and waste management operations. The basic design and materials of construction selected to solve the anticipated corrosion problems from hot acidic gases are reviewed. Problems surfacing during trial runs prior to radioactive operations are discussed

  1. Radionuclide transport modelling for a buried near surface low level radioactive waste

    International Nuclear Information System (INIS)

    Terzi, R.

    2004-01-01

    The disposal of radioactive waste, which is the last step of any radioactive waste management policy, has not yet been developed in Turkey. The existing legislation states only the discharge limits for the radioactive wastes to be discharged to the environment. The objective of this modelling study is to assist in safety assessment and selecting disposal site for gradually increasing non-nuclear radioactive wastes. This mathematical model has been developed for the environmental radiological assessment of near surface disposal sites for the low and intermediate level radioactive wastes. The model comprised of three main components: source term, geosphere transport and radiological assessment. Radiation dose for the babies (1 years age) and adults (≥17 years age) have been computed for the radionuclides Cesium 137 (Cs-137) and Strontium 90 (Sr-90), having the activity of 1.10 12 Becquerel(Bq), in radioactive waste through transport of radionuclide in liquid phase with the various pathways. The model consisted of first order ordinary differential equations was coded as a TCODE file in MATLAB program. The radiation dose to man for the realist case and low probability case have been calculated by using Runge-Kutta solution method in MATLAB programme for radionuclide transport from repository to soil layer and then to the ground water(saturated zone) through drinking water directly and consuming agricultural and animal products pathways in one year period. Also, the fatal cancer risk assessment has been made by taking into account the annual dose received by people. Various dose values for both radionuclides have been found which depended on distribution coefficient, retardation factor and dose conversion factors. The most important critical parameters on radiological safety assessment are the distribution coefficient in soil layer, seepage velocity in unsaturated zone and thickness of the unsaturated zone (soil zone). The highest radiation dose and average dose to

  2. Specific corrosion product on interior surface of a bronze wine vessel with loop-handle and its growth mechanism, Shang Dynasty, China

    International Nuclear Information System (INIS)

    Li Yang; Bao Zhirong; Wu Taotao; Jiang, Junchun; Chen Guantao; Pan Chunxu

    2012-01-01

    In this paper, a kind of specific stalactitic product was found on the interior surface of a covered bronze wine vessel with loop-handle (Chinese name is you), which was fabricated in Shang Dynasty (1700 B.C.–1100 B.C.) and now is collected in Xiaogan Museum, Hubei province of China. The microstructures of the product were characterized systematically by using optical microscopy, scanning electron microscope, transmission electron microscope, X-ray diffraction, and Raman microscopy. The experimental results revealed that the product belonged to a kind of malachite with high purity and high crystallinity. The growth of the product was considered to be a possible reason that the vessel was overly airtight within a museum display cabinet besides a lid of the vessel, which made the excess of H 2 O and CO 2 gas concentrations inside the vessel during long-term storage. This corrosion product is very harmful to bronze cultural relics, because of a large amount of copper consumption from the matrix which will reduce its life. The growth mechanism of the specific stalactitic product and the suggestions for preservation of the similar bronze relics in museum were proposed. - Highlights: ► The stalactitic product was the high purity and good crystallinity malachite. ► Its growth was related to the excess of H 2 O and CO 2 gas concentrations in museum. ► It is harmful to the bronzes, because copper will be consumed from the matrix. ► The suggestions for preservation of the similar bronzes in museum were proposed.

  3. Specific corrosion product on interior surface of a bronze wine vessel with loop-handle and its growth mechanism, Shang Dynasty, China

    Energy Technology Data Exchange (ETDEWEB)

    Li Yang; Bao Zhirong; Wu Taotao [School of Physics and Technology, Center for Electron Microscopy and MOE Key Laboratory of Artificial Micro- and Nano-structures, Wuhan University, Wuhan 430072 (China); Jiang, Junchun [Xiaogan Museum, Xiaogan 432000 (China); Chen Guantao [Center for Archaeometry, Wuhan University, Wuhan 430072 (China); Pan Chunxu, E-mail: cxpan@whu.edu.cn [School of Physics and Technology, Center for Electron Microscopy and MOE Key Laboratory of Artificial Micro- and Nano-structures, Wuhan University, Wuhan 430072 (China); Center for Archaeometry, Wuhan University, Wuhan 430072 (China)

    2012-06-15

    In this paper, a kind of specific stalactitic product was found on the interior surface of a covered bronze wine vessel with loop-handle (Chinese name is you), which was fabricated in Shang Dynasty (1700 B.C.-1100 B.C.) and now is collected in Xiaogan Museum, Hubei province of China. The microstructures of the product were characterized systematically by using optical microscopy, scanning electron microscope, transmission electron microscope, X-ray diffraction, and Raman microscopy. The experimental results revealed that the product belonged to a kind of malachite with high purity and high crystallinity. The growth of the product was considered to be a possible reason that the vessel was overly airtight within a museum display cabinet besides a lid of the vessel, which made the excess of H{sub 2}O and CO{sub 2} gas concentrations inside the vessel during long-term storage. This corrosion product is very harmful to bronze cultural relics, because of a large amount of copper consumption from the matrix which will reduce its life. The growth mechanism of the specific stalactitic product and the suggestions for preservation of the similar bronze relics in museum were proposed. - Highlights: Black-Right-Pointing-Pointer The stalactitic product was the high purity and good crystallinity malachite. Black-Right-Pointing-Pointer Its growth was related to the excess of H{sub 2}O and CO{sub 2} gas concentrations in museum. Black-Right-Pointing-Pointer It is harmful to the bronzes, because copper will be consumed from the matrix. Black-Right-Pointing-Pointer The suggestions for preservation of the similar bronzes in museum were proposed.

  4. Development of commercial robots for radwaste handling

    International Nuclear Information System (INIS)

    Colborn, K.A.

    1988-01-01

    The cost and dose burden associated with low level radwaste handling activities is a matter of increasing concern to the commercial nuclear power industry. This concern is evidenced by the fact that many utilities have begun to revaluate waste generation, handling, and disposal activities at their plants in an effort to improve their overall radwaste handling operations. This paper reports on the project Robots for Radwaste Handling, to identify the potential of robots to improve radwaste handling operations. The project has focussed on the potential of remote or automated technology to improve well defined, recognizable radwaste operations. The project focussed on repetitive, low skill level radwaste handling and decontamination tasks which involve significant radiation exposure

  5. Site selection report basalt waste isolation program near-surface test facility

    International Nuclear Information System (INIS)

    Sharpe, S.D.

    1978-01-01

    A site selection committee was established to review the information gathered on potential sites and to select a site for the Near-Surface Test Facility Phase I. A decision was made to use a site on the north face of Gable Mountain located on the Hanford Site. This site provided convenient access to the Pomona Basalt Flow. This flow was selected for use at this site because it exhibited the characteristics established in the primary criteria. These criteria were: the flows thickness; its dryness; its nearness to the surface; and, its similarities to basalt units which are candidates for the repository. After the selection of the Near-Surface Test Facility Phase I Site, the need arose for an additional facility to demonstrate safe handling, storage techniques, and the physical effects of radioactive materials on an in situ basalt formation. The committee reviewed the sites selected for Phase I and chose the same site for locating Phase II of the Near-Surface Test Facility

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  7. Condensed listing of surface boreholes at the Waste Isolation Pilot Plant Project through 31 December 1995

    International Nuclear Information System (INIS)

    Hill, L.R.; Aguilar, R.; Mercer, J.W.; Newman, G.

    1997-01-01

    This report contains a condensed listing of Waste Isolation Pilot Plant (WIPP) project surface boreholes drilled for the purpose of site selection and characterization through 31 December 1995. The US Department of Energy (DOE) sponsored the drilling activities, which were conducted primarily by Sandia National Laboratories. The listing provides physical attributes such as location (township, range, section, and state-plane coordinates), elevation, and total borehole depth, as well as the purpose for the borehole, drilling dates, and information about extracted cores. The report also presents the hole status (plugged, testing, monitoring, etc.) and includes salient findings and references. Maps with borehole locations and times-of-drilling charts are included

  8. On release of radionuclides from a near-surface radioactive waste repository to the environment

    Directory of Open Access Journals (Sweden)

    Gudelis Arūnas

    2015-09-01

    Full Text Available A closed near-surface radioactive waste repository is the source of various radionuclides causing the human exposure. Recent investigations confirm an effectiveness of the engineering barriers installed in 2006 to prevent the penetration of radionuclides to the environment. The tritium activity concentration in groundwater decreased from tens of kBq/l to below hundreds of Bq/l. The monitoring and groundwater level data suggest the leaching of tritium from previously contaminated layers of unsaturated zone by rising groundwater while 210Pb may disperse as a decay product of 226Ra daughters.

  9. Procedures and techniques for closure of near surface disposal facilities for radioactive waste

    International Nuclear Information System (INIS)

    2001-12-01

    The overall objective of this report is to provide Member States with guidance on planning and implementation of closure of near surface disposal facilities for low and intermediate level radioactive waste. The specific objectives are to review closure concepts, requirements, and components of closure systems; to discuss issues and approaches to closure, including regulatory, economic, and technical aspects; and to present major examples of closure techniques used and/or considered by Member States. Some examples of closure experience from Member States are presented in the Appendix and were indexed separately

  10. Heavy metal contamination of surface soil in electronic waste dismantling area: site investigation and source-apportionment analysis.

    Science.gov (United States)

    Jinhui Li; Huabo Duan; Pixing Shi

    2011-07-01

    The dismantling and disposal of electronic waste (e-waste) in developing countries is causing increasing concern because of its impacts on the environment and risks to human health. Heavy-metal concentrations in the surface soils of Guiyu (Guangdong Province, China) were monitored to determine the status of heavy-metal contamination on e-waste dismantling area with a more than 20 years history. Two metalloids and nine metals were selected for investigation. This paper also attempts to compare the data among a variety of e-waste dismantling areas, after reviewing a number of heavy-metal contamination-related studies in such areas in China over the past decade. In addition, source apportionment of heavy metal in the surface soil of these areas has been analysed. Both the MSW open-burning sites probably contained invaluable e-waste and abandoned sites formerly involved in informal recycling activities are the new sources of soil-based environmental pollution in Guiyu. Although printed circuit board waste is thought to be the main source of heavy-metal emissions during e-waste processing, requirement is necessary to soundly manage the plastic separated from e-waste, which mostly contains heavy metals and other toxic substances.

  11. Radioactive wastes: the challenge of volumes reduction

    International Nuclear Information System (INIS)

    Lepetit, V.

    2005-01-01

    The reduction of radioactive waste volumes is a priority for the French atomic energy commission (CEA) and for the Areva group. This article gives a rapid overview of the equipments and processes used to separate the valorizable materials from the ultimate wastes: pulsed separation columns and evaporators for the liquid-liquid extraction, compactification of spent fuel hulls, remote handling systems, recoverable colloid for surface decontamination, decontaminating foam, hydrothermal oxidation of organic and aqueous effluents, cold crucible vitrification etc. (J.S.)

  12. Operations Program Plan for the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    1990-09-01

    This document, Revision 4 of the Operations Program Plan, has been developed as the seven-year master plan for operating of the Waste Isolation Pilot Plant (WIPP). Subjects covered include public and technical communications; regulatory and environmental programs; startup engineering; radiation handling, surface operations, and underground operations; waste certification and waste handling; transportation development; geotechnical engineering; experimental operations; engineering program; general maintenance; security program; safety, radiation, and regulatory assurance; quality assurance program; training program; administration activities; management systems program; and decommissioning. 243 refs., 19 figs., 25 tabs. (SM)

  13. Contact-handled transuranic transportation system structural analysis

    International Nuclear Information System (INIS)

    Lamoreaux, G.H.; Romesberg, L.E.; Sutherland, S.H.; Duffey, T.A.

    1980-01-01

    The Transuranic Package Transporter (TRUPACT) is a Type B overpack being developed for contact-handled transuranic waste. End-on, side-on, and corner impacts of the loaded TRUPACT due to a 9 m drop onto an unyielding surface have been analyzed. In each case the analyses progressed from simplified hand approaches to successively more complex finite element calculations. The first analysis of each series represents the hand calculations which were carried out to obtain initial thicknesses of foam. The remaining analyses were performed using the dynamic and nonlinear analysis capabilities of ADINA, a structural analysis finite element computer program

  14. Final Hanford Site Transuranic (TRU) Waste Characterization QA Project Plan

    International Nuclear Information System (INIS)

    GREAGER, T.M.

    2000-01-01

    The Quality Assurance Project Plan (QAPjP) has been prepared for waste characterization activities to be conducted by the Transuranic (TRU) Project at the Hanford Site to meet requirements set forth in the Waste Isolation Pilot Plan (WIPP) Hazardous Waste Facility Permit, 4890139088-TSDF, Attachment B, including Attachments B1 through B6 (WAP) (DOE, 1999a). The QAPjP describes the waste characterization requirements and includes test methods, details of planned waste sampling and analysis, and a description of the waste characterization and verification process. In addition, the QAPjP includes a description of the quality assurance/quality control (QA/QC) requirements for the waste characterization program. Before TRU waste is shipped to the WIPP site by the TRU Project, all applicable requirements of the QAPjP shall be implemented. Additional requirements necessary for transportation to waste disposal at WIPP can be found in the ''Quality Assurance Program Document'' (DOE 1999b) and HNF-2600, ''Hanford Site Transuranic Waste Certification Plan.'' TRU mixed waste contains both TRU radioactive and hazardous components, as defined in the WLPP-WAP. The waste is designated and separately packaged as either contact-handled (CH) or remote-handled (RH), based on the radiological dose rate at the surface of the waste container. RH TRU wastes are not currently shipped to the WIPP facility

  15. A relationship between leach rate of nuclear waste glass and residual amount of sodium on the glass surface

    International Nuclear Information System (INIS)

    Kamizono, Hiroshi; Banba, Tsunetaka

    1984-12-01

    Leach tests of simulated high-level waste glass were carried out in order to examine the quantitative relationship between the amount of elements on the sample surface and that in the leachate. An experimental equation was obtained expressing the relationship between the amount of Na on the sample surface and that in the leachate. This shows that it is possible in some cases to estimate the amount of Na in the leachate by measuring the amount of Na on the sample surface. One example of such an estimation was observed with the simulated high-level waste glass leached at 100 0 C in the presence of a backfill material. (author)

  16. Impacts of Solid Waste Leachate on Groundwater and Surface Water Quality

    International Nuclear Information System (INIS)

    Karim, S.

    2010-01-01

    The present investigation was carried out to assess the impacts of solid waste leachate on groundwater and surface water quality at unlined dumping site. Six leachate samples collected from different locations have average values of COD and BOD 2563 mg/L and 442 mg/L, respectively. Surface water samples were collected in two different seasons (rainy and non- rainy). Samples collected during non-rainy season were found to be more contaminated than rainy season. Soil samples collected from the depth of 1.5 m are contaminated with heavy metals (Cd, Cr, Fe and Zn) and E.coli. Presence of E.coli shows that leachate has deteriorated groundwater quality. (author)

  17. Engineered surface barriers for waste disposal sites: lysimeter facility design and construction

    International Nuclear Information System (INIS)

    Phillips, S.J.; Ruben, M.S.; Kirkham, R.R.

    1988-01-01

    A facility to evaluate performance of engineered surface carriers for confinement of buried wastes has been designed, constructed, and operations initiated. The Field Lysimeter Test Facility is located at the US Department of Energy's Hanford Site in Richland, Washington. The facility consists of 18 one-dimensional drainage and weighing lysimeters used to evaluate 7 replicated barrier treatments. Distinct layers of natural earth materials were used to construct layered soil and rock barriers in each lysimeter. These barrier designs are capable in principal of significantly reducing or precluding infiltration of meteoric water through barriers into underlying contaminated zones. This paper summarizes salient facility design and construction features used in testing of the Hanford Site's engineered surface barriers

  18. WASTE TREATMENT BUILDING SYSTEM DESCRIPTION DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    F. Habashi

    2000-06-22

    The Waste Treatment Building System provides the space, layout, structures, and embedded subsystems that support the processing of low-level liquid and solid radioactive waste generated within the Monitored Geologic Repository (MGR). The activities conducted in the Waste Treatment Building include sorting, volume reduction, and packaging of dry waste, and collecting, processing, solidification, and packaging of liquid waste. The Waste Treatment Building System is located on the surface within the protected area of the MGR. The Waste Treatment Building System helps maintain a suitable environment for the waste processing and protects the systems within the Waste Treatment Building (WTB) from most of the natural and induced environments. The WTB also confines contaminants and provides radiological protection to personnel. In addition to the waste processing operations, the Waste Treatment Building System provides space and layout for staging of packaged waste for shipment, industrial and radiological safety systems, control and monitoring of operations, safeguards and security systems, and fire protection, ventilation and utilities systems. The Waste Treatment Building System also provides the required space and layout for maintenance activities, tool storage, and administrative facilities. The Waste Treatment Building System integrates waste processing systems within its protective structure to support the throughput rates established for the MGR. The Waste Treatment Building System also provides shielding, layout, and other design features to help limit personnel radiation exposures to levels which are as low as is reasonably achievable (ALARA). The Waste Treatment Building System interfaces with the Site Generated Radiological Waste Handling System, and with other MGR systems that support the waste processing operations. The Waste Treatment Building System interfaces with the General Site Transportation System, Site Communications System, Site Water System, MGR

  19. WASTE TREATMENT BUILDING SYSTEM DESCRIPTION DOCUMENT

    International Nuclear Information System (INIS)

    Habashi, F.

    2000-01-01

    The Waste Treatment Building System provides the space, layout, structures, and embedded subsystems that support the processing of low-level liquid and solid radioactive waste generated within the Monitored Geologic Repository (MGR). The activities conducted in the Waste Treatment Building include sorting, volume reduction, and packaging of dry waste, and collecting, processing, solidification, and packaging of liquid waste. The Waste Treatment Building System is located on the surface within the protected area of the MGR. The Waste Treatment Building System helps maintain a suitable environment for the waste processing and protects the systems within the Waste Treatment Building (WTB) from most of the natural and induced environments. The WTB also confines contaminants and provides radiological protection to personnel. In addition to the waste processing operations, the Waste Treatment Building System provides space and layout for staging of packaged waste for shipment, industrial and radiological safety systems, control and monitoring of operations, safeguards and security systems, and fire protection, ventilation and utilities systems. The Waste Treatment Building System also provides the required space and layout for maintenance activities, tool storage, and administrative facilities. The Waste Treatment Building System integrates waste processing systems within its protective structure to support the throughput rates established for the MGR. The Waste Treatment Building System also provides shielding, layout, and other design features to help limit personnel radiation exposures to levels which are as low as is reasonably achievable (ALARA). The Waste Treatment Building System interfaces with the Site Generated Radiological Waste Handling System, and with other MGR systems that support the waste processing operations. The Waste Treatment Building System interfaces with the General Site Transportation System, Site Communications System, Site Water System, MGR

  20. 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

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

    International Nuclear Information System (INIS)

    T. Wolery

    2005-01-01

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

  2. Preparation and Surface Sizing Application of Sizing Agent Based on Collagen from Leather Waste

    Directory of Open Access Journals (Sweden)

    Xuechuan Wang

    2015-09-01

    Full Text Available Collagen extracted from leather waste was modified with maleic anhydride. Then, using ammonium persulfate as an initiator, by pre-modifying collagen reacted with styrene and ethyl acrylate monomers, a vinyl-grafted collagen sizing agent (VGCSA for paper was prepared. Before the experiment, the performance of VGCSA was tested and VGCSA emulsion was applied to the surface sizing of the corrugated paper. Effects of the amount of VGCSA, the compound proportion of VGCSA, and starch and styrene-acrylic emulsion were studied relative to paper properties. The morphological changes of the paper before and after sizing were characterized by SEM. It was found that the collagen reacted with styrene and ethyl acrylate monomers. Through the grafting of vinyl and collagen, the crystallinity and thermal stability of VGCSA increased. The structure of VGCSA was spherical with a uniform size, and the average particle size was approximately 350 to 400 nm. After being sized, the surface fibers of paper became smooth and orderly. The optimal sizing of VGCSA was 8 g/m2. The optimal proportion of VGCSA with starch was 4:6, and the optimal proportion of VGCSA with SAE was 2:8. The research indicates that collagen extracted from leather waste could be used as a biomaterial, and environmental and economic benefits could be created as well.

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

    Energy Technology Data Exchange (ETDEWEB)

    T. Wolery

    2005-02-22

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

  4. Control of water infiltration into near surface LLW [low-level radioactive waste] disposal units

    International Nuclear Information System (INIS)

    Schulz, R.K.; Ridky, R.W.; O'Donnell, E.O.

    1990-12-01

    Three kinds of waste disposal unit covers a barriers to water infiltration are being investigated. They are: (1) resistive layer barrier, (2) conductive layer barrier, and (3) bioengineering management. The resistive layer barrier consists of compacted earthen material (e.g. clay). The conductive layer barrier consists of a conductive layer in conjunction with a capillary break. As long as unsaturated flow conditions are maintained the conductive layer will wick water around the capillary break. Below grade layered covers such as (1) and (2) will fail if there is appreciable subsidence of the cover. Remedial action for this kind of failure will be difficult. A surface cover, called bioengineering management, is meant to overcome this problem. The bioengineering management surface barrier is easily repairable if damaged by subsidence; therefore, it could be the system of choice under active subsidence conditions. The bioengineering management procedure also has been shown to be effective in dewatering saturated trenches and could be used for remedial action efforts. After cessation of subsidence, that procedure could be replaced by a resistive layer barrier, or perhaps even better, a resistive layer barrier/conductive layer barrier system. This latter system would then give long-term effective protection against water entry to waste and without institutional care. These various concepts are being assessed in six large (70ft x 45ft x 10ft each) lysimeters at Beltsville, Maryland. 6 refs., 20 figs.,

  5. Seismic surface wave tomography of waste sites. 1997 annual progress report

    International Nuclear Information System (INIS)

    Long, T.L.

    1997-01-01

    'The objective of the Seismic Surface Wave Tomography of Waste Sites is to develop a robust technique for field acquisition and analysis of surface wave data for the interpretation of shallow structures, such as those associated with the burial of wastes. The analysis technique is to be developed and tested on an existing set of seismic data covering the K-901 burial site at the East Tennessee Technology Park. Also, a portable prototype for a field acquisition system will be designed and developed to obtain additional data for analysis and testing of the technique. The K-901 data have been examined and a preliminary Single Valued Decomposition inversion has been obtained. The preliminary data indicates a need for additional seismic data to ground-truth the inversion. The originally proposed gravity data acquisition has been dropped because sufficient gravity data are now available for a preliminary analysis and because the seismic data are considered more critical to the interpretation. The proposed prototype for the portable acquisition and analysis system was developed during the first year and will be used in part of the acquisition of additional seismic data.'

  6. Types of safety assessments of near surface repository for radioactive waste

    International Nuclear Information System (INIS)

    Mateeva, M.

    2004-01-01

    The purpose of this article is to presents the classification of different types safety assessments of near surface repository for low and intermediate level radioactive waste substantiated with results of safety assessments generated in Bulgaria. The different approach of safety assessments applied for old existing repository as well as for site selection for construction new repository is outlined. The regulatory requirements in Bulgaria define three main types of assessments: Safety assessment; Technical substation of repository safety; Assessment of repository influence on environment that is in form of report prepared from the Ministry of environment and waters on the base of results obtained in two first types of assessments. Additionally first type is subdivided in three categories - preliminary safety assessment, safety assessment and post closure safety assessment, which are generated using deterministic approach. The technical substation of repository safety is generated using probabilistic approach. Safety assessment results that are presented here are based on evaluation of existing old repository type 'Radon' in Novi Han and real site selection procedure for new near surface repository for low and intermediate level radioactive waste from nuclear power station in Kozloduy. The important role of safety assessment for improvement the repository safety as well as for repository licensing, correct site selection and right choice of engineer barriers and repository design is discussed using generated results. (author)

  7. Unvented Drum Handling Plan

    International Nuclear Information System (INIS)

    MCDONALD, K.M.

    2000-01-01

    This drum-handling plan proposes a method to deal with unvented transuranic drums encountered during retrieval of drums. Finding unvented drums during retrieval activities was expected, as identified in the Transuranic (TRU) Phase I Retrieval Plan (HNF-4781). However, significant numbers of unvented drums were not expected until excavation of buried drums began. This plan represents accelerated planning for management of unvented drums. A plan is proposed that manages unvented drums differently based on three categories. The first category of drums is any that visually appear to be pressurized. These will be vented immediately, using either the Hanford Fire Department Hazardous Materials (Haz. Mat.) team, if such are encountered before the facilities' capabilities are established, or using internal capabilities, once established. To date, no drums have been retrieved that showed signs of pressurization. The second category consists of drums that contain a minimal amount of Pu isotopes. This minimal amount is typically less than 1 gram of Pu, but may be waste-stream dependent. Drums in this category are assayed to determine if they are low-level waste (LLW). LLW drums are typically disposed of without venting. Any unvented drums that assay as TRU will be staged for a future venting campaign, using appropriate safety precautions in their handling. The third category of drums is those for which records show larger amounts of Pu isotopes (typically greater than or equal to 1 gram of Pu). These are assumed to be TRU and are not assayed at this point, but are staged for a future venting campaign. Any of these drums that do not have a visible venting device will be staged awaiting venting, and will be managed under appropriate controls, including covering the drums to protect from direct solar exposure, minimizing of container movement, and placement of a barrier to restrict vehicle access. There are a number of equipment options available to perform the venting. The

  8. Long-term management USDOE transuranic waste

    International Nuclear Information System (INIS)

    Bennett, W.S.; Gilbert, K.V.; Lowrey, R.Y.

    1982-01-01

    Activities for permanent disposal of US DOE TRU waste are presently focused on newly generated and stored waste. The buried waste and contaminated soils pose no near term problem. Decisions on any possible actions for these wastes will be deferred until the newly generated and stored wastes are being placed into disposal on a routine basis. Several elements must be in place before such disposal can become routine. These elements consist of: a disposal facility; waste acceptance criteria; waste certification mechanisms; waste processing facilities; and a waste transportation system. Each of these elements has been the subject of considerable activity in the recent past. Progress and plans for each element are summarized. As of January 1981, DOE has 60,500 m 3 of waste classified as Transuranic waste (TRU) in retrievable storage, and projects that additional TRU waste will be generated at an average rate of 4500 m 3 per year for the next 10 years. Over 99% of this waste is contact handled, with the remainder being remote handled, i.e., surface radiation dose levels exceeding 200 mrem/h. An estimated 273,000 m 3 of TRU waste were placed in shallow land burial prior to establishment of the 1970 policy. In addition, large quantities of soil at DOE sites are contaminated with TRU elements due to disposal of liquid wastes and by contact of soil with solid, buried waste whose original containers are now badly degraded. Possibly as much as 10,000,000 m 3 of soil are contaminated above 10 nCi/gm. Less than 1,000,000 m 3 are estimated to be contaminated above 100 nCi/gm

  9. Coal combustion waste management at landfills and surface impoundments 1994-2004.

    Energy Technology Data Exchange (ETDEWEB)

    Elcock, D.; Ranek, N. L.; Environmental Science Division

    2006-09-08

    On May 22, 2000, as required by Congress in its 1980 Amendments to the Resource Conservation and Recovery Act (RCRA), the U.S. Environmental Protection Agency (EPA) issued a Regulatory Determination on Wastes from the Combustion of Fossil Fuels. On the basis of information contained in its 1999 Report to Congress: Wastes from the Combustion of Fossil Fuels, the EPA concluded that coal combustion wastes (CCWs), also known as coal combustion by-products (CCBs), did not warrant regulation under Subtitle C of RCRA, and it retained the existing hazardous waste exemption for these materials under RCRA Section 3001(b)(3)(C). However, the EPA also determined that national regulations under Subtitle D of RCRA were warranted for CCWs that are disposed of in landfills or surface impoundments. The EPA made this determination in part on the basis of its findings that 'present disposal practices are such that, in 1995, these wastes were being managed in 40 percent to 70 percent of landfills and surface impoundments without reasonable controls in place, particularly in the area of groundwater monitoring; and while there have been substantive improvements in state regulatory programs, we have also identified gaps in State oversight' (EPA 2000). The 1999 Report to Congress (RTC), however, may not have reflected the changes in CCW disposal practices that occurred since the cutoff date (1995) of its database and subsequent developments. The U.S. Department of Energy (DOE) and the EPA discussed this issue and decided to conduct a joint DOE/EPA study to collect new information on the recent CCW management practices by the power industry. It was agreed that such information would provide a perspective on the chronological adoption of control measures in CCW units based on State regulations. A team of experts from the EPA, industry, and DOE (with support from Argonne National Laboratory) was established to develop a mutually acceptable approach for collecting and analyzing data

  10. Remaining Sites Verification Package for the 100-B-1 Surface Chemical and Solid Waste Dumping Area,. Attachment to Waste Site Reclassification Form 2006-003

    International Nuclear Information System (INIS)

    Carlson, R.A.

    2006-01-01

    The 100-B-1 waste site was a dumping site that was divided into two areas. One area was used as a laydown area for construction materials, and the other area was used as a chemical dumping area. The 100-B-1 Surface Chemical and Solid Waste Dumping Area site meets the remedial action objectives specified in the Remaining Sites ROD. The results demonstrate that residual contaminant concentrations support future unrestricted land uses that can be represented by a rural-residential scenario. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River

  11. Origin and evolution of cup-shaped structures on leached nuclear waste containment glass surfaces

    International Nuclear Information System (INIS)

    Dubois, C.; Villa, F.; Chambaudet, A.; Vernaz, E.

    1994-01-01

    A three-dimensional surface microanalysis system equipped with a sensitive topographical probe was used to quantify the evolution of cup-shaped structures formed by aqueous leaching of nuclear waste containment glass. A model of the dissolution phenomenon provides satisfactory correlation between calculated and measured cup radius and depth. Dissolution cups form from cracks on the initially cut glass surface. Large cracks control the phenomenon by forming the largest cups, which gradually absorb smaller ones. The evolution of the size and shape of the dissolution cups was described by a model that assumes a constant dissolution rate on the surface, diminishing with crack depth. The best fit with the experimental data was obtained with a dissolution rate one hundred times lower at the bottom of the crack than at the surface. Moreover, it is predictable that all the cups will gradually disappear as they grow larger and flatter over a leaching period of some 2 years, for the glass composition and experimental leaching procedures used in this work

  12. Mass transfer of CO2 to groundwaters from a near-surface waste disposal site

    International Nuclear Information System (INIS)

    Caron, F.; Wilkinson, S.R.; Manni, G.; Torok, J.

    1995-01-01

    Gaseous 14 CO 2 originating from buried low-level radioactive wastes (LLRW) in a near-surface disposal site can be released to the environment via two major paths: gas-phase diffusion through soils to the atmosphere, and dissolution in groundwater, followed by aqueous migration. Aqueous migration would give the highest dose to an individual, especially if C-14 was converted to an organic form and ingested. Gaseous diffusion would give a lower dose, largely because of atmospheric dispersion and dilution. The objective of this study was to develop the capability to estimate which of the two paths will likely be dominant for typical near-surface disposal facilities. The main missing parameter for making this estimate was a mass-transfer coefficient (K L ) of 14 CO 2 to groundwaters, which was determined experimentally using a large sand box. The K L thus determined was approximately 10 to 20 times smaller than for an open liquid surface. This suggests that there is a potential resistance to mass transfer, probably caused by the capillary fringe. The value obtained was incorporated into a simple model of CO 2 transport around a typical near-surface disposal site. The model suggests that CO 2 transport via both gaseous release and aqueous migration paths are of similar magnitude for a repository located ∼2 m above the water table. (author). 11 refs., 2 tabs., 2 figs

  13. The surface water model for assessing Canada's nuclear fuel waste disposal concept

    International Nuclear Information System (INIS)

    Bird, G.A.; Stephenson, M.; Cornett, R.J.

    1993-01-01

    Canada's Nuclear Fuel Waste Management Program (NFWMP) is investigating the concept of disposal of nuclear fuel waste in a vault excavated deep in crystalline rock on the Canadian Shield. Probabilistic vault, geosphere, and biosphere models are implemented using Monte Carlo simulation techniques to trace nuclides transported in groundwater to the surface environment and humans far into the future. This paper describes the surface water submodel and its parameter values, sensitivity analysis, and validation. The surface water model is a simple, time-dependent, mass balance model of a lake that calculates radioactive and stable isotope contaminant concentrations in lake water and sediment. These concentrations are input to the other submodels and used to predict the radiological dose to humans and other biota. Parameter values in the model are based on the literature and the author's own data, and are generic to Canadian Shield lakes. Most parameters are represented by log normally distributed probability density functions. Sensitivity analysis indicates that nuclide concentrations in lake water and sediment are governed primarily by hydrological flushing with catchment area being the most important parameter. When catchment area is held constant lake area and nuclide transfer rate from water to sediment strongly influence concentrations in both water and sediment. For volatile nuclides, gaseous evasion also has a marked influence on concentrations in both water and sediment, whereas sedimentation rate strongly influences sediment nuclide concentrations. Validation tests demonstrate that the models predictions for 60 Co, 134 Cs, 3 H, P, Cd and Ca are consistent with empirical data when uncertainties are taken into account

  14. ENSI's view on NTB 08-02 'Report on how to handle the recommendations in expert reports and statements of views on the demonstration of the nuclear waste disposal programme'

    International Nuclear Information System (INIS)

    2012-03-01

    In the framework of a review of the project Opalinus clay for the disposal of spent fuel assemblies, high-level radioactive waste and long living intermediate-level waste, a team of experts worked out numerous recommendations to be considered in the future. In 2006, the Swiss Federal Council stated that the realization of a radioactive waste disposal had been demonstrated to be possible. At the same time, the operators of the Swiss nuclear power plants were instructed to answer the open questions and to decide how they would proceed in the future. Similarly, the National Cooperative for the Disposal of Radioactive Waste (NAGRA) presented a report which ENSI had to review and evaluate. This review is reported in the present document. The NAGRA report, as well as ENSI's comments, is structured in accordance with the 200 recommendations made by the experts' team. ENSI states that the recommendations and all open questions of the experts' team were treated completely and transparently by NAGRA; they were considered in the elaboration of proposals for underground storage sites. ENSI declares that the recommendations for the demonstration of the feasibility of the waste disposal were given the correct priority. An important mile stone for more precise indications about site specific data will be the future request for the granting of the framework license, at the end of the site selection process, according to the Sectoral Plan for Geologic Deep Repositories. The recommendations about Opalinus clay as host rock are of general nature. They concern a repository for high-level (HLW) waste as well as for low- and intermediate-level (ILW) wastes. They also concern techniques independent of the host rock, like how to handle filling and sealing material. In step 1 of the Sectoral Plan, besides the Opalinus clay NAGRA proposed exclusively clay-rich host rocks with similar properties. Opalinus clay was proposed for 5 of the 6 potential disposal sites for a ILW repository. Some

  15. Welding method by remote handling

    International Nuclear Information System (INIS)

    Hashinokuchi, Minoru.

    1994-01-01

    Water is charged into a pit (or a water reservoir) and an article to be welded is placed on a support in the pit by remote handling. A steel plate is disposed so as to cover the article to be welded by remote handling. The welding device is positioned to the portion to be welded and fixed in a state where the article to be welded is shielded from radiation by water and the steel plate. Water in the pit is drained till the portion to be welded is exposed to the atmosphere. Then, welding is conducted. After completion of the welding, water is charged again to the pit and the welding device and fixing jigs are decomposed in a state where the article to be welded is shielded again from radiation by water and the steel plate. Subsequently, the steel plate is removed by remote handling. Then, the article to be welded is returned from the pit to a temporary placing pool by remote handling. This can reduce operator's exposure. Further, since the amount of the shielding materials can be minimized, the amount of radioactive wastes can be decreased. (I.N.)

  16. Kinetic and mechanism studies of the adsorption of lead onto waste cow bone powder (WCBP) surfaces.

    Science.gov (United States)

    Cha, Jihoon; Cui, Mingcan; Jang, Min; Cho, Sang-Hyun; Moon, Deok Hyun; Khim, Jeehyeong

    2011-01-01

    This study examines the adsorption isotherms, kinetics and mechanisms of Pb²(+) sorption onto waste cow bone powder (WCBP) surfaces. The concentrations of Pb²(+) in the study range from 10 to 90 mg/L. Although the sorption data follow the Langmuir and Freundlich isotherm, a detailed examination reveals that surface sorption or complexation and co-precipitation are the most important mechanisms, along with possibly ion exchange and solid diffusion also contributing to the overall sorption process. The co-precipitation of Pb²(+) with the calcium hydroxyapatite (Ca-HAP) is implied by significant changes in Ca²(+) and PO₄³⁻ concentrations during the metal sorption processes. The Pb²(+) sorption onto the WCBP surface by metal complexation with surface functional groups such as ≡ POH. The major metal surface species are likely to be ≡ POPb(+). The sorption isotherm results indicated that Pb²(+) sorption onto the Langmuir and Freundlich constant q(max) and K( F ) is 9.52 and 8.18 mg g⁻¹, respectively. Sorption kinetics results indicated that Pb²(+) sorption onto WCBP was pseudo-second-order rate constants K₂ was 1.12 g mg⁻¹ h⁻¹. The main mechanism is adsorption or surface complexation (≡POPb(+): 61.6%), co-precipitation or ion exchange [Ca₃(.)₉₃ Pb₁(.)₀₇ (PO₄)₃ (OH): 21.4%] and other precipitation [Pb 50 mg L⁻¹ and natural pH: 17%). Sorption isotherms showed that WCBP has a much higher Pb²(+) removal rate in an aqueous solution; the greater capability of WCBP to remove aqueous Pb²(+) indicates its potential as another promising way to remediate Pb²(+)-contaminated media.

  17. Autogenous Tumbling Media Assessment to Clean Weathered Surfaces of Waste-Rock Particles from a Basalt Quarry

    Directory of Open Access Journals (Sweden)

    Baran Tufan

    2015-06-01

    Full Text Available In this study, the optimum feed composition in autogenous tumbling of basalt waste-rock particles to clean their weathered surface was determined. The weathered surfaces of basalt are generally cut out consequent to extraction of basalt columns in quarry operations. The inefficiently cut out portions of basalt cause formation of huge quarry waste dumps causing visual pollution on roadsides. Mixtures of different particle size fractions of basalt waste-rock particles were experimented to achieve the optimum feed material composition. The minimum loss of commercially available basalt particles and maximum clear surface was intended. The results were compared with respect to weight loss (% and reflectance values of used and generated samples.

  18. Development of performance assessment methodology for establishment of quantitative acceptance criteria of near-surface radioactive waste disposal

    Energy Technology Data Exchange (ETDEWEB)

    Kim, C. R.; Lee, E. Y.; Park, J. W.; Chang, G. M.; Park, H. Y.; Yeom, Y. S. [Korea Hydro and Nuclear Power Co., Ltd., Seoul (Korea, Republic of)

    2002-03-15

    The contents and the scope of this study are as follows : review of state-of-the-art on the establishment of waste acceptance criteria in foreign near-surface radioactive waste disposal facilities, investigation of radiological assessment methodologies and scenarios, investigation of existing models and computer codes used in performance/safety assessment, development of a performance assessment methodology(draft) to derive quantitatively radionuclide acceptance criteria of domestic near-surface disposal facility, preliminary performance/safety assessment in accordance with the developed methodology.

  19. Investigation of waste biomass co-pyrolysis with petroleum sludge using a response surface methodology.

    Science.gov (United States)

    Hu, Guangji; Li, Jianbing; Zhang, Xinying; Li, Yubao

    2017-05-01

    The treatment of waste biomass (sawdust) through co-pyrolysis with refinery oily sludge was carried out in a fixed-bed reactor. Response surface method was applied to evaluate the main and interaction effects of three experimental factors (sawdust percentage in feedstock, temperature, and heating rate) on pyrolysis oil and char yields. It was found that the oil and char yields increased with sawdust percentage in feedstock. The interaction between heating rate and sawdust percentage as well as between heating rate and temperature was significant on the pyrolysis oil yield. The higher heating value of oil originated from sawdust during co-pyrolysis at a sawdust/oily sludge ratio of 3:1 increased by 5 MJ/kg as compared to that during sawdust pyrolysis alone, indicating a synergistic effect of co-pyrolysis. As a result, petroleum sludge can be used as an effective additive in the pyrolysis of waste biomass for improving its energy recovery. Copyright © 2017 Elsevier Ltd. All rights reserved.

  20. Trench 'bathtubbing' and surface plutonium contamination at a legacy radioactive waste site.

    Science.gov (United States)

    Payne, Timothy E; Harrison, Jennifer J; Hughes, Catherine E; Johansen, Mathew P; Thiruvoth, Sangeeth; Wilsher, Kerry L; Cendón, Dioni I; Hankin, Stuart I; Rowling, Brett; Zawadzki, Atun

    2013-01-01

    Radioactive waste containing a few grams of plutonium (Pu) was disposed between 1960 and 1968 in trenches at the Little Forest Burial Ground (LFBG), near Sydney, Australia. A water sampling point installed in a former trench has enabled the radionuclide content of trench water and the response of the water level to rainfall to be studied. The trench water contains readily measurable Pu activity (~12 Bq/L of (239+240)Pu in 0.45 μm-filtered water), and there is an associated contamination of Pu in surface soils. The highest (239+240)Pu soil activity was 829 Bq/kg in a shallow sample (0-1 cm depth) near the trench sampling point. Away from the trenches, the elevated concentrations of Pu in surface soils extend for tens of meters down-slope. The broader contamination may be partly attributable to dispersion events in the first decade after disposal, after which a layer of soil was added above the trenched area. Since this time, further Pu contamination has occurred near the trench-sampler within this added layer. The water level in the trench-sampler responds quickly to rainfall and intermittently reaches the surface, hence the Pu dispersion is attributed to saturation and overflow of the trenches during extreme rainfall events, referred to as the 'bathtub' effect.

  1. Trench ‘Bathtubbing’ and Surface Plutonium Contamination at a Legacy Radioactive Waste Site

    Science.gov (United States)

    2013-01-01

    Radioactive waste containing a few grams of plutonium (Pu) was disposed between 1960 and 1968 in trenches at the Little Forest Burial Ground (LFBG), near Sydney, Australia. A water sampling point installed in a former trench has enabled the radionuclide content of trench water and the response of the water level to rainfall to be studied. The trench water contains readily measurable Pu activity (∼12 Bq/L of 239+240Pu in 0.45 μm-filtered water), and there is an associated contamination of Pu in surface soils. The highest 239+240Pu soil activity was 829 Bq/kg in a shallow sample (0–1 cm depth) near the trench sampling point. Away from the trenches, the elevated concentrations of Pu in surface soils extend for tens of meters down-slope. The broader contamination may be partly attributable to dispersion events in the first decade after disposal, after which a layer of soil was added above the trenched area. Since this time, further Pu contamination has occurred near the trench-sampler within this added layer. The water level in the trench-sampler responds quickly to rainfall and intermittently reaches the surface, hence the Pu dispersion is attributed to saturation and overflow of the trenches during extreme rainfall events, referred to as the ‘bathtub’ effect. PMID:24256473

  2. Surface complexation modeling of uranyl adsorption on corrensite from the Waste Isolation Pilot Plant Site

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang-Won; Leckie, J.O. [Stanford Univ., CA (United States); Siegel, M.D. [Sandia National Labs., Albuquerque, NM (United States)

    1995-09-01

    Corrensite is the dominant clay mineral in the Culebra Dolomite at the Waste Isolation Pilot Plant. The surface characteristics of corrensite, a mixed chlorite/smectite clay mineral, have been studied. Zeta potential measurements and titration experiments suggest that the corrensite surface contains a mixture of permanent charge sites on the basal plane and SiOH and AlOH sites with a net pH-dependent charge at the edge of the clay platelets. Triple-layer model parameters were determined by the double extrapolation technique for use in chemical speciation calculations of adsorption reactions using the computer program HYDRAQL. Batch adsorption studies showed that corrensite is an effective adsorbent for uranyl. The pH-dependent adsorption behavior indicates that adsorption occurs at the edge sites. Adsorption studies were also conducted in the presence of competing cations and complexing ligands. The cations did not affect uranyl adsorption in the range studied. This observation lends support to the hypothesis that uranyl adsorption occurs at the edge sites. Uranyl adsorption was significantly hindered by carbonate. It is proposed that the formation of carbonate uranyl complexes inhibits uranyl adsorption and that only the carbonate-free species adsorb to the corrensite surface. The presence of the organic complexing agents EDTA and oxine also inhibits uranyl sorption.

  3. Surface complexation modeling of uranyl adsorption on corrensite from the Waste Isolation Pilot Plant Site

    International Nuclear Information System (INIS)

    Park, Sang-Won; Leckie, J.O.; Siegel, M.D.

    1995-09-01

    Corrensite is the dominant clay mineral in the Culebra Dolomite at the Waste Isolation Pilot Plant. The surface characteristics of corrensite, a mixed chlorite/smectite clay mineral, have been studied. Zeta potential measurements and titration experiments suggest that the corrensite surface contains a mixture of permanent charge sites on the basal plane and SiOH and AlOH sites with a net pH-dependent charge at the edge of the clay platelets. Triple-layer model parameters were determined by the double extrapolation technique for use in chemical speciation calculations of adsorption reactions using the computer program HYDRAQL. Batch adsorption studies showed that corrensite is an effective adsorbent for uranyl. The pH-dependent adsorption behavior indicates that adsorption occurs at the edge sites. Adsorption studies were also conducted in the presence of competing cations and complexing ligands. The cations did not affect uranyl adsorption in the range studied. This observation lends support to the hypothesis that uranyl adsorption occurs at the edge sites. Uranyl adsorption was significantly hindered by carbonate. It is proposed that the formation of carbonate uranyl complexes inhibits uranyl adsorption and that only the carbonate-free species adsorb to the corrensite surface. The presence of the organic complexing agents EDTA and oxine also inhibits uranyl sorption

  4. Defense waste transportation: cost and logistics studies

    International Nuclear Information System (INIS)

    Andrews, W.B.; Cole, B.M.; Engel, R.L.; Oylear, J.M.

    1982-08-01

    Transportation of nuclear wastes from defense programs is expected to significantly increase in the 1980s and 1990s as permanent waste disposal facilities come into operation. This report uses models of the defense waste transportation system to quantify potential transportation requirements for treated and untreated contact-handled transuranic (CH-TRU) wastes and high-level defense wastes (HLDW). Alternative waste management strategies in repository siting, waste retrieval and treatment, treatment facility siting, waste packaging and transportation system configurations were examined to determine their effect on transportation cost and hardware requirements. All cost estimates used 1980 costs. No adjustments were made for future changes in these costs relative to inflation. All costs are reported in 1980 dollars. If a single repository is used for defense wastes, transportation costs for CH-TRU waste currently in surface storage and similar wastes expected to be generated by the year 2000 were estimated to be 109 million dollars. Recovery and transport of the larger buried volumes of CH-TRU waste will increase CH-TRU waste transportation costs by a factor of 70. Emphasis of truck transportation and siting of multiple repositories would reduce CH-TRU transportation costs. Transportation of HLDW to repositories for 25 years beginning in 1997 is estimated to cost $229 M in 1980 costs and dollars. HLDW transportation costs could either increase or decrease with the selection of a final canister configuration. HLDW transportation costs are reduced when multiple repositories exist and emphasis is placed on truck transport

  5. Formation of filtration fields close to near-surface radioactive waste storages

    International Nuclear Information System (INIS)

    Mart'yanov, V.V.

    2008-01-01

    Data on the formation of filtration fields in the location of near-surface radioactive waste storages for the conditions of uniformly isotropic properties of bearing strata are demonstrated. The possibility for changing parameters of mean-caused ground flow depending on water permeability of the storages and their dimensions in plan is noted. Comparison of different filtration fields permits to determine a state of its isolating properties. Assessment criteria of the storage engineering barriers integrity are given. Conditions for uniformly isotropic properties of bearing strata by three scenarios, when engineering barriers of the storage are waterproof, distracted or lost protective properties in full, have been determined. Changing filtration field, geochemical and radiochemical situations in bearing strata are noted to represent one of basic characteristics of the integrity of the storage [ru

  6. Conceptual model to determine maximum activity of radioactive waste in near-surface disposal facilities

    International Nuclear Information System (INIS)

    Iarmosh, I.; Olkhovyk, Yu.

    2016-01-01

    For development of the management strategy for radioactive waste to be placed in near - surface disposal facilities (NSDF), it is necessary to justify long - term safety of such facilities. Use of mathematical modelling methods for long - term forecasts of radwaste radiation impact and assessment of radiation risks from radionuclides migration can help to resolve this issue. The purpose of the research was to develop the conceptual model for determining the maximum activity of radwaste to be safely disposed in the NSDF and to test it in the case of Lot 3 Vector NSDF (Chornobyl exclusion zone). This paper describes an approach to the development of such a model. The conceptual model of "9"0 Sr migration from Lot 3 through aeration zone and aquifer soils was developed. The results of modelling are shown. The proposals on further steps for the model improvement were developed

  7. Enhanced Production of Cellulase from Pineapple Waste by Response Surface Methodology

    Directory of Open Access Journals (Sweden)

    P. Saravanan

    2013-01-01

    Full Text Available Optimization of the media components for cellulase production using Trichoderma reesei was carried out. The optimization of cellulase production using pineapple waste as substrate was performed with statistical methodology based on experimental designs. The screening of nutrients and their influence on the cellulase production was studied using a Plackett-Burman design. Avicel, soybean cake flour, KH2PO4, and yeast extract were found to have the positive influence for the production of cellulase. The selected components were optimized using response surface methodology. The optimum concentrations are avicel: 26.5 g/L, soybean cake flour: 22.5 g/L, KH2PO4: 4.5 g/L, and yeast extract: 12.3 g/L. A maximum cellulase activity of 8.61 IU/mL was obtained under the optimized medium in the validation experiment.

  8. The dispersal and impact of salt from surface storage piles the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Reith, C.C.; Louderbough, E.T.

    1986-01-01

    A comprehensive program of ecological studies occurs at the Waste Isolation Pilot Plant (WIPP) in an effort to detect and quantify impacts of excavated salt which is stored on the surface in two piles: one having originated in 1980, the other in 1984. Both piles are surrounded by berms which channel runoff to holding ponds, so nearly all dispersal is due to the resuspension, transport, and deposition of salt particles by wind. Ecological parameters which have been monitored since 1984 include: visual evidence (via photography), soil properties, microbial activity, leaf-litter decomposition, seedling emergence, plant foliar cover, and plant species diversity. These are periodically assessed at experimental plots near the salt piles, and at control plots several kilometers away

  9. Use of artificial barriers in a site for surface storage of radioactive waste

    International Nuclear Information System (INIS)

    Gros, J.C.; Madoz-Escande, C.; Metivier, J.M.; Grimaud, P.

    1990-01-01

    The objective is the on site study of the influence of an injection screen on the flow in a water table of a porous medium, in order to improve the safety of a surface radioactive waste storage site. A hydrodispersive study has provided information for the definition of the role of the screen: the transfer times of the pollutant in the water table are increased by a factor of 2 and, in comparison, the concentration are clearly reduced by a factor of 10. The implantation of an injection screen in the ground should result in an improvement in the restrictive quality of the barrier and the contamination of an aquifer should be slower without interruption to the flow

  10. Scientific and technical basis for the near surface disposal of low and intermediate level waste

    International Nuclear Information System (INIS)

    2002-01-01

    This report presents an overview of the scientific and technical basis for the disposal of low- and intermediate-level radioactive waste in near surface repositories. The focus is on basic principles, approaches, methodologies and technical criteria that can be used to develop and assess the performance of a disposal facility, and for building confidence in repository safety. This includes consideration of the multiple barrier concept, the performance of engineered barriers, the role of natural barriers and the development of a safety case. The emphasis is on defining the conditions relevant to the containment of the radionuclides in the repository and the processes that may affect the integrity of the engineered barriers. Both generic and specific data requirements for repository development and the assurance of safety are addressed. A large number of bibliographical references are given to support the information provided in this report

  11. Egg shell waste as heterogeneous nanocatalyst for biodiesel production: Optimized by response surface methodology.

    Science.gov (United States)

    Pandit, Priti R; Fulekar, M H

    2017-08-01

    Worldwide consumption of hen eggs results in availability of large amount of discarded egg waste particularly egg shells. In the present study, the waste shells were utilized for the synthesis of highly active heterogeneous calcium oxide (CaO) nanocatalyst to transesterify dry biomass into methyl esters (biodiesel). The CaO nanocatalyst was synthesied by calcination-hydration-dehydration technique and fully characterized by infrared spectroscopy, X-ray powder diffraction (XRD), scanning electron microscope (SEM), transmission electron microscope (TEM), brunauer-emmett-teller (BET) elemental and thermogravimetric analysis. TEM image showed that the nano catalyst had spherical shape with average particle size of 75 nm. BET analysis indicated that the catalyst specific surface area was 16.4 m 2  g -1 with average pore diameter of 5.07 nm. The effect of nano CaO catalyst was investigated by direct transesterification of dry biomass into biodiesel along with other reaction parameters such as catalyst ratio, reaction time and stirring rate. The impact of the transesterification reaction parameters and microalgal biodiesel yield were analyzed by response surface methodology based on a full factorial, central composite design. The significance of the predicted mode was verified and 86.41% microalgal biodiesel yield was reported at optimal parameter conditions 1.7% (w/w), catalyst ratio, 3.6 h reaction time and stirring rate of 140.6 rpm. The biodiesel conversion was determined by 1 H nuclear magnetic resonance spectroscopy (NMR). The fuel properties of prepared biodiesel were found to be highly comply with the biodiesel standard ASTMD6751 and EN14214. Copyright © 2017 Elsevier Ltd. All rights reserved.

  12. Test phase plan for the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    1993-03-01

    The US Department of Energy (DOE) has prepared this Test Phase Plan for the Waste Isolation Pilot Plant to satisfy the requirements of Public Law 102-579, the Waste Isolation Pilot Plant (WIPP) Land Withdrawal Act (LWA). The Act provides seven months after its enactment for the DOE to submit this Plan to the Environmental Protection Agency (EPA) for review. A potential geologic repository for transuranic wastes, including transuranic mixed wastes, generated in national-defense activities, the WIPP is being constructed in southeastern New Mexico. Because these wastes remain radioactive and chemically hazardous for a very long time, the WIPP must provide safe disposal for thousands of years. The DOE is developing the facility in phases. Surface facilities for receiving waste have been built and considerable underground excavations (2150 feet below the surface) that are appropriate for in-situ testing, have been completed. Additional excavations will be completed when they are required for waste disposal. The next step is to conduct a test phase. The purpose of the test phase is to develop pertinent information and assess whether the disposal of transuranic waste and transuranic mixed waste in the planned WIPP repository can be conducted in compliance with the environmental standards for disposal and with the Solid Waste Disposal Act (SWDA) (as amended by RCRA, 42 USC. 6901 et. seq.). The test phase includes laboratory experiments and underground tests using contact-handled transuranic waste. Waste-related tests at WIPP will be limited to contact-handled transuranic and simulated wastes since the LWA prohibits the transport to or emplacement of remote-handled transuranic waste at WIPP during the test phase

  13. Definition of intrusion scenarios and example concentration ranges for the disposal of near-surface waste at the Hanford Site

    International Nuclear Information System (INIS)

    Aaberg, R.L.; Kennedy, W.E. Jr.

    1990-10-01

    The US Department of Energy (DOE) is in the process of conducting performance assessments of its radioactive waste sites and disposal systems to ensure that public health and safety are protected, the environment is preserved, and that no remedial actions after disposal are required. Hanford Site low-level waste performance assessments are technical evaluations of waste sites or disposal systems that provide a basis for making decisions using established criteria. The purpose of this document is to provide a family of scenarios to be considered when calculating radionuclide exposure to individuals who may inadvertently intrude into near-surface waste disposal sites. Specific performance assessments will use modifications of the general scenarios described here to include additional site/system details concerning the engineering design, waste form, inventory, and environmental setting. This document also describes and example application of the Hanford-specific scenarios in the development of example concentration ranges for the disposal of near-surface wastes. The overall goal of the example calculations is to illustrate the application of the scenarios in a performance assessment to assure that people in the future cannot receive a dose greater than an established limit. 24 refs., 2 figs., 5 tabs

  14. Handling Procedures of Vegetable Crops

    Science.gov (United States)

    Perchonok, Michele; French, Stephen J.

    2004-01-01

    The National Aeronautics and Space Administration (NASA) is working towards future long duration manned space flights beyond low earth orbit. The duration of these missions may be as long as 2.5 years and will likely include a stay on a lunar or planetary surface. The primary goal of the Advanced Food System in these long duration exploratory missions is to provide the crew with a palatable, nutritious, and safe food system while minimizing volume, mass, and waste. Vegetable crops can provide the crew with added nutrition and variety. These crops do not require any cooking or food processing prior to consumption. The vegetable crops, unlike prepackaged foods, will provide bright colors, textures (crispy), and fresh aromas. Ten vegetable crops have been identified for possible use in long duration missions. They are lettuce, spinach, carrot, tomato, green onion, radish, bell pepper, strawberries, fresh herbs, and cabbage. Whether these crops are grown on a transit vehicle (e.g., International Space Station) or on the lunar or planetary surface, it will be necessary to determine how to safely handle the vegetables while maintaining acceptability. Since hydrogen peroxide degrades into water and oxygen and is generally recognized as safe (GRAS), hydrogen peroxide has been recommended as the sanitizer. The objective of th is research is to determine the required effective concentration of hydrogen peroxide. In addition, it will be determined whether the use of hydrogen peroxide, although a viable sanitizer, adversely affects the quality of the vegetables. Vegetables will be dipped in 1 % hydrogen peroxide, 3% hydrogen peroxide, or 5% hydrogen peroxide. Treated produce and controls will be stored in plastic bags at 5 C for up to 14 days. Sensory, color, texture, and total plate count will be measured. The effect on several vegetables including lettuce, radish, tomato and strawberries has been completed. Although each vegetable reacts to hydrogen peroxide differently, the

  15. Survey of tritiated oil sources and handling practices

    International Nuclear Information System (INIS)

    Miller, J.M.

    1994-08-01

    Tritium interactions with oil sources (primarily associated with pumps) in tritium-handling facilities can lead to the incorporation of tritium in the oil and the production of tritiated hydrocarbons. This results in a source of radiological hazard and the need for special handling considerations during maintenance, decontamination, decommissioning and waste packaging and storage. The results of a general survey of tritiated-oil sources and their associated characteristics, handling practices, analysis techniques and waste treatment/storage methods are summarized here. Information was obtained from various tritium-handling laboratories, fusion devices, and CANDU plants. 38 refs., 1 fig

  16. Purifying waste waters in the surface treatment industry; Depuracion de las aguas residuales en la industria de tratamiento de superficies

    Energy Technology Data Exchange (ETDEWEB)

    Queralt Torrell, R.; Martinez Hidalgo, E. [Generalitat de Catalunya. Barcelona (Spain)

    1999-07-01

    The electrolytic coating industry in Spain comprises some 2,000 firms whose waste waters are highly problematic, mainly due to the presence of heavy metals and cyanides. This article sets out series of internal measures for conserving the baths, reducing entrainment and optimising washes, thereby minimising the volume and concentration of the waste waters. It also discusses the advantages and disadvantages of the different chemical reagents employed in the physicochemical treatment of waste waters during the oxidation/reduction and metal precipitation processes. In addition, it mentions other techniques such as the use of ion exchange resins, membranes and electro-coagulation, which are becoming increasingly widespread. Finally, it offers a summary report on 11 waste water treatment facilities installed in different surface treatment factories. (Author) 21 refs.

  17. Handling and transport problems (1960)

    International Nuclear Information System (INIS)

    Pomarola, J.; Savouyaud, J.

    1960-01-01

    I. The handling and transport of radioactive wastes involves the danger of irradiation and contamination. It is indispensable: - to lay down a special set of rules governing the removal and transport of wastes within centres or from one centre to another; - to give charge of this transportation to a group containing teams of specialists. The organisation, equipment and output of these teams is being examined. II. Certain materials are particularly dangerous to transport, and for these special vehicles and fixed installations are necessary. This is the case especially for the evacuation of very active liquids. A transport vehicle is described, consisting of a trailer tractor and a recipient holding 500 litres of liquid of which the activity can reach 1000 C/l; the decanting operation, the route to be followed by the vehicle, and the precautions taken are also described. (author) [fr

  18. Handling of radioactive sources in Ecuador

    International Nuclear Information System (INIS)

    Benitez, Manuel

    2000-01-01

    This document describes the following aspects: sealed and unsealed radioactive sources, radiation detectors, personnel and area monitoring, surface pollution, radioactive wastes control and radioactive sources transferring. (The author)

  19. Estimation of mean time to failure of a near surface radioactive waste repository for PWR power stations

    International Nuclear Information System (INIS)

    Aguiar, Lais A. de; Frutuoso e Melo, P.F.; Alvim, Antonio C.M.

    2007-01-01

    This work aims at estimating the mean time to failure (MTTF) of each barrier of a near surface radioactive waste repository. It is assumed that surface water infiltrates through the barriers, reaching the matrix where radionuclides are contained, releasing them to the environment. Radioactive wastes considered in this work are low and medium level wastes (produced during operation of a PWR nuclear power station) fixed on cement. The repository consists of 6 saturated porous media barriers (top cover, upper layer, packages, basis, repository walls and geosphere). It has been verified that the mean time to failure (MTTF) of each barrier increases for radionuclides having higher retardation factor (Fr) and also that the MTTF for concrete is larger for Nickel , while for the geosphere, Plutonium gives the largest MTTF. (author)

  20. Surface erosion and hydrology of earth covers used in shallow land burial of low-level radioactive waste

    International Nuclear Information System (INIS)

    Bent, G.C.

    1988-01-01

    Shallow land burial is the current method of disposal of low-level radioactive waste in the United States. The most serious technical problems encountered in shallow land burial are water-related. Water is reported to come into contact with the waste by erosion of earth covers or through infiltration of precipitation through the earth covers. The objectives of this study were to: compare and evaluate the effects of crested wheatgrass and streambank wheatgrass on surface erosion of simulated earth covers at Idaho National Engineering Laboratory (INEL), characterize the surface hydrology, and estimate cumulative soil loss for average and extreme rainfall events and determine if the waste will become exposed during its burial life due to erosion. 30 refs., 26 figs., 21 tabs

  1. Surface displacements and pillar stresses associated with nuclear waste disposal in salt

    International Nuclear Information System (INIS)

    Hardy, M.P.; St John, C.M.

    1977-01-01

    A numerical model for regional analysis of stresses and displacement, resulting from heat generating waste placement in underground salt excavations, is presented. The model, which is an extension of that described by McClain and Starfield (1971), is based upon the displacement discontinuity method of stress analysis. It incorporates an empirical characterization of creep behavior of material on the excavation horizon and accounts for thermally induced stresses and displacements. The versatility of this approach is illustrated by the results of three relatively short simulations of test scale disposal facilities at shallow and greater depths. In addition, a three-dimensional code was used to evaluate the surface displacement history for a full-scale repository. This latter code, a thermoelastic analysis, gives an upper bound for the surface movements. It is concluded that the pillar stresses are the result of a complex non-linear interaction of many variables, and the maximum pillar stress can reach several multiples of the tributory-area pillar stress

  2. Development of a methodology for the safety assessment of near surface disposal facilities for radioactive waste

    International Nuclear Information System (INIS)

    Simon, I.; Cancio, D.; Alonso, L.F.; Agueero, A.; Lopez de la Higuera, J.; Gil, E.; Garcia, E.

    2000-01-01

    The Project on the Environmental Radiological Impact in CIEMAT is developing, for the Spanish regulatory body Consejo de Seguridad Nuclear (CSN), a methodology for the Safety Assessment of near surface disposal facilities. This method has been developed incorporating some elements developed through the participation in the IAEA's ISAM Programme (Improving Long Term Safety Assessment Methodologies for Near Surface Radioactive Waste Disposal Facilities). The first step of the approach is the consideration of the assessment context, including the purpose of the assessment, the end-Points, philosophy, disposal system, source term and temporal scales as well as the hypothesis about the critical group. Once the context has been established, and considering the peculiarities of the system, an specific list of features, events and processes (FEPs) is produced. These will be incorporated into the assessment scenarios. The set of scenarios will be represented in the conceptual and mathematical models. By the use of mathematical codes, calculations are performed to obtain results (i.e. in terms of doses) to be analysed and compared against the criteria. The methodology is being tested by the application to an hypothetical engineered disposal system based on an exercise within the ISAM Programme, and will finally be applied to the Spanish case. (author)

  3. Seismic surface-wave tomography of waste sites. 1998 annual progress report

    International Nuclear Information System (INIS)

    Long, T.L.

    1998-01-01

    'The objective of the Seismic Surface Wave Tomography of Waste Sites is to develop a robust technique for field acquisition and analysis of surface wave data for the interpretation of shallow structures, such as those associated with the burial of wastes. The analysis technique is to be developed and tested on an existing set of seismic data covering the K-901 burial site at the East Tennessee Technology Park. Also, a portable prototype for a field acquisition system will be designed and developed to obtain additional data for analysis and testing of the technique. The portable analysis system will display an image representing the shear-wave velocity structure. The image would be developed in the field from successive data samples. As of May 1998, the author established compatibility with computer programs at Georgia Tech and computed a preliminary singular value decomposition solution for the K-901 data. The analysis included modeling of surface wave dispersion and analysis of velocity structure. The analysis demonstrated that the authors needed additional field data to verify the conclusions and provide independent confirmation of velocity structure. The K-901 site data were obtained with 8 Hz geophones. The frequencies below 8 Hz are strongly attenuated in such recording instruments and are difficult to analyze. In particular, group velocities can have multiple answers for a given frequency. Consequently, without a record of the low-frequency energy, the authors found it difficult to identify the portion of the dispersion curve responsible for the seismogram. In particular, it was difficult to determine if the reverse dispersion observed in the frequencies above 8 Hz was caused by a low velocity layer or caused by observing only the frequencies above the group velocity minimum. In either model, synthetic seismograms can be made to match the observed data for the higher frequencies. The contract for the proposed work was completed in December. The field work was

  4. Radionuclide transport from near-surface repository for radioactive waste - The unsaturated zone approach

    Energy Technology Data Exchange (ETDEWEB)

    Jakimaviciute-Maseliene, V. [Vilnius University (Lithuania); Mazeika, J. [Nature Research Centre (Lithuania); Motiejunas, S. [Radioactive Waste Management Agency (Lithuania)

    2014-07-01

    About 100 000 m{sup 3} of solid conditioned Low and Intermediate Level Waste (LILW), generated during operation and decommissioning of the Ignalina nuclear power plant (INPP), are to be disposed of in a near-surface repository (NSR) - a 'hill'-type repository with reinforced concrete vaults and with engineered and natural barriers. The northeastern Lithuania and the environment of the INPP in particular were recognized as the areas most suitable for a near-surface repository (Stabatiske Site). The engineered barriers of the repository consist of concrete cells surrounded by clay-based material of low permeability with about the same isolating capacity in all directions. The clay materials must be effectively compactable so that required hydraulic conductivity is reached. The Lithuanian Triassic clay turned out to be sufficiently rich in smectites and was proposed as main candidate for sealing of the repository. When the concrete vaults are filled, the repository cover will be constructed. The surface of the mound will be planted with grass. In this study a computer code FEFLOW 5.0 was applied for simulating the transport of the most mobile radionuclides ({sup 3}H, {sup 14}C, {sup 59}Ni and {sup 94}Nb) with moisture through an unsaturated vault of the near-surface repository in Stabatiske Site. The HYDRUS-1D analysis was used to assess the radionuclide transport in the repository and to estimate initial activity concentrations of radionuclides transported from the cemented waste matrix. Radionuclide release from the vault in the unsaturated conditions after closure of the repository and consequent contaminant plume transport has been assessed taking into account site-specific natural and engineering conditions and based on a normal evolution scenario. The highest peak radionuclide activity concentrations were estimated applying the FEFLOW code. The highest value of {sup 14}C activity concentration(about 1.3x10{sup 8} Bq/m{sup 3}) at the groundwater table

  5. Municipal Solid Waste management

    OpenAIRE

    Mirakovski, Dejan; Hadzi-Nikolova, Marija; Doneva, Nikolinka

    2010-01-01

    Waste management covers newly generated waste or waste from an onging process. When steps to reduce or even eliminate waste are to be considered, it is imperative that considerations should include total oversight, technical and management services of the total process.From raw material to the final product this includes technical project management expertise, technical project review and pollution prevention technical support and advocacy.Waste management also includes handling of waste, in...

  6. Aerobot Sampling and Handling System, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Honeybee Robotics proposes to: ?Derive and document the functional and technical requirements for Aerobot surface sampling and sample handling across a range of...

  7. Criticality safety of transuranic storage arrays at the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Boyd, W.A.; Fecteau, M.W.

    1993-01-01

    The Waste Isolation Pilot Plant (WIPP) facility is designed to store transuranic waste that will consist mainly of surface contaminate articles and sludge. The fissile material in the waste is predominantly 239 Pu. The waste is grouped into two categories: contact-handled waste, which will be stored in 55-gal steel drums or in steel boxes, and remote-handled waste, which will be stored in specially designed cylindrical steel canisters. To show that criticality safety will be acceptable, criticality analyses were performed to demonstrate that a large number of containers with limiting loadings of fissile material could be stored at the site and meet a k eff limit of 0.95. Criticality analyses based on the classic worst-case moderated plutonium sphere approach would severely limit the capacity for storage of waste at the facility. Therefore, these analyses use realistic or credible worst-case assumptions to better represent the actual storage situation without compromising the margin of safety. Numerous sensitivity studies were performed to determine the importance of various parameters on the criticality of the configuration. It was determined that the plutonium loading has the dominant effect on the system reactivity. Nearly all other reactivity variations from the sensitivity studies were found to be relatively small. The analysis shows that criticality of the contact-handled waste storage drums and boxes and the remote-handled canisters is prevented by restrictions on maximum fissile loading per container and on the size of handling/storage areas

  8. The surface water submodel for the assessment of Canada's nuclear fuel waste management concept

    International Nuclear Information System (INIS)

    Bird, G.A.; Stephenson, M.; Cornett, R.J.

    1992-12-01

    A requirement in assessing the safety of Canada's nuclear fuel waste management concept is the prediction of radiological doses to humans and other biota, which may occur far in the future as a result of releases of nuclides to the biosphere. A biosphere model has been developed, consisting of four integrated submodels describing surface water, soil, atmosphere, and food-dose components. This report documents the surface water submodel, which is a simple, generic mass balance model of a Canadian Shield lake. Nuclide input to the lake is the time-dependent mass output from the geosphere model. Nuclides enter the lake from compacted sediments. The surface water submodel calculates nuclide concentrations in lake water and sediment. These concentrations are used in the other biosphere submodels to predict the radiological dose to biota. Selection of parameter values for the model is based on the literature, our own data, and conservative assumptions to ensure that doses are not underestimated. MOst parameters are represented by log normal. This probabilistic approach of using distributed parameter values accounts for variability and uncertainty in parameter values, and short-term environmental fluctuations. Long-term environmental changes, such as glaciation, are not considered in the model. Sensitivity analysis indicates that nuclide concentrations in lake water and sediment are governed primarily by hydrological flushing, with lake catchment area being the most important parameter. When catchment area is held constant, as would occur at a specific site, lake area and nuclide transfer rate from water to sediment strongly influence concentrations in both water and sediment. Sediment accumulation rate also strongly influences sediment nuclide concentrations. Validation of model predictions using published studies and other data demonstrates that our model is realistic and suitable for assessing Canada's disposal concept. (Author)

  9. Safety assessment on the human intrusion scenarios of near surface disposal facility for low and very low level radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Sung Wook; Park, Jin Baek [Korea Radioactive Waste Agency, Daejeon (Korea, Republic of); Park, Sang Ho [Chungnam National University, Daejeon (Korea, Republic of)

    2016-03-15

    The second-stage near surface disposal facility for low and very low level radioactive waste's permanent disposal is to be built. During the institutional control period, the inadvertent intrusion of the general public is limited. But after the institutional control period, the access to the general public is not restricted. Therefore human who has purpose of residence and resource exploration can intrude the disposal facility. In this case, radioactive effects to the intruder should be limited within regulatory dose limits. This study conducted the safety assessment of human intrusion on the second-stage surface disposal facility through drilling and post drilling scenario. Results of drilling and post drilling scenario were satisfied with regulatory dose limits. The result showed that post-drilling scenario was more significant than drilling scenario. According to the human intrusion time and behavior after the closure of the facility, dominant radionuclide contributing to the intruder was different. Sensitivity analyses on the parameters about the human behavior were also satisfied with regulatory dose limits. Especially, manual redistribution factor was the most sensitive parameter on exposure dose. A loading plan of spent filter waste and dry active waste was more effective than a loading plan of spent filter waste and other wastes for the radiological point of view. These results can be expected to provide both robustness and defense in depth for the development of safety case further.

  10. Charging for waste motivates generators to optimize waste control at the source

    International Nuclear Information System (INIS)

    Berry, J.B.; Homan, F.J.

    1988-01-01

    The Department of Energy (DOE) has recognized the need for waste management that incorporates improved waste-handling techniques and more stringent regulatory requirements to prevent future liabilities such as Superfund sites. DOE-Oak Ridge Operations (DOE-ORO) has recognized that an effective waste management program focuses on control at the source and that the burden for responsible waste management can be placed on generators by charging for waste management costs. The principle of including the waste management costs in the total cost of the product, even when the product is research and development, is being implemented at Oak Ridge National Laboratory (ORNL). Charging waste management costs to generators creates an incentive to optimize processes so that less waste is produced, and it provides a basis for determining the cost effectiveness of capital improvements so that the mature phase of waste management can be attained. Improving waste management practices requires a long-range commitment and consistent administration. Making this commitment and providing adequate funding for proper waste disposal are most cost-effective measures than the alternative of paying for remedial actions after improper disposal. This paper summarizes a plan to charge waste generators, the administrative structure of the plan, a comparison between the rate structure and changes in waste disposal operations, and issues that have surfaced as the plan is implemented

  11. Nuclear fuel handling apparatus

    International Nuclear Information System (INIS)

    Andrea, C.; Dupen, C.F.G.; Noyes, R.C.

    1977-01-01

    A fuel handling machine for a liquid metal cooled nuclear reactor in which a retractable handling tube and gripper are lowered into the reactor to withdraw a spent fuel assembly into the handling tube. The handling tube containing the fuel assembly immersed in liquid sodium is then withdrawn completely from the reactor into the outer barrel of the handling machine. The machine is then used to transport the spent fuel assembly directly to a remotely located decay tank. The fuel handling machine includes a decay heat removal system which continuously removes heat from the interior of the handling tube and which is capable of operating at its full cooling capacity at all times. The handling tube is supported in the machine from an articulated joint which enables it to readily align itself with the correct position in the core. An emergency sodium supply is carried directly by the machine to provide make up in the event of a loss of sodium from the handling tube during transport to the decay tank. 5 claims, 32 drawing figures

  12. Other Special Waste

    DEFF Research Database (Denmark)

    Brogaard, Line Kai-Sørensen; Christensen, Thomas Højlund

    2011-01-01

    In addition to the main types of special waste related to municipal solid waste (MSW) mentioned in the previous chapters (health care risk waste, WEEE, impregnated wood, hazardous waste) a range of other fractions of waste have in some countries been defined as special waste that must be handled...... separately from MSW. Some of these other special wastes are briefly described in this chapter with respect to their definition, quantity and composition, and management options. The special wastes mentioned here are batteries, tires, polyvinylchloride (PVC) and food waste....

  13. Design and operational considerations of United States commercial near-surface low-level radioactive waste disposal facilities

    International Nuclear Information System (INIS)

    Birk, S.M.

    1997-10-01

    In accordance with the Low-Level Radioactive Waste Policy Amendments Act of 1985, states are responsible for providing for disposal of commercially generated low-level radioactive waste (LLW) within their borders. LLW in the US is defined as all radioactive waste that is not classified as spent nuclear fuel, high-level radioactive waste, transuranic waste, or by-product material resulting from the extraction of uranium from ore. Commercial waste includes LLW generated by hospitals, universities, industry, pharmaceutical companies, and power utilities. LLW generated by the country''s defense operations is the responsibility of the Federal government and its agency, the Department of Energy. The commercial LLRW disposal sites discussed in this report are located near: Sheffield, Illinois (closed); Maxey Flats, Kentucky (closed); Beatty, Nevada (closed); West Valley, New York (closed); Barnwell, South Carolina (operating); Richland, Washington (operating); Ward Valley, California, (proposed); Sierra Blanca, Texas (proposed); Wake County, North Carolina (proposed); and Boyd County, Nebraska (proposed). While some comparisons between the sites described in this report are appropriate, this must be done with caution. In addition to differences in climate and geology between sites, LLW facilities in the past were not designed and operated to today''s standards. This report summarizes each site''s design and operational considerations for near-surface disposal of low-level radioactive waste. The report includes: a description of waste characteristics; design and operational features; post closure measures and plans; cost and duration of site characterization, construction, and operation; recent related R and D activities for LLW treatment and disposal; and the status of the LLW system in the US

  14. A summary of the sources of input parameter values for the Waste Isolation Pilot Plant final porosity surface calculations