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Sample records for waste package project

  1. Yucca Mountain Site Characterization Project Waste Package Plan

    International Nuclear Information System (INIS)

    Harrison-Giesler, D.J.; Jardine, L.J.

    1991-02-01

    The goal of the US Department of Energy's (DOE) Yucca Mountain Site Characterization Project (YMP) waste package program is to develop, confirm the effectiveness of, and document a design for a waste package and associated engineered barrier system (EBS) for spent nuclear fuel and solidified high-level nuclear waste (HLW) that meets the applicable regulatory requirements for a geologic repository. The Waste Package Plan describes the waste package program and establishes the technical approach against which overall progress can be measured. It provides guidance for execution and describes the essential elements of the program, including the objectives, technical plan, and management approach. The plan covers the time period up to the submission of a repository license application to the US Nuclear Regulatory Commission (NRC). 1 fig

  2. Salt Repository Project Waste Package Program Plan: Draft

    International Nuclear Information System (INIS)

    Carr, J.A.; Cunnane, J.C.

    1986-01-01

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

  3. Waste Package Project quarterly report, July 1, 1995--September 30, 1995

    International Nuclear Information System (INIS)

    Ladkany, S.G.

    1995-01-01

    The following tasks are reported: overview and progress of nuclear waste package project and container design; nuclear waste container design considerations; structural investigation of multi purpose nuclear waste package canister; and design requirements of rock tunnel drift for long-term storage of high-level waste (faulted tunnel model study by photoelasticity/finite element analysis)

  4. Overview of hydrothermal testing of waste-package barrier materials at the Basalt Waste Isolation Project

    International Nuclear Information System (INIS)

    1982-01-01

    The current Waste Package Department (WPD) hydrothermal testing program for the Basalt Waste Isolation Project (BWIP) has followed a systematic approach for the testing of waste-barrier-basalt interactions based on sequential penetration of barriers by intruding groundwaters. Present test activities in the WPD program have focused on determining radionuclide solubility limits (or steady-state conditions) of simulated waste forms and the long-term stability of waste package barriers under site-specific hydrothermal conditions. The resulting data on solution compositions and solid alteration products have been used to evaluate waste form degradation under conditions specific to a nuclear waste repository located in basalt (NWRB). Isothermal, time-invariant compositional data on sampled solutions have been coupled with realistic hydrologic flow data for near-field and far-field modeling for the calculation of meaningful radionuclide release rates. Radionuclides that are not strongly sorbed or precipitated from solution and that, therefore, may require special attention to ensure their isolation within the waste package have been identified. Taken together, these hydrothermal test data have been used to establish design requirements for waste packages located in basalt

  5. Waste package performance assessment for the Yucca Mountain project

    International Nuclear Information System (INIS)

    O'Connell, W.J.; Lappa, D.A.; Thatcher, R.M.

    1989-01-01

    The authors completed a first cycle of model development from a specification to a computer program, PANDORA-1, for long-term performance assessment of waste packages. The model for one waste package at a time incorporates processes specific to the unsaturated environment at the proposed Yucca Mountain, NV, site. PANDORA-1 models the most likely processes and several modes of waste alteration and release. The development identified information needs for future models; many processes, local details, and combinations will have to be examined. Integration of ensemble performance and quantification of uncertainties are modeling steps at higher aggregation. Methodologies for these steps include sampling, which is well studied; we have focused on several open questions. The authors can now calculate the amount of variance reduction available from Latin hypercube sampling; it is a limited reduction. A new method, uncertainty analysis test-bed program compares the new with old sampling methods

  6. Salt Repository Project: Waste Package Program (WPP) modeling activiteis: FY 1984 annual report

    International Nuclear Information System (INIS)

    Kuhn, W.L.; Simonson, S.A.; Pulsipher, B.A.

    1987-03-01

    The Pacific Northwest Laboratory (PNL) is supporting the US Department of Energy's (DOE) Salt Repository Project (SRP) through its Waste Package Program (WPP). During FY 1984, the WPP continued its program of waste package component development and interactions testing and application of the resulting data base to develop predictive models describing waste package degradation and radionuclide release. Within the WPP, the Modeling Task (Task 04 during FY 1984) was conducted to interpret the tests in such a way that scientifically defensible models can be developed for use in qualification of the waste package

  7. Waste package performance assessment for the Yucca Mountain Project

    International Nuclear Information System (INIS)

    O'Connell, W.J.; Lappa, D.A.; Thatcher, R.M.

    1989-02-01

    We completed a first cycle of model development from a specification to a computer program, PANDORA-1, for long-term performance assessment of waste packages. The model for one waste package at a time incorporates processes specific to the unsaturated environment at the proposed Yucca Mountain, NV, site. PANDORA-1 models the most likely processes and several modes of waste alteration and release. The development identified information needs for future models; many processes, local details, and combinations will have to be examined. Integration of ensemble performance and quantification of uncertainties are modeling steps at higher aggregation. Methodologies for these steps include sampling, which is well studied; we have focused on several open questions. We can now calculate the amount of variance reduction available from Latin hypercube sampling; it is a limited reduction. A new method, controlled sampling, provides substantial variance reduction for a broad range of model functions. An uncertainty analysis test-bed program compares the new with old sampling methods. 7 refs., 1 tab

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

    International Nuclear Information System (INIS)

    Nelson, T.; Russell, E.; Johnson, G.L.; Morissette, R.; Stahl, D.; LaMonica, L.; Hertel, G.

    1989-04-01

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

  9. Waste Package Lifting Calculation

    International Nuclear Information System (INIS)

    H. Marr

    2000-01-01

    The objective of this calculation is to evaluate the structural response of the waste package during the horizontal and vertical lifting operations in order to support the waste package lifting feature design. The scope of this calculation includes the evaluation of the 21 PWR UCF (pressurized water reactor uncanistered fuel) waste package, naval waste package, 5 DHLW/DOE SNF (defense high-level waste/Department of Energy spent nuclear fuel)--short waste package, and 44 BWR (boiling water reactor) UCF waste package. Procedure AP-3.12Q, Revision 0, ICN 0, calculations, is used to develop and document this calculation

  10. Preliminary selection criteria for the Yucca Mountain Project waste package container material

    International Nuclear Information System (INIS)

    Halsey, W.G.

    1991-01-01

    The Department of Energy's Yucca Mountain Project (YMP) is evaluating a site at Yucca Mountain in Nevada for construction of a geologic repository for the storage of high-level nuclear waste. Lawrence Livermore National Laboratory's (LLNL) Nuclear Waste Management Project (NWMP) has the responsibility for design, testing, and performance analysis of the waste packages. The design is performed in an iterative manner in three sequential phases (conceptual design, advanced conceptual design, and license application design). An important input to the start of the advanced conceptual design is the selection of the material for the waste containers. The container material is referred to as the 'metal barrier' portion of the waste package, and is the responsibility of the Metal Barrier Selection and Testing task at LLNL. The selection will consist of several steps. First, preliminary, material-independent selection criteria will be established based on the performance goals for the container. Second, a variety of engineering materials will be evaluated against these criteria in a screening process to identify candidate materials. Third, information will be obtained on the performance of the candidate materials, and final selection criteria and quantitative weighting factors will be established based on the waste package design requirements. Finally, the candidate materials will be ranked against these criteria to determine whether they meet the mandated performance requirements, and to provide a comparative score to choose the material for advanced conceptual design activities. This document sets forth the preliminary container material selection criteria to be used in screening candidate materials. 5 refs

  11. International co-ordinated research project on low and intermediate level waste package performance

    International Nuclear Information System (INIS)

    Dayal, R.

    2001-01-01

    As part of IAEA's mandate to facilitate the transfer and exchange of information amongst Member States, the Agency is currently coordinating an international R and D project, involving 12 developed and developing countries, on Performance of Low and Intermediate Level Waste Packages under Disposal Conditions. This paper will review the current status of the Coordinated Research Project (CRP) and summarize the key findings of the work completed to date within the context of the CRP in the participating Member States. (author)

  12. Naval Waste Package Design Report

    International Nuclear Information System (INIS)

    M.M. Lewis

    2004-01-01

    A design methodology for the waste packages and ancillary components, viz., the emplacement pallets and drip shields, has been developed to provide designs that satisfy the safety and operational requirements of the Yucca Mountain Project. This methodology is described in the ''Waste Package Design Methodology Report'' Mecham 2004 [DIRS 166168]. To demonstrate the practicability of this design methodology, four waste package design configurations have been selected to illustrate the application of the methodology. These four design configurations are the 21-pressurized water reactor (PWR) Absorber Plate waste package, the 44-boiling water reactor (BWR) waste package, the 5-defense high-level waste (DHLW)/United States (U.S.) Department of Energy (DOE) spent nuclear fuel (SNF) Co-disposal Short waste package, and the Naval Canistered SNF Long waste package. Also included in this demonstration is the emplacement pallet and continuous drip shield. The purpose of this report is to document how that design methodology has been applied to the waste package design configurations intended to accommodate naval canistered SNF. This demonstrates that the design methodology can be applied successfully to this waste package design configuration and support the License Application for construction of the repository

  13. International co-ordinated research project on low and intermediate level waste package performance

    Energy Technology Data Exchange (ETDEWEB)

    Dayal, R. [International Atomic Energy Agency IAEA, Vienna (Austria)

    2001-07-01

    As part of IAEA's mandate to facilitate the transfer and exchange of information amongst Member States, the Agency is currently coordinating an international R and D project, involving 12 developed and developing countries, on Performance of Low and Intermediate Level Waste Packages under Disposal Conditions. This paper will review the current status of the Coordinated Research Project (CRP) and summarize the key findings of the work completed to date within the context of the CRP in the participating Member States. (author)

  14. Packaging design criteria (onsite) project W-520 immobilized low-activity waste transportation system

    International Nuclear Information System (INIS)

    BOEHNKE, W.M.

    2001-01-01

    A plan is currently in place to process the high-level radioactive wastes that resulted from uranium and plutonium recovery operations from Spent Nuclear Fuel at the Hanford Site, Richland, Washington. Currently, millions of gallons of high-level radioactive waste in the form of liquids, sludges, and saltcake are stored in many large underground tanks onsite. This waste will be processed and separated into high-level and low-activity fractions. Both fractions will then be vitrified (i.e., blended with molten borosilicate glass) in order to encapsulate the toxic radionuclides. The immobilized low-activity waste (ILAW) glass will be poured into LAW canisters, allowed to cool and harden to solid form, sealed by welding, and then transported to a double-lined trench in the 200 East Area for permanent disposal. This document presents the packaging design criteria (PDC) for an onsite LAW transportation system, which includes the ILAW canister, ILAW package, and transport vehicle and defines normal and accident conditions. This PDC provides the basis for the ILAW onsite transportation system design and fabrication and establishes the transportation safety criteria that the design will be evaluated against in the Package Specific Safety Document (PSSD). It provides the criteria for the ILAW canister, cask and transport vehicles and defines normal and accident conditions. The LAW transportation system is designed to transport stabilized waste from the vitrification facility to the ILAW disposal facility developed by Project W-520. All ILAW transport will take place within the 200 East Area (all within the Hanford Site)

  15. WASTE PACKAGE TRANSPORTER DESIGN

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  16. WASTE PACKAGE TRANSPORTER DESIGN

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-09-23

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

  17. Waste package performance assessment

    International Nuclear Information System (INIS)

    Lester, D.H.

    1981-01-01

    This paper describes work undertaken to assess the life-expectancy and post-failure nuclide release behavior of high-level and waste packages in a geologic repository. The work involved integrating models of individual phenomena (such as heat transfer, corrosion, package deformation, and nuclide transport) and using existing data to make estimates of post-emplacement behavior of waste packages. A package performance assessment code was developed to predict time to package failure in a flooded repository and subsequent transport of nuclides out of the leaking package. The model has been used to evaluate preliminary package designs. The results indicate, that within the limitation of model assumptions and data base, packages lasting a few hundreds of years could be developed. Very long lived packages may be possible but more comprehensive data are needed to confirm this

  18. Final status of the salt repository project waste package program experimental database

    International Nuclear Information System (INIS)

    Thornton, B.M.; Reimus, P.W.

    1988-03-01

    This report describes the final status of the Salt Repository Project Waste Package Program Experimental Database. The data base serves as a clearinghouse for all data collected within the Waste Package Program (WPP) and its predecessor programs at Pacific Northwest Laboratory (PNL). The database was maintained using RS/1 database management software. Documented assurance that the entries in the database were consistent with experimental records was provided by having each experimentalist inspect the entries and signify that they were in agreement with the records. The inspection and signoff were done per PNL technical procedures. Data for which it was impossible to obtain the experimentalist's inspection and signature were segregated from the rest of the database, although they could still be accessed by WPP staff. The WPPED contains two groups of subdirectories. One group contains data taken prior to the installation of quality assurance procedures at PNL. The other group of subdirectories contains data taken under the NQA-1 procedures since their installation in April 1985. As part of closeout activities in the Salt Repository Project, the WPP database has been archived onto magnetic media. The data in the database are available by request on magnetic media or in hardcopy form. 2 refs

  19. Package materials, waste form

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    The schedules for waste package development for the various host rocks were presented. The waste form subtask activities were reviewed, with the papers focusing on high-level waste, transuranic waste, and spent fuel. The following ten papers were presented: (1) Waste Package Development Approach; (2) Borosilicate Glass as a Matrix for Savannah River Plant Waste; (3) Development of Alternative High-Level Waste Forms; (4) Overview of the Transuranic Waste Management Program; (5) Assessment of the Impacts of Spent Fuel Disassembly - Alternatives on the Nuclear Waste Isolation System; (6) Reactions of Spent Fuel and Reprocessing Waste Forms with Water in the Presence of Basalt; (7) Spent Fuel Stabilizer Screening Studies; (8) Chemical Interactions of Shale Rock, Prototype Waste Forms, and Prototype Canister Metals in a Simulated Wet Repository Environment; (9) Impact of Fission Gas and Volatiles on Spent Fuel During Geologic Disposal; and (10) Spent Fuel Assembly Decay Heat Measurement and Analysis

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

    International Nuclear Information System (INIS)

    1978-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1978-04-01

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

  2. Waste disposal package

    Science.gov (United States)

    Smith, M.J.

    1985-06-19

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

  3. Waste package materials selection process

    International Nuclear Information System (INIS)

    Roy, A.K.; Fish, R.L.; McCright, R.D.

    1994-01-01

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

  4. Radioactive waste disposal package

    Science.gov (United States)

    Lampe, Robert F.

    1986-11-04

    A radioactive waste disposal package comprising a canister for containing vitrified radioactive waste material and a sealed outer shell encapsulating the canister. A solid block of filler material is supported in said shell and convertible into a liquid state for flow into the space between the canister and outer shell and subsequently hardened to form a solid, impervious layer occupying such space.

  5. Waste package characterisation

    Energy Technology Data Exchange (ETDEWEB)

    Sannen, L.; Bruggeman, M.; Wannijn, J.P

    1998-09-01

    Radioactive wastes originating from the hot labs of the Belgian Nuclear Research Centre SCK-CEN contain a wide variety of radiotoxic substances. The accurate characterisation of the short- and long-term radiotoxic components is extremely difficult but required in view of geological disposal. This paper describes the methodology which was developed and adopted to characterise the high- and medium-level waste packages at the SCK-CEN hot laboratories. The proposed method is based on the estimation of the fuel inventory evacuated in a particular waste package; a calculation of the relative fission product contribution on the fuel fabrication and irradiation footing; a comparison of the calculated, as expected, dose rate and the real measured dose rate of the waste package. To cope with the daily practice an appropriate fuel inventory estimation route, a user friendly computer programme for fission product and corresponding dose rate calculation, and a simple dose rate measurement method have been developed and implemented.

  6. Waste package characterisation

    International Nuclear Information System (INIS)

    Sannen, L.; Bruggeman, M.; Wannijn, J.P.

    1998-09-01

    Radioactive wastes originating from the hot labs of the Belgian Nuclear Research Centre SCK-CEN contain a wide variety of radiotoxic substances. The accurate characterisation of the short- and long-term radiotoxic components is extremely difficult but required in view of geological disposal. This paper describes the methodology which was developed and adopted to characterise the high- and medium-level waste packages at the SCK-CEN hot laboratories. The proposed method is based on the estimation of the fuel inventory evacuated in a particular waste package; a calculation of the relative fission product contribution on the fuel fabrication and irradiation footing; a comparison of the calculated, as expected, dose rate and the real measured dose rate of the waste package. To cope with the daily practice an appropriate fuel inventory estimation route, a user friendly computer programme for fission product and corresponding dose rate calculation, and a simple dose rate measurement method have been developed and implemented

  7. Waste package reliability analysis

    International Nuclear Information System (INIS)

    Pescatore, C.; Sastre, C.

    1983-01-01

    Proof of future performance of a complex system such as a high-level nuclear waste package over a period of hundreds to thousands of years cannot be had in the ordinary sense of the word. The general method of probabilistic reliability analysis could provide an acceptable framework to identify, organize, and convey the information necessary to satisfy the criterion of reasonable assurance of waste package performance according to the regulatory requirements set forth in 10 CFR 60. General principles which may be used to evaluate the qualitative and quantitative reliability of a waste package design are indicated and illustrated with a sample calculation of a repository concept in basalt. 8 references, 1 table

  8. Waste package performance analysis

    International Nuclear Information System (INIS)

    Lester, D.H.; Stula, R.T.; Kirstein, B.E.

    1982-01-01

    A performance assessment model for multiple barrier packages containing unreprocessed spent fuel has been applied to several package designs. The resulting preliminary assessments were intended for use in making decisions about package development programs. A computer model called BARIER estimates the package life and subsequent rate of release of selected nuclides. The model accounts for temperature, pressure (and resulting stresses), bulk and localized corrosion, and nuclide retardation by the backfill after water intrusion into the waste form. The assessment model assumes a post-closure, flooded, geologic repository. Calculations indicated that, within the bounds of model assumptions, packages could last for several hundred years. Intact backfills of appropriate design may be capable of nuclide release delay times on the order of 10 7 yr for uranium, plutonium, and americium. 8 references, 6 figures, 9 tables

  9. Objectives for radioactive waste packaging

    International Nuclear Information System (INIS)

    Flowers, R.H.

    1982-04-01

    The report falls under the headings: introduction; the nature of radioactive wastes; how to manage radioactive wastes; packaging of radioactive wastes (supervised storage; disposal); waste form evaluation and test requirements (supervised storage; disposal); conclusions. (U.K.)

  10. Packages for radiactive waste disposal

    International Nuclear Information System (INIS)

    Oliveira, R. de.

    1983-01-01

    The development of multi-stage type package for sea disposal of compactable nuclear wastes, is presented. The basic requirements for the project followed the NEA and IAEA recommendations and observations of the solutions adopted by others countries. The packages of preliminary design was analysed, by computer, under several conditions arising out of its nature, as well as their conditions descent, dumping and durability in the deep of sea. The designed pressure equalization mechanic and the effect compacting on the package, by prototypes and specific tests, were studied. These prototypes were also submitted to the transport tests of the 'Regulament for the Safe Transport of Radioactive Materials'. Based on results of the testes and the re-evaluation of the preliminary design, final indications and specifications for excuting the package design, are presented. (M.C.K.) [pt

  11. Current status of waste package designs for the Yucca Mountain Project

    International Nuclear Information System (INIS)

    Ballou, L.B.

    1989-07-01

    Conceptual designs for waste packages containing spent fuel or high-level waste glass have been developed for use in a repository at Yucca Mountain. The basis for these designs reflects the unique nature of the expected service environment associated with disposal in welded tuff in the unsaturated zone. In addition to a set of reference designs, alternative design concepts are being considered that would contain and isolate the waste radionuclides in a more aggressive service environment. Consideration is also being given to the feasibility of a concept known as ''heat tailoring'' that employs the thermal energy released by the wasteforms to enhance and extend the performance of the containers. 5 refs., 3 figs

  12. Tritium waste package

    Science.gov (United States)

    Rossmassler, Rich; Ciebiera, Lloyd; Tulipano, Francis J.; Vinson, Sylvester; Walters, R. Thomas

    1995-01-01

    A containment and waste package system for processing and shipping tritium xide waste received from a process gas includes an outer drum and an inner drum containing a disposable molecular sieve bed (DMSB) seated within outer drum. The DMSB includes an inlet diffuser assembly, an outlet diffuser assembly, and a hydrogen catalytic recombiner. The DMSB absorbs tritium oxide from the process gas and converts it to a solid form so that the tritium is contained during shipment to a disposal site. The DMSB is filled with type 4A molecular sieve pellets capable of adsorbing up to 1000 curies of tritium. The recombiner contains a sufficient amount of catalyst to cause any hydrogen add oxygen present in the process gas to recombine to form water vapor, which is then adsorbed onto the DMSB.

  13. Waste Package Design Methodology Report

    Energy Technology Data Exchange (ETDEWEB)

    D.A. Brownson

    2001-09-28

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

  14. Waste Package Design Methodology Report

    International Nuclear Information System (INIS)

    D.A. Brownson

    2001-01-01

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

  15. Scoping corrosion tests on candidate waste package basket materials for the Yucca Mountain project

    International Nuclear Information System (INIS)

    Konynenburg, R.A. van; Curtis, P.G.; Summers, T.S.E.

    1998-03-01

    A scoping corrosion test was performed on candidate waste package basket materials. The corrosion medium was a pH-buffered solution of chemical species expected to be produced by radiolysis. The test was conducted at 90 C for 96 hours. Samples included aluminum-, copper-, stainless steel- and zirconium-based metallic materials and several ceramics, incorporating neutron-absorbing elements. Sample weight losses and solution chemical changes were measured. Both corrosion of the host materials and dissolution of the neutron-absorbing elements were studied. The ceramics and the zirconium-based materials underwent only minor corrosion. The stainless steel-based materials performed well except for a welded sample. The aluminum- and copper-based materials exhibited the highest corrosion rates. Boron dissolution depends on its chemical form. Boron oxide and many metal borides dissolve readily in acidic solutions while high-chromium borides and boron carbide, though thermodynamically unstable, exhibit little dissolution in short times. The results of solution chemical analyses were consistent with this. Gadolinium did not dissolve significantly from monazite, and hafnium showed little dissolution from a variety of host materials, in keeping with its low solubility

  16. Naval Waste Package Design Sensitivity

    International Nuclear Information System (INIS)

    T. Schmitt

    2006-01-01

    The purpose of this calculation is to determine the sensitivity of the structural response of the Naval waste packages to varying inner cavity dimensions when subjected to a comer drop and tip-over from elevated surface. This calculation will also determine the sensitivity of the structural response of the Naval waste packages to the upper bound of the naval canister masses. The scope of this document is limited to reporting the calculation results in terms of through-wall stress intensities in the outer corrosion barrier. This calculation is intended for use in support of the preliminary design activities for the license application design of the Naval waste package. It examines the effects of small changes between the naval canister and the inner vessel, and in these dimensions, the Naval Long waste package and Naval Short waste package are similar. Therefore, only the Naval Long waste package is used in this calculation and is based on the proposed potential designs presented by the drawings and sketches in References 2.1.10 to 2.1.17 and 2.1.20. All conclusions are valid for both the Naval Long and Naval Short waste packages

  17. Waste Package Component Design Methodology Report

    International Nuclear Information System (INIS)

    D.C. Mecham

    2004-01-01

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

  18. Waste Package Component Design Methodology Report

    Energy Technology Data Exchange (ETDEWEB)

    D.C. Mecham

    2004-07-12

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

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

    International Nuclear Information System (INIS)

    2004-06-01

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

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

    International Nuclear Information System (INIS)

    Pool, K.N.; Jones, T.E.; McKinley, S.G.; Tingey, J.M.; Longaker, T.M.; Gibson, J.A.

    1990-01-01

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

  1. TRU waste transportation package development

    International Nuclear Information System (INIS)

    Eakes, R.G.; Lamoreaux, G.H.; Romesberg, L.E.; Sutherland, S.H.; Duffey, T.A.

    1980-01-01

    Inventories of the transuranic wastes buried or stored at various US DOE sites are tabulated. The leading conceptual design of Type-B packaging for contact-handled transuranic waste is the Transuranic Package Transporter (TRUPACT), a large metal container comprising inner and outer tubular steel frameworks which are separated by rigid polyurethane foam and sheathed with steel plate. Testing of TRUPACT is reported. The schedule for its development is given. 6 figures

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

    Energy Technology Data Exchange (ETDEWEB)

    Gdowski, G.E.

    1997-12-01

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

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

    International Nuclear Information System (INIS)

    Shelton-Davis, C.V.

    2003-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Colleen Shelton-Davis

    2003-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Shelton-Davis, C.V.

    2003-09-26

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

  6. Waste package project quarterly report, April 1, 1995--June 30, 1995

    International Nuclear Information System (INIS)

    Ladkany, S.G.

    1995-01-01

    Research activities spanned multidisciplinary areas: structural and stress analysis of the container including nonlinear yield and damage assessment and structural stability studies of the dynamic behavior of the steel multipurpose container under normal and accidental handling conditions; nuclear fission criticality studies in the canisters; investigation of novel canister design concepts and corrosion studies; heat transfer studies of the waste canisters and in adjacent rock drifts; fluid flow in porous media and radionuclide transport in near field rock; stresses and stability of the rock formations resulting from thermal loading of fuel elements and multi-tunnel concept being analyzed; faulted rock tunnel model; and three link steel robot for remote handling of high level nuclear waste

  7. Second Generation Waste Package Design Study

    International Nuclear Information System (INIS)

    Armijo, J.S.; Misra, M.; Kar, Piyush

    2007-01-01

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

  8. The Garbage Project Revisited: From a 20th Century Archaeology of Food Waste to a Contemporary Study of Food Packaging Waste

    Directory of Open Access Journals (Sweden)

    La Vergne Lehmann

    2015-06-01

    Full Text Available In 1973, Dr. Bill Rathje and his students at the University of Arizona began what was to become a two decade long investigation into American consumer waste habits. An archaeologist by profession, Rathje decided to adapt traditional archaeological methods and apply them to contemporary archaeological situations. This provided a platform for improving the understanding of what was really happening with, amongst other forms of waste, food at the consumer household level. The Garbage Project was able to study consumer behaviours directly from the material realities they left behind rather than from self-conscious self-reports of surveys and interviews. Using the same rationale, this study developed a profile of the packaged and processed food consumption in three regional Victorian municipalities. The main findings identified that consumers were limited to the food retail opportunities closest to their home and that they took greater care to wash out recyclables if they were placed in the recycling bin compared to the same item placed in a kerbside landfill bin. There was also an apparent lack of understanding about appropriate food storage and buying for purpose, especially with regard to the volume of the item they purchased, which appears to result in partially used recyclable containers being put in the kerbside landfill bin. By understanding the nature of the packaging and food that has been thrown away, it is possible to develop a narrative around what people understand about food purchasing practices, longevity, storage and how they use it at home. This in turn can assist community engagement and education around nutrition, meal planning and purchasing as well as community waste education.

  9. Reference waste package environment report

    International Nuclear Information System (INIS)

    Glassley, W.E.

    1986-01-01

    One of three candidate repository sites for high-level radioactive waste packages is located at Yucca Mountain, Nevada, in rhyolitic tuff 700 to 1400 ft above the static water table. Calculations indicate that the package environment will experience a maximum temperature of ∼230 0 C at 9 years after emplacement. For the next 300 years the rock within 1 m of the waste packages will remain dehydrated. Preliminary results suggest that the waste package radiation field will have very little effect on the mechanical properties of the rock. Radiolysis products will have a negligible effect on the rock even after rehydration. Unfractured specimens of repository rock show no change in hydrologic characteristics during repeated dehydration-rehydration cycles. Fractured samples with initially high permeabilities show a striking permeability decrease during dehydration-rehydration cycling, which may be due to fracture healing via deposition of silica. Rock-water interaction studies demonstrate low and benign levels of anions and most cations. The development of sorptive secondary phases such as zeolites and clays suggests that anticipated rock-water interaction may produce beneficial changes in the package environment

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

    Energy Technology Data Exchange (ETDEWEB)

    BOEHNKE, W.M.

    2000-07-13

    The steel waste package is used primarily for the shipment of remote-handled radioactive waste from the 324 Building to the 200 Area for interim storage. The steel waste package is authorized for shipment of transuranic isotopes. The maximum allowable radioactive material that is authorized is 500,000 Ci. This exceeds the highway route controlled quantity (3,000 A{sub 2}s) and is a type B packaging.

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

    International Nuclear Information System (INIS)

    BOEHNKE, W.M.

    2000-01-01

    The steel waste package is used primarily for the shipment of remote-handled radioactive waste from the 324 Building to the 200 Area for interim storage. The steel waste package is authorized for shipment of transuranic isotopes. The maximum allowable radioactive material that is authorized is 500,000 Ci. This exceeds the highway route controlled quantity (3,000 A 2 s) and is a type B packaging

  12. Generalized waste package containment model

    International Nuclear Information System (INIS)

    Liebetrau, A.M.; Apted, M.J.

    1985-02-01

    The US Department of Energy (DOE) is developing a performance assessment strategy to demonstrate compliance with standards and technical requirements of the Environmental Protection Agency (EPA) and the Nuclear Regulatory Commission (NRC) for the permanent disposal of high-level nuclear wastes in geologic repositories. One aspect of this strategy is the development of a unified performance model of the entire geologic repository system. Details of a generalized waste package containment (WPC) model and its relationship with other components of an overall repository model are presented in this paper. The WPC model provides stochastically determined estimates of the distributions of times-to-failure of the barriers of a waste package by various corrosion mechanisms and degradation processes. The model consists of a series of modules which employ various combinations of stochastic (probabilistic) and mechanistic process models, and which are individually designed to reflect the current state of knowledge. The WPC model is designed not only to take account of various site-specific conditions and processes, but also to deal with a wide range of site, repository, and waste package configurations. 11 refs., 3 figs., 2 tabs

  13. Safety evaluation for packaging (onsite) concrete-lined waste packaging

    Energy Technology Data Exchange (ETDEWEB)

    Romano, T.

    1997-09-25

    The Pacific Northwest National Laboratory developed a package to ship Type A, non-transuranic, fissile excepted quantities of liquid or solid radioactive material and radioactive mixed waste to the Central Waste Complex for storage on the Hanford Site.

  14. Waste package/repository impact study: Final report

    Energy Technology Data Exchange (ETDEWEB)

    1985-09-01

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

  15. Waste package/repository impact study: Final report

    International Nuclear Information System (INIS)

    1985-09-01

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

  16. Waste package performance in unsaturated rock

    International Nuclear Information System (INIS)

    Pigford, T.H.; Lee, W.W.-L.

    1989-03-01

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

  17. Transport packages for nuclear material and waste

    International Nuclear Information System (INIS)

    1997-01-01

    The regulations and responsibilities concerning the transport packages of nuclear materials and waste are given in the guide. The approval procedure, control of manufacturing, commissioning of the packaging and the control of use are specified. (13 refs.)

  18. [The Waste Package Project. Final report, July 1, 1995--February 27, 1996]: Volume 3, Stress study in faulted tunnel models by combined photoelastic measurements and finite element analysis

    International Nuclear Information System (INIS)

    Ladkany, S.G.; Huang, Yuping.

    1996-01-01

    The aim of this part of the Nuclear Waste Package Project research at UNLV is to investigate the stresses in a model of a faulted mountain and the effect of the fault on the stability of drifts in a proposed High Level Nuclear Waste Repository. An investigation was performed to develop a proper technique for analyzing the stresses in and around three adjacent scaled tunnel models, along with the stress concentration factors resulting from the existence of a fault that penetrates two of the three tunnels, at an inclined angel of 44 degrees to the horizontal plane. The results and experience gained from this investigation will be used in a future project in which a full-size repository drift and a penetrating fault will be modeled and analyzed

  19. Interim storage of radioactive waste packages

    International Nuclear Information System (INIS)

    1998-01-01

    This report covers all the principal aspects of production and interim storage of radioactive waste packages. The latest design solutions of waste storage facilities and the operational experiences of developed countries are described and evaluated in order to assist developing Member States in decision making and design and construction of their own storage facilities. This report is applicable to any category of radioactive waste package prepared for interim storage, including conditioned spent fuel, high level waste and sealed radiation sources. This report addresses the following issues: safety principles and requirements for storage of waste packages; treatment and conditioning methods for the main categories of radioactive waste; examples of existing interim storage facilities for LILW, spent fuel and high level waste; operational experience of Member States in waste storage operations including control of storage conditions, surveillance of waste packages and observation of the behaviour of waste packages during storage; retrieval of waste packages from storage facilities; technical and administrative measures that will ensure optimal performance of waste packages subject to various periods of interim storage

  20. Waste package performance allocation system study report

    International Nuclear Information System (INIS)

    Memory, R.D.

    1994-01-01

    The Waste Package Performance Allocation system study was performed in order to provide a technical basis for the selection of the waste package period of substantially complete containment and its resultant contribution to the overall total system performance. This study began with a reference case based on the current Mined Geologic Disposal System (MGDS) baseline design and added a number of alternative designs. The waste package designs were selected from the designs being considered in detail during Advanced Conceptual Design (ACD). The waste packages considered were multi-barrier packages with a 0.95 cm Alloy 825 inner barrier and a 10, 20, or 45 cm thick carbon steel outer barrier. The waste package capacities varied from 6 to 12 to 21 Pressurized Water Reactor (PWR) fuel assemblies. The vertical borehole and in-drift emplacement modes were also considered, as were thermal loadings of 25, 57, and 114 kW/acre. The repository cost analysis indicated that the 21 PWR in-drift emplacement mode option with the 10 cm and 20 cm outer barrier thicknesses are the least expensive and that the 12 PWR in-drift case has approximately the same cost as the 6 PWR vertical borehole. It was also found that the cost increase from the 10 cm outer barrier waste package to the 20 cm waste package was less per centimeter than the increase from the 20 cm outer barrier waste package to the 45 cm outer barrier waste package. However, the repository cost was nearly linear with the outer barrier thickness for the 21 PWR in-drift case. Finally, corrosion rate estimates are provided and the relationship of repository cost versus waste package lifetime is discussed as is cumulative radionuclide release from the waste package and to the accessible environment for time periods of 10,000 years and 100,000 years

  1. Packaged low-level waste verification system

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

  2. Deliberated opinion of the Environment Authority on the project of an installation for the packaging and warehousing of activated wastes (ICEDA) in Saint-Vulbas (01)

    International Nuclear Information System (INIS)

    2013-01-01

    This report proposes an environmental review of the project of an installation of packaging and warehousing of activated wastes (ICEDA) which aims at the warehousing for a maximum duration of 50 years of radioactive wastes produced during the dismantling of the first nuclear reactors, notably in expectation of a definitive solution of deep geological storage which is currently being studied. The Environment Authority has notably examined the design modifications aimed at improving soil bearing capacity, the relationship between the estimation of waste volumes to be warehoused and their emissions on the one hand, and the design and dimensions of buildings on the other hand, and the taking of the flooding and seismic risks into account in building design. While formulating critics and recommendations on the content of the submitted documents, the report presents the context and the project, indicates the concerned administrative procedures, evokes the main environmental issues related to the building permit, gives an overview of the program to which this project belongs, and briefly addresses the issues of building design and sizing. The next part analyses, comments and discusses the content of the impact study (analysis of the initial condition notably regarding flooding and seismic risks, radiological and geological impacts of the project in its exploitation phase)

  3. Nuclear waste package design for the Vadose zone in tuff

    International Nuclear Information System (INIS)

    O'Neal, W.C.; Ballou, L.B.; Gregg, D.W.; Russell, E.W.

    1984-02-01

    This report presents an overview of the selection and analysis of conceptual waste package designs that will be used by the Nevada Nuclear Waste Storage Investigations (NNWSI) project for disposal of high-level nuclear waste (HLW) at the proposed Yucca Mountain, Nevada Site. The design requirements that the waste packages are required to meet are listed. Concept drawings for the reference designs and one alternative package design are shown. Four metal alloys; 304L SS, 321 SS, 316L SS and Incoloy 825 have been selected for candidate canister/overpack materials, and 1020 carbon steel has been selected as the reference metal for the borehole liners. A summary of the results of technical and economic analysis supporting the selection of the conceptual waste package designs is included. Post-closure containment and release rates are not discussed in this paper. 17 references, 2 figures, 2 tables

  4. NWTS waste package program plan. Volume II. Program logic networks

    International Nuclear Information System (INIS)

    1981-10-01

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

  5. Development of Specifications for Radioactive Waste Packages

    International Nuclear Information System (INIS)

    2006-10-01

    The main objective of this publication is to provide guidelines for the development of waste package specifications that comply with waste acceptance requirements for storage and disposal of radioactive waste. It will assist waste generators and waste package producers in selecting the most significant parameters and in developing and implementing specifications for each individual type of waste and waste package. This publication also identifies and reviews the activities and technical provisions that are necessary to meet safety requirements; in particular, selection of the significant safety parameters and preparation of specifications for waste forms, waste containers and waste packages using proven approaches, methods and technologies. This report provides guidance using a systematic, stepwise approach, integrating the technical, organizational and administrative factors that need to be considered at each step of planning and implementing waste package design, fabrication, approval, quality assurance and control. The report reflects the considerable experience and knowledge that has been accumulated in the IAEA Member States and is consistent with the current international requirements, principles, standards and guidance for the safe management of radioactive waste

  6. Development of Specifications for Radioactive Waste Packages

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2006-10-15

    The main objective of this publication is to provide guidelines for the development of waste package specifications that comply with waste acceptance requirements for storage and disposal of radioactive waste. It will assist waste generators and waste package producers in selecting the most significant parameters and in developing and implementing specifications for each individual type of waste and waste package. This publication also identifies and reviews the activities and technical provisions that are necessary to meet safety requirements; in particular, selection of the significant safety parameters and preparation of specifications for waste forms, waste containers and waste packages using proven approaches, methods and technologies. This report provides guidance using a systematic, stepwise approach, integrating the technical, organizational and administrative factors that need to be considered at each step of planning and implementing waste package design, fabrication, approval, quality assurance and control. The report reflects the considerable experience and knowledge that has been accumulated in the IAEA Member States and is consistent with the current international requirements, principles, standards and guidance for the safe management of radioactive waste.

  7. CERAMIC WASTE FORM DATA PACKAGE

    Energy Technology Data Exchange (ETDEWEB)

    Amoroso, J.; Marra, J.

    2014-06-13

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

  8. Project monitoring package (PMP) : A package for project activity monitoring

    International Nuclear Information System (INIS)

    Vyas, K.N.; Kannan, A.; Susandhi, R.; Basu, S.

    1987-01-01

    A package for preparing PERT/CPM network diagrams has been written for PDP-11/34. The program uses PLOT-10 library calls for device interfacing. The package is essentially non-interactive in nature, and reads input data in the form of activity description and duration. It calculates the critical path time and performs time scaling of the events. The report gives a brief outline of the logic used, a sample plot and tabular output for reference. An additional facility for performing project activity monitoring has also been implemented. Activity monitoring generally requires various reports such as feed back reports from various group co-ordinators, information report for project co-ordinator and brief periodical reports for management. A package 'DATATRIEVE' (DTR) on PDP-11/34 system is utilized for generating the above mentioned reports. As DTR can also use normal sequential files, an interfacing program has been written which reformats the files accepted by PERT program acceptable to DTR. Various types of reports as generated by DTR are included. However this part of the package is not transportable and can be implemented only on systems having DTR. 6 figures. (author)

  9. Packaging radioactive wastes for geologic disposal

    International Nuclear Information System (INIS)

    Benton, H.A.

    1996-01-01

    The M ampersand O contractor for the DOE Office of Civilian Radioactive Waste Management is developing designs of waste packages that will contain the spent nuclear fuel assemblies from commercial and Navy reactor plants and various civilian and government research reactor plants, as well as high-level wastes vitrified in glass. The safe and cost effective disposal of the large and growing stockpile of nuclear waste is of national concern and has generated political and technical debate. This paper addresses the technical aspects of disposing of these wastes in large and robust waste packages. The paper discusses the evolution of waste package design and describes the current concepts. In addition, the engineering and regulatory issues that have governed the development are summarized and the expected performance in meeting the requirements are discussed

  10. Engineered waste-package-system design specification

    International Nuclear Information System (INIS)

    1983-05-01

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

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

    Science.gov (United States)

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

    2016-10-01

    Packaging waste is a major issue in several countries. Representing in industrialized countries around 30-35% of municipal solid waste yearly generated, this waste stream has steadily grown over the years even if, especially in Europe, specific recycling and recovery targets have been fixed. Therefore, an increasing attention starts to be devoted to prevention measures and interventions. Filling a gap in the current literature, this explorative paper is a first attempt to map the increasingly important phenomenon of prevention policies in the packaging sector. Through a theoretical sampling, 11 countries/states (7 in and 4 outside Europe) have been selected and analyzed by gathering and studying primary and secondary data. Results show evidence of three specific trends in packaging waste prevention policies: fostering the adoption of measures directed at improving packaging design and production through an extensive use of the life cycle assessment; raising the awareness of final consumers by increasing the accountability of firms; promoting collaborative efforts along the packaging supply chains. Copyright © 2016 Elsevier Ltd. All rights reserved.

  12. Development of waste packages for tuff

    International Nuclear Information System (INIS)

    Rothman, A.J.

    1982-01-01

    The objective of this program is to develop nuclear waste packages that meet the Nuclear Regulatory Commission's requirements for a licensed repository in tuff at the Nevada Test Site. Selected accomplishments for FY82 are: (1) Selection, collection of rock, and characterization of suitable outcrops (for lab experiments); (2) Rock-water interactions (Bullfrog Tuff); (3) Corrosion tests of ferrous metals; (4) Thermal modeling of waste package in host rock; (5) Preliminary fabrication tests of alternate backfills (crushed tuff); (6) Reviewed Westinghouse conceptual waste package designs for tuff and began modification for unsaturated zone; and (7) Waste Package Codes (BARIER and WAPPA) now running on our computer. Brief discussions are presented for rock-water interactions, corrosion tests of ferrous metals, and thermal and radionuclide migration modelling

  13. Evaluation and compilation of DOE waste package test data

    International Nuclear Information System (INIS)

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

    1990-11-01

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

  14. A waste package strategy for regulatory compliance

    International Nuclear Information System (INIS)

    Stahl, D.; Cloninger, M.O.

    1990-01-01

    This paper summarizes the strategy given in the Site Characterization Plan for demonstrating compliance with the post closure performance objectives for the waste package and the Engineered Barrier System contained in the Code of Federal Regulations. The strategy consists of the development of a conservative waste package design that will meet the regulatory requirements with sufficient margin for uncertainty using a multi-barrier approach that takes advantage of the unsaturated nature of the Yucca Mountain site. 7 refs., 1 fig

  15. Status of ERDA TRU waste packaging study

    International Nuclear Information System (INIS)

    Doty, J.W. Jr.

    1977-01-01

    This paper discusses the status of Task 3 of the TRU Waste Cyclone Drum Incinerator and Treatment System program. This task covers acceptable TRU packaging for interim storage and terminal isolation. The kind of TRU wastes generated by contractors and its transport are discussed. Both drum and box systems are desirable

  16. ERG review of waste package corrosion mechanisms

    International Nuclear Information System (INIS)

    Geisert, R.E.

    1988-01-01

    The Engineering Review Group (ERG) was established by the Office of Nuclear Waste Isolation (ONWI) to help evaluate engineering-related issues in the US Department of Energy's nuclear waste repository program. The ERG reviewed the waste package corrosion mechanisms. This report documents the ERG's comments and recommendations on these subjects and the ONWI response to the specific points raised by the ERG. 1 ref

  17. Aqueous Corrosion Rates for Waste Package Materials

    Energy Technology Data Exchange (ETDEWEB)

    S. Arthur

    2004-10-08

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

  18. Aqueous Corrosion Rates for Waste Package Materials

    International Nuclear Information System (INIS)

    Arthur, S.

    2004-01-01

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

  19. Large transport packages for decommissioning waste

    International Nuclear Information System (INIS)

    Price, M.S.T.

    1988-03-01

    The main tasks performed during the period related to the influence of manufacture, transport and disposal on the design of such packages. It is deduced that decommissioning wastes will be transported under the IAEA Transport Regulations under either the Type B or Low Specific Activity (LSA) categories. If the LSA packages are self-shielded, reinforced concrete is the preferred material of construction. But the high cost of disposal implies that there is a strong reason to investigate the use of returnable shields for LSA packages and in such cases they are likely to be made of ferrous metal. Economic considerations favour the use of spheroidal graphite cast iron for this purpose. Transport operating hazards have been investigated using a mixture of desk studies, routes surveys and operations data from the railway organisations. Reference routes were chosen in the Federal Republic of Germany, France and the United Kingdom. This work has led to a description of ten accident scenarios and an evaluation of the associated accident probabilities. The effect of disposal on design of packages has been assessed in terms of the radiological impact of decommissioning wastes, an in addition corrosion and gas evolution have been examined. The inventory of radionuclides in a decommissioning waste package has low environmental impact. If metal clad reinforced concrete packages are to be used, the amount of gas evolution is such that a vent would need to be included in the design. Similar unclad packages would be sufficiently permeable to gases to prevent a pressure build-up. (author)

  20. Symmetric Rock Fall on Waste Package

    International Nuclear Information System (INIS)

    Sreten Mastilovic

    2001-01-01

    The objective of this calculation is to determine the structural response of the Naval SNF (spent nuclear fuel) Waste Package (WP) and the emplacement pallet (EP) subjected to the rock fall DBE (design basis event) dynamic loads. The scope of this calculation is limited to reporting the calculation results in terms of stress intensities and residual stresses in the WP, and stress intensities and maximum permanent downward displacements of the EP-lifting surface. The information provided by the sketches (Attachment I) is that of the potential design of the type of WP and EP considered in this calculation, and all obtained results are valid for those designs only. This calculation is associated with the waste package design and is performed by the Waste Package Design Section in accordance with Reference 24. AP-3.124, ''Calculations'', is used to perform the calculation and develop the document

  1. Large transport packages for decommissioning waste

    International Nuclear Information System (INIS)

    Price, M.S.T.

    1988-08-01

    This document reports progress on a study of large transport packages for decommissioning waste and is the semi-annual report for the period 1 January - 30 June 1988. The main tasks performed during the period related to the assembly of package design criteria ie those aspects of manufacture, handling, storage, transport and disposal which impose constraints on design. This work was synthesised into a design specification for packages which formed the conclusion of that task and was the entry into the final task - the development of package design concepts. The design specifications, which concentrated on the Industrial Package category of the IAEA Transport Regulations, has been interpreted for the two main concepts (a) a self-shielded package disposed of in its entirety and (b) a package with returnable shielding. Preliminary information has been prepared on the cost of providing the package as well as transport to a repository and disposal. There is considerable uncertainty about the cost of disposal and variations of over a factor of 10 are possible. Under these circumstances there is merit in choosing a design concept which is relatively insensitive to disposal cost variations. The initial results indicate that on these grounds the package with returnable shielding is preferred. (author)

  2. Methodologies for certification of transuranic waste packages

    International Nuclear Information System (INIS)

    Christensen, R.N.; Kok, K.D.

    1980-10-01

    The objective of this study was to postulate methodologies for certification that a waste package is acceptable for disposal in a licensed geologic repository. Within the context of this report, certification means the overall process which verifies that a waste package meets the criteria or specifications established for acceptance for disposal in a repository. The overall methodology for certification will include (1) certifying authorities, (2) tests and procedures, and (3) documentation and quality assurance programs. Each criterion will require a methodology that is specific to that criterion. In some cases, different waste forms will require a different methodology. The purpose of predicting certification methodologies is to provide additional information as to what changes, if any, are needed for the TRU waste in storage

  3. EPRI waste processing projects

    International Nuclear Information System (INIS)

    Shaw, R.A.

    1987-01-01

    The Electric Power Research Institute (EPRI) manages research for its sponsoring electric utilities in the United States. Research in the area of low level radioactive waste (LLRW) from light water reactors focuses primarily on waste processing within the nuclear power plants, monitoring of the waste packages, and assessments of disposal technologies. Accompanying these areas and complimentary to them is the determination and evaluation of the sources of nuclear power plants radioactive waste. This paper focuses on source characterization of nuclear power plant waste, LLRW processing within nuclear power plants, and the monitoring of these wastes. EPRI's work in waste disposal technology is described in another paper in this proceeding by the same author. 1 reference, 5 figures

  4. Operational Waste Volume Projection

    Energy Technology Data Exchange (ETDEWEB)

    STRODE, J.N.

    2000-08-28

    Waste receipts to the double-shell tank system are analyzed and wastes through the year 2015 are projected based on generation trends of the past 12 months. A computer simulation of site operations is performed, which results in projections of tank fill schedules, tank transfers, evaporator operations, tank retrieval, and aging waste tank usage. This projection incorporates current budget planning and the clean-up schedule of the Tri-Party Agreement. Assumptions were current as of June. 2000.

  5. Operational waste volume projection

    International Nuclear Information System (INIS)

    Koreski, G.M.; Strode, J.N.

    1995-06-01

    Waste receipts to the double-shell tank system are analyzed and wastes through the year 2015 are projected based on generation trends of the past 12 months. A computer simulation of site operations is performed, which results in projections of tank fill schedules, tank transfers, evaporator operations, tank retrieval, and aging waste tank usage. This projection incorporates current budget planning and the clean-up schedule of the tri-party agreement. Assumptions are current as of June 1995

  6. Operational Waste Volume Projection

    International Nuclear Information System (INIS)

    STRODE, J.N.

    2000-01-01

    Waste receipts to the double-shell tank system are analyzed and wastes through the year 2015 are projected based on generation trends of the past 12 months. A computer simulation of site operations is performed, which results in projections of tank fill schedules, tank transfers, evaporator operations, tank retrieval, and aging waste tank usage. This projection incorporates current budget planning and the clean-up schedule of the Tri-Party Agreement. Assumptions were current as of June. 2000

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

    International Nuclear Information System (INIS)

    O'Neal, W.C.; Gregg, D.W.; Hockman, J.N.; Russell, E.W.; Stein, W.

    1984-01-01

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

  8. Nuclear waste package fabricated from concrete

    International Nuclear Information System (INIS)

    Pfeiffer, P.A.; Kennedy, J.M.

    1987-03-01

    After the United States enacted the Nuclear Waste Policy Act in 1983, the Department of Energy must design, site, build and operate permanent geologic repositories for high-level nuclear waste. The Department of Energy has recently selected three sites, one being the Hanford Site in the state of Washington. At this particular site, the repository will be located in basalt at a depth of approximately 3000 feet deep. The main concern of this site, is contamination of the groundwater by release of radionuclides from the waste package. The waste package basically has three components: the containment barrier (metal or concrete container, in this study concrete will be considered), the waste form, and other materials (such as packing material, emplacement hole liners, etc.). The containment barriers are the primary waste container structural materials and are intended to provide containment of the nuclear waste up to a thousand years after emplacement. After the containment barriers are breached by groundwater, the packing material (expanding sodium bentonite clay) is expected to provide the primary control of release of radionuclide into the immediate repository environment. The loading conditions on the concrete container (from emplacement to approximately 1000 years), will be twofold; (1) internal heat of the high-level waste which could be up to 400 0 C; (2) external hydrostatic pressure up to 1300 psi after the seepage of groundwater has occurred in the emplacement tunnel. A suggested container is a hollow plain concrete cylinder with both ends capped. 7 refs

  9. Igneous Intrusion Impacts on Waste Packages and Waste Forms

    International Nuclear Information System (INIS)

    P. Bernot

    2004-01-01

    The purpose of this model report is to assess the potential impacts of igneous intrusion on waste packages and waste forms in the emplacement drifts at the Yucca Mountain Repository. The model is based on conceptual models and includes an assessment of deleterious dynamic, thermal, hydrologic, and chemical impacts. This constitutes the waste package and waste form impacts submodel of the Total System Performance Assessment for the License Application (TSPA-LA) model assessing the impacts of a hypothetical igneous intrusion event on the repository total system performance. This submodel is carried out in accordance with Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of SR and LA (BSC 2003a) and Total System Performance Assessment-License Application Methods and Approaches (BSC 2002a). The technical work plan is governed by the procedures of AP-SIII.10Q, Models. Any deviations from the technical work plan are documented in the TSPA-LA approach to implementing the models for waste package and waste form response during igneous intrusion is based on identification of damage zones. Zone 1 includes all emplacement drifts intruded by the basalt dike, and Zone 2 includes all other emplacement drifts in the repository that are not in Zone 1. This model report will document the following model: (1) Impacts of magma intrusion on the components of engineered barrier system (e.g., drip shields and cladding) of emplacement drifts in Zone 1, and the fate of waste forms. (2) Impacts of conducting magma heat and diffusing magma gases on the drip shields, waste packages, and cladding in the Zone 2 emplacement drifts adjacent to the intruded drifts. (3) Impacts of intrusion on Zone 1 in-drift thermal and geochemical environments, including seepage hydrochemistry. The scope of this model only includes impacts to the components stated above, and does not include impacts to other engineered barrier system (EBS) components such as the invert and

  10. Igneous Intrusion Impacts on Waste Packages and Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    P. Bernot

    2004-08-16

    The purpose of this model report is to assess the potential impacts of igneous intrusion on waste packages and waste forms in the emplacement drifts at the Yucca Mountain Repository. The model is based on conceptual models and includes an assessment of deleterious dynamic, thermal, hydrologic, and chemical impacts. This constitutes the waste package and waste form impacts submodel of the Total System Performance Assessment for the License Application (TSPA-LA) model assessing the impacts of a hypothetical igneous intrusion event on the repository total system performance. This submodel is carried out in accordance with Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of SR and LA (BSC 2003a) and Total System Performance Assessment-License Application Methods and Approaches (BSC 2002a). The technical work plan is governed by the procedures of AP-SIII.10Q, Models. Any deviations from the technical work plan are documented in the TSPA-LA approach to implementing the models for waste package and waste form response during igneous intrusion is based on identification of damage zones. Zone 1 includes all emplacement drifts intruded by the basalt dike, and Zone 2 includes all other emplacement drifts in the repository that are not in Zone 1. This model report will document the following model: (1) Impacts of magma intrusion on the components of engineered barrier system (e.g., drip shields and cladding) of emplacement drifts in Zone 1, and the fate of waste forms. (2) Impacts of conducting magma heat and diffusing magma gases on the drip shields, waste packages, and cladding in the Zone 2 emplacement drifts adjacent to the intruded drifts. (3) Impacts of intrusion on Zone 1 in-drift thermal and geochemical environments, including seepage hydrochemistry. The scope of this model only includes impacts to the components stated above, and does not include impacts to other engineered barrier system (EBS) components such as the invert and

  11. 10 CFR 60.143 - Monitoring and testing waste packages.

    Science.gov (United States)

    2010-01-01

    ....143 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTES IN... repository operations area, the environment of the waste packages selected for the waste package monitoring program shall be representative of the environment in which the wastes are to be emplaced. (c) The waste...

  12. Quality control concept for radioactive waste packages

    International Nuclear Information System (INIS)

    Warnecke, E.; Martens, B.R.; Odoj, R.

    1990-01-01

    In the Federal Republic of Germany a contract with the BfS for the performance of quality control measures is necessary. It is principally possible to apply two alternative methods: random checks on waste packages or qualification of conditioning processes with subsequent inspections. Priority is given to the control by the process qualification. Both methods have successfully been developed in the Federal Republic of Germany and can be applied. In the course of the qualification of conditioning processes it must be demonstrated by inactive and/or active runs that waste packages are produced which fulfil the waste acceptance requirements. The qualification results in the fixation of a handbook for the operation of the respective conditioning process including the process instrumentation and the operational margins. The qualified process will be inspected to assure the compliance of the actual operation with the conditions fixed in the handbook. (orig./DG)

  13. Improved permeation barriers for tritiated waste packaging

    International Nuclear Information System (INIS)

    Vassallo, G.; Van Den Bergh, R.; Forcey, K.S.; Perujo, A.

    1994-01-01

    High-density polyethylene (HDPE) is extensively used as flexible bagging or packaging for soft tritiated waste in the tritium community because of its low permeability to the more radiotoxic form of tritium, i.e., tritiated water (HTO). However, HDPE does not represent a perfect barrier to HTO nor does it effectively hinder the permeation of elemental tritium, i.e, HT. This latter drawback is particularly important considering that the elemental form may readily convert to HTO outside of the waste package. The possible use of a multilayer film as packing material for the conditioning of tritiated waste is assessed, and its capability to hinder the permeation of elemental tritium is measured and compared with that of bare HDPE. The material investigated is readily available from the food industry. 5 refs., 1 tab

  14. Packaged low-level waste verification system

    International Nuclear Information System (INIS)

    Tuite, K.T.; Winberg, M.; Flores, A.Y.; Killian, E.W.; McIsaac, C.V.

    1996-01-01

    Currently, states and low-level radioactive waste (LLW) disposal site operators have no method of independently verifying the radionuclide content of packaged LLW that arrive at disposal sites for disposal. At this time, disposal sites rely on LLW generator shipping manifests and accompanying records to insure that LLW received meets the waste acceptance criteria. An independent verification system would provide a method of checking generator LLW characterization methods and help ensure that LLW disposed of at disposal facilities meets requirements. The Mobile Low-Level Waste Verification System (MLLWVS) provides the equipment, software, and methods to enable the independent verification of LLW shipping records to insure that disposal site waste acceptance criteria are being met. The MLLWVS system was developed under a cost share subcontract between WMG, Inc., and Lockheed Martin Idaho Technologies through the Department of Energy's National Low-Level Waste Management Program at the Idaho National Engineering Laboratory (INEL)

  15. Characterization of radioactive waste forms and packages

    International Nuclear Information System (INIS)

    1997-01-01

    This publication provides a compendium of waste form, container and waste package properties which are potential importance for waste characterization to support approval for treatment/conditioning, storage and disposal methods and for predicting both short and long term waste behaviour in the repository environment. The properties to be characterized are defined in terms of the technical rationale for their control and characterization. Characterization methods for each property are described in general with reference to detailed discussions existing in the literature. Guidance as to the advantages and disadvantages of individual methods from a technical perspective is also provided where appropriate. This report deals with the characterization of all types of radioactive wastes except spent fuel intended for direct disposal. 115 refs, 17 figs, 12 tabs

  16. IGNEOUS INTRUSION IMPACTS ON WASTE PACKAGES AND WASTE FORMS

    International Nuclear Information System (INIS)

    Bernot, P.

    2004-01-01

    The purpose of this model report is to assess the potential impacts of igneous intrusion on waste packages and waste forms in the emplacement drifts at the Yucca Mountain Repository. The models are based on conceptual models and includes an assessment of deleterious dynamic, thermal, hydrologic, and chemical impacts. The models described in this report constitute the waste package and waste form impacts submodel of the Total System Performance Assessment for the License Application (TSPA-LA) model assessing the impacts of a hypothetical igneous intrusion event on the repository total system performance. This submodel is carried out in accordance with Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of LA (BSC 2004 [DIRS:167796]) and Total System Performance Assessment-License Application Methods and Approaches (BSC 2003 [DIRS: 166296]). The technical work plan was prepared in accordance with AP-2.27Q, Planning for Science Activities. Any deviations from the technical work plan are documented in the following sections as they occur. The TSPA-LA approach to implementing the models for waste package and waste form response during igneous intrusion is based on identification of damage zones. Zone 1 includes all emplacement drifts intruded by the basalt dike, and Zone 2 includes all other emplacement drifts in the repository that are not in Zone 1. This model report will document the following model assessments: (1) Mechanical and thermal impacts of basalt magma intrusion on the invert, waste packages and waste forms of the intersected emplacement drifts of Zone 1. (2) Temperature and pressure trends of basaltic magma intrusion intersecting Zone 1 and their potential effects on waste packages and waste forms in Zone 2 emplacement drifts. (3) Deleterious volatile gases, exsolving from the intruded basalt magma and their potential effects on waste packages of Zone 2 emplacement drifts. (4) Post-intrusive physical

  17. Cleanup Verification Package for the 118-B-1, 105-B Solid Waste Burial Ground

    International Nuclear Information System (INIS)

    Capron, J.M.

    2008-01-01

    This cleanup verification package documents completion of remedial action, sampling activities, and compliance criteria for the 118-B-1, 105-B Solid Waste Burial Ground. This waste site was the primary burial ground for general wastes from the operation of the 105-B Reactor and P-10 Tritium Separation Project and also received waste from the 105-N Reactor. The burial ground received reactor hardware, process piping and tubing, fuel spacers, glassware, electrical components, tritium process wastes, soft wastes and other miscellaneous debris

  18. EQ6 Calculations for Chemical Degradation of Navy Waste Packages

    International Nuclear Information System (INIS)

    S. LeStrange

    1999-01-01

    The Monitored Geologic Repository Waste Package Operations of the Civilian Radioactive Waste Management System Management and Operating Contractor (CRWMS M and O) performed calculations to provide input for disposal of spent nuclear fuel (SNF) from the Navy (Refs. 1 and 2). The Navy SNF has been considered for disposal at the potential Yucca Mountain site. For some waste packages, the containment may breach (Ref. 3), allowing the influx of water. Water in the waste package may moderate neutrons, increasing the likelihood of a criticality event within the waste package. The water may gradually leach the fissile components and neutron absorbers out of the waste package. In addition, the accumulation of silica (SiO 2 ) in the waste package over time may further affect the neutronics of the system. This study presents calculations of the long-term geochemical behavior of waste packages containing the Enhanced Design Alternative (EDA) II inner shell, Navy canister, and basket components. The calculations do not include the Navy SNF in the waste package. The specific study objectives were to determine the chemical composition of the water and the quantity of silicon (Si) and other solid corrosion products in the waste package during the first million years after the waste package is breached. The results of this calculation will be used to ensure that the type and amount of criticality control material used in the waste package design will prevent criticality

  19. EQ6 Calculations for Chemical Degradation of Navy Waste Packages

    Energy Technology Data Exchange (ETDEWEB)

    S. LeStrange

    1999-11-15

    The Monitored Geologic Repository Waste Package Operations of the Civilian Radioactive Waste Management System Management & Operating Contractor (CRWMS M&O) performed calculations to provide input for disposal of spent nuclear fuel (SNF) from the Navy (Refs. 1 and 2). The Navy SNF has been considered for disposal at the potential Yucca Mountain site. For some waste packages, the containment may breach (Ref. 3), allowing the influx of water. Water in the waste package may moderate neutrons, increasing the likelihood of a criticality event within the waste package. The water may gradually leach the fissile components and neutron absorbers out of the waste package. In addition, the accumulation of silica (SiO{sub 2}) in the waste package over time may further affect the neutronics of the system. This study presents calculations of the long-term geochemical behavior of waste packages containing the Enhanced Design Alternative (EDA) II inner shell, Navy canister, and basket components. The calculations do not include the Navy SNF in the waste package. The specific study objectives were to determine the chemical composition of the water and the quantity of silicon (Si) and other solid corrosion products in the waste package during the first million years after the waste package is breached. The results of this calculation will be used to ensure that the type and amount of criticality control material used in the waste package design will prevent criticality.

  20. Waste package emplacement borehole option study

    International Nuclear Information System (INIS)

    Streeter, W.S.

    1992-03-01

    This study evaluates the cost and thermal effects of various waste package emplacement configurations that differ in emplacement orientation, number of containers per borehole, and standoff distance at the potential Yucca Mountain nuclear waste repository. In this study, eight additional alternatives to the vertical and horizontal orientation options presented in the Site Characterization Plan Conceptual Design Report are considered. Typical panel layout configurations based on thermal analysis of the waste and cost estimates for design and construction, operations, and closure and decommissioning were made for each emplacement option. For the thermal analysis average waste 10 years out of reactor and the SIM code were used to determine whether the various configurations temperatures would exceed the design criteria for temperature. This study does not make a recommendation for emplacement configuration, but does provide information for comparison of alternatives

  1. Mixed waste chemical compatibility with packaging components

    International Nuclear Information System (INIS)

    Nigrey, P.J.; Conroy, M.; Blalock, L.B.

    1994-01-01

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

  2. Release of powdered material from waste packages

    International Nuclear Information System (INIS)

    Berg, H.P.; Gruendler, D.; Peiffer, F.; Seehars, H.D.

    1990-01-01

    Possible incidents in the operational phase of the planned German repository KONRAD for radioactive waste with negligible heat production were investigated to assess the radiological consequences. For these investigations release fractions of the radioactive materials are required. This paper deals with the determination of the release of powdered material from waste packages under mechanical stress. These determinations were based on experiments. The experimental procedure and the process parameters chosen in accordance with the conditions in the planned repository will be described. The significance of the experimental results is discussed with respect to incidents in the planned repository. 8 figs., 3 tabs

  3. 10 CFR 63.134 - Monitoring and testing waste packages.

    Science.gov (United States)

    2010-01-01

    ....134 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTES IN A... geologic repository operations area, the environment of the waste packages selected for the waste package monitoring program must be representative of the environment in which the wastes are to be emplaced. (c) The...

  4. Thermal analysis of NNWSI conceptual waste package designs

    International Nuclear Information System (INIS)

    Stein, W.; Hockman, J.N.; O'Neal, W.C.

    1984-04-01

    Lawrence Livermore National Laboratory is involved in the design and testing of high-level nuclear waste packages. Many of the aspects of waste package design and testing (e.g., corrosion and leaching) depend in part on the temperature history of the emplaced packages. This report discusses thermal modeling and analysis of various emplaced waste package conceptual designs including the models used, the assumptions and approximations made, and the results obtained. 16 references

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

    International Nuclear Information System (INIS)

    1981-09-01

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

  6. Industrial Waste Landfill IV upgrade package

    International Nuclear Information System (INIS)

    1994-01-01

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

  7. Application of systems engineering to determine performance requirements for repository waste packages

    International Nuclear Information System (INIS)

    Aitken, E.A.; Stimmell, G.L.

    1987-01-01

    The waste package for a nuclear waste repository in salt must contribute substantially to the performance objectives defined by the Salt Repository Project (SRP) general requirements document governing disposal of high-level waste. The waste package is one of the engineered barriers providing containment. In establishing the performance requirements for a project focused on design and fabrication of the waste package, the systems engineering methodology has been used to translate the hierarchy requirements for the repository system to specific performance requirements for design and fabrication of the waste package, a subsystem of the repository. This activity is ongoing and requires a methodology that provides traceability and is capable of iteration as baseline requirements are refined or changed. The purpose of this summary is to describe the methodology being used and the way it can be applied to similar activities in the nuclear industry

  8. Large packages for reactor decommissioning waste

    International Nuclear Information System (INIS)

    Price, M.S.T.

    1991-01-01

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

  9. Packaging of radioactive wastes for sea disposal

    International Nuclear Information System (INIS)

    1981-02-01

    The Convention on the Prevention of Marine Pollution by the Dumping of Wastes and Other Matter, known as the London Dumping Convention was adopted by an inter-governmental conference in London in 1972 and came into force in 1975. In 1977, the IAEA Board of Governors agreed that there is a continuing responsibility for the IAEA to contribute to the effectiveness of the London Dumping Conventions by providing guidance relevant to the various aspects of dumping radioactive wastes at sea. In the light of the above responsibilities, the IAEA organized a Technical Committee Meeting from 3 to 7 December 1979 to assess the current situation concerning the requirements and the practices of packaging radioactive wastes for dumping at sea with a view to providing further guidance on this subject. The present report summarizes the results of this meeting

  10. Repository Waste Package Transporter Shielding Weight Optimization

    International Nuclear Information System (INIS)

    C.E. Sanders; Shiaw-Der Su

    2005-01-01

    The Yucca Mountain repository requires the use of a waste package (WP) transporter to transport a WP from a process facility on the surface to the subsurface for underground emplacement. The transporter is a part of the waste emplacement transport systems, which includes a primary locomotive at the front end and a secondary locomotive at the rear end. The overall system with a WP on board weights over 350 metric tons (MT). With the shielding mass constituting approximately one-third of the total system weight, shielding optimization for minimal weight will benefit the overall transport system with reduced axle requirements and improved maneuverability. With a high contact dose rate on the WP external surface and minimal personnel shielding afforded by the WP, the transporter provides radiation shielding to workers during waste emplacement and retrieval operations. This paper presents the design approach and optimization method used in achieving a shielding configuration with minimal weight

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

    International Nuclear Information System (INIS)

    Oliveira Santos, P. de; Miaw, S.T.W.

    1990-01-01

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

  12. Low-level waste packaging--a managerial perspective

    International Nuclear Information System (INIS)

    Motl, G.P.; Hebbard, L.B. Jr.

    1980-01-01

    This paper emphasizes managerial responsibility for assuring that facility waste is properly packaged. Specifically, existing packaging regulations are summarized, several actual violations are reviewed and, lastly, some recommendations are made to assist managerial personnel in fulfilling their responsibility to ensure that low-level waste is packaged safely and properly before shipment to the disposal site

  13. Basalt waste isolation project overview

    International Nuclear Information System (INIS)

    Dahlem, D.H.

    1987-01-01

    The proposed candidate site for a high-level nuclear waste repository is located beneath the Hanford Nuclear Reservation in southeastern Washington State. At this point, the Hanford Reservation has been selected as one of three preferred candidates in the draft environmental assessment. Project activities have concentrated on understanding the site location with respect to the 10CFR60, 40CFR191, and 10CFR960, identifying critical parameters for design of water package and repository seals, and identifying parameters for repository design. This paper describes the program to evaluate the site and to identify the natural processes that would effect isolation

  14. DHLW Glass Waste Package Criticality Analysis (SCPB:N/A)

    International Nuclear Information System (INIS)

    Davis, J.W.

    1996-01-01

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to determine the viability of the Defense High-Level Waste (DHLW) Glass waste package concept with respect to criticality regulatory requirements in compliance with the goals of the Waste Package Implementation Plan (Ref. 5.1) for conceptual design. These design calculations are performed in sufficient detail to provide a comprehensive comparison base with other design alternatives. The objective of this evaluation is to show to what extent the concept meets the regulatory requirements or indicate additional measures that are required for the intact waste package

  15. Number of Waste Package Hit by Igneous Intrusion

    International Nuclear Information System (INIS)

    M. Wallace

    2004-01-01

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

  16. Transport concept of new waste management system (inner packaging system)

    International Nuclear Information System (INIS)

    Hakozaki, K.; Wada, R.

    2004-01-01

    Kobe Steel, Ltd. (KSL) and Transnuclear Tokyo (TNT) have jointly developed a new waste management system concept (called ''Inner packaging system'') for high dose rate wastes generated from nuclear power plants under cooperation with Tokyo Electric Power Company (TEPCO). The inner packaging system is designed as a total management system dedicated to the wastes from nuclear plants in Japan, covering from the wastes conditioning in power plants up to the disposal in final repository. This paper presents the new waste management system concept

  17. Performance implications of waste package emplacement orientation

    International Nuclear Information System (INIS)

    Wilder, D.G.

    1991-05-01

    Emplacement borehole orientation directly impacts many aspects of the Engineered Barrier System (EBS) and interactions with the near field environment. This paper considers the impacts of orientation on the hydrologic portion of the environment and its interactions with the EBS. The hydrologic environment is considered from a conceptual standpoint, the numerical analyses are left for subsequent work. As reported in this paper, several aspects of the hydrological environment are more favorable for long term performance of vertically oriented rather than horizontally oriented Waste Packages. 19 refs., 15 figs

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

    Energy Technology Data Exchange (ETDEWEB)

    Russell, E.W.; McCright, R.D.; O`Neal, W.C.

    1983-10-12

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

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

    International Nuclear Information System (INIS)

    Russell, E.W.; McCright, R.D.; O'Neal, W.C.

    1983-01-01

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

  20. Initial waste package interaction tests: status report

    International Nuclear Information System (INIS)

    Shade, J.W.; Bradley, D.J.

    1980-12-01

    This report describes the results of some initial investigations of the effects of rock media on the release of simulated fission products from a sngle waste form, PNL reference glass 76-68. All tests assemblies contained a minicanister prepared by pouring molten, U-doped 76-68 glass into a 2-cm-dia stanless steel tube closed at one end. The tubes were cut to 2.5 to 7.5 cm in length to expose a flat glass surface rimmed by the canister wall. A cylindrical, whole rock pellet, cut from one of the rock materials used, was placed on the glass surface then both the canister and rock pellet were packed in the same type of rock media ground to about 75 μm to complete the package. Rock materials used were a quartz monzonite basalt and bedded salt. These packages were run from 4 to 6 weeks in either 125 ml digestion bombs or 850 ml autoclaves capable of direct solution sampling, at either 250 or 150 0 C. Digestion bomb pressures were the vapor pressure of water, 600 psig at 250 0 C, and the autoclaves were pressurized at 2000 psig with an argon overpressure. In general, the solution chemistry of these initial package tests suggests that the rock media is the dominant controlling factor and that rock-water interaction may be similar to that observed in some geothermal areas. In no case was uranium observed in solution above 15 ppB. The observed leach rates of U glass not in contact with potential sinks (rock surfaces and alteration products) have been observed to be considerably higher. Thus the use of leach rates and U concentrations observed from binary leach experiments (waste-form water only) to ascertain long-term environmental consequences appear to be quite conservative compared to actual U release in the waste package experiments. Further evaluation, however, of fission product transport behavior and the role of alteration phases as fission product sinks is required

  1. Commercial Spent Nuclear Fuel Waste Package Misload Analysis

    International Nuclear Information System (INIS)

    J.K. Knudson

    2003-01-01

    The purpose of this calculation is to estimate the probability of misloading a commercial spent nuclear fuel waste package with a fuel assembly(s) that has a reactivity (i.e., enrichment and/or burnup) outside the waste package design. The waste package designs are based on the expected commercial spent nuclear fuel assemblies and previous analyses (Macheret, P. 2001, Section 4.1 and Table 1). For this calculation, a misloaded waste package is defined as a waste package that has a fuel assembly(s) loaded into it with an enrichment and/or burnup outside the waste package design. An example of this type of misload is a fuel assembly designated for the 21-PWR Control Rod waste package being incorrectly loaded into a 21-PWR Absorber Plate waste package. This constitutes a misloaded 21-PWR Absorber Plate waste package, because the reactivity (i.e., enrichment and/or burnup) of a 21-PWR Control Rod waste package fuel assembly is outside the design of a 21-PWR Absorber Plate waste package. These types of misloads (i.e., fuel assembly with enrichment and/or burnup outside waste package design) are the only types that are evaluated in this calculation. This calculation utilizes information from ''Frequency of SNF Misload for Uncanistered Fuel Waste Package'' (CRWMS M and O 1998) as the starting point. The scope of this calculation is limited to the information available. The information is based on the whole population of fuel assemblies and the whole population of waste packages, because there is no information about the arrival of the waste stream at this time. The scope of this calculation deviates from that specified in ''Technical Work Plan for: Risk and Criticality Department'' (BSC 2002a, Section 2.1.30) in that only waste package misload is evaluated. The remaining issues identified (i.e., flooding and geometry reconfiguration) will be addressed elsewhere. The intended use of the calculation is to provide information and inputs to the Preclosure Safety Analysis

  2. 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

  3. Safety evaluation report for packaging (onsite) concrete-lined waste packaging

    International Nuclear Information System (INIS)

    Romano, T.

    1997-01-01

    The Pacific Northwest National Laboratory developed a package to ship Type A, non-transuranic, fissile excepted quantities of liquid or solid radioactive material and radioactive mixed waste to the Central Waste Complex for storage on the Hanford Site

  4. Isotopic analysis of radioactive waste packages (an inexpensive approach)

    International Nuclear Information System (INIS)

    Padula, D.A.; Richmond, J.S.

    1983-01-01

    A computer printout of the isotopic analysis for all radioactive waste packages containing resins, or other aqueous filter media is now required at the disposal sites at Barnwell, South Carolina, and Beatty, Nevada. Richland, Washington requires an isotopic analysis for all radioactive waste packages. The NRC (Nuclear Regulatory Commission), through 10 CFR 61, will require shippers of radioactive waste to classify and label for disposal all radioactive waste forms. These forms include resins, filters, sludges, and dry active waste (trash). The waste classification is to be based upon 10 CFR 61 (Section 1-7). The isotopes upon which waste classification is to be based are tabulated. 7 references, 8 tables

  5. EASYTRACK Project: Work package 3

    DEFF Research Database (Denmark)

    Savio, Enrico; De Chiffre, Leonardo

    ). For this project, the Centre for Geometrical Metrology (CGM) at the Technical University of Denmark (DTU) has taken care of freeform measurements, in collaboration with the Dipartimento di Innovazione Meccanica e Gestionale (DIMEG) - University of Padova - Italy. The present report describes feasibility...... and experimental results of calibrations of freeform parts on CMMs as contribution to development of the ISO 15530 series of standards currently under consideration by the ISO TC 213/WG10....

  6. Post emplacement environment of waste packages

    International Nuclear Information System (INIS)

    Knauss, K.G.; Oversby, V.M.; Wolery, T.J.

    1983-01-01

    Experiments have been conducted as part of the Nevada Nuclear Waste Storage Investigations Project to determine the changes in water chemistry due to reaction of the Topopah Spring tuff with natural groundwater at temperatures up to 150 0 C. The reaction extent has been investigated as a function of rock-to-water ratio, temperature, reaction time, physical state of the samples, and geographic location of the samples within the tuff unit. Results of these experiments will be used to provide information on the water chemistry to be expected if a high-level waste repository were to be constructed in the Topopah Spring tuff. 6 references, 5 figures, 1 table

  7. Progress in waste package and engineered barrier system performance assessment and design

    International Nuclear Information System (INIS)

    Van Luik, A.; Stahl, D.; Harrison, D.

    1993-01-01

    As part of the U.S. Department of Energy's evaluation of site suitability for a potential high-level radioactive waste repository, long-term interactions between the engineered barrier system and the site must be determined. This requires a waste-package/engineered-system design, a description of the environment around the emplacement zone, and models that simulate operative processes describing these engineered/natural systems interactions. Candidate designs are being evaluated, including a more robust, multi-barrier waste package, and a drift emplacement mode. Tools for evaluating designs, and emplacement mode are the currently available waste-package/engineered-system performance assessment codes development for the project. For assessments that support site suitability, environmental impact, or licensing decisions, more capable codes are needed. Code capability requirements are being written, and existing codes are to be evaluated against those requirements. Recommendations are being made to focus waste-packaging/engineered-system code-development

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

    International Nuclear Information System (INIS)

    1981-10-01

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

  9. Is radioactive mixed waste packaging and transportation really a problem

    International Nuclear Information System (INIS)

    McCall, D.L.; Calihan, T.W. III.

    1992-01-01

    Recently, there has been significant concern expressed in the nuclear community over the packaging and transportation of radioactive mixed waste under US Department of Transportation regulation. This concern has grown more intense over the last 5 to 10 years. Generators and regulators have realized that much of the waste shipped as ''low-level radioactive waste'' was in fact ''radioactive mixed waste'' and that these wastes pose unique transportation and disposal problems. Radioactive mixed wastes must, therefore, be correctly identified and classed for shipment. If must also be packaged, marked, labeled, and otherwise prepared to ensure safe transportation and meet applicable storage and disposal requirements, when established. This paper discusses regulations applicable to the packaging and transportation of radioactive mixed waste and identifies effective methods that waste shippers can adopt to meet the current transportation requirements. This paper will include a characterization and description of the waste, authorized packaging, and hazard communication requirements during transportation. Case studies will be sued to assist generators in understanding mixed waste shipment requirements and clarify the requirements necessary to establish a waste shipment program. Although management and disposal of radioactive mixed waste is clearly a critical issue, packaging and transportation of these waste materials is well defined in existing US Department of Transportation hazardous material regulations

  10. Radiaoctive waste packaging for transport and final disposal

    International Nuclear Information System (INIS)

    Suarez, A.A.

    1989-01-01

    Prior and after the conditioning of radioactive wastes is the packaging design of uppermost importance since it will be the first barrier against water and human intrusion. The choice of the proper package according waste category as well criteria utilized for final disposal are shown. (author) [pt

  11. Release rates from waste packages in a salt repository

    International Nuclear Information System (INIS)

    Chambre, P.L.; Hwang, Y.; Lee, W.W.L.; Pigford, T.H.

    1987-06-01

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

  12. Assessment of plastic packaging waste : material origin, methods, properties

    NARCIS (Netherlands)

    Luijsterburg, B.J.; Goossens, J.G.P.

    2014-01-01

    The global plastics production has increased annually and a substantial part is used for packaging (in Europe 39%). Most plastic packages are discarded after a relatively short service life and the resulting plastic packaging waste is subsequently landfilled, incinerated or recycled. Laws of several

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

    Energy Technology Data Exchange (ETDEWEB)

    M. Gross

    2004-10-14

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

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

    International Nuclear Information System (INIS)

    M. Gross

    2004-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-02-15

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

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

    International Nuclear Information System (INIS)

    Proctor, M.L.

    2006-01-01

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

  17. Packaging and transportation manual. Chapter on the packaging and transportation of hazardous and radioactive waste

    International Nuclear Information System (INIS)

    1998-03-01

    The purpose of this chapter is to outline the requirements that Los Alamos National Laboratory employees and contractors must follow when they package and ship hazardous and radioactive waste. This chapter is applied to on-site, intra-Laboratory, and off-site transportation of hazardous and radioactive waste. The chapter contains sections on definitions, responsibilities, written procedures, authorized packaging, quality assurance, documentation for waste shipments, loading and tiedown of waste shipments, on-site routing, packaging and transportation assessment and oversight program, nonconformance reporting, training of personnel, emergency response information, and incident and occurrence reporting. Appendices provide additional detail, references, and guidance on packaging for hazardous and radioactive waste, and guidance for the on-site transport of these wastes

  18. Packaging and transportation manual. Chapter on the packaging and transportation of hazardous and radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

    The purpose of this chapter is to outline the requirements that Los Alamos National Laboratory employees and contractors must follow when they package and ship hazardous and radioactive waste. This chapter is applied to on-site, intra-Laboratory, and off-site transportation of hazardous and radioactive waste. The chapter contains sections on definitions, responsibilities, written procedures, authorized packaging, quality assurance, documentation for waste shipments, loading and tiedown of waste shipments, on-site routing, packaging and transportation assessment and oversight program, nonconformance reporting, training of personnel, emergency response information, and incident and occurrence reporting. Appendices provide additional detail, references, and guidance on packaging for hazardous and radioactive waste, and guidance for the on-site transport of these wastes.

  19. Thermomechanical scoping calculations for the waste package environment tests

    International Nuclear Information System (INIS)

    Butkovich, T.R.; Yow, J.L. Jr.

    1986-03-01

    During the site characterization phase of the Nevada Nuclear Waste Storage Investigation Project, tests are planned to provide field information on the hydrological and thermomechanical environment. These results are needed for assessing performance of stored waste packages emplaced at depth in excavations in a rock mass. Scoping calculations were performed to provide information on displacements and stress levels attained around excavations in the rock mass from imposing a thermal load designed to simulate the heat produced by radioactive decay. In this way, approximate levels of stresses and displacements are available for choosing instrumentation type and sensitivity as well as providing indications for optimizing instrument emplacement during the test. 7 refs., 9 figs., 1 tab

  20. Review of DOE waste package program. Subtask 1.1. National waste package program, April-September 1982

    International Nuclear Information System (INIS)

    Soo, P.

    1983-03-01

    The current effort is part of an ongoing task to evaluate the national high-level waste package effort. It includes evaluations of reference waste form, container, and packing material components with respect to determining how they may contribute to the containment and controlled release of radionuclides after waste packages have been emplaced in salt and basalt repositories. Chemical and mechanical failure/degradation modes for the waste package have been reviewed and the licensing data requirements to demonstrate compliance with NRC performance objectives specified

  1. Cleanup Verification Package for the 600-259 Waste Site

    Energy Technology Data Exchange (ETDEWEB)

    J. M. Capron

    2006-02-09

    This cleanup verification package documents completion of remedial action for the 600-259 waste site. The site was the former site of the Special Waste Form Lysimeter, consisting of commercial reactor isotope waste forms in contact with soils within engineered caissons, and was used by Pacific Northwest National Laboratory to collect data regarding leaching behavior for target analytes. A Grout Waste Test Facility also operated at the site, designed to test leaching rates of grout-solidified low-level radioactive waste.

  2. Cleanup Verification Package for the 600-259 Waste Site

    International Nuclear Information System (INIS)

    Capron, J.M.

    2006-01-01

    This cleanup verification package documents completion of remedial action for the 600-259 waste site. The site was the former site of the Special Waste Form Lysimeter, consisting of commercial reactor isotope waste forms in contact with soils within engineered caissons, and was used by Pacific Northwest National Laboratory to collect data regarding leaching behavior for target analytes. A Grout Waste Test Facility also operated at the site, designed to test leaching rates of grout-solidified low-level radioactive waste

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

    Energy Technology Data Exchange (ETDEWEB)

    K.G. Mon

    2004-10-01

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

  4. Shielding design of radioactive contaminated metal waste packaging

    International Nuclear Information System (INIS)

    Zou Wenhua; Dong Zhiqiang; Yao Zhenyu; Xu Shuhe; Wang Wen

    2015-01-01

    Focusing on the cylindrical source model to calculate γ dose field of waste packages with the relative formulae then derived. By comparing the calculated data of waste packages of type Ⅷ steel box with the monitoring data, it is found that the cylinder source model could accurately reflect the distributions of γ dose of the waste package. Based on the results of the cylindrical source model, a reasonable shielding technology applicable to waste package containers was designed to meet relevant requirements prescribed in standards about the transport and disposal of radioactive materials. The cylinder source model calculated dose distributions for single package in this paper is simple and easy to implement but slightly larger than the monitoring data providing a certain safety margin for the shielding design. It is suitable for radiological engineering practices. (authors)

  5. A history of solid waste packaging at the Hanford Site

    International Nuclear Information System (INIS)

    Duncan, D.R.; Weyns-Rollosson, D.I.; Pottmeyer, J.A.; Stratton, T.J.

    1995-02-01

    Since the initiation of the defense materials product mission, a total of more than 600,000 m 3 of radioactive solid waste has been stored or disposed at the US Department of Energy's (DOE) Hanford Site, located in southeastern Washington State. As the DOE complex prepares for its increasing role in environmental restoration and waste remediation, the characterization of buried and retrievably stored waste will become increasingly important. Key to this characterization is an understanding of the standards and specifications to which waste was packaged; the regulations that mandated these standards and specifications; the practices used for handling and packaging different waste types; and the changes in these practices with time

  6. WIPP waste package testing on simulated DHLW: emplacement

    International Nuclear Information System (INIS)

    Molecke, M.A.

    1984-01-01

    Several series of simulated (nonradioactive) defense high-level waste (DHLW) package tests have been emplaced in the WIPP, a research and development facility authorized to demonstrate the safe disposal of defense-related wastes. The primary purpose of these 3-to-7 year duration tests is to evaluate the in situ materials performance of waste package barriers (canisters, overpacks, backfills, and nonradioactive DHLW glass waste form) for possible future application to a licensed waste repository in salt. This paper describes all test materials, instrumentation, and emplacement and testing techniques, and discusses progress of the various tests. These tests are intended to provide information on materials behavior (i.e., corrosion, metallurgical and geochemical alterations, waste form durability, surface interactions, etc.), as well as comparison between several waste package designs, fabrications details, and actual costs. These experiments involve 18 full-size simulated DHLW packages (approximately 3.0 m x 0.6 m diameter) emplaced in vertical boreholes in the salt drift floor. Six of the test packages contain internal electrical heaters (470 W/canister), and were emplace under approximately reference DHLW repository conditions. Twelve other simulated DHLW packages were emplaced under accelerated-aging or overtest conditions, including the artificial introduction of brine, and a thermal loading approximately three to four times higher than reference. Eight of these 12 test packages contain 1500 W/canister electrical heaters; the other four are filled with DHLW glass. 9 refs., 1 fig

  7. Commercial and ERDA waste packaging criteria: possible similarities and differences

    International Nuclear Information System (INIS)

    Lowrie, B.

    1977-01-01

    The schedule calls for hot operation of two waste repositories by the end of 1985, and these facilities will have to be licensed. This licensing requirement sets the commercial program apart from the ERDA defense waste program. Packaging criteria are discussed for commercial and ERDA wastes. The different NRC, DOT, and EPA criteria are considered

  8. Nuclear Waste Education Project

    International Nuclear Information System (INIS)

    1989-01-01

    In summary, both the Atlanta and Albuquerque pilot seminars achieved the Nuclear Waste Education Project's goal of informing citizens on both the substance and the process of nuclear waste policy so that they can better participate in future nuclear waste decisions. Nuclear waste issues are controversial, and the seminars exposed the nature of the controversy, and utilized the policy debates to create lively and provocative sessions. The format and content of any citizen education curriculum must be made to fit the particular goal that has been chosen. If the Department of Energy and the LWVEF decide to continue to foster an informed dialogue among presenters and participants, the principles of controversial issues education would serve this goal well. If, however, the Department of Energy and/or the LWVEF decide to go beyond imparting information and promoting a lively discussion of the issues, towards some kind of consensus-building process, it would be appropriate to integrate more interactive sessions into the format. As one evaluator wrote, ''In-depth participation in finding solutions or establishing policy -- small group discussion'' would have been preferable to the plenary sessions that mostly were in the form of lectures and expert panel discussion. The evaluator continued by saying, ''Since these [small group discussions] would require more time commitment, they might be part of follow-up workshops focused on particular topics.''

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

    International Nuclear Information System (INIS)

    McGrail, B.P.

    1984-01-01

    Hydrothermal waste package interaction tests were conducted with a mixture of crushed glass, basalt, and steel in methane-containing synthetic basalt groundwater. In the absence of gamma radiolysis, methane was found to have little influence on the corrosion behavior of the waste package constituents. Under gamma radiolysis, methane was found to significantly lower the solution oxidation potential when compared to identical tests without methane. In addition, colloidal hydrocarbon polymers that have been produced under the irradiation conditions of these experiments were not formed. The presence of the waste package constituents apparently inhibited the formation of the polymers. However, the mechanism which prevented their formation was not determined

  10. Uncertainty analysis of nuclear waste package corrosion

    International Nuclear Information System (INIS)

    Kurth, R.E.; Nicolosi, S.L.

    1986-01-01

    This paper describes the results of an evaluation of three uncertainty analysis methods for assessing the possible variability in calculating the corrosion process in a nuclear waste package. The purpose of the study is the determination of how each of three uncertainty analysis methods, Monte Carlo, Latin hypercube sampling (LHS) and a modified discrete probability distribution method, perform in such calculations. The purpose is not to examine the absolute magnitude of the numbers but rather to rank the performance of each of the uncertainty methods in assessing the model variability. In this context it was found that the Monte Carlo method provided the most accurate assessment but at a prohibitively high cost. The modified discrete probability method provided accuracy close to that of the Monte Carlo for a fraction of the cost. The LHS method was found to be too inaccurate for this calculation although it would be appropriate for use in a model which requires substantially more computer time than the one studied in this paper

  11. Hanford Waste Vitrification Project overview and status

    International Nuclear Information System (INIS)

    Swenson, L.D.; Smets, J.L.

    1993-01-01

    The Hanford Waste Vitrification Project (HWVP) is being constructed at the US DOE's Hanford Site in Richland, WA. Engineering and design are being accomplished by Fluor Daniel Inc. in Irvine, CA. Technical input is furnished by Westinghouse Hanford Co. and construction management services by UE ampersand C-Catalytic Inc. The HWVP will immobilize high level nuclear waste in a glass matrix for eventual disposal in the federal repository. The HWVP consists of several structures, the major ones being the Vitrification Building, the Canister Storage Building, fan house, sand filter, waste hold tank, pump house, and administration and construction facilities. Construction started in April 1992 with the clearing and grubbing activities that prepared the site for fencing and construction preparation. Several design packages have been released for procurement activities. The most significant package release is for the Canister Storage Building, which will be the first major structure to be constructed

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

    International Nuclear Information System (INIS)

    Hakl, A.R.

    1981-10-01

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

  13. Overview of the waste/barrier/rock interactions program of the basalt waste isolation project

    International Nuclear Information System (INIS)

    Salter, P.F.; Burnell, J.R.; Lane, D.L.

    1986-01-01

    The waste package waste/barrier/rock interactions testing program of the Basalt Waste Isolation Project is designed to assess the interactions between nuclear waste forms, other waste package components, and the environment in order to evaluate long-term waste package isolation (radionuclide release) behavior. The program involves reacting fully radioactive waste forms with combinations of steel or copper container material and basalt/bentonite packing material in site-specific ground water under anticipated repository conditions to obtain the steady state radionuclide concentrations required to predictively model waste package radionuclide concentrations required to predictively model waste package radionuclide releases. Both static and flow-through autoclaves are being used in the test program to determine radionuclide concentrations as a function of time and groundwater flow rate, and to evaluate the solid phase and water chemistry changes that control those concentrations. This test program, when combined with project hydrologic and geochemical testing and modeling efforts, and natural analog studies, provides the information required to evaluate long-term radionuclide mobility within a waste package emplaced in a basalt repository

  14. Cleanup Verification Package for the 300-18 Waste Site

    International Nuclear Information System (INIS)

    Capron, J.M.

    2005-01-01

    This cleanup verification package documents completion of remedial action for the 300-18 waste site. This site was identified as containing radiologically contaminated soil, metal shavings, nuts, bolts, and concrete

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

    International Nuclear Information System (INIS)

    Smith, M.J.S.; Streatfield, R.E.

    1987-02-01

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

  16. Monitoring and inspection techniques for long term storage of higher activity waste packages

    International Nuclear Information System (INIS)

    Bolton, Gary

    2013-01-01

    In 2009, following recent changes in United Kingdom (UK) Government Policy, the Nuclear Decommissioning Authority (NDA) identified a knowledge gap in the area of long term interim storage of waste packages. A cross-industry Integrated Project Team (IPT) for Interim Storage was created with responsibility for delivering Industry Guidance on the storage of packaged Higher Activity Waste (HAW) for the current UK civil decommissioning and clean-up programmes. This included a remit to direct research and development projects via the NDA's Direct Research Portfolio (DRP) to fill the knowledge gap. The IPT for Interim Storage published Industry Guidance in 2012 which established a method to define generic package performance criteria and made recommendations on monitoring and inspection. The package performance method consists of the following steps; identification of the package safety function, identification of evolutionary processes that may affect safety function performance, determination of measurable indicators of these evolutionary processes and calibration of the indicators into package performance zones. This article provides an overview of three projects funded by the NDA's DRP that the UK National Nuclear Laboratory (NNL) have completed to address monitoring and inspection needs of waste packages in interim storage. (orig.)

  17. Waste Management Process Improvement Project

    International Nuclear Information System (INIS)

    Atwood, J.; Borden, G.; Rangel, G. R.

    2002-01-01

    The Bechtel Hanford-led Environmental Restoration Contractor team's Waste Management Process Improvement Project is working diligently with the U.S. Department of Energy's (DOE) Richland Operations Office to improve the waste management process to meet DOE's need for an efficient, cost-effective program for the management of dangerous, low-level and mixed-low-level waste. Additionally the program must meet all applicable regulatory requirements. The need for improvement was highlighted when a change in the Groundwater/Vadose Zone Integration Project's waste management practices resulted in a larger amount of waste being generated than the waste management organization had been set up to handle

  18. Characterization of silicoaluminates for low and medium activity wastes packaging

    International Nuclear Information System (INIS)

    Rivoallan, A.; Berson, X.

    1996-01-01

    Studies are done in order to demonstrate many advantages (as an important volume reduction and a greater chemical stability) of packaging low and medium activity wastes in crystal structures compared with concrete and bitumen. In order to understand the consequences of hazardous chemical composition (especially anions) in the waste on the characteristics of the mineral packaging, a simulation study is developed with inactive concentrates. It leads to well crystallized structures which have not the same major crystallized phase. (authors)

  19. Non-Destructive Testing for Control of Radioactive Waste Package

    Science.gov (United States)

    Plumeri, S.; Carrel, F.

    2015-10-01

    Characterization and control of radioactive waste packages are important issues in the management of a radioactive waste repository. Therefore, Andra performs quality control inspection on radwaste package before disposal to ensure the compliance of the radwast characteristics with Andra waste disposal specifications and to check the consistency between Andra measurements results and producer declared properties. Objectives of this quality control are: assessment and improvement of producer radwaste packages quality mastery, guarantee of the radwaste disposal safety, maintain of the public confidence. To control radiological characteristics of radwaste package, non-destructive passive methods (gamma spectrometry and neutrons counting) are commonly used. These passive methods may not be sufficient, for instance to control the mass of fissile material contained inside radwaste package. This is particularly true for large concrete hull of heterogeneous radwaste containing several actinides mixed with fission products like 137Cs. Non-destructive active methods, like measurement of photofission delayed neutrons, allow to quantify the global mass of actinides and is a promising method to quantify mass of fissile material. Andra has performed different non-destructive measurements on concrete intermediate-level short lived nuclear waste (ILW-SL) package to control its nuclear material content. These tests have allowed Andra to have a first evaluation of the performance of photofission delayed neutron measurement and to identify development needed to have a reliable method, especially for fissile material mass control in intermediate-level long lived waste package.

  20. Study on retrievability of waste package in geological disposal

    International Nuclear Information System (INIS)

    Hasegawa, Hiroshi; Noda, Masaru

    2002-02-01

    Retrievability of waste packages in geological disposal of high-level radioactive waste has been investigated from a technical aspect in various foreign countries, reflecting a social concern while retrievability is not provided as a technical requirement. This study investigates the concept of reversibility and retrievability in foreign countries and a technical feasibility on retrievability of waste packages in the geological disposal concept shown in the H12 report. The conclusion obtained through this study is as follows: 1. Concept of reversibility and retrievability in foreign countries. Many organizations have reconsidered the retrievability as one option in the geological disposal to improve the reversibility of the stepwise decision-making process and provide the flexibility, even based upon the principle of the geological disposal that retrieval of waste from the repository is not intended. 2. Technical feasibility on the retrievability in disposal concept in the H12 report. It is confirmed to be able to remove the buffer and to retrieve the waste packages by currently available technologies even after the stages following emplacement of the buffer. It must be noted that a large effort and expense would be required for some activities such as the reconstruction of access route if the activities started after a stage of backfilling disposal tunnels. 3. Evaluation of feasibility on the retrievability and extraction of the issues. In the near future, it is necessary to study and confirm the practical workability and economical efficiency for the retrieving method of waste packages proposed in this study, the handling and processing method of removed buffer materials, and the retrieving method of waste packages in the case of degrading the integrity of waste packages or not emplacing the waste packages in the assumed attitude, etc. (author)

  1. 21-PWR Waste Package Side and End Impacts

    International Nuclear Information System (INIS)

    V. Delabrosse

    2003-01-01

    The objective of this calculation is to determine the structural response of a 21-Pressurized Water Reactor (PWR) spent nuclear fuel waste package impacting an unyielding surface. A range of initial velocities and initial angles between the waste package and the unyielding surface is studied. The scope of this calculation is limited to estimating the area of the outer shell (OS) where the residual stress exceeds a given limit (hereafter ''damaged area''). The stress limit is defined as a fraction of the yield strength of the OS material, Alloy 22 (SB-575 N06022), at the appropriate temperature. The design of the 21-PWR waste package used in this calculation is that defined in Reference 8. However, a value of 4 mm was used for the gap between the inner shell and the OS, and the thickness of the OS was reduced by 2 mm. The sketch in Attachment I provides additional information not included in Reference 8. All obtained results are valid for this design only. This calculation is associated with the waste package design and was performed by the Specialty Analyses and Waste Package Design Section. The waste package (i.e. uncanistered spent nuclear fuel disposal container) is classified as Quality Level 1

  2. 21-PWR Waste Package Side and End Impacts

    International Nuclear Information System (INIS)

    T. Schmitt

    2005-01-01

    The objective of this calculation is to determine the structural response of a 21-Pressurized Water Reactor (PWR) spent nuclear fuel waste package impacting an unyielding surface. A range of initial velocities and initial angles between the waste package and the unyielding surface is studied. The scope of this calculation is limited to estimating the area of the outer shell (OS) where the residual stress exceeds a given limit (hereafter ''damaged area''). The stress limit is defined as a fraction of the yield strength of the OS material, Alloy 22 (SB-575 N06022), at the appropriate temperature. The design of the 21-PWR waste package used in this calculation is that defined in Reference 8. However, a value of 4 mm was used for the gap between the inner shell and the OS, and the thickness of the OS was reduced by 2 mm. The sketch in Attachment I provides additional information not included in Reference 8. All obtained results are valid for this design only. This calculation is associated with the waste package design and was performed by the Specialty Analyses and Waste Package Design Section. The waste package (i.e. uncanistered spent nuclear fuel disposal container) is classified as Quality Level 1

  3. STUDY ON PACKAGING WASTE PREVENTION IN ROMANIA

    Directory of Open Access Journals (Sweden)

    Scortar Lucia-Monica

    2013-07-01

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

  4. Polyethylene liners in radioactive mixed waste packages: An engineering study

    International Nuclear Information System (INIS)

    Whitney, G.A.

    1991-05-01

    Westinghouse Hanford Company manages and operates the Hanford Site 200 Area radioactive solid waste treatment, storage, and disposal facilities for the US Department of Energy-Richland Operations Office under contract AC06-87RL10930. These facilities include solid waste disposal sites and radioactive solid waste storage areas. This document is 1 in a series of 25 reports or actions identified in a Solid Waste Management Event Fact Sheet and critique report (Appendix E) to address the problem of stored, leaking 183-H Solar Evaporation Basin waste drums. It specifically addresses the adequacy of polyethylene liners used as internal packaging of radioactive mixed waste. This document is to be used by solid waste generators preparing solid waste for storage at Hanford Site facilities. This document is also intended for use by Westinghouse Hanford Company solid waste technical staff involved with approval and acceptance of radioactive solid waste

  5. 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

  6. CH Packaging Operations for High Wattage Waste at LANL

    International Nuclear Information System (INIS)

    Washington TRU Solutions LLC

    2002-01-01

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

  7. CH Packaging Operations for High Wattage Waste at LANL

    International Nuclear Information System (INIS)

    Washington TRU Solutions LLC

    2003-01-01

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

  8. Computerized waste-accountability shipping and packaging system

    International Nuclear Information System (INIS)

    Jackson, J.A.; Baston, M. Jr.; DeVer, E.A.

    1981-01-01

    The Waste Accountability, Shipping and Packaging System (WASP) is a real-time computerized system designed and implemented by Mound Facility to meet the stringent packaging and reporting requirements of radioactive waste being shipped to burial sites. The system stores packaging data and inspection results for each unit and prepares all necessary documents at the time of shipment. Shipping data specific for each burial site are automatically prepared on magnetic tape for transmission to the computing center at that site. WASP has enabled Mound Facility to effectively meet the requirements of the burial sites, diminishing the possibility of being rejected from a site because of noncompliance

  9. CH Packaging Operations for High Wattage Waste at LANL

    International Nuclear Information System (INIS)

    Washington TRU Solutions LLC

    2002-01-01

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

  10. DESIGN ANALYSIS FOR THE NAVAL SNF WASTE PACKAGE

    International Nuclear Information System (INIS)

    T.L. Mitchell

    2000-01-01

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

  11. Waste package for a repository located in salt

    International Nuclear Information System (INIS)

    Basham, S.J. Jr.

    1983-01-01

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

  12. Integrity of radioactive waste packages at the Yucca mountain repository

    International Nuclear Information System (INIS)

    Sandquist, G.; Biaglow, A.; Huber, M.; Jagmin, C.

    2004-01-01

    Several of the important physical and chemical processes that impact the integrity of the radioactive waste packages planned for disposal at the proposed Repository at Yucca Mountain are examined. These processes are described by the aerodynamic, thermodynamic, and chemical interactions associated with the waste packages. The effects of chemical corrosion, mechanical erosion, temperature distributions throughout the repository environs, interactions of air, water, and solid particles, and radiological and biological influences are addressed. Materials will be exposed to at least 3 conditions threatening the integrity of the waste package: 1) accumulated dust and particles on the package surface and suspended in the air, 2) chemical reactions from deposits on the waste package infrastructure materials and tight contact areas, and crevices, and 3) environmental factors affecting chemical reactions such as moisture, pH, Eh, and radiolysis. All 3 of these conditions can combine and produce damaging impacts upon the thin protective layer on the alloy surface of the waste package. There are certain benefits from the low-temperature operating mode with ambient temperature below 85 Celsius degrees, but the materials could be subjected to a maximum temperature of 180 Celsius degrees which might introduce stress corrosion cracking and high temperature effects

  13. One project's waste is another project's resource

    International Nuclear Information System (INIS)

    Short, J.

    1997-01-01

    The author describes the efforts being made toward pollution prevention within the DOE complex, as a way to reduce overall project costs, in addition to decreasing the amount of waste to be handled. Pollution prevention is a concept which is trying to be ingrained into project planning. Part of the program involves the concept that ultimately the responsibility for waste comes back to the generator. Parts of the program involve efforts to reuse materials and equipment on new projects, to recycle wastes to generate offsetting revenue, and to increase awareness, accountability and incentives so as to stimulate action on this plan. Summaries of examples are presented in tables

  14. Methods for maintaining a record of waste packages during waste processing and storage

    International Nuclear Information System (INIS)

    2005-01-01

    During processing, radioactive waste is converted into waste packages, and then sent for storage and ultimately for disposal. A principal condition for acceptance of a waste package is its full compliance with waste acceptance criteria for disposal or storage. These criteria define the radiological, mechanical, physical, chemical and biological properties of radioactive waste that can, in principle, be changed during waste processing. To declare compliance of a waste package with waste acceptance criteria, a system for generating and maintaining records should be established to record and track all relevant information, from raw waste characteristics, through changes related to waste processing, to final checking and verification of waste package parameters. In parallel, records on processing technology and the operational parameters of technological facilities should adhere to established and approved quality assurance systems. A records system for waste management should be in place, defining the data to be collected and stored at each step of waste processing and using a reliable selection process carried over into the individual steps of the waste processing flow stream. The waste management records system must at the same time ensure selection and maintenance of all the main information, not only providing evidence of compliance of waste package parameters with waste acceptance criteria but also serving as an information source in the case of any future operations involving the stored or disposed waste. Records generated during waste processing are a constituent part of the more complex system of waste management record keeping, covering the entire life cycle of radioactive waste from generation to disposal and even the post-closure period of a disposal facility. The IAEA is systematically working on the preparation of a set of publications to assist its Member States in the development and implementation of such a system. This report covers all the principal

  15. Conceptual waste packaging options for deep borehole disposal

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

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

    International Nuclear Information System (INIS)

    Wick, E.A.

    1985-01-01

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

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

    International Nuclear Information System (INIS)

    Moak, D.P.; Perrin, J.S.

    1986-07-01

    The Engineering Review Group (ERG) was established by the Office of Nuclear Waste Isolation (ONWI) to help evaluate engineering-related issues in the US Department of Energy's nuclear waste repository program. The October 1984 meeting of the ERG reviewed the waste package container materials selection and corrosion. This report documents the ERG's comments and recommendations on these subjects and the ONWI response to the specific points raised by the ERG

  18. From waste packages acceptance criteria to waste packages acceptance process at the Centre de l'Aube disposal facility

    International Nuclear Information System (INIS)

    Dutzer, M.

    2003-01-01

    The Centre de l'Aube disposal facility has now been operated for 10 years. At the end of 2001, about 124,000 m3 of low and intermediate level short lived waste packages, representing 180,000 packages, have been disposed, for a total capacity of 1,000,000 m3. The flow of waste packages is now between 12 and 15,000 m3 per year, that is one third of the flow that was taken into account for the design of the repository. It confirms the efforts by waste generators to minimise waste production. This flow represents 25 to 30,000 packages, 50% are conditioned into the compaction facility of the repository, so that 17,000 packages are disposed per year. 54 disposal vaults have been closed. In 1996-1999, the safety assessment of the repository have been reviewed, taking into account the experience of operation. This assessment was investigated by the regulatory body and, subsequently, a so-called 'definitive license' to operate was granted to ANDRA on September 2, 1999 with updated licensing requirements. Another review will be performed in 2004. To ensure a better consistency with the safety assessment of the facility, Andra issued new technical requirements for waste packages at the end of 2000. Discussions with waste generators also showed that the waste package acceptance process should be improved to provide a more precise definition of operational criteria to comply with in waste conditioning facilities. Consequently, a new approach has been implemented since 2000. (orig.)

  19. DOE progress in assessing the long term performance of waste package materials

    International Nuclear Information System (INIS)

    Berusch, A.; Gause, E.

    1987-01-01

    Under the Nuclear Waste Policy Act of 1982 (NWPA)[1], the US Dept. of Energy (DOE) is conducting activities to select and characterize candidate sites suitable for the construction and operation of a geologic repository for the disposal of high-level nuclear wastes. DOE is funding three first repository projects: Basalt Waste Isolation Project, BWIP; Nevada Nuclear Waste Isolation Project, NNWSI; and Salt Repository Project Office, SRPO. It is essential in the licensing process that DOE demonstrate to the NRC that the long-term performance of the materials and design will be in compliance with the requirements of 10 CFR 60.113 on substantially complete containment within the waste packages for 300 to 1000 years and a controlled release rate from the engineered barrier system (EBS) for 10,000 years of 1 part in 10 5 per year for radionuclides present in defined quantities 100 years after permanent closure. Obviously, the time spans involved make it impractical to base the assessment of the long term performance of waste package materials on real time, prototypical testing. The assessment of performance will be implemented by the use of models that are supported by real time field and laboratory tests, monitoring, and natural analog studies. Each of the repository projects is developing a plan for demonstrating long-term waste package material performance depending on the particular materials and the package-perturbed, time-dependent environment under which the materials must function. An overview of progress in each of these activities for each of the projects is provided in the following

  20. Repository documentation rethought. A comprehensive approach from untreated waste to waste packages for final disposal

    Energy Technology Data Exchange (ETDEWEB)

    Anthofer, Anton Philipp; Schubert, Johannes [VPC GmbH, Dresden (Germany)

    2017-11-15

    The German Act on Reorganization of Responsibility for Nuclear Disposal (Entsorgungsuebergangsgesetz (EntsorgUebG)) adopted in June 2017 provides the energy utilities with the new option of transferring responsibility for their waste packages to the Federal Government. This is conditional on the waste packages being approved for delivery to the Konrad final repository. A comprehensive approach starts with the dismantling of nuclear facilities and extends from waste disposal and packaging planning to final repository documentation. Waste package quality control measures are planned and implemented as early as in the process qualification stage so that the production of waste packages that are suitable for final deposition can be ensured. Optimization of cask and loading configuration can save container and repository volume. Workflow planning also saves time, expenditure and exposure time for personnel at the facilities. VPC has evaluated this experience and developed it into a comprehensive approach.

  1. Peer Review of the Waste Package Material Performance Interim Report

    International Nuclear Information System (INIS)

    J. A. Beavers; T. M. Devine, Jr.; G. S. Frankel; R. H. Jones; R. G. Kelly; R. M. Latanision; J. H. Payer

    2001-01-01

    At the request of the U.S. Department of Energy, Bechtel SAIC Company, LLC, formed the Waste Package Materials Performance Peer Review Panel (the Panel) to review the technical basis for evaluating the long-term performance of waste package materials in a proposed repository at Yucca Mountain, Nevada. This is the interim report of the Panel; a final report will be issued in February 2002. In its work to date, the Panel has identified important issues regarding waste package materials performance. In the remainder of its work, the Panel will address approaches and plans to resolve these issues. In its review to date, the Panel has not found a technical basis to conclude that the waste package materials are unsuitable for long-term containment at the proposed Yucca Mountain Repository. Nevertheless, significant technical issues remain unsettled and, primarily because of the extremely long life required for the waste packages, there will always be some uncertainty in the assessment. A significant base of scientific and engineering knowledge for assessing materials performance does exist and, therefore, the likelihood is great that uncertainty about the long-term performance can be substantially reduced through further experiments and analysis

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

    International Nuclear Information System (INIS)

    Soo, P.

    1985-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Soo, P. (ed.)

    1984-08-01

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

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

    International Nuclear Information System (INIS)

    Soo, P.

    1984-08-01

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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

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

    International Nuclear Information System (INIS)

    Doering, Thomas; Pasupathi, V.

    2002-01-01

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

  7. Initial specifications for nuclear waste package external dimensions and materials

    International Nuclear Information System (INIS)

    Gregg, D.W.; O'Neal, W.C.

    1983-09-01

    Initial specifications of external dimensions and materials for waste package conceptual designs are given for Defense High Level Waste (DHLW), Commercial High Level Waste (CHLW) and Spent Fuel (SF). The designs have been developed for use in a high-level waste repository sited in a tuff media in the unsaturated zone. Drawings for reference and alternative package conceptual designs are presented for each waste form for both vertical and horizontal emplacement configurations. Four metal alloys: 304L SS, 321 SS, 316L SS and Incoloy 825 are considered for the canister or overpack; 1020 carbon steel was selected for horizontal borehole liners, and a preliminary packing material selection is either compressed tuff or compressed tuff containing iron bearing smectite clay as a binder

  8. Measurement of radionuclides in waste packages

    Science.gov (United States)

    Brodzinski, R.L.; Perkins, R.W.; Rieck, H.G.; Wogman, N.A.

    1984-09-12

    A method is described for non-destructively assaying the radionuclide content of solid waste in a sealed container by analysis of the waste's gamma-ray spectrum and neutron emissions. Some radionuclides are measured by characteristic photopeaks in the gamma-ray spectrum; transuranic nuclides are measured by neutron emission rate; other radionuclides are measured by correlation with those already measured.

  9. TRU waste form and package criteria meeting

    Energy Technology Data Exchange (ETDEWEB)

    None

    1977-08-01

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

  10. The Los Alamos National Laboratory Transuranic Waste Retireval Project

    International Nuclear Information System (INIS)

    Montoya, G.M.; Christensen, D.V.; Stanford, A.R.

    1997-01-01

    This paper presents the status of the Los Alamos National Laboratory (LANL) project for remediation of transuranic (TRU) and TRU mixed waste from Pads 1, 2, and 4. Some of the TRU waste packages retrieved from Pad I are anticipated to be part of LANL's initial inventory to be shipped to the Waste Isolation Pilot Plant (WIPP) in April 1998. The TRU Waste Inspectable Storage Project (TWISP) was initiated in February 1993 in response to the New Mexico Environment Department's (NMED's) Consent Agreement for Compliance Order, ''New Mexico Hazardous Waste Agreement (NMHWA) 93-03.'' The TWISP involves the recovery of approximately 16,865 TRU and TRU-mixed waste containers currently under earthen cover on Pads 1, 2, and 4 at Technical Area 54, Area G, and placement of that waste into inspectable storage. All waste will be moved into inspectable storage by September 30, 2003. Waste recovery and storage operations emphasize protection of worker safety, public health, and the environment

  11. Defense Waste Processing Facility Process Simulation Package Life Cycle

    International Nuclear Information System (INIS)

    Reuter, K.

    1991-01-01

    The Defense Waste Processing Facility (DWPF) will be used to immobilize high level liquid radioactive waste into safe, stable, and manageable solid form. The complexity and classification of the facility requires that a performance based operator training to satisfy Department of Energy orders and guidelines. A major portion of the training program will be the application and utilization of Process Simulation Packages to assist in training the Control Room Operators on the fluctionality of the process and the application of the Distribution Control System (DCS) in operating and managing the DWPF process. The packages are being developed by the DWPF Computer and Information Systems Simulation Group. This paper will describe the DWPF Process Simulation Package Life Cycle. The areas of package scope, development, validation, and configuration management will be reviewed and discussed in detail

  12. Mass Transfer Model for a Breached Waste Package

    International Nuclear Information System (INIS)

    Hsu, C.; McClure, J.

    2004-01-01

    The degradation of waste packages, which are used for the disposal of spent nuclear fuel in the repository, can result in configurations that may increase the probability of criticality. A mass transfer model is developed for a breached waste package to account for the entrainment of insoluble particles. In combination with radionuclide decay, soluble advection, and colloidal transport, a complete mass balance of nuclides in the waste package becomes available. The entrainment equations are derived from dimensionless parameters such as drag coefficient and Reynolds number and based on the assumption that insoluble particles are subjected to buoyant force, gravitational force, and drag force only. Particle size distributions are utilized to calculate entrainment concentration along with geochemistry model abstraction to calculate soluble concentration, and colloid model abstraction to calculate colloid concentration and radionuclide sorption. Results are compared with base case geochemistry model, which only considers soluble advection loss

  13. The packaging of intermediate and low level radioactive wastes

    International Nuclear Information System (INIS)

    Flowers, R.H.

    1985-01-01

    Solid radioactive wastes will generally require some kind of packaging to prepare them for a period of storage followed probably by a land burial. In this Paper the specification of the package is discussed in relation to the properties which will facilitate those two phases of the management of the waste. It is concluded that, by adopting the philosophy of redundant barriers for the disposal phase, a suitable package can be specified for any particular waste product even before the repository site has been selected. Low water flow and an appropriate depth to reduce the risk of accidental re-exposure are the technical site parameters for which particular values will have to be assured at that stage. (author)

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

    International Nuclear Information System (INIS)

    McGrail, B.P.

    1984-11-01

    Hydrothermal waste package interaction tests with methane-containing synthetic basalt groundwater have shown that in the absence of gamma radiolysis, methane has little influence on the glass dissolution rate. Gamma radiolysis tests at fluxes of 5.5 x 10 5 and 4.4 x 10 4 R/hr showed that methane-saturated groundwater was more reducing than identical experiments where Ar was substituted for CH 4 . Dissolved methane, therefore, may be beneficial to the waste package in limiting the solubility of redox sensitive radionuclides such a 99 Tc. Hydrocarbon polymers known to form under the irradiation conditions of these tests were not produced. The presence of the waste package constituents apparently inhibited the formation of the polymers, however, the mechanism which prevented their formation was not determined

  15. Swing-Down of 21-PWR Waste Package

    International Nuclear Information System (INIS)

    A.K. Scheider

    2001-01-01

    The objective of this calculation is to determine the structural response of the waste package (WP) swinging down from a horizontally suspended height. The WP used for that purpose is the 21-Pressurized Water Reactor (PWR) WP. The scope of this document is limited to reporting the calculation results in terms of stress intensities. This calculation is associated with the WP design and was performed by the Waste Package Design group in accordance with the ''Technical Work Plan for: Waste Package Design Description for LA'' (Ref. 13). AP-3.12Q, ''Calculations'' (Ref. 18) is used to perform the calculation and develop the document. The information provided by the sketches attached to this calculation is that of the potential design of the type of 21-PWR WP design considered in this calculation and provides the potential dimensions and materials for the 21-PWR WP design

  16. Optimization of an impact limiter for radioactive waste packaging

    International Nuclear Information System (INIS)

    Mourao, Rogerio Pimenta; Mattar Neto, Miguel

    1999-01-01

    A certain class of packages for the transportation of radioactive wastes - type B packages in the transport jargon - is supposed to resist to a series of postulated tests, the most severe for the majority of the packages being the 9 m height drop test. To improve the performance of the packages under this test, impact limiters are added to them, normally as a removable overpack, with the primary goal of reducing the deceleration loads transmitted to the packages and their contents. The first impact limiter concept, developed during the '70s, used a shell-type impact limiter attached to both ends of the package. Later on, wood was tested as impact limiter filling, which improved the package's mechanical performance, but not its thermal resistance. The popularization of the polymeric materials and their growing use in engineer applications have led to the use of these materials in impact limiters, with the extra advantage of the polymers good thermal properties. This paper proposes a methodology for the optimization of an impact limiter for a package for the conditioning of spent sealed sources. Two simplified methods for the design of impact limiters are presented. Finally, a brief discussion is presented on the methodology usually employed in the design of accident-resisting packages. (author)

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

    International Nuclear Information System (INIS)

    1986-02-01

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

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

    International Nuclear Information System (INIS)

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

    1983-03-01

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

  19. Phosphates as packaging materials for separated nuclear wastes

    International Nuclear Information System (INIS)

    Audubert, F.

    2006-10-01

    The author gives an overview of fifteen years of research activities performed within the context of the so-called Bataille bill which recommended in 1991 new investigations on the management of nuclear wastes. She presents studies aimed at the elaboration of phosphates with an apatite structure, and outlines the determination of compositions adapted to iodine, caesium and tri- or tetravalent actinide incorporation. She reports the synthesis of phosphates with a monazite structure for caesium and actinide confinement. Finally, she reports studies dealing with the waste packaging issue (elaboration of packaging matrices, properties)

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

    International Nuclear Information System (INIS)

    1990-10-01

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

  1. Performance analysis of conceptual waste package designs in salt repositories

    International Nuclear Information System (INIS)

    Jansen, G. Jr.; Raines, G.E.; Kircher, J.F.

    1984-01-01

    A performance analysis of commercial high-level waste and spent fuel conceptual package designs in reference repositories in three salt formations was conducted with the WAPPA waste package code. Expected conditions for temperature, stress, brine composition, radiation level, and brine flow rate were used as boundary conditions to compute expected corrosion of a thick-walled overpack of 1025 wrought steel. In all salt formations corrosion by low Mg salt-dissolution brines typical of intrusion scenarios was too slow to cause the package to fail for thousands of years after burial. In high Mg brines judged typical of thermally migrating brines in bedded salt formations, corrosion rates which would otherwise have caused the packages to fail within a few hundred years were limited by brine availability. All of the brine reaching the package was consumed by reaction with the iron in the overpack, thus preventing further corrosion. Uniform brine distribution over the package surface was an important factor in predicting long package lifetimes for the high Mg brines. 14 references, 15 figures

  2. Preliminary assessment of the controlled release of radionuclides from waste packages containing borosilicate waste glass

    International Nuclear Information System (INIS)

    Strachan, D.M.; McGrail, B.P.; Apted, M.J.; Engle, D.W.; Eslinger, P.W.

    1990-06-01

    The purpose of this report is to provide a preliminary assessment of the release-rate for an engineered barriers subsystem (EBS) containing waste packages of defense high-level waste borosilicate glass at geochemical and hydrological conditions similar to the those at Yucca Mountain. The relationship between the proposed Waste Acceptance Preliminary Specifications (WAPS) test of glass- dissolution rate and compliance with the NRC's release-rate criterion is also evaluated. Calculations are reported for three hierarchical levels: EBS analysis, waste-package analysis, and waste-glass analysis. The following conclusions identify those factors that most acutely affect the magnitude of, or uncertainty in, release-rate performance

  3. WASTE PACKAGE OPERATIONS FY99 CLOSURE METHODS REPORT

    Energy Technology Data Exchange (ETDEWEB)

    M. C. Knapp

    1999-09-23

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

  4. WASTE PACKAGE OPERATIONS FY-99 CLOSURE METHODS REPORT

    International Nuclear Information System (INIS)

    M. C. Knapp

    1999-01-01

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

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

    NARCIS (Netherlands)

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

    2014-01-01

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

  6. Concept for waste package environment tests in the Yucca Mountain exploratory shaft

    International Nuclear Information System (INIS)

    Yow, J.L. Jr.

    1985-05-01

    The Nevada Nuclear Waste Storage Investigations (NNWSI) project is studying a tuffaceous rock unit located at Yucca Mountain on the western boundary of the Nevada Test Site, Nye County, Nevada. The objective is to evaluate the suitability of the volcanic rocks located above the water table at Yucca Mountain as a potential location for a repository for high level radioactive waste. As part of the NNWSI project, Lawrence Livermore National Laboratory is responsible for the design of the waste package and for determining the expected performance of the waste package in the repository environment. To design an optimal waste package system for the unsaturated emplacement environment, the mechanisms by which liquid water can return to contact the metal canister after peaking of the thermal load must be established. Definition of these flux and flow mechanisms is essential for estimating canister corrosion modes and rates. Therefore, three waste package environment tests are being designed for the in situ phase of exploratory shaft testing. These tests emphasize measurement techniques that offer the possibility of characterizing the movement of water into and through the pores and fractures of the densely welded Topopah Spring Member. Other measurement techniques will be used to examine the interactions between moisture migration and the thermomechanical rock mass behavior. Three reduced-scale heater tests will use electrical resistive heaters in a horizontal configuration. All three tests are designed to investigate moisture conditions in the rock during heating and cooling phases of a thermal cycle so that the effects of these moisture conditions on the performance of the waste package system may be established. 28 refs., 4 figs., 3 tabs

  7. Testing of the permissible inventories in radioactive waste packages

    International Nuclear Information System (INIS)

    Stegmaier, W.

    1988-01-01

    The inventories of radionuclides in waste packages which are to be stored in repositories are determined in the Waste Acceptance Requirements of the repository and in the Act on Transport of Dangerous Goods. In this report limiting values of relevant radionuclides are given in such a way that it is possible to use them in a standardized manner. The limiting values apply to single radionuclides, for handling mixtures of nuclides it is necessary to use the sum formula. The minimized number of waste packages which must be produced from a given quantity of raw waste and an inventory of radionuclides keeping all parameters can be calculated with the help of the shown calculating sheet. (orig.) [de

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

    International Nuclear Information System (INIS)

    1993-01-01

    Low and intermediate level radioactive wastes are generated at all stages in the nuclear fuel cycle and also from the medical, industrial and research applications of radiation. These wastes can potentially present risks to health and the environment if they are not managed adequately. Their effective management will require the wastes to be safely stored, transported and ultimately disposed of. The waste container, which may be defined as any vessel, drum or box, made from metals, concrete, polymers or composite materials, in which the waste form is placed for interim storage, for transport and/or for final disposal, is an integral part of the whole package for the management of low and intermediate level wastes. It has key roles to play in several stages of the waste management process, starting from the storage of raw wastes and ending with the disposal of conditioned wastes. This report provides an overview of the various roles that a container may play and the factors that are important in each of these roles. This report has two main objectives. The first is to review the main requirements for the design of waste containers. The second is to provide advice on the design, fabrication and handling of different types of containers used in the management of low and intermediate level radioactive solid wastes. Recommendations for design and testing are given, based on the extensive experience available worldwide in waste management. This report is not intended to have any regulatory status or objectives. 56 refs, 16 figs, 10 tabs

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

    International Nuclear Information System (INIS)

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

    1987-10-01

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

  10. Characterization of Old Nuclear Waste Packages Coupling Photon Activation Analysis and Complementary Non-Destructive Techniques

    International Nuclear Information System (INIS)

    Carrel, Frederick; Coulon, Romain; Laine, Frederic; Normand, Stephane; Sari, Adrien; Charbonnier, Bruno; Salmon, Corine

    2013-06-01

    Radiological characterization of nuclear waste packages is an industrial issue in order to select the best mode of storage. The characterization becomes crucial particularly for waste packages produced at the beginning of the French nuclear industry. For the latter, available information is often incomplete and some key parameters are sometimes missing (content of the package, alpha-activity, fissile mass...) In this case, the use of non-destructive methods, both passive and active, is an appropriate solution to characterize nuclear waste packages and to obtain all the information of interest. In this article, we present the results of a complete characterization carried out on the TE 1060 block, which is a nuclear waste package produced during the 1960's in Saclay. This characterization is part of the DEMSAC (Dismantling of Saclay's facilities) project (ICPE part). It has been carried out in the SAPHIR facility, located in Saclay and housing a linear electron accelerator. This work enables to show the great interest of active methods (photon activation analysis and high-energy imaging) as soon as passive techniques encounter severe limitations. (authors)

  11. WAPDEG Analysis of Waste Package and Drip shield Degradation

    International Nuclear Information System (INIS)

    K. Mon

    2004-01-01

    As directed by ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]), an analysis of the degradation of the engineered barrier system (EBS) drip shields and waste packages at the Yucca Mountain repository is developed. The purpose of this activity is to provide the TSPA with inputs and methodologies used to evaluate waste package and drip shield degradation as a function of exposure time under exposure conditions anticipated in the repository. This analysis provides information useful to satisfy ''Yucca Mountain Review Plan, Final Report'' (NRC 2003 [DIRS 163274]) requirements. Several features, events, and processes (FEPs) are also discussed (Section 6.2, Table 15). The previous revision of this report was prepared as a model report in accordance with AP-SIII.10Q, Models. Due to changes in the role of this report since the site recommendation, it no longer contains model development. This revision is prepared as a scientific analysis in accordance with AP-SIII.9Q, ''Scientific Analyses'' and uses models previously validated in (1) ''Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural Material'' (BSC 2004 [DIRS 169985]); (2) ''General Corrosion and Localized Corrosion of Waste Package Outer Barrier'' (BSC 2004 [DIRS 169984]); and (3) ''General Corrosion and Localized Corrosion of Drip Shield'' (BSC 2004 [DIRS 169845]). The integrated waste package degradation (IWPD) analysis presented in this report treats several implementation-related issues, such as defining the number and size of patches per waste package that undergo stress corrosion cracking; recasting the weld flaw analysis in a form as implemented in the Closure Weld Defects (CWD) software; and, general corrosion rate manipulations (e.g., change of scale in Section 6.3.4). The weld flaw portion of this report takes input from an engineering calculation (BSC 2004

  12. Evaluation of alternative spent fuel waste package concepts for a repository in Basalt

    International Nuclear Information System (INIS)

    Hall, G.V.B.; Nair, B.R.

    1986-01-01

    The United States government has established a program for the disposal of spent nuclear fuel and high-level radioactive waste. The Nuclear Waste Policy Act (NWPA) of 1982 requires the first nuclear waste repository to begin receiving high-level radioactive waste in 1998. One of the potentially acceptable sites currently being evaluated is the Hanford Site in the Pasco Basin in the state of Washington where the host rock is basalt. Under the direction of the United States Department of Energy (DOE), Rockwell International's Rockwell Hanford Operations (RHO) has initiated the Basalt Waste Isolation Project (BWIP). The BWIP must design waste packages for emplacement in the repository. As part of the BWIP waste package development program, several alternative designs were considered for the disposal of spent nuclear fuel. This paper describes the concepts that were evaluated, the criteria that was developed for judging their relative merits, and the methodology that was employed. The results of the evaluation show that a Pipe-In-Tunnel design, which uses a long carbon steel pipe for the containment barrier for multiple packages of consolidated spent fuel, has the highest rating. Other designs which had high ratings are also discussed

  13. Industrial Waste Landfill IV upgrade package

    International Nuclear Information System (INIS)

    1994-01-01

    This document consists of page replacements for the Y-12 industrial waste landfill. The cover page is to replace the old page, and a new set of text pages are to replace the old ones. A replacement design drawing is also included

  14. Secondary Waste Form Down Selection Data Package – Ceramicrete

    Energy Technology Data Exchange (ETDEWEB)

    Cantrell, Kirk J.; Westsik, Joseph H.

    2011-08-31

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

  15. Cleanup Verification Package for the 600-47 Waste Site

    International Nuclear Information System (INIS)

    Cutlip, M.J.

    2005-01-01

    This cleanup verification package documents completion of interim remedial action for the 600-47 waste site. This site consisted of several areas of surface debris and contamination near the banks of the Columbia River across from Johnson Island. Contaminated material identified in field surveys included four areas of soil, wood, nuts, bolts, and other metal debris

  16. Cleanup Verification Package for the 300 VTS Waste Site

    International Nuclear Information System (INIS)

    Clark, S.W.; Mitchell, T.H.

    2006-01-01

    This cleanup verification package documents completion of remedial action for the 300 Area Vitrification Test Site, also known as the 300 VTS site. The site was used by Pacific Northwest National Laboratory as a field demonstration site for in situ vitrification of soils containing simulated waste

  17. Cleanup Verification Package for the 300 VTS Waste Site

    Energy Technology Data Exchange (ETDEWEB)

    S. W. Clark and T. H. Mitchell

    2006-03-13

    This cleanup verification package documents completion of remedial action for the 300 Area Vitrification Test Site, also known as the 300 VTS site. The site was used by Pacific Northwest National Laboratory as a field demonstration site for in situ vitrification of soils containing simulated waste.

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  19. Preliminary waste acceptance requirements - Konrad repository project

    International Nuclear Information System (INIS)

    Brennecke, P.W.; Warnecke, E.H.

    1991-01-01

    In Germany, the planned Konrad repository is proposed for the disposal of all types of radioactive wastes whose thermal influence upon the host rock is negligible. The Bundesamt fuer Strahlenschutz has established Preliminary Waste Acceptance Requirements (as of April 1990) for this facility. The respective requirements were developed on the basis of the results of site-specific safety assessments. They include general requirements on the waste packages to be disposed of as well as more specific requirements on the waste forms, the packaging and the radionuclide inventory per waste package. In addition, the delivery of waste packages was regulated. An outline of the structure and the elements of the Preliminary Waste Acceptance Requirements of April 1990 is given including comments on their legal status. (Author)

  20. Selection of barrier metals for a waste package in tuff

    International Nuclear Information System (INIS)

    Russell, E.W.; McCright, R.D.; O'Neal, W.C.

    1983-10-01

    The Nevada Nuclear Waste Storage Investigations (NNWSI) project under the Civilian Radioactive Waste Management Program is planning a repository at Yucca Mountain at the Nevada Test Site for isolation of high-level nuclear waste. Lawrence Livermore National Laboratory is developing designs for an engineered barrier system containing several barriers such as the waste form, a canister and/or an overpack, packing, and near field host rock. In this paper we address the selection of metal containment barriers. 13 references, 4 tables

  1. Selection of barrier metals for a waste package in tuff

    International Nuclear Information System (INIS)

    Russell, E.W.; McCright, R.D.; O'Neal, W.C.

    1983-09-01

    The Nevada Nuclear Waste Storage Investigation (NNWSI) project under the Civilian Radioactive Waste Management Program is planning a repository at Yucca Mountain at the Nevada Test Site for isolation of high-level nuclear waste. LLNL is developing designs for an engineered barrier system containing several barriers such as the waste form, a canister and/or an overpack, packing, and near field host rock. The selection of metal containment barriers is addressed. 13 references

  2. 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

  3. Gas formation in drum waste packages of Paks NPP

    International Nuclear Information System (INIS)

    Molnar, M.; Palcsu, L.; Svingor, E.; Szanto, Z.; Futo, I.; Ormai, P.

    2000-01-01

    Gas composition measurements have been carried out by mass spectrometry analysis of samples taken from the headspace of ten drum waste packages generated and temporarily stored at Paks NPP. Four drums contained compacted solid waste, three drums were filled with grouted (solidified) sludge and three drums contained solid waste without compaction. The drums have been equipped with a special gas outlet system to make repeated sampling possible. Based on the first measurements significant differences in the gas composition and the rate of gas generation among the drums were found. (author)

  4. Guidelines for sea dumping packages of radioactive waste. Revised version.

    International Nuclear Information System (INIS)

    Anon.

    1979-04-01

    The purpose of these Guidelines is to establish general requirements and provide practical information for the design and manufacture of packages for sea dumping of radioactive waste, in accordance with the terms of the OECD Council Decision establishing a Multilateral Consultation and Surveillance Mechanism for Sea Dumping of Radioactive Waste. These Guidelines are in compliance with the IAEA Revised Definition and Recommendations of 1978, for applying the London Dumping Convention to radioactive waste, and are intended for application under the responsibility of the appropriate national authorities of countries participating in the NEA Mechanism

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

    International Nuclear Information System (INIS)

    MANN, F.M.

    2000-01-01

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

  6. 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

  7. Expected environment for waste packages in a salt repository

    International Nuclear Information System (INIS)

    Pederson, L.R.; Clark, D.E.; Hodges, F.N.; McVay, G.L.; Rai, D.

    1983-01-01

    This paper discusses results of recent efforts to define the very near-field (within approximately 2 m) environmental conditions to which waste packages will be exposed in a salt repository. These conditions must be considered in the experimental design for waste package materials testing, which includes corrosion of barrier materials and leaching of waste forms. Site-specific brine compositions have been determined, and standard brine compositions have been selected for testing purposes. Actual brine compositions will vary depending on origin, temperature, irradiation history, and contact with irradiated rock salt. Results of irradiating rock salt, synthetic brines, rock salt/brine mixtures, and reactions of irradiated rock salt with brine solutions are reported. 38 references, 3 figures, 2 tables

  8. Operational considerations in drift emplacement of waste packages

    International Nuclear Information System (INIS)

    Benton, H.A.

    1993-01-01

    This paper discusses the operational considerations as well as the advantages and disadvantages of emplacing waste packages in drifts in a repository. The considerations apply particularly to the potential repository for spent nuclear fuel and high-level waste glass at Yucca Mountain, although most of the considerations and the advantages and disadvantages discussed in this paper do not necessarily represent the official views of the DOE or of the Management and Operations Contractor, since most of these considerations are still under active discussion and the final decisions will not be made for some time - perhaps years. This paper describes the issues, suggests some principles upon which decisions should be based, and states some of the most significant advantages and disadvantages of the emplacement modes, and the associated waste package types and thermal loadings

  9. Environmental monitoring and deep ocean disposal of packaged radioactive waste

    International Nuclear Information System (INIS)

    Mitchell, N.T.; Preston, A.

    1980-01-01

    Environmental monitoring in the context of the dumping of packaged radioactive waste in the deep ocean is discussed in detail. The principles and objectives laid down by the IAEA and the ICRP are reviewed. Monitoring and its relationships to radiation exposure, research, control measures and public information are described. Finally, the actual practice in the UK of environmental monitoring is detailed for the measurable case of liquid wastes in coastal waters and also for package waste in deep oceans which has to be calculated. It is concluded that better mathematical models are needed to predict the dose to man and that more research into oceanographic and biological transfer processes should be carried out. (UK)

  10. Effects of sorption hysteresis on radionuclide releases from waste packages

    International Nuclear Information System (INIS)

    Barney, G.S.; Reed, D.T.

    1985-01-01

    A one-dimensional, numerical transport model was used to calculate radionuclide releases from waste packages emplaced in a nuclear waste repository in basalt. The model incorporates both sorption and desorption isotherm parameters measured previously for sorption of key radionuclides on the packing material component of the waste package. Sorption hysteresis as described by these isotherms lowered releases of some radionuclides by as much as two orders of magnitude. Radionuclides that have low molar inventories (relative to uranium), high solubility, and strongly sorbed, are most affected by sorption hysteresis. In these cases, almost the entire radionuclide inventory is sorbed on the packing material. The model can be used to help optimize the thickness of the packing material layer by comparing release rate versus packing material thickness curves with Nuclear Regulatory Commission (NRC) and Environmental Protection Agency (EPA) release limits

  11. Interim performance specifications for conceptual waste-package designs for geologic isolation in salt repositories

    International Nuclear Information System (INIS)

    1983-06-01

    The interim performance specifications and data requirements presented apply to conceptual waste package designs for all waste forms which will be isolated in salt geologic repositories. The waste package performance specifications and data requirements respond to the waste package performance criteria. Subject areas treated include: containment and controlled release, operational period safety, criticality control, identification, and waste package performance testing requirements. This document was generated for use in the development of conceptual waste package designs in salt. It will be revised as additional data, analyses, and regulatory requirements become available

  12. PROBABILISTIC ANALYSES OF WASTE PACKAGE QUANTITIES IMPACTED BY POTENTIAL IGNEOUS DISRUPTION AT YUCCA MOUNTAIN

    International Nuclear Information System (INIS)

    M.G. Wallace

    2005-01-01

    A probabilistic analysis was conducted to estimate ranges for the numbers of waste packages that could be damaged in a potential future igneous event through a repository at Yucca Mountain. The analyses include disruption from an intrusive igneous event and from an extrusive volcanic event. This analysis supports the evaluation of the potential consequences of future igneous activity as part of the total system performance assessment for the license application for the Yucca Mountain Project (YMP). The first scenario, igneous intrusion, investigated the case where one or more igneous dikes intersect the repository. A swarm of dikes was characterized by distributions of length, width, azimuth, and number of dikes and the spacings between them. Through the use in part of a latin hypercube simulator and a modified video game engine, mathematical relationships were built between those parameters and the number of waste packages hit. Corresponding cumulative distribution function curves (CDFs) for the number of waste packages hit under several different scenarios were calculated. Variations in dike thickness ranges, as well as in repository magma bulkhead positions were examined through sensitivity studies. It was assumed that all waste packages in an emplacement drift would be impacted if that drift were intersected by a dike. Over 10,000 individual simulations were performed. Based on these calculations, out of a total of over 11,000 planned waste packages distributed over an area of approximately 5.5 km 2 , the median number of waste packages impacted was roughly 1/10 of the total. Individual cases ranged from 0 waste packages to the entire inventory being impacted. The igneous intrusion analysis involved an explicit characterization of dike-drift intersections, built upon various distributions that reflect the uncertainties associated with the inputs. The second igneous scenario, volcanic eruption (eruptive conduits), considered the effects of conduits formed in

  13. 77 FR 23751 - Certain Food Waste Disposers and Components and Packaging Thereof; Institution of Investigation...

    Science.gov (United States)

    2012-04-20

    ... States after importation of certain food waste disposers and components and packaging thereof by reason... States after importation of certain food waste disposers and components and packaging thereof by reason... importation of certain food waste disposers and components and packaging thereof that infringe the claim of U...

  14. Quality assurance of radioactive waste packages by computerized tomography

    International Nuclear Information System (INIS)

    Reimers, P.

    1992-01-01

    According to task 3 'Testing and Evaluation of Conditioned Waste and Technical Barriers' quality assurance is a main scope of research concerned with the handling of radioactive waste. It was provided to characterize medium and high active waste by standard test methods which have been developed and experienced in this contract. Quality evaluation of radioactive waste packages is preferentially done by non-destructive testing methods. The main task of this contract was the elaboration of specific non-destructive testing methods for conditioned and sealed waste packages as well as for the matrix materials themselves (e.g. bitumen, concrete, ceramics and glass). CT with X-rays turned out to be one of the best methods for the comprehensive non-destructive characterization of the physical and technical properties of the above described test objects. The method is especially suitable for the non-destructive evaluation of the absolute density value, of the density distribution, of the gamma activity distribution, of the localization of voids, cracks and inclusions, of the visualization of swelling, shrinkage and phase precipitations, as well as the detection of liquid phases in bentonite and cemented waste. 9 refs., 10 figs., 2 tabs

  15. Demonstration tests for low level radioactive waste packaging safety

    International Nuclear Information System (INIS)

    Nagano, I.; Shimura, S.; Miki, T.; Tamamura, T.; Kunitomi, K.

    1993-01-01

    The transport packaging for low level radioactive waste (so-called the LLW packaging) has been developed to be utilized for transportation of LLW in 200 liter-drums from Japanese nuclear power stations to the LLW Disposal Center at Rokkashomura in Aomori Prefecture. Transportation is expected to start from December in 1992. We will explain the brief history of the development, technical features and specifications as well as two kinds of safety demonstration tests, namely one is '1.2 meter free drop test' and the other is 'ISO container standard test'. (J.P.N.)

  16. Development of waste packages for TRU-disposal. 5. Development of cylindrical metal package for TRU wastes

    International Nuclear Information System (INIS)

    Mine, Tatsuya; Mizubayashi, Hiroshi; Asano, Hidekazu; Owada, Hitoshi; Otsuki, Akiyoshi

    2005-01-01

    Development of the TRU waste package for hulls and endpieces compression canisters, which include long-lived and low sorption nuclides like C-14 is essential and will contribute a lot to a reasonable enhancement of safety and economy of the TRU-disposal system. The cylindrical metal package made of carbon steel for canisters to enhance the efficiency of the TRU-disposal system and to economically improve their stacking conditions was developed. The package is a welded cylindrical construction with a deep drawn upper cover and a disc plate for a bottom cover. Since the welding is mainly made only for an upper cover and a bottom disc plate, this package has a better containment performance for radioactive nuclide and can reduce the cost for construction and manufacturing including its welding control. Furthermore, this package can be laid down in pile for stacking in the circular cross section disposal tunnel for the sedimentary rock, which can drastically minimize the space for disposal tunnel as mentioned previously in TRU report. This paper reports the results of the study for application of newly developed metal package into the previous TRU-disposal system and for the stacking equipment for the package. (author)

  17. Effect of alpha and gamma radiation on the near-field chemistry and geochemistry of high-level waste packages

    International Nuclear Information System (INIS)

    Reed, D.T.

    1985-12-01

    Ionizing radiation can potentially alter geochemical and chemical processes in a geologic system. These effects can either enhance or reduce the performance of the waste package in a deep geologic repository. Current indications are that, in a repository located in basalt, ionizing radiation significantly affects geochemical/chemical processes but does not appear to significantly affect factors important to the long-term performance of the repository. The experimental results presented in this paper were obtained as part of an ongoing effort by the Basalt Waste Isolation Project to determine the effect of ionizing radiation on chemical and geochemical processes in the environment of the waste package. Gamma radiolysis experiments were done by subjecting samples of synthetic basalt groundwater in the presence of various waste package components (basalt/packing/low-carbon steel) to high levels of gamma radiation from a 60 Co source. Post-irradiation analysis was done on the gas, liquid, and solid components of the basalt system. The results obtained are important in evaluating waste package performance during the containment period. The effect of alpha radiation on the basalt groundwater system in the presence of waste package components is important in evaluating waste package performance during the isolation period. The experimental work in this area is in a very preliminary stage. Results from two experiments are reported. 9 refs., 4 figs., 7 tabs

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

    Science.gov (United States)

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

    2015-09-01

    This work reports the oxidation behaviour and waste-to-energy output of different semi-rigid and flexible aluminium packagings when incinerated at 850°C in an air atmosphere enriched with 6% oxygen, in the laboratory setting. The physical properties of the different packagings were determined, including their metallic aluminium contents. The ash contents of their combustion products were determined according to standard BS ISO 1171:2010. The net calorific value, the required energy, and the calorific gain associated with each packaging type were determined following standard BS EN 13431:2004. Packagings with an aluminium lamina thickness of >50μm did not fully oxidise. During incineration, the weight-for-weight waste-to-energy output of the packagings with thick aluminium lamina was lower than that of packagings with thin lamina. The calorific gain depended on the degree of oxidation of the metallic aluminium, but was greater than zero for all the packagings studied. Waste aluminium may therefore be said to act as an energy source in municipal solid waste incineration systems. Copyright © 2015 Elsevier Ltd. All rights reserved.

  19. Waste package transfer, emplacement and retrievability in the French deep geological repository

    Energy Technology Data Exchange (ETDEWEB)

    Roulet, Alain; Delort, Daniel; Herve, Jean Francois; Bosgiraud, Jean Michel; Guenin, Jean Jacques [Technical Department ANDRA (France)

    2009-06-15

    Safe, reliable and reversible handling of waste is a significant issue related to the design and safety assessment of deep geological repository in France. The first step taken was to study various waste handling solutions. ANDRA also decided to fabricate and demonstrate industrial scale handling equipment for HLW (since 2003) and for ILW-LL wastes (since 2008). We will review the main equipment developed for the transfer process in the repository, for both types of waste, and underline the benefits of developing industrial demonstrators within the framework of international cooperation agreements. Waste retrieval capability will be simultaneously examined. Two types of waste have to be handled underground in Andra's repository. The HLW disposal package for vitrified waste is a 2 ton carbon steel cylindrical canister with a diameter of 600 mm. The weight of ILW-LL concrete disposal packages range from a minimum of 6 tonnes to over 20 tonnes, and their volume from approximately 5 to 10 m3. The underground transfer to the disposal drift requires moving the disposal package within a shielded transfer cask placed on a trailer. Transfer cask design has evolved since 2005, due to optimisation studies and as a result of industrial feedback from SKB. For HLW handling equipment two design options have been studied. In the first solution (Andra's Dossier 2005), the waste package are emplaced, one at a time, in the disposal drift by a pushing robot. Successive steps in design and proto-typing have lead to improve the design of the equipment and to gain confidence. Recently a fully integrated process has been successfully demonstrated, at full scale, (in a 100 m long mock up drift) as part of the EC funded ESDRED Project. This demonstrator is now on display in Andra's Technology Centre at Saudron, near the Bure Underground Laboratory. The second disposal option which has been investigated is based on a concept of utilising an external apparatus to push a row of

  20. Impacts of cathodic protection on waste package performance

    International Nuclear Information System (INIS)

    Atkins, J.E.; Lee, J.H.; Andrews, R.W.

    1996-01-01

    The current design concept for a multi-barrier waste container for the potential repository at Yucca Mountain, Nevada, calls for an outer barrier of 100 mm thick corrosion-allowance material (CAM) (carbon steel) and an inner barrier of 20 mm thick corrosion-resistant material (CRM) (Alloy 825). Fulfillment of the NRC subsystem requirements (10 CFR 60.113) of substantially complete containment and controlled release of radionuclides from the engineered barrier system (EBS) will rely mostly upon the robust waste container design, among other EBS components. In the current waste container design, some degree of cathodic protection of CRM will be provided by CAM. This paper discusses a sensitivity case study for the impacts of cathodic protection of the inner barrier by the outer barrier on the performance of waste package

  1. Low-Level Radioactive Waste siting simulation information package

    International Nuclear Information System (INIS)

    1985-12-01

    The Department of Energy's National Low-Level Radioactive Waste Management Program has developed a simulation exercise designed to facilitate the process of siting and licensing disposal facilities for low-level radioactive waste. The siting simulation can be conducted at a workshop or conference, can involve 14-70 participants (or more), and requires approximately eight hours to complete. The exercise is available for use by states, regional compacts, or other organizations for use as part of the planning process for low-level waste disposal facilities. This information package describes the development, content, and use of the Low-Level Radioactive Waste Siting Simulation. Information is provided on how to organize a workshop for conducting the simulation. 1 ref., 1 fig

  2. Transportation packagings for high-level wastes and unprocessed transuranic wastes

    International Nuclear Information System (INIS)

    Wilmot, E.L.; Romesberg, L.E.

    1982-01-01

    Packagings used for nuclear waste transport are varied in size, shape, and weight because they must accommodate a wide variety of waste forms and types. However, this paper will discuss the common characteristics among the packagings in order to provide a broad understanding of packaging designs. The paper then discusses, in some detail, a design that has been under development recently at Sandia National Laboratories (SNL) for handling unprocessed, contact-handled transuranic (CHTRU) wastes as well as a cask design for defense high-level wastes (HLW). As presently conceived, the design of the transuranic package transporter (TRUPACT) calls for inner and outer boxes that are separated by a rigid polyurethane foam. The inner box has a steel frame with stainless steel surfaces; the outer box is similarly constructed except that carbon steel is used for the outside surfaces. The access to each box is through hinged doors that are sealed after loading. To meet another waste management need, a cask is being developed to transport defense HLW. The cask, which is at the preliminary design stage, is being developed by General Atomic under the direction of the TTC. The cask design relies heavily on state-of-the-art spent-fuel cask designs though it can be much simpler due to the characteristics of the HLW. A primary purpose of this paper is to show that CHTRU waste and defense HLW currently are and will be transported in packagings designed to meet the hazards of transportation that are present in general commerce

  3. Evaluation and compilation of DOE waste package test data

    International Nuclear Information System (INIS)

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

    1993-06-01

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

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

    DEFF Research Database (Denmark)

    Cimpan, Ciprian; Wenzel, Henrik; Maul, Anja

    2015-01-01

    of economies of scale and discussed complementary relations occurring between capacity size, technology level and operational practice. Processing costs (capital and operational expenditure) per unit waste input were found to decrease from above 100 € for small plants with a basic technology level to 60......This contribution presents the results of a techno-economic analysis performed for German Materials Recovery Facilities (MRFs) which sort commingled lightweight packaging waste (consisting of plastics, metals, beverage cartons and other composite packaging). The study addressed the importance......-70 € for large plants employing advanced process flows. Typical operational practice, often riddled with inadequate process parameters was compared with planned or designed operation. The former was found to significantly influence plant efficiency and therefore possible revenue streams from the sale of output...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-05-23

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

  6. Nanotechnology for the Solid Waste Reduction of Military Food Packaging

    Science.gov (United States)

    2016-06-01

    WP-200816) Nanotechnology for the Solid Waste Reduction of Military Food Packaging June 2016 This document has been cleared for public release...NAME OF RESPONSIBLE PERSON 19b. TELEPHONE NUMBER (Include area code) 01/06/2016 Cost and Performance Report 04/01/2008 - 01/01/2015 Nanotechnology for...Soldier Research, Development and Engineering Center Robin Altmeyer - AmeriQual U.S. Army Natick Soldier Research, Development and Engineering

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

    OpenAIRE

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

    2014-01-01

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

  8. Compatibility of packaging components with simulant mixed waste

    International Nuclear Information System (INIS)

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

    1996-01-01

    The purpose of hazardous and radioactive materials packaging is to enable these materials to be transported without posing a threat to the health or property of the general public. To achieve this aim, regulations in the US have been written establishing general design requirements for such packagings. While no regulations have been written specifically for mixed waste packaging, regulations for the constituents of mixed wastes, i.e., hazardous and radioactive substances, have been codified by the US Department of Transportation (US DOT, 49 CFR 173) and the US Nuclear Regulatory Commission (NRC, 10 CFR 71). Based on these national requirements, a Chemical Compatibility Testing Program was developed in the Transportation Systems Department at Sandia National Laboratories (SNL). The program provides a basis to assure any regulatory body that the issue of packaging material compatibility towards hazardous and radioactive materials has been addressed. In this paper, the authors present the results of the second phase of this testing program. The first phase screened five liner materials and six seal materials towards four simulant mixed wastes. This phase involved the comprehensive testing of five candidate liner materials to an aqueous Hanford Tank simulant mixed waste. The comprehensive testing protocol involved exposing the respective materials a matrix of four gamma radiation doses (∼ 1, 3, 6, and 40 kGy), three temperatures (18, 50, and 60 C), and four exposure times (7, 14, 28, and 180 days). Following their exposure to these combinations of conditions, the materials were evaluated by measuring five material properties. These properties were specific gravity, dimensional changes, hardness, stress cracking, and mechanical properties

  9. Waste package designs for disposal of high-level waste in salt formations

    International Nuclear Information System (INIS)

    Basham, S.J. Jr.; Carr, J.A.

    1984-01-01

    In the United States of America the selected method for disposal of radioactive waste is mined repositories located in suitable geohydrological settings. Currently four types of host rocks are under consideration: tuff, basalt, crystalline rock and salt. Development of waste package designs for incorporation in mined salt repositories is discussed. The three pertinent high-level waste forms are: spent fuel, as disassembled and close-packed fuel pins in a mild steel canister; commercial high-level waste (CHLW), as borosilicate glass in stainless-steel canisters; defence high-level waste (DHLW), as borosilicate glass in stainless-steel canisters. The canisters are production and handling items only. They have no planned long-term isolation function. Each waste form requires a different approach in package design. However, the general geometry and the materials of the three designs are identical. The selected waste package design is an overpack of low carbon steel with a welded closure. This container surrounds the waste forms. Studies to better define brine quantity and composition, radiation effects on the salt and brines, long-term corrosion behaviour of the low carbon steel, and the leaching behaviour of the spent fuel and borosilicate glass waste forms are continuing. (author)

  10. Full-scale testing of waste package inspection system

    International Nuclear Information System (INIS)

    Yagi, T.; Kuribayashi, H.; Moriya, Y.; Fujisawa, H.; Takebayashi, N.

    1989-01-01

    In land disposal of low-level radioactive waste (LLW) in Japan, it is legally required that the waste packages to be disposed of be inspected for conformance to applicable technical regulations prior to shipment from each existing power station. JGC has constructed a fully automatic waste package inspection system for the purpose of obtaining the required design data and proving the performance of the system. This system consists of three inspection units (for visual inspection, surface contamination/dose rate measurement and radioactivity/weight measurement), a labelling unit, a centralized control unit and a drum handling unit. The outstanding features of the system are as follows: The equipment and components are modularized and designed to be of the most compact size and the quality control functions are performed by an advanced centralized control system. The authors discuss how, as a result of the full-scale testing, it has been confirmed that this system satisfies all the performance requirements for the inspection of disposal packages

  11. Aging and Phase Stability of Waste Package Outer Barrier

    Energy Technology Data Exchange (ETDEWEB)

    Tammy S. Edgecumble Summers

    2001-08-23

    This Analysis Model Report (AMR) was prepared in accordance with the Work Direction and Planning Document, ''Aging and Phase Stability of Waste Package Outer Barrier'' (CRWMS M&O 1999a). ICN 01 of this AMR was developed following guidelines provided in TWP-MGR-MD-000004 REV 01, ''Technical Work Plan for: Integrated Management of Technical Product Input Department'' (BSC 2001, Addendum B). It takes into consideration the Enhanced Design Alternative II (EDA II), which has been selected as the preferred design for the Engineered Barrier System (EBS) by the License Application Design Selection (LADS) program team (CRWMS M&O 1999b). The salient features of the EDA II design for this model are a waste package (WP) consisting of an outer barrier of Alloy 22 and an inner barrier of Type 316L stainless steel. This report provides information on the phase stability of Alloy 22l, the current waste-package-outer-barrier (WPOB) material. These phase stability studies are currently divided into three general areas: (1) Long-range order reactions; (2) Intermetallic and carbide precipitation in the base metal; and (3) Intermetallic and carbide precipitation in welded samples.

  12. Role of statistics in characterizing nuclear waste package behavior

    International Nuclear Information System (INIS)

    Bowen, W.M.

    1984-11-01

    The characterization of nuclear waste package behavior is primarily based on the outcome of laboratory tests, where components of a proposed waste package are either individually or simultaneously subjected to simulated repository conditions. At each step of a testing method, both controllable and uncontrollable factors contribute to the overall uncertainty in the final outcome of the test. If not dealt with correctly, these sources of uncertainty could obscure or distort important information that might otherwise be gleaned from the test data. This could result in misleading or erroneous conclusions about the behavior characteristic being studied. It could also preclude estimation of the individual contributions of the major sources of uncertainty to the overall uncertainty. Statistically designed experiments and sampling plans, followed by correctly applied statistical analysis and estimation methods will yield the most information possible for the time and resources spent on experimentation, and they can eliminate the above concerns. Conclusions reached on the basis of such information will be sound and defensible. This presentation is intended to emphasize the importance of correctly applied, theoretically sound statistical methodology in characterizing nuclear waste package behavior. 8 references, 1 table

  13. Role of statistics in characterizing nuclear waste package behavior

    International Nuclear Information System (INIS)

    Bowen, W.M.

    1984-01-01

    The characterization of nuclear waste package behavior is primarily based on the outcome of laboratory tests, where components of a proposed waste package are either individually or simultaneously subjected to simulated repository conditions. At each step of a testing method, both controllable and uncontrollable factors contribute to the overall uncertainty in the final outcome of the test. If not dealt with correctly, these sources of uncertainty could obscure or distort important information that might otherwise be gleaned form the test data. This could result in misleading or erroneous conclusions about the behavior characteristic being studied. It could also preclude estimation of the individual contributions of the major sources of uncertainty to the overall uncertainty. Statistically designed experiments and sampling plans, followed by correctly applied statistical analysis and estimation methods will yield the most information possible for the time and resources spent on experimentation, and they can eliminate the above concerns. Conclusions reached on the basis of such information will be sound and defensible. This presentation is intended to emphasize the importance of correctly applied, theoretically sound statistical methodology in characterizing nuclear waste package behavior

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

    International Nuclear Information System (INIS)

    Manaktala, H.K.; Interrante, C.G.

    1990-12-01

    This report deals with technical information that is considered essential for demonstrating the ability of the high-level radioactive waste package to provide ''substantially complete containment'' of its contents (vitrified waste form or spent light-water reactor fuel) for a period of 300 to 1000 years in a geological repository environment. The discussion is centered around technical considerations of the repository environment, materials and fabrication processes for the waste package components, various degradation modes of the materials of construction of the waste packages, and inspection and monitoring of the waste package during the preclosure and retrievability period, which could begin up to 50 years after initiation of waste emplacement. The emphasis in this report is on metallic materials. However, brief references have been made to other materials such as ceramics, graphite, bonded ceramic-metal systems, and other types of composites. The content of this report was presented to an external peer review panel of nine members at a workshop held at the Center for Nuclear Waste Regulatory Analyses (CNWRA), Southwest Research Institute, San Antonio, Texas, April 2--4, 1990. The recommendations of the peer review panel have been incorporated in this report. There are two companion reports; the second report in the series provides state-of-the-art techniques for uncertainty evaluations. 97 refs., 1 fig

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

    Energy Technology Data Exchange (ETDEWEB)

    Manaktala, H.K. (Southwest Research Inst., San Antonio, TX (USA). Center for Nuclear Waste Regulatory Analyses); Interrante, C.G. (Nuclear Regulatory Commission, Washington, DC (USA). Div. of High-Level Waste Management)

    1990-12-01

    This report deals with technical information that is considered essential for demonstrating the ability of the high-level radioactive waste package to provide substantially complete containment'' of its contents (vitrified waste form or spent light-water reactor fuel) for a period of 300 to 1000 years in a geological repository environment. The discussion is centered around technical considerations of the repository environment, materials and fabrication processes for the waste package components, various degradation modes of the materials of construction of the waste packages, and inspection and monitoring of the waste package during the preclosure and retrievability period, which could begin up to 50 years after initiation of waste emplacement. The emphasis in this report is on metallic materials. However, brief references have been made to other materials such as ceramics, graphite, bonded ceramic-metal systems, and other types of composites. The content of this report was presented to an external peer review panel of nine members at a workshop held at the Center for Nuclear Waste Regulatory Analyses (CNWRA), Southwest Research Institute, San Antonio, Texas, April 2--4, 1990. The recommendations of the peer review panel have been incorporated in this report. There are two companion reports; the second report in the series provides state-of-the-art techniques for uncertainty evaluations. 97 refs., 1 fig.

  16. Waste Management Project Contingency Analysis

    International Nuclear Information System (INIS)

    Edward L. Parsons, Jr.

    1999-01-01

    The purpose of this report is to provide the office of Waste Management (WM) with recommended contingency calculation procedures for typical WM projects. Typical projects were defined as conventional construction-type activities that use innovative elements when necessary to meet the project objectives. Projects involve treatment, storage, and disposal of low level, mixed low level, hazardous, transuranic, and high level waste. Cost contingencies are an essential part of Total Cost Management. A contingency is an amount added to a cost estimate to compensate for unexpected expenses resulting from incomplete design, unforeseen and unpredictable conditions, or uncertainties in the project scope (DOE 1994, AACE 1998). Contingency allowances are expressed as percentages of estimated cost and improve cost estimates by accounting for uncertainties. The contingency allowance is large at the beginning of a project because there are more uncertainties, but as a project develops, the allowance shrinks to adjust for costs already incurred. Ideally, the total estimated cost remains the same throughout a project. Project contingency reflects the degree of uncertainty caused by lack of project definition, and process contingency reflects the degree of uncertainty caused by use of new technology. Different cost estimation methods were reviewed and compared with respect to terminology, accuracy, and Cost Guide standards. The Association for the Advancement of Cost Engineering (AACE) methods for cost estimation were selected to represent best industry practice. AACE methodology for contingency analysis can be readily applied to WM Projects, accounts for uncertainties associated with different stages of a project, and considers both project and process contingencies and the stage of technical readiness. As recommended, AACE contingency allowances taper off linearly as a project nears completion

  17. Thermal modeling of nuclear waste package designs for disposal in tuff

    International Nuclear Information System (INIS)

    Hockman, J.N.; O'Neal, W.C.

    1983-09-01

    Lawrence Livermore National Laboratory is involved in the design and testing of high-level nuclear waste packages. Many of the aspects of waste package design and testing (e.g., corrosion and leaching) depend in part on the temperature history of the emplaced packages. This paper discusses thermal modeling and analysis of various emplaced waste package conceptual designs including the models used, the assumptions and approximations made, and the results obtained. 6 references, 6 figures, 3 tables

  18. Thermal modeling of nuclear waste package designs for disposal in tuff

    International Nuclear Information System (INIS)

    Hockman, J.N.; O'Neal, W.C.

    1984-02-01

    Lawrence Livermore National Laboratory is involved in the design and testing of high-level nuclear waste packages. Many of the aspects of waste package design and testing (e.g., corrosion and leaching) depend in part on the temperature history of the emplaced packages. This paper discusses thermal modeling and analysis of various emplaced waste package conceptual designs including the models used, the assumptions and approximations made, and the results obtained. 6 references, 6 figures, 4 tables

  19. Materials of Criticality Safety Concern in Waste Packages

    International Nuclear Information System (INIS)

    Larson, S.L.; Day, B.A.

    2006-01-01

    10 CFR 71.55 requires in part that the fissile material package remain subcritical when considering 'the most reactive credible configuration consistent with the chemical and physical form of the material'. As waste drums and packages may contain unlimited types of materials, determination of the appropriately bounding moderator and reflector materials to ensure compliance with 71.55 requires a comprehensive analysis. Such an analysis was performed to determine the materials or elements that produce the most reactive configuration with regards to both moderation and reflection of a Pu-239 system. The study was originally performed for the TRUPACT-II shipping package and thus the historical fissile mass limit for the package, 325 g Pu-239, was used [1]. Reactivity calculations were performed with the SCALE package to numerically assess the moderation or reflection merits of the materials [2]. Additional details and results are given in SAIC-1322-001 [3]. The development of payload controls utilizing process knowledge to determine the classification of special moderator and/or reflector materials and the associated fissile mass limit is also addressed. (authors)

  20. Secondary Waste Cementitious Waste Form Data Package for the Integrated Disposal Facility Performance Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Cantrell, Kirk J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Westsik, Joseph H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Serne, R Jeffrey [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Um, Wooyong [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Cozzi, Alex D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-05-16

    A review of the most up-to-date and relevant data currently available was conducted to develop a set of recommended values for use in the Integrated Disposal Facility (IDF) performance assessment (PA) to model contaminant release from a cementitious waste form for aqueous wastes treated at the Hanford Effluent Treatment Facility (ETF). This data package relies primarily upon recent data collected on Cast Stone formulations fabricated with simulants of low-activity waste (LAW) and liquid secondary wastes expected to be produced at Hanford. These data were supplemented, when necessary, with data developed for saltstone (a similar grout waste form used at the Savannah River Site). Work is currently underway to collect data on cementitious waste forms that are similar to Cast Stone and saltstone but are tailored to the characteristics of ETF-treated liquid secondary wastes. Recommended values for key parameters to conduct PA modeling of contaminant release from ETF-treated liquid waste are provided.

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

    International Nuclear Information System (INIS)

    Davis, M.S.

    1985-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Davis, M.S. (ed.)

    1985-12-01

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

  3. Uncertainty evaluation methods for waste package performance assessment

    International Nuclear Information System (INIS)

    Wu, Y.T.; Nair, P.K.; Journel, A.G.; Abramson, L.R.

    1991-01-01

    This report identifies and investigates methodologies to deal with uncertainties in assessing high-level nuclear waste package performance. Four uncertainty evaluation methods (probability-distribution approach, bounding approach, expert judgment, and sensitivity analysis) are suggested as the elements of a methodology that, without either diminishing or enhancing the input uncertainties, can evaluate performance uncertainty. Such a methodology can also help identify critical inputs as a guide to reducing uncertainty so as to provide reasonable assurance that the risk objectives are met. This report examines the current qualitative waste containment regulation and shows how, in conjunction with the identified uncertainty evaluation methodology, a framework for a quantitative probability-based rule can be developed that takes account of the uncertainties. Current US Nuclear Regulatory Commission (NRC) regulation requires that the waste packages provide ''substantially complete containment'' (SCC) during the containment period. The term ''SCC'' is ambiguous and subject to interpretation. This report, together with an accompanying report that describes the technical considerations that must be addressed to satisfy high-level waste containment requirements, provides a basis for a third report to develop recommendations for regulatory uncertainty reduction in the ''containment''requirement of 10 CFR Part 60. 25 refs., 3 figs., 2 tabs

  4. PRIDE Surveillance Projects Data Packaging Project Information Package Specification Version 1.1

    Energy Technology Data Exchange (ETDEWEB)

    Kelleher, D. M.; Shipp, R. L.; Mason, J. D.

    2010-08-31

    Information Package Specification version 1.1 describes an XML document format called an information package that can be used to store information in information management systems and other information archives. An information package consists of package information, the context required to understand and use that information, package metadata that describes the information, and XML signatures that protect the information. The information package described in this specification was designed to store Department of Energy (DOE) and National Nuclear Security Administration (NNSA) information and includes the metadata required for that information: a unique package identifier, information marking that conforms to DOE and NNSA requirements, and access control metadata. It is an implementation of the Open Archival Information System (OAIS) Reference Model archival information package tailored to meet NNSA information storage requirements and designed to be used in the computing environments at the Y-12 National Security Complex and at other NNSA sites.

  5. Radioactive waste package assay facility. Volume 3. Data processing

    International Nuclear Information System (INIS)

    Creamer, S.C.; Lalies, A.A.; Wise, M.O.

    1992-01-01

    This report, in three volumes, covers the work carried out by Taylor Woodrow Construction Ltd, and two major sub-contractors: Harwell Laboratory (AEA Technology) and Siemens Plessey Controls Ltd, on the development of a radioactive waste package assay facility, for cemented 500 litre intermediate level waste drums. Volume 3, describes the work carried out by Siemens Plessey Controls Ltd on the data-processing aspects of an integrated waste assay facility. It introduces the need for a mathematical model of the assay process and develops a deterministic model which could be tested using Harwell experimental data. Relevant nuclear reactions are identified. Full implementation of the model was not possible within the scope of the Harwell experimental work, although calculations suggested that the model behaved as predicted by theory. 34 figs., 52 refs., 5 tabs

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

    Science.gov (United States)

    Darnay, Arsen; Franklin, William E.

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

  7. PRIDE Surveillance Projects Data Packaging Project, Information Package Specification Version 1.0

    Energy Technology Data Exchange (ETDEWEB)

    Kelleher, D.M.; Shipp, R. L.; Mason, J. D.

    2009-09-28

    This document contains a specification for a standard XML document format called an information package that can be used to store information and the context required to understand and use that information in information management systems and other types of information archives. An information package consists of packaged information, a set of information metadata that describes the packaged information, and an XML signature that protects the packaged information. The information package described in this specification was designed to be used to store Department of Energy (DOE) and National Nuclear Security Administration (NNSA) information and includes the metadata required for that information: a unique package identifier, information marking that conforms to DOE and NNSA requirements, and access control metadata. Information package metadata can also include information search terms, package history, and notes. Packaged information can be text content, binary content, and the contents of files and other containers. A single information package can contain multiple types of information. All content not in a text form compatible with XML must be in a text encoding such as base64. Package information is protected by a digital XML signature that can be used to determine whether the information has changed since it was signed and to identify the source of the information. This specification has been tested but has not been used to create production information packages. The authors expect that gaps and unclear requirements in this specification will be identified as this specification is used to create information packages and as information stored in information packages is used. The authors expect to issue revised versions of this specification as needed to address these issues.

  8. Spent fuel and high level waste: Chemical durability and performance under simulated repository conditions. Results of a coordinated research project 1998-2004. Part 2: Results of a previously unpublished CRP: Performance of high level waste forms and packages under repository conditions. Results of a co-ordinated research project 1991-1998

    International Nuclear Information System (INIS)

    2007-07-01

    The objective of the CRP (Coordinated Research Projekt) on the 'Performance of High Level Waste Forms and Packages under Repository Conditions' was to contribute to the development and implementation of proper and sound technologies for HLW and spent fuel management. Special emphasis was given to the identification of various waste form properties and the study of their long term durability in simulated repository conditions. Another objective was to promote the co-operation and exchange of information between Member States on experimental concerning behaviour of the waste form. The CRP was composed of research contracts and agreements with Argentina, Australia, Belgium, Canada, China, Czech Republic, Finland, France, Germany, India, Japan, Russia, and the United States of America. The publication includes 14 individual contributions of the participants to the CRP, which are indexed separately.

  9. Shielding Calculations on Waste Packages – The Limits and Possibilities of different Calculation Methods by the example of homogeneous and inhomogeneous Waste Packages

    OpenAIRE

    Adams Mike; Smalian Silva

    2017-01-01

    For nuclear waste packages the expected dose rates and nuclide inventory are beforehand calculated. Depending on the package of the nuclear waste deterministic programs like MicroShield® provide a range of results for each type of packaging. Stochastic programs like “Monte-Carlo N-Particle Transport Code System” (MCNP®) on the other hand provide reliable results for complex geometries. However this type of program requires a fully trained operator and calculations are time consuming. The prob...

  10. West Valley Demonstration Project low-level and transuranic waste assay and methodology

    International Nuclear Information System (INIS)

    McVay, C.W.

    1987-03-01

    In the decontamination and decommissioning of the West Valley Nuclear Facility, waste materials are being removed and packaged in a variety of waste containers which require classification in accordance with USNRC 10 CFR 61 and DOE 5820.2 criteria. Low-Level and Transuranic waste assay systems have been developed to efficiently assay and classify the waste packages. The waste is assayed by segmented gamma scanning, passive neutron techniques, dose rate conversion, and/or radiochemical laboratory analysis. The systems are capable of handling all the waste forms currently packaged as part of the Project. The above systems produce a list of nuclides present with their concentrations and determines the classification of the waste packages based on criteria outlined in DOE Order 5820.2 and USNRC 10 CFR 61.55. 9 refs., 12 figs., 8 tabs

  11. Waste package performance criteria for deepsea disposal of low-level radioactive wastes

    International Nuclear Information System (INIS)

    Colombo, P.; Fuhrmann, M.

    1988-07-01

    Sea disposal of low-level radioactive waste began in the United States in 1946, and was placed under the licensing authority of the Atomic Energy Commission (AEC). The practice stopped completely in 1970. Most of the waste disposed of at sea was packaged in second- hand or reconditioned 55-gallon drums filled with cement so that the average package density was sufficiently greater than that of sea water to ensure sinking. It was assumed that all the contents would eventually be released since the packages were not designed or required to remain intact for sustained periods of time after descent to the ocean bottom. Recently, there has been renewed interest in ocean disposal, both in this country and abroad, as a waste management alternative to land burial. The Marine Protection, Research and Sanctuaries Act of 1972 (PL 92-532) gives EPA the regulatory responsibility for ocean dumping of all materials, including radioactive waste. This act prohibits the ocean disposal of high-level radioactive waste and requires EPA to control the ocean disposal of all other radioactive waste through the issuance of permits. In implementing its permit authorities, EPA issued on initial set of regulations and criteria in 1973 to control the disposal of material into the ocean waters. It was in these regulations that EPA initially introduced the general requirement of isolation and containment of radioactive waste as the basic operating philosophy. 37 refs

  12. Alternatives for packaging and transport of greater-than-class C low-level waste

    International Nuclear Information System (INIS)

    Smith, R.I.

    1990-06-01

    Viable methods for packaging greater-than-class C (GTCC) low-level wastes and for transporting those wastes from the waste generator sites or from an eastern interim storage site to the Yucca Mountain repository site have been identified and evaluated. Estimated costs for packaging and transporting the population of GTCC wastes expected to be accumulated through the year 2040 have been developed for three waste volume scenarios, for two preferred packaging methods for activated metals from reactor operations and from reactor decommissioning, and for two packaging density assumptions for the activated metals from reactor decommissioning. 7 refs. 7 tabs

  13. EPA's high-level waste standards and waste package performance

    International Nuclear Information System (INIS)

    Meyers, S.

    1985-01-01

    The seven assurance requirements EPA was considering were these: (1) Disposal systems shall not depend on active institutional controls for more than 100 years after disposal; (2) Long-term disposal system performance should be monitored for a reasonable time as a supplement to other types of protection; (3) Disposal systems shall be marked and their locations recorded in all appropriate government records; (4) Disposal systems shall be designed with several different types of barriers, both natural and engineered; (5) Sites should not be located where scarce or easily accessible resources are located; (6) Site selection should consider the relative isolation offered by potential alternatives; and (7) Wastes shall be recoverable for a reasonable time after disposal. (orig./PW)

  14. Gamma and Neutron Radiolysis in the 21-PWR Waste Package

    Energy Technology Data Exchange (ETDEWEB)

    J.S. Tang

    2001-05-03

    The objective of this calculation is to compute gamma and neutron dose rates in order to determine the maximum radiolytic production of nitric acid and other chemical species inside the 21-PWR (pressurized-water reactor) waste package (WP). The scope of this calculation is limited to the time period between 5,000 and 100,000 years after emplacement. The information provided by the sketches attached to this calculation is that of the potential design for the type of WP considered in this calculation. The results of this calculation will be used to evaluate nitric acid corrosion of fuel cladding from radiolysis in the 21-PWR WP. This calculation was performed in accordance with the Technical Work Plan for: Waste Package Design Description for LA (Civilian Radioactive Waste Management System (CRWMS) Management and Operating Contractor (M&O) 2000a). AP-3.124, Calculations, is used to perform the calculation and develop the document. This calculation is associated with the total system performance assessment (TSPA) of which the spent fuel cladding integrity is to be evaluated.

  15. Gamma and Neutron Radiolysis in the 21-PWR Waste Package

    International Nuclear Information System (INIS)

    J.S. Tang

    2001-01-01

    The objective of this calculation is to compute gamma and neutron dose rates in order to determine the maximum radiolytic production of nitric acid and other chemical species inside the 21-PWR (pressurized-water reactor) waste package (WP). The scope of this calculation is limited to the time period between 5,000 and 100,000 years after emplacement. The information provided by the sketches attached to this calculation is that of the potential design for the type of WP considered in this calculation. The results of this calculation will be used to evaluate nitric acid corrosion of fuel cladding from radiolysis in the 21-PWR WP. This calculation was performed in accordance with the Technical Work Plan for: Waste Package Design Description for LA (Civilian Radioactive Waste Management System (CRWMS) Management and Operating Contractor (M and O) 2000a). AP-3.124, Calculations, is used to perform the calculation and develop the document. This calculation is associated with the total system performance assessment (TSPA) of which the spent fuel cladding integrity is to be evaluated

  16. Radioactive waste inventories and projections

    International Nuclear Information System (INIS)

    McLaren, L.H.

    1982-11-01

    This bibliography contains information on radioactive waste inventories and projections included in the Department of Energy's Energy Data Base from January 1981 through September 1982. The arrangement is by report number for reports, followed by nonreports in reverse chronological order. These citations are to research reports, journal articles, books, patents, theses, and conference papers from worldwide sources. Five indexes, each preceded by a brief description, are provided: Corporate Author, Personal Author, Subject, Contract Number, and Report Number. (25 abstracts)

  17. EVALUATION OF WASTE PACKAGE EXTERNAL ENVIRONMENTAL CONDITION STUDY

    International Nuclear Information System (INIS)

    E. N. Lindner and E. F. Dembowski

    1998-01-01

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

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

    NARCIS (Netherlands)

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

    2011-01-01

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

  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. Parametric study of the effects of thermal environment on a waste package for a tuff repository

    Energy Technology Data Exchange (ETDEWEB)

    Johnstone, J K; Sundberg, W D; Krumhansl, J L [Sandia National Laboratories Albuquerque, NM, (USA)

    1982-12-31

    The thermal environment has been modeled in a simple reference waste package in a tuff repository for a variety of variables. The waste package was composed of the waste form, canister, overpack and backfill. The emplacement hole was 122cm dia. Waste forms used in the calculations were commercial high level waste (CHLW) and spent fuel (SF). Canister loadings varied from 50 to 100 kW/acre. Primary attention was focused on the backfill behavior in the thermal and chemical environment. Results are related to the maximum temperature calculated for the backfill. These calculations raise serious concerns about the effectiveness of the backfill within the context of the total waste package.

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

    International Nuclear Information System (INIS)

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

    1988-08-01

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

  2. Basalt waste isolation project. Quarterly report, April 1, 1981-June 30, 1981

    Energy Technology Data Exchange (ETDEWEB)

    Deju, R.A.

    1981-08-01

    This document reports progress made in the Basalt Waste Isolation Project during the third quarter of fiscal year 1981. Efforts are described for the following programs of the project work breakdown structure: systems; waste package; site; repository; regulatory and institutional; test facilities; in situ test facilities.

  3. Evaluation and compilation of DOE waste package test data

    International Nuclear Information System (INIS)

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

    1991-12-01

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

  4. Vertical Drop of 44-BWR Waste Package With Lifting Collars

    Energy Technology Data Exchange (ETDEWEB)

    A.K. Scheider

    2005-08-23

    The objective of this calculation is to determine the structural response of a waste package (WP) dropped flat on its bottom from a specified height. The WP used for that purpose is the 44-Boiling Water Reactor (BWR) WP. The scope of this document is limited to reporting the calculation results in terms of stress intensities. The Uncanistered Waste Disposal Container System is classified as Quality Level 1 (Ref. 4, page 7). Therefore, this calculation is subject to the requirements of the Quality Assurance Requirements and Description (Ref. 16). AP-3. 12Q, Design Calculations and Analyses (Ref. 11) is used to perform the calculation and develop the document. The information provided by the sketches attached to this calculation is that of the potential design of the type of 44-BWR WP considered in this calculation and provides the potential dimensions and materials for that design.

  5. Radioactive waste packages stored at the Aube facility for low-intermediate activity wastes. A selective and controlled storage

    International Nuclear Information System (INIS)

    2005-01-01

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

  6. Behaviour Test with the Leaching of a Waste package

    International Nuclear Information System (INIS)

    Fischer, G.R

    1999-01-01

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

  7. Estimation of waste package performance requirements for a nuclear waste repository in basalt

    International Nuclear Information System (INIS)

    Wood, B.J.

    1980-07-01

    A method of developing waste package performance requirements for specific nuclides is described, and based on federal regulations concerning permissible concentrations in solution at the point of discharge to the accessible environment, a simple and conservative transport model, and baseline and potential worst-case release scenarios

  8. TRANSPORT LOCOMOTIVE AND WASTE PACKAGE TRANSPORTER ITS STANDARDS IDENTIFICATION STUDY

    International Nuclear Information System (INIS)

    Draper, K.D.

    2005-01-01

    To date, the project has established important to safety (ITS) performance requirements for structures, systems and components (SSCs) based on identification and categorization of event sequences that may result in a radiological release. These performance requirements are defined within the ''Nuclear Safety Design Basis for License Application'' (NSDB) (BSC 2005). Further, SSCs credited with performing safe functions are classified as ITS. In turn, performance confirmation for these SSCs is sought through the use of consensus code and standards. The purpose of this study is to identify applicable codes and standards for the waste package (WP) transporter and transport locomotive ITS SSCs. Further, this study will form the basis for selection and the extent of applicability of each code and standard. This study is based on the design development completed for License Application only. Accordingly, identification of ITS SSCs beyond those defined within the NSDB are based on designs that may be subject to further development during detail design. Furthermore, several design alternatives may still be under consideration to satisfy certain safety functions, and that final selection will not be determined until further design development has occurred. Therefore, for completeness, throughout this study alternative designs currently under consideration will be discussed. Further, the results of this study will be subject to evaluation as part of a follow-on gap analysis study. Based on the results of this study the gap analysis will evaluate each code and standard to ensure each ITS performance requirement is fully satisfied. When a performance requirement is not fully satisfied a ''gap'' is highlighted. Thereafter, the study will identify supplemental requirements to augment the code or standard to meet performance requirements. Further, the gap analysis will identify non-standard areas of the design that will be subject to a Development Plan. Non-standard components and

  9. Methodologies for assessing long-term performance of high-level radioactive waste packages

    International Nuclear Information System (INIS)

    Stephens, K.; Boesch, L.; Crane, B.; Johnson, R.; Moler, R.; Smith, S.; Zaremba, L.

    1986-01-01

    Several methods the Nuclear Regulatory Commission (NRC) can use to independently assess Department of Energy (DOE) waste package performance were identified by The Aerospace Corporation. The report includes an overview of the necessary attributes of performance assessment, followed by discussions of DOE methods, probabilistic methods capable of predicting waste package lifetime and radionuclide releases, process modeling of waste package barriers, sufficiency of the necessary input data, and the applicability of probability density functions. It is recommended that the initial NRC performance assessment (for the basalt conceptual waste package design) should apply modular simulation, using available process models and data, to demonstrate this assessment method

  10. Thermal Response of the 21-PWR Waste Package to a Fire Accident

    International Nuclear Information System (INIS)

    F.P. Faucher; H. Marr; M.J. Anderson

    2000-01-01

    The objective of this calculation is to evaluate the thermal response of the 21-PWR WP (pressurized water reactor waste package) to the regulatory fire event. The scope of this calculation is limited to the two-dimensional waste package temperature calculations to support the waste package design. The information provided by the sketches attached to this calculation (Attachment IV) is that of the potential design of the type of waste package considered in this calculation. The procedure AP-3.12Q.Calculations (Reference 1), and the Development Plan (Reference 24) are used to develop this calculation

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

    International Nuclear Information System (INIS)

    Elliott, M.L.; Shafer, P.J.; Lamar, D.A.; Merrill, R.A.; Grunewald, W.; Roth, G.; Tobie, W.

    1996-03-01

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

  12. Waste package for Yucca Mountain repository: Strategy for regulatory compliance

    International Nuclear Information System (INIS)

    Cloninger, M.; Short, D.; Stahl, D.

    1989-02-01

    This document summarizes the strategy given in the Site Characterization Plan (1) for demonstrating compliance with the post closure performance objectives for the waste package and the Engineered Barrier System (EBS) contained in the Code of Federal Regulations. The strategy consists of the development of a conservative waste package design that will meet the regulatory requirements with sufficient margin for uncertainty using a multi-barrier approach that takes advantage of the unsaturated nature of the Yucca Mountain site. This strategy involves an iterative process designed to achieve compliance with the requirements for substantially complete containment and EBS release. The strategy will be implemented in such a manner that sufficient evidence will be provided for presentation to the Nuclear Regulatory Commission (NRC) so that it may make a finding that there is ''reasonable assurance'' that these performance requirements will indeed be met. In implementing the strategy, DOE recognizes four fundamental goals: (1) protect public health and safety; (2) minimize financial and other resource commitments; (3) comply with applicable laws and regulations; and (4) maintain an aggressive schedule. The strategy is intended to be a reasonable balance of these competing goals. 7 refs., 3 figs., 1 tab

  13. WASTE PACKAGE CORROSION STUDIES USING SMALL MOCKUP EXPERIMENTS

    International Nuclear Information System (INIS)

    B.E. Anderson; K.B. Helean; C.R. Bryan; P.V. Brady; R.C. Ewing

    2005-01-01

    The corrosion of spent nuclear fuel and subsequent mobilization of radionuclides is of great concern in a geologic repository, particularly if conditions are oxidizing. Corroding A516 steel may offset these transport processes within the proposed waste packages at the Yucca Mountain Repository (YMR) by retaining radionuclides, creating locally reducing conditions, and reducing porosity. Ferrous iron, Fe 2+ , has been shown to reduce UO 2 2+ to UO 2(s) [1], and some ferrous iron-bearing ion-exchange materials adsorb radionuclides and heavy metals [2]. Of particular interest is magnetite, a potential corrosion product that has been shown to remove TcO 4 - from solution [3]. Furthermore, if Fe 2+ minerals, rather than fully oxidized minerals such as goethite, are produced during corrosion, then locally reducing conditions may be present. High electron availability leads to the reduction and subsequent immobilization of problematic dissolved species such as TcO 4 - , NpO 2 + , and UO 2 2+ and can also inhibit corrosion of spent nuclear fuel. Finally, because the molar volume of iron material increases during corrosion due to oxygen and water incorporation, pore space may be significantly reduced over long time periods. The more water is occluded, the bulkier the corrosion products, and the less porosity is available for water and radionuclide transport. The focus of this paper is on the nature of Yucca Mountain waste package steel corrosion products and their effects on local redox state, radionuclide transport, and porosity

  14. FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation

    Energy Technology Data Exchange (ETDEWEB)

    K. Mon

    2004-10-11

    The purpose of this report is to evaluate and document the inclusion or exclusion of features, events and processes (FEPs) with respect to drip shield and waste package modeling used to support the Total System Performance Assessment for License Application (TSPA-LA). Thirty-three FEPs associated with the waste package and drip shield performance have been identified (DTN: MO0407SEPFEPLA.000 [DIRS 170760]). A screening decision, either ''included'' or ''excluded,'' has been assigned to each FEP, with the technical bases for screening decisions, as required by the Nuclear Regulatory Commission (NRC) in 10 CFR 63.114 (d, e, and f) [DIRS 156605]. The FEPs analyses in this report address issues related to the degradation and potential failure of the drip shield and waste package over the post closure regulatory period of 10,000 years after permanent closure. For included FEPs, this report summarizes the disposition of the FEP in TSPA-LA. For excluded FEPs, this report provides the technical bases for the screening arguments for exclusion from TSPA-LA. The analyses are for the TSPA-LA base-case design (BSC 2004 [DIRS 168489]), where a drip shield is placed over the waste package without backfill over the drip shield (BSC 2004 [DIRS 168489]). Each FEP includes one or more specific issues, collectively described by a FEP name and description. The FEP description encompasses a single feature, event, or process, or a few closely related or coupled processes, provided the entire FEP can be addressed by a single specific screening argument or TSPA-LA disposition. The FEPs were assigned to associated Project reports, so the screening decisions reside with the relevant subject-matter experts.

  15. FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation

    International Nuclear Information System (INIS)

    K. Mon

    2004-01-01

    The purpose of this report is to evaluate and document the inclusion or exclusion of features, events and processes (FEPs) with respect to drip shield and waste package modeling used to support the Total System Performance Assessment for License Application (TSPA-LA). Thirty-three FEPs associated with the waste package and drip shield performance have been identified (DTN: MO0407SEPFEPLA.000 [DIRS 170760]). A screening decision, either ''included'' or ''excluded,'' has been assigned to each FEP, with the technical bases for screening decisions, as required by the Nuclear Regulatory Commission (NRC) in 10 CFR 63.114 (d, e, and f) [DIRS 156605]. The FEPs analyses in this report address issues related to the degradation and potential failure of the drip shield and waste package over the post closure regulatory period of 10,000 years after permanent closure. For included FEPs, this report summarizes the disposition of the FEP in TSPA-LA. For excluded FEPs, this report provides the technical bases for the screening arguments for exclusion from TSPA-LA. The analyses are for the TSPA-LA base-case design (BSC 2004 [DIRS 168489]), where a drip shield is placed over the waste package without backfill over the drip shield (BSC 2004 [DIRS 168489]). Each FEP includes one or more specific issues, collectively described by a FEP name and description. The FEP description encompasses a single feature, event, or process, or a few closely related or coupled processes, provided the entire FEP can be addressed by a single specific screening argument or TSPA-LA disposition. The FEPs were assigned to associated Project reports, so the screening decisions reside with the relevant subject-matter experts

  16. The use of performance assessments in Yucca Mountain repository waste package design activities

    International Nuclear Information System (INIS)

    Jardine, L.J.

    1990-01-01

    The Yucca Mountain Project is developing performance assessment approaches as part of the evaluations of the suitability of Yucca Mountain as a repository site. Lawrence Livermore National Laboratory is developing design concepts and the scientific performance assessment methodologies and techniques used for the waste package and engineered barrier system components. This paper presents an overview of the approach under development for postclosure performance assessments that will guide the conceptual design activities and assist in the site suitability evaluations. This approach includes establishing and modeling for the long time periods required by regulations: near-field environment characteristics surrounding the emplaced wastes; container materials performance responses; and waste form properties. All technical work is being done under a fully qualified quality assurance program

  17. Evolution of waste-package design at the potential U.S. geologic repository

    International Nuclear Information System (INIS)

    Benton, H.; Harkins, B.

    2000-01-01

    This paper describes the evolution of the waste-package design at the potential geologic repository for spent nuclear fuel and high-level waste at Yucca Mountain in Nevada. Because the potential repository is the first of its kind, the design of its components must be flexible and capable of evolving in response to continuing scientific study, development efforts, and changes to performance criteria. The team of scientists and engineers at the Yucca Mountain Project has utilized a systematic, scientific approach to design the potential geologic nuclear-waste repository. As a result of continuing development efforts, the design has incorporated a growing base of scientific and engineering information to ensure that regulatory and performance requirements are met. (authors)

  18. LLNL/YMP Waste Container Fabrication and Closure Project

    International Nuclear Information System (INIS)

    1990-10-01

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

  19. The Hanford Site solid waste treatment project; Waste Receiving and Processing (WRAP) Facility

    International Nuclear Information System (INIS)

    Roberts, R.J.

    1991-01-01

    The Waste Receiving and Processing (WRAP) Facility will provide treatment and temporary storage (consisting of in-process storage) for radioactive and radioactive/hazardous mixed waste. This facility must be constructed and operated in compliance with all appropriate US Department of Energy (DOE) orders and Resource Conservation and Recovery Act (RCRA) regulations. The WRAP Facility will examine and certify, segregate/sort, and treat for disposal suspect transuranic (TRU) wastes in drums and boxes placed in 20-yr retrievable storage since 1970; low-level radioactive mixed waste (RMW) generated and placed into storage at the Hanford Site since 1987; designated remote-handled wastes; and newly generated TRU and RMW wastes from high-level waste (HLW) recovery and processing operations. In order to accelerated the WRAP Project, a partitioning of the facility functions was done in two phases as a means to expedite those parts of the WRAP duties that were well understood and used established technology, while allowing more time to better define the processing functions needed for the remainder of WRAP. The WRAP Module 1 phase one, is to provide the necessary nondestructive examination and nondestructive assay services, as well as all transuranic package transporter (TRUPACT-2) shipping for both WRAP Project phases, with heating, ventilation, and air conditioning; change rooms; and administrative services. Phase two of the project, WRAP Module 2, will provide all necessary waste treatment facilities for disposal of solid wastes. 1 tab

  20. The role of waste package specifications as a forerunner to ILW repository conditions for acceptance

    International Nuclear Information System (INIS)

    Barlow, S.V.; Palmer, J.D.

    1998-01-01

    In the absence of a finalized repository site, design or associated safety case, Nirex is not in a position to issue conditions for acceptance. Nirex has therefore developed a strategy which facilitates packaging of intermediate level waste by providing guidance through waste package specifications, supported by the formal assessment of specific packaging proposals on a case-by-case basis. The waste package specifications are comprehensive and cover all aspects of the waste package including dimensions and other key features, performance standards, wasteform, quality assurance, and data recording requirements. The waste package specifications will be subject to periodic review as repository design and safety cases are finalized and will progressively become site- and design-specific. The waste package specifications will eventually form the basis for conditions for acceptance. The strategy described in this paper has been successfully followed by Nirex and customers for the past ten years and has permitted wastes to be packaged for a deep repository with confidence in the absence of a finalized site and safety cases for the repository. Because the process has its basis in a generic repository concept, it remains robust, despite the increased uncertainty following the March 1997 Secretary of State's decision, as to the siting and time-scale of a deep waste repository, and continues to be an important component of the UK's waste management strategy. (author)

  1. Basalt Waste Isolation Project. Annual report, fiscal year 1980

    International Nuclear Information System (INIS)

    1980-11-01

    During this fiscal year the information available in the fields of geology and hydrology of the Columbia Plateau was consolidated and two reports were issued summarizing this information. In addition, the information on engineered barriers was consolidated and a report summarizing the research to date on waste package development and design of borehole seals was prepared. The waste package studies, when combined with the hydrologic integration, revealed that even under extreme disruptive conditions, a repository in basalt with appropriately designed waste packages can serve as an excellent barrier for containment of radionuclides for the long periods of time required for waste isolation. On July 1, 1980, the first two heater tests at the Near-Surface Test Facility were started and have been successfully operated to this date. The papers on the Near-Surface Test Facility section of this report present the results of the equipment installed and the preliminary results of the testing. In October 1979, the US Department of Energy selected the joint venture of Kaiser Engineers/Parsons Brinckerhoff Quade and Douglas, Inc., to be the architect-engineer to produce a conceptual design of a repository in basalt. During the year, this design has progressed and concept selection has now been completed. This annual report presents a summary of the highlights of the work completed during fiscal year 1980. It is intended to supplement and summarize the nearly 200 papers and reports that have been distributed to date as a part of the Basalt Waste Isolation Project studies

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-11-15

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    M. J. Appel and J. M. Capron

    2007-07-25

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

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

    International Nuclear Information System (INIS)

    Crow, S.R.

    1996-01-01

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

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

    International Nuclear Information System (INIS)

    Appel, M.J.; Capron, J.M.

    2007-01-01

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

  7. Operational waste volume projection. Revision 20

    International Nuclear Information System (INIS)

    Koreski, G.M.; Strode, J.N.

    1994-01-01

    Waste receipts to the double-shell tank system are analyzed and wastes through the year 2015 are projected based on generation trends of the past 12 months. A computer simulation of site operations is performed, which results in projections of tank fill schedules, tank transfers, evaporator operations, tank retrieval, and aging waste tank usage. This projection incorporates current budget planning and the clean-up schedule of the Tri-Party Agreement. Assumptions were current as of July 1994

  8. Hanford Waste Vitrification Plant Project Waste Form Qualification Program Plan

    International Nuclear Information System (INIS)

    Randklev, E.H.

    1993-06-01

    The US Department of Energy has created a waste acceptance process to help guide the overall program for the disposal of high-level nuclear waste in a federal repository. This Waste Form Qualification Program Plan describes the hierarchy of strategies used by the Hanford Waste Vitrification Plant Project to satisfy the waste form qualification obligations of that waste acceptance process. A description of the functional relationship of the participants contributing to completing this objective is provided. The major activities, products, providers, and associated scheduling for implementing the strategies also are presented

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-02-01

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

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

    International Nuclear Information System (INIS)

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

    1982-02-01

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

  11. Depleted uranium oxides as spent-nuclear-fuel waste-package fill materials

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    1997-01-01

    Depleted uranium dioxide fill inside the waste package creates the potential for significant improvements in package performance based on uranium geochemistry, reduces the potential for criticality in a repository, and consumes DU inventory. As a new concept, significant uncertainties exist: fill properties, impacts on package design, post- closure performance

  12. Salt Repository Project waste emplacement mode decision paper: Revison 1

    International Nuclear Information System (INIS)

    1987-08-01

    This paper provides a recommendation as to the mode of waste emplacement to be used as the current basis for site characterization activity for the Deaf Smith County, Texas, high level nuclear waste repository site. It also presents a plan for implementing the recommendation so as to provide a high level of confidence in the project's success. Since evaluations of high-level waste disposal in geologic repositories began in the 1950s, most studies emplacement in salt formations employed the vertical orientation for emplacing waste packages in boreholes in the floor of the underground facility. This orientation was used in trials at Project Salt Vault in the 1960s. The Waste Isolation Pilot Plant (WIPP) has recently settled on a combination of vertical and horizontal modes for various waste types. This paper analyzes the information available and develops a project position upon which to base current site characterization activities. The position recommended is that the SRP should continue to use the vertical waste emplacement mode as the reference design and to carry the horizontal mode as a ''passive'' alternative. This position was developed based upon the conclusions of a decision analysis, risk assessment, and cost/schedule impact assessment. 52 refs., 6 figs., 1 tab

  13. FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation

    International Nuclear Information System (INIS)

    K.G. Mon; L.A. Rottinghaus

    2004-01-01

    closely related or coupled processes provided that the entire FEP can be addressed by a single specific screening argument or TSPA disposition. Descriptor phrases provide additional detail about the subject and content of the FEP beyond the FEP name and description. The FEPs have been assigned to associated scientific analyses. The assignments were based on the nature of the FEPs so that the analysis and resolution for screening decisions reside with the subject-matter experts in the relevant disciplines. This scientific analysis addresses the screening decisions associated with the FEPs for the Performance Assessment Project Waste Package Degradation Modeling group

  14. FEPs Screening of Processes and Issues in Drip Shield and Waste Package Degradation

    Energy Technology Data Exchange (ETDEWEB)

    K.G. Mon; L.A. Rottinghaus

    2004-03-26

    or event, or a few closely related or coupled processes provided that the entire FEP can be addressed by a single specific screening argument or TSPA disposition. Descriptor phrases provide additional detail about the subject and content of the FEP beyond the FEP name and description. The FEPs have been assigned to associated scientific analyses. The assignments were based on the nature of the FEPs so that the analysis and resolution for screening decisions reside with the subject-matter experts in the relevant disciplines. This scientific analysis addresses the screening decisions associated with the FEPs for the Performance Assessment Project Waste Package Degradation Modeling group.

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

    International Nuclear Information System (INIS)

    Colleen Shelton-Davis

    2005-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Colleen Shelton-Davis

    2005-11-01

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

  17. The ATB-8K packaging for transport of radioactive waste in Sweden

    International Nuclear Information System (INIS)

    Michels, L.; Dybeck, P.

    1998-01-01

    The ATB-8K container has been developed on behalf of SKB, the Swedish nuclear fuel and waste management organization, to transport large volumes of radioactive waste conditioned in moulds and drums, or large size scrap components, from nuclear facilities to the Swedish Final Repository for radioactive waste (SFR). In most cases the waste is under LSA form, but when the dose rate at 3 meters from the unshielded object exceeds 10 mSv/h, the transport packaging must been the regulatory requirements applicable to type B(U) packages, with no fissile content. Considering the dose rate around the package, it will be transported under exclusive use. The ATB-8k packaging is therefore a type B(U) packaging, specially designed for the transportation of high activity conditioned waste. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    McGrail, B. Peter; Icenhower, Jonathan P.; Martin, Paul F.; Schaef, Herbert T.; O' Hara, Matthew J.; Rodriguez, Eugenio; Steele, Jackie L.

    2001-02-01

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

  19. Safety analysis report for packaging (onsite) for the Waste Encapsulation and Storage Facility ion exchange module

    International Nuclear Information System (INIS)

    Romano, T.

    1997-01-01

    The Waste Encapsulation and Storage Facility (WESF) is in need of providing an emergency ion exchange system to remove cesium or strontium from the pool cell in the event of a capsule failure. The emergency system is call the WESF Emergency Ion Exchange System and the packaging is called the WESF ion exchange module (WIXM). The packaging system will meet the onsite transportation requirements for a Type B, highway route controlled quantity package as well as disposal requirements for Category 3 waste

  20. A model for quantifying construction waste in projects according to the European waste list.

    Science.gov (United States)

    Llatas, C

    2011-06-01

    The new EU challenge is to recover 70% by weight of C&D waste in 2020. Literature reveals that one major barrier is the lack of data. Therefore, this paper presents a model which allows technicians to estimate C&D waste during the design stage in order to promote prevention and recovery. The types and quantities of CW are estimated and managed according to EU guidelines, by building elements and specifically for each project. The model would allow detection of the source of the waste and to adopt other alternative procedures which delete hazardous waste and reduce CW. Likewise, it develops a systematic structure of the construction process, a waste classification system and some analytical expressions which are based on factors. These factors depend on technology and represent a standard on site. It would allow to develop a database of waste anywhere. A Spanish case study is covered. Factors were obtained by studying over 20 dwellings. The source and types of packaging waste, remains, soil and hazardous waste were estimated in detail and were compared with other studies. Results reveal that the model can be implemented in projects and the chances of reducing and recovery C&D waste could be increased, well above the EU challenge. Copyright © 2011 Elsevier Ltd. All rights reserved.

  1. Modelling approach to evaluate safety of LILW-SL disposal in slovenia considering different waste packaging options

    International Nuclear Information System (INIS)

    Perko, J.; Mallants, D.

    2007-01-01

    The long-term safety of radioactive waste repositories is usually demonstrated by means of a safety assessment which normally includes modelling of radionuclide release from a multi-barrier surface or deep repository to the geosphere and biosphere. The present quantitative evaluation performed emphasizes on contrasting disposal options under consideration in Slovenia and concerns siting, disposal concept (deep versus surface), and waste packaging. The assessment has identified a number of conditions that would lead to acceptable waste disposal solutions, while at the same time results also revealed options that would result in exceeding the radiological criteria. Results presented are the output of a collective effort of a Quintessa-led Consortium with SCK-CEN and Belgatom, in the framework of a recent PHARE project. The key objective of this work was to identify the preferred disposal concept and packaging option from a number of alternatives being considered by the Slovenian radioactive waste management agency (ARAO) for low and intermediate level short-lived waste (LILW-SL). The emphasis of the assessment was the consideration of several waste treatment and packaging options in an attempt to identify the minimum required containment characteristics which would result in safe disposal and the cost-benefit of additional safety measures. Waste streams for which alternative treatment and packaging solutions were developed and evaluated include decommissioning waste and NPP operational wastes containing drums with unconditioned ion exchange resins in overpacked tube type containers (TTCs). For the former the disposal options under consideration were either direct disposal of loose pieces grouted into a vault or use of high integrity containers. For the latter three options were foreseen. The first is overpacking of resin containing TTCs grouted into high integrity containers, the second option is complete treatment with hydration, neutralisation, and cementation of

  2. Radionuclides difficult to measure in waste packages. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Thierfeldt, S; Deckert, A [Brenk Systemplanung, Aachen (Germany)

    1995-11-01

    In this study nuclide specific correlation analyses between key nuclides that can be easily measured and nuclides that are difficult to measure are presented. Data are taken from studies and data compilations from various countries. The results of this study can serve to perform assays of the nuclide specific radionuclide contents in waste packages by gamma measurements of {sup 60}Co and {sup 137}Cs and calculation of the contents of other nuclides via the correlation analyses, sometimes referred to as `scaling factor method`. It can thus be avoided to have to take samples from the waste for separate analysis. An attempt is made to also investigate the physical and chemical backgrounds behind the proposed correlations. For example, a formation pathway common to the two nuclides to be correlated can be regarded as an explanation, if a good correlation is found. On the other hand, if the observed correlation is of poor quality, reasons may possibly lie in different behaviour of the two nuclides in the water system of the nuclear plant. This implies not only chemical solubility, transfer constants etc. in the water system, which would not only affect the proportionality between the two nuclides, but a different behavior in different parts of the water system must be assumed (e.g. different filter efficiencies etc). 47 refs, 57 figs, 40 tabs.

  3. Development plan. High activity-long living wastes project. Abstract

    International Nuclear Information System (INIS)

    2007-01-01

    This brochure presents the actions that the ANDRA (the French national agency of radioactive wastes) has to implement in the framework of the project of high activity-long living (HALL) radioactive wastes (HAVL project) conformably to the requirements of the program defined in the law from June 28, 2006 (law no 2006-739). This law precises the three, complementary, research paths to explore for the management of this type of wastes: separation and transmutation of long-living radioactive elements, reversible disposal in deep geologic underground, and long duration storage. The ANDRA's action concerns the geologic disposal aspect. The following points are presented: the HALL wastes and their containers, the reversible disposal procedure, the HAVL project: financing of researches, storage concepts, development plan of the project (dynamics, information and dialogue approach, input data, main steps, schedule); the nine programs of the HAVL project (laboratory experiments and demonstration tests, surface survey, scientific program, simulation program, surface engineering studies and technological tests, information and communication program, program of environment and facilities surface observation and monitoring, waste packages management, monitoring and transport program, disposal program); the five transverse technical and scientific activities (safety, reversibility, cost, health and occupational safety, impact study). (J.S.)

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

    Energy Technology Data Exchange (ETDEWEB)

    MANN, F.M.

    2000-03-02

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

  5. Project Management Plan 105-KE Basin sludge retrieval and packaging

    International Nuclear Information System (INIS)

    McWethy, L.M.

    1994-01-01

    The KE Basin contains over 1,100 metric tons of spent nuclear fuel (SNF). The bulk of this inventory consists of over 50,000 zircaloy clad, uranium metal N-Reactor fuel element assemblies, along with less than half a metric ton of single-pass reactor fuel elements, stored in over 3,600 open top canister assemblies. In addition, sludge containing fissile and fission product material from damaged/degraded fuel has accumulated in the basin. The sludge, particularly the fines, impacts basin operations by clouding the water and making activities requiring a clear view impossible to complete until after sludge settles. Packaging would get the sludge out of the operator's way and allow it to be moved within the basin in a more manageable state. The primary project objective is to develop, procure, and quality the equipment needed to remove all sludge from the KE Basin with minimal dose commitment, minimal cost, and on schedule. The project will provide: (1) the development, testing, and installation of equipment for sludge retrieval and packaging; (2) understanding of and experience with actual sludge through near-term sludge packaging feature tests in the KE Basin; (3) sludge removal and handling equipment required to support debris removal, fuel handling, and other activities involving sludge within the KE Basin; and (4) enlist industry expertise in all phases of the project. This Project Management Plant establishes the organizational responsibilities, control systems, and procedures for the execution of project activities for KE Basin sludge retrieval packaging, to meet programmatic requirements within authorized funding and approved schedules

  6. Determination of Radioisotope Content by Measurement of Waste Package Dose Rates - 13394

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Daiane Cristini B.; Gimenes Tessaro, Ana Paula; Vicente, Roberto [Nuclear and Energy Research Institute Brazil, Radioactive Waste Management Department IPEN/GRR, Sao Paulo. SP. (Brazil)

    2013-07-01

    The objective of this communication is to report the observed correlation between the calculated air kerma rates produced by radioactive waste drums containing untreated ion-exchange resin and activated charcoal slurries with the measured radiation field of each package. Air kerma rates at different distances from the drum surface were calculated with the activity concentrations previously determined by gamma spectrometry of waste samples and the estimated mass, volume and geometry of solid and liquid phases of each waste package. The water content of each waste drum varies widely between different packages. Results will allow determining the total activity of wastes and are intended to complete the previous steps taken to characterize the radioisotope content of wastes packages. (authors)

  7. Management of waste from packaging of construction materials in building construction works

    OpenAIRE

    González Pericot, Natalia; Río Merino, Mercedes del

    2011-01-01

    Every material arriving at the construction site comes protected in some type of packaging, fundamentally cardboard, plastic or wood, and presently the great majority of these packagings finish in a container mixed with the rest of waste of the construction work. The increasing tendency to use prefabricated materials increases the volume of packaging necessary in product transport; in addition, the traditional materials also arrive more protected with packaging. A specific management for ...

  8. Aging and Phase Stability of Waste Package Outer Barrier

    Energy Technology Data Exchange (ETDEWEB)

    F. Wong

    2004-09-28

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

  9. Aging and Phase Stability of Waste Package Outer Barrier

    International Nuclear Information System (INIS)

    F. Wong

    2004-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NA

    2005-10-25

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

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

    International Nuclear Information System (INIS)

    2005-01-01

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

  12. Calculation of projected waste loads for transuranic waste management alternatives

    International Nuclear Information System (INIS)

    Hong, K.; Kotek, T.; Koebnick, B.; Wang, Y.; Kaicher, C.

    1995-01-01

    The level of treatment and the treatment and interim storage site configurations (decentralized, regional, or centralized) impact transuranic (TRU) waste loads at and en route to sites in the US Department of Energy (DOE) complex. Other elements that impact waste loads are the volume and characteristics of the waste and the unit operation parameters of the technologies used to treat it. Projected annual complexwide TRU waste loads under various TRU waste management alternatives were calculated using the WASTEunderscoreMGMT computational model. WASTEunderscoreMGMT accepts as input three types of data: (1) the waste stream inventory volume, mass, and contaminant characteristics by generating site and waste stream category; (2) unit operation parameters of treatment technologies; and (3) waste management alternative definitions. Results indicate that the designed capacity of the Waste Isolation Pilot Plant, identified under all waste management alternatives as the permanent disposal facility for DOE-generated TRU waste, is sufficient for the projected complexwide TRU waste load under any of the alternatives

  13. Development of characterization methods applied to radioactive wastes and waste packages

    International Nuclear Information System (INIS)

    Guy, C.; Bienvenu, Ph.; Comte, J.; Excoffier, E.; Dodi, A.; Gal, O.; Gmar, M.; Jeanneau, F.; Poumarede, B.; Tola, F.; Moulin, V.; Jallu, F.; Lyoussi, A.; Ma, J.L.; Oriol, L.; Passard, Ch.; Perot, B.; Pettier, J.L.; Raoux, A.C.; Thierry, R.

    2004-01-01

    This document is a compilation of R and D studies carried out in the framework of the axis 3 of the December 1991 law about the conditioning and storage of high-level and long lived radioactive wastes and waste packages, and relative to the methods of characterization of these wastes. This R and D work has permitted to implement and qualify new methods (characterization of long-lived radioelements, high energy imaging..) and also to improve the existing methods by lowering detection limits and reducing uncertainties of measured data. This document is the result of the scientific production of several CEA laboratories that use complementary techniques: destructive methods and radiochemical analyses, photo-fission and active photonic interrogation, high energy imaging systems, neutron interrogation, gamma spectroscopy and active and passive imaging techniques. (J.S.)

  14. Nuclear Waste Package Mockups: A Study of In-situ Redox State

    Science.gov (United States)

    Helean, K.; Anderson, B.; Brady, P. V.

    2006-05-01

    The Yucca Mountain Repository (YMR), located in southern Nevada, is to be the first facility in the U.S. for the permanent disposal of high-level radioactive waste and spent nuclear fuels. Total system performance assessment(TSPA) has indicated that among the major radionuclides contributing to dose are Np, Tc, and I. These three radionuclides are mobile in most geochemical settings, and therefore sequestering them within the repository horizon is a priority for the Yucca Mountain Project (YMP). Corroding steel may offset radionuclide transport processes within the proposed waste packages at YMR by retaining radionuclides, creating locally reducing conditions, and reducing porosity. Ferrous iron has been shown to reduce UO22+ to UO2s, and some ferrous iron-bearing ion-exchange materials have been shown to adsorb radionuclides and heavy metals. Locally reducing conditions may lead to the reduction and subsequent immobilization of problematic dissolved species such as TcO4-, NpO2+, and UO22+ and can also inhibit corrosion of spent nuclear fuel. Water occluded during corrosion produces bulky corrosion products, and consequently less porosity is available for water and radionuclide transport. The focus of this study is on the nature of Yucca Mountain waste package corrosion products and their effects on local redox conditions, radionuclide transport, and porosity. In order to measure in-situ redox, six small-scale (1:40) waste package mockups were constructed using A516 and 316 stainless steel, the same materials as the proposed Yucca Mountain waste packages. The mockups are periodically injected with a simulated groundwater and the accumulated effluent and corrosion products are evaluated for their Fe(II)/Fe(III) content and mineralogy. Oxygen fugacities are then calculated and, thus, in-situ redox conditions are determined. Early results indicate that corrosion products are largely amorphous Fe-oxyhydroxides, goethite and magnetite. That information together with the

  15. Integrated waste hydrogen utilization project

    International Nuclear Information System (INIS)

    Armstrong, C.

    2004-01-01

    'Full text:' The BC Hydrogen Highway's, Integrated Waste Hydrogen Utilization Project (IWHUP) is a multi-faceted, synergistic collaboration that will capture waste hydrogen and promote its use through the demonstration of 'Hydrogen Economy' enabling technologies developed by Canadian companies. IWHUP involves capturing and purifying a small portion of the 600 kg/hr of by-product hydrogen vented to the atmosphere at the ERCO's electrochemical sodium chlorate plant in North Vancouver, BC. The captured hydrogen will then be compressed so it is suitable for transportation on roadways and can be used as a fuel in transportation and stationary fuel cell demonstrations. In summary, IWHUP invests in the following; Facilities to produce up to 20kg/hr of 99.999% pure 6250psig hydrogen using QuestAir's leading edge Pressure Swing Absorption technology; Ultra high-pressure transportable hydrogen storage systems developed by Dynetek Industries, Powertech Labs and Sacre-Davey Engineering; A Mobile Hydrogen Fuelling Station to create Instant Hydrogen Infrastructure for light-duty vehicles; Natural gas and hydrogen (H-CNG) blending and compression facilities by Clean Energy for fueling heavy-duty vehicles; Ten hydrogen, internal combustion engine (H-ICE), powered light duty pick-up vehicles and a specialized vehicle training, maintenance, and emissions monitoring program with BC Hydro, GVRD and the District of North Vancouver; The demonstration of Westport's H-CNG technology for heavy-duty vehicles in conjunction with local transit properties and a specialized vehicle training, maintenance, and emissions monitoring program; The demonstration of stationary fuel cell systems that will provide clean power for reducing peak-load power demands (peak shaving), grid independence and water heating; A comprehensive communications and outreach program designed to educate stakeholders, the public, regulatory bodies and emergency response teams in the local community, Supported by industry

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

    International Nuclear Information System (INIS)

    Calmus, D.B.

    1994-01-01

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

  17. 10 CFR 60.135 - Criteria for the waste package and its components.

    Science.gov (United States)

    2010-01-01

    ... Section 60.135 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTES... for HLW shall be designed so that the in situ chemical, physical, and nuclear properties of the waste package and its interactions with the emplacement environment do not compromise the function of the waste...

  18. Spent fuel and waste inventories and projections

    International Nuclear Information System (INIS)

    Carter, W.L.; Finney, B.C.; Alexander, C.W.; Blomeke, J.O.; McNair, J.M.

    1980-08-01

    Current inventories of commercial spent fuels and both commercial and US Department of Energy radioactive wastes were compiled, based on judgments of the most reliable information available from Government sources and the open literature. Future waste generation rates and quantities to be accumulated over the remainder of this century are also presented, based on a present projection of US commercial nuclear power growth and expected defense-related activities. Spent fuel projections are based on the current DOE/EIA estimate of nuclear growth, which projects 180 GW(e) in the year 2000. It is recognized that the calculated spent fuel discharges are probably high in view of recent reactor cancellations; hence adjustments will be made in future updates of this report. Wastes considered, on a chapter-by-chapter basis, are: spent fuel, high-level wastes, transuranic wastes, low-level wastes, mill tailings (active sites), and remedial action wastes. The latter category includes mill tailings (inactive sites), surplus facilities, formerly utilized sites, and the Grand Junction Project. For each category, waste volume inventories and projections are given through the year 2000. The land usage requirements are given for storage/disposal of low-level and transuranic wastes, and for present inventories of mill tailings

  19. Criticality Potential of Waste Packages Containing DOE SNF Affected by Igneous Intrusion

    International Nuclear Information System (INIS)

    D.S. Kimball; C.E. Sanders

    2006-01-01

    The Department of Energy (DOE) is currently preparing an application to submit to the U.S. Nuclear Regulatory Commission for a construction authorization for a monitored geologic repository. The repository will contain spent nuclear fuel (SNF) and defense high-level waste (DHLW) in waste packages placed in underground tunnels, or drifts. The primary objective of this paper is to perform a criticality analysis for waste packages containing DOE SNF affected by a disruptive igneous intrusion event in the emplacement drifts. The waste packages feature one DOE SNF canister placed in the center and surrounded by five High-Level Waste (HLW) glass canisters. The effective neutron multiplication factor (k eff ) is determined for potential configurations of the waste package during and after an intrusive igneous event. Due to the complexity of the potential scenarios following an igneous intrusion, finding conservative and bounding configurations with respect to criticality requires some additional considerations. In particular, the geometry of a slumped and damaged waste package must be examined, drift conditions must be modeled over a range of parameters, and the chemical degradation of DOE SNF and waste package materials must be considered for the expected high temperatures. The secondary intent of this calculation is to present a method for selecting conservative and bounding configurations for a wide range of end conditions

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  1. Draft Technical Position Subtask 1.1: waste package performance after repository closure. Volume 1

    International Nuclear Information System (INIS)

    Davis, M.S.; Schweitzer, D.G.

    1983-08-01

    This document provides guidance to the DOE on the issues and information necessary for the NRC to evaluate waste package performance after repository closure. Minimal performance objectives of the waste package are required by proposed 10 CFR 60. This Draft Technical Position describes the various options available to the DOE for compliance and discusses advantages and disadvantages of various choices. Examples are discussed dealing with demonstrability, predictability and reasonable assurance. The types of performance are considered. The document summarizes presently identified high priority issues needed to evaluate waste package performance after repository closure. 20 references, 7 tables

  2. What are the communication challenges for politicians, experts and stakeholders in order to enhance transparency in nuclear waste management decisions? Report from a Team Syntegrity Meeting. The European Project RISCOM-II. Work Package 3

    International Nuclear Information System (INIS)

    Andersson, Kjell; Espejo, Raul; Wene, Clas-Otto

    2003-09-01

    The Team Syntegrity Meeting is a special part of the project. It aims for increased awareness among key stakeholder groups in Europe about how nuclear waste decision processes should be developed in order to increase transparency and trust. Team Syntegrity is conducted with a special meeting format. The self-organisation of the meeting is a strong positive feature of the format. Instead of having a project leader setting the agenda, the participants formulate their own topics of relevance starting from an opening question. This report documents the meeting that was held in Lanaken, Belgium on 14-17 May 2002. The opening question for the meeting was: What are communication challenges for politicians, experts and stakeholders in order to enhance transparency in nuclear waste management decisions? There are different opinions about how communication on nuclear waste issues should be done. There are differences between stakeholder groups, and there are different approaches taken in various countries. Still it should be possible to reach a deeper understanding of social communications, that is, understanding the requirements to have effective communications between policy makers, experts and stakeholders. The aim was thus not to develop common views on the nuclear waste problem as such, but rather common grounds for developing procedures for effective communication. Hopefully, this meeting made some progress in this direction. The call for the Team Syntegrity (TS) Meeting resulted in 105 Statements of Importance given in Appendix 2. Following the TS format the meeting then formed its own agenda by first producing 30 Aggregated Statements of Importance (Appendix 3), which were grouped into 12 Consolidated Statements of Importance or topics. The group discussions were thus held under the twelve topics of: Consultation, communication and participation; Mutual learning; Roles and arenas; Heritage; Transparency; Wider context; Process; Risk; Institutional cultures

  3. Environmental monitoring and deep ocean disposal of packaged radioactive waste

    International Nuclear Information System (INIS)

    Mitchell, N.T.; Preston, A.

    1980-01-01

    The aims and objectives of environmental monitoring as laid down, for example by the ICRP and the IAEA include the assessment of actual or potential radiation exposure of man and the requirements of scientific investigations. The fulfillment of these aims is discussed in the context of the disposal of packaged radioactive waste in the deep Atlantic Ocean within the terms of the London Dumping Convention and within a regional agreement, the consultation/surveillance mechanism of the Nuclear Energy Agency. The paper discusses UK attitudes to such environmental monitoring, concentrates on the first of these ICRP objectives and shows how this is unlikely to be achieved by direct measurement in view of the small quantities of radioactive material involved relative to the scale of the receiving environment, and the timescale on which return to man can be conceived. Whilst meaningful environmental measurement is very unlikely to facilitate direct estimation of public radiation exposure by monitoring, it is still held that the basic objective of environmental monitoring can be met. A means by which this may be achieved is by oceanographic models. These procedures are discussed, illustrating the application of this philosophy in practice. (H.K.)

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

    International Nuclear Information System (INIS)

    Molesworth, T.V.; Strachan, N.R.; Findlay, D.J.S.; Wise, M.O.; Forrest, K.R.; Rogers, J.D.

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    Taylor, R.F.

    1988-07-01

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

  6. Corrosion of Metal Inclusions In Bulk Vitrification Waste Packages

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-31

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

  7. Management of packaging waste in Poland--development agenda and accession to the EU.

    Science.gov (United States)

    Grodzińska-Jurczak, Małgorzata; Zakowska, Hanna; Read, Adam

    2004-06-01

    In recent years the issue of the municipal waste in Poland has become increasingly topical, with a considerable rise in the waste generation, much of which can be attributed to a boom in product packaging (mainly plastic). The annual production of plastics packaging has been constantly increasing over the last 20 to 30 years, and now exceeds 3.7 million tons. Due to a lack of processing technologies and poorly developed selective segregation system, packaging waste is still treated as a part of the municipal solid waste (MSW) stream, most of which is landfilled. As a result of Poland's access to the European Union, previous legal regulations governing municipal waste management have been harmonized with those binding on the member countries. One of the main changes, the most revolutionary one, is to make entrepreneurs liable for environmental risks resulting from the introduction of packaging to the market, and for its recycling. In practice, all entrepreneurs are to ensure recovery, and recycling, of used packaging from products introduced to the market at the required level. In recent year, the required recycling levels were fulfilled for all types of materials but mainly by large institutions using grouped and transport packaging waste for that matter. Household packaging gathered in the selective segregation system at the municipalities was practically left alone. This paper is an attempt to describe the system and assess the first year of functioning of the new, revamped system of packaging waste management in Poland. Recommendations are made relating to those features that need to be included in packaging waste management systems in order to maximize their sustainability and harmonization with the EU legal system.

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

    International Nuclear Information System (INIS)

    Ipatti, A.

    1993-03-01

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

  9. Waste Treatment & Immobilization Plant Project

    Data.gov (United States)

    Federal Laboratory Consortium — In southeastern Washington State, Bechtel National, Inc. is designing, constructing and commissioning the world's largest radioactive waste treatment plant for the...

  10. Low-Level Legacy Waste Processing Experience at the West Valley Demonstration Project

    International Nuclear Information System (INIS)

    Valenti, P.J.; Rowell, L.E.; Kurasch, D.H.; Moore, H.R.

    2006-01-01

    This paper presents detailed results and lessons learned from the very challenging and highly successful 2005 low level radioactive waste sorting, packaging, and shipping campaign that removed over 95% of the available inventory of 350,000 ft 3 of legacy low level waste at the West Valley Demonstration Project near West Valley, New York. First some programmatic perspective and site history is provided to provide pertinent context for DOE's waste disposal mandates at the site. This is followed by a detailed description of the waste types, the storage locations, the containers, and the varied sorting and packaging facilities used to accomplish the campaign. The overall sorting and packaging protocols for this inventory of wastes are defined. This is followed by detailed sorting data and results concluding with lessons learned. (authors)

  11. Development of an air flow calorimeter prototype for the measurement of thermal power released by large radioactive waste packages.

    Science.gov (United States)

    Razouk, R; Beaumont, O; Failleau, G; Hay, B; Plumeri, S

    2018-03-01

    The estimation and control of the thermal power released by the radioactive waste packages are a key parameter in the management of radioactive waste geological repository sites. In the framework of the European project "Metrology for decommissioning nuclear facilities," the French National Agency of Radioactive Waste Management (ANDRA) collaborates with Laboratoire National de Métrologie et D'essais in order to measure the thermal power up to 500 W of typical real size radioactive waste packages (of at least 0.175 m 3 ) with an uncertainty better than 5% by using a measurement method traceable to the international system of units. One of the selected metrological approaches is based on the principles of air flow calorimetry. This paper describes in detail the development of the air flow calorimeter prototype as well as the design of a radioactive waste package simulator used for its calibration. Results obtained from the calibration of the calorimeter and from the determination of thermal powers are presented here with an investigation of the measurement uncertainties.

  12. Development of an air flow calorimeter prototype for the measurement of thermal power released by large radioactive waste packages

    Science.gov (United States)

    Razouk, R.; Beaumont, O.; Failleau, G.; Hay, B.; Plumeri, S.

    2018-03-01

    The estimation and control of the thermal power released by the radioactive waste packages are a key parameter in the management of radioactive waste geological repository sites. In the framework of the European project "Metrology for decommissioning nuclear facilities," the French National Agency of Radioactive Waste Management (ANDRA) collaborates with Laboratoire National de Métrologie et D'essais in order to measure the thermal power up to 500 W of typical real size radioactive waste packages (of at least 0.175 m3) with an uncertainty better than 5% by using a measurement method traceable to the international system of units. One of the selected metrological approaches is based on the principles of air flow calorimetry. This paper describes in detail the development of the air flow calorimeter prototype as well as the design of a radioactive waste package simulator used for its calibration. Results obtained from the calibration of the calorimeter and from the determination of thermal powers are presented here with an investigation of the measurement uncertainties.

  13. HORIZONTAL LIFTING OF 5 DHLW/DOE LONG, 12-PWR LONG AND 24-BWR WASTE PACKAGES

    International Nuclear Information System (INIS)

    V. de la Brosse

    2001-01-01

    The objective of this calculation was to determine the structural response of a 12-Pressurized Water Reactor (PWR) Long, a 24-Boiling Water Reactor (BWR) and a 5-Defense High Level Waste/Department of Energy (DHLW/DOE)--Long spent nuclear fuel waste packages lifted in a horizontal position. The scope of this calculation was limited to reporting the calculation results in terms of maximum stress intensities in the trunnion collar sleeves. In addition, the maximum stress intensities in the inner and outer shells of the waste packages were presented for illustrative purposes. The information provided by the sketches (Attachments I, II and III) is that of the potential design of the types of waste packages considered in this calculation, and all obtained results are valid for these designs only. This calculation is associated with the waste package design and was performed by the Waste Package Design Section in accordance with the ''Technical work plan for: Waste Package Design Description for LA'' (Ref. 7). AP-3.12Q, Calculations (Ref. 13), was used to perform the calculation and develop the document

  14. Structural and Thermal Safety Analysis Report for the Type B Radioactive Waste Transport Package

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. H.; Seo, K. S.; Lee, J. C.; Bang, K. S

    2007-09-15

    We carried out structural safety evaluation for the type B radioactive waste transport package. Requirements for type B packages according to the related regulations such as IAEA Safety Standard Series No. TS-R-1, Korea Most Act. 2001-23 and US 10 CFR Part 71 were evaluated. General requirements for packages such as those for a lifting attachment, a tie-down attachment and pressure condition were considered. For the type B radioactive waste transport package, the structural, thermal and containment analyses were carried out under the normal transport conditions. Also the safety analysis were conducted under the accidental transport conditions. The 9 m drop test, 1 m puncture test, fire test and water immersion test under the accidental transport conditions were consecutively done. The type B radioactive waste transport packages were maintained the structural and thermal integrities.

  15. Scientific investigation plan for NNWSI WBS element 1.2.2.5.L: NNWSI waste package performance assessment: Revision 1

    International Nuclear Information System (INIS)

    Eggert, K.G.; O'Connell, W.J.; Lappa, D.A.

    1986-01-01

    Waste package performance assessment contains three broad categories of activities. These activities are: (1) development of a hydrothermal flow and transport model to test concepts to be used in establishing boundary conditions for performance calculations, and to interface EBS release calculations with total system performance calculations; (2) development of a waste package systems model to provide integrated deterministic assessments of performance and analyses of waste package designs; and (3) development of an uncertainty methodology for combination with the system model to perform probabilistic reliability and performance analysis waste package designs. The first category contains activities that aid in determining the scope of a separate, simplified set of hydrologic calculations needed to characterize the waste package environment for performance assessment calculations. The last two activity categories are directly concerned with waste package performance calculations. A rationale for each activity under these groups is presented. All of the activities of performance assessment are either code development or analyses of waste package problems

  16. Solid Waste Projection Model: Model user's guide

    International Nuclear Information System (INIS)

    Stiles, D.L.; Crow, V.L.

    1990-08-01

    The Solid Waste Projection Model (SWPM) system is an analytical tool developed by Pacific Northwest Laboratory (PNL) for Westinghouse Hanford company (WHC) specifically to address solid waste management issues at the Hanford Central Waste Complex (HCWC). This document, one of six documents supporting the SWPM system, contains a description of the system and instructions for preparing to use SWPM and operating Version 1 of the model. 4 figs., 1 tab

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

    Energy Technology Data Exchange (ETDEWEB)

    Horn, J. M., LLNL

    1998-01-01

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

  18. 78 FR 1881 - Certain Food Waste Disposers and Components and Packaging Thereof; Notice of the Commission's...

    Science.gov (United States)

    2013-01-09

    ... INTERNATIONAL TRADE COMMISSION [Investigation No. 337-TA-838] Certain Food Waste Disposers and Components and Packaging Thereof; Notice of the Commission's Determination Not To Review Initial Determinations Granting Complainant's Motions To Partially Terminate the Investigation and To Withdraw the...

  19. Integrated performance assessment model for waste policy package behavior and radionuclide release

    International Nuclear Information System (INIS)

    Kossik, R.; Miller, I.; Cunnane, M.

    1992-01-01

    Golder Associates Inc. (GAI) has developed a probabilistic total system performance assessment and strategy evaluation model (RIP) which can be applied in an iterative manner to evaluate repository site suitability and guide site characterization. This paper describes one component of the RIP software, the waste package behavior and radionuclide release model. The waste package component model considers waste package failure by various modes, matrix alteration/dissolution, and radionuclide mass transfer. Model parameters can be described as functions of local environmental conditions. The waste package component model is coupled to component models for far-field radionuclide transport and disruptive events. The model has recently been applied to the proposed repository at Yucca Mountain

  20. Integrated performance assessment model for waste package behavior and radionuclide release

    International Nuclear Information System (INIS)

    Kossik, R.; Miller, I.; Cunnane, M.

    1992-01-01

    Golder Associates Inc. (GAI) has developed a probabilistic total system performance assessment and strategy evaluation model (RIP) which can be applied in an iterative manner to evaluate repository site suitability and guide site characterization. This paper describes one component of the RIP software, the waste package behavior and radionuclide release model. The waste package component model considers waste package failure by various modes, matrix alteration/dissolution, and radionuclide mass transfer. Model parameters can be described as functions of local environmental conditions. The waste package component model is coupled to component models for far-field radionuclide transport and disruptive events. The model has recently been applied to the proposed repository at Yucca Mountain

  1. Thermal analysis of Yucca Mountain commercial high-level waste packages

    International Nuclear Information System (INIS)

    Altenhofen, M.K.; Eslinger, P.W.

    1992-10-01

    The thermal performance of commercial high-level waste packages was evaluated on a preliminary basis for the candidate Yucca Mountain repository site. The purpose of this study is to provide an estimate for waste package component temperatures as a function of isolation time in tuff. Several recommendations are made concerning the additional information and modeling needed to evaluate the thermal performance of the Yucca Mountain repository system

  2. What are the communication challenges for politicians, experts and stakeholders in order to enhance transparency in nuclear waste management decisions? Report from a Team Syntegrity Meeting. The European Project RISCOM-II. Work Package 3

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Kjell [Karinta-Konsult, Taeby (Sweden); Espejo, Raul [Syncho Ltd., Birmingham (United Kingdom); Wene, Clas-Otto [Wenergy, Lund (Sweden)

    2003-09-01

    The Team Syntegrity Meeting is a special part of the project. It aims for increased awareness among key stakeholder groups in Europe about how nuclear waste decision processes should be developed in order to increase transparency and trust. Team Syntegrity is conducted with a special meeting format. The self-organisation of the meeting is a strong positive feature of the format. Instead of having a project leader setting the agenda, the participants formulate their own topics of relevance starting from an opening question. This report documents the meeting that was held in Lanaken, Belgium on 14-17 May 2002. The opening question for the meeting was: What are communication challenges for politicians, experts and stakeholders in order to enhance transparency in nuclear waste management decisions? There are different opinions about how communication on nuclear waste issues should be done. There are differences between stakeholder groups, and there are different approaches taken in various countries. Still it should be possible to reach a deeper understanding of social communications, that is, understanding the requirements to have effective communications between policy makers, experts and stakeholders. The aim was thus not to develop common views on the nuclear waste problem as such, but rather common grounds for developing procedures for effective communication. Hopefully, this meeting made some progress in this direction. The call for the Team Syntegrity (TS) Meeting resulted in 105 Statements of Importance given in Appendix 2. Following the TS format the meeting then formed its own agenda by first producing 30 Aggregated Statements of Importance (Appendix 3), which were grouped into 12 Consolidated Statements of Importance or topics. The group discussions were thus held under the twelve topics of: Consultation, communication and participation; Mutual learning; Roles and arenas; Heritage; Transparency; Wider context; Process; Risk; Institutional cultures

  3. In-Package Chemistry Abstraction

    Energy Technology Data Exchange (ETDEWEB)

    P.S. Domski

    2003-07-21

    The work associated with the development of this model report was performed in accordance with the requirements established in ''Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of SR and LA'' (BSC 2002a). The in-package chemistry model and in-package chemistry model abstraction are developed to predict the bulk chemistry inside of a failed waste package and to provide simplified expressions of that chemistry. The purpose of this work is to provide the abstraction model to the Performance Assessment Project and the Waste Form Department for development of geochemical models of the waste package interior. The scope of this model report is to describe the development and validation of the in-package chemistry model and in-package chemistry model abstraction. The in-package chemistry model will consider chemical interactions of water with the waste package materials and the waste form for commercial spent nuclear fuel (CSNF) and codisposed high-level waste glass (HLWG) and N Reactor spent fuel (CDNR). The in-package chemistry model includes two sub-models, the first a water vapor condensation (WVC) model, where water enters a waste package as vapor and forms a film on the waste package components with subsequent film reactions with the waste package materials and waste form--this is a no-flow model, the reacted fluids do not exit the waste package via advection. The second sub-model of the in-package chemistry model is the seepage dripping model (SDM), where water, water that may have seeped into the repository from the surrounding rock, enters a failed waste package and reacts with the waste package components and waste form, and then exits the waste package with no accumulation of reacted water in the waste package. Both of the submodels of the in-package chemistry model are film models in contrast to past in-package chemistry models where all of the waste package pore space was filled with water. The

  4. In-Package Chemistry Abstraction

    International Nuclear Information System (INIS)

    P.S. Domski

    2003-01-01

    The work associated with the development of this model report was performed in accordance with the requirements established in ''Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of SR and LA'' (BSC 2002a). The in-package chemistry model and in-package chemistry model abstraction are developed to predict the bulk chemistry inside of a failed waste package and to provide simplified expressions of that chemistry. The purpose of this work is to provide the abstraction model to the Performance Assessment Project and the Waste Form Department for development of geochemical models of the waste package interior. The scope of this model report is to describe the development and validation of the in-package chemistry model and in-package chemistry model abstraction. The in-package chemistry model will consider chemical interactions of water with the waste package materials and the waste form for commercial spent nuclear fuel (CSNF) and codisposed high-level waste glass (HLWG) and N Reactor spent fuel (CDNR). The in-package chemistry model includes two sub-models, the first a water vapor condensation (WVC) model, where water enters a waste package as vapor and forms a film on the waste package components with subsequent film reactions with the waste package materials and waste form--this is a no-flow model, the reacted fluids do not exit the waste package via advection. The second sub-model of the in-package chemistry model is the seepage dripping model (SDM), where water, water that may have seeped into the repository from the surrounding rock, enters a failed waste package and reacts with the waste package components and waste form, and then exits the waste package with no accumulation of reacted water in the waste package. Both of the submodels of the in-package chemistry model are film models in contrast to past in-package chemistry models where all of the waste package pore space was filled with water. The current in-package

  5. Greater-than-Class C low-level radioactive waste characterization. Appendix E-4: Packaging factors for greater-than-Class C low-level radioactive waste

    International Nuclear Information System (INIS)

    Quinn, G.; Grant, P.; Winberg, M.; Williams, K.

    1994-09-01

    This report estimates packaging factors for several waste types that are potential greater-than-Class C (GTCC) low-level radioactive waste (LLW). The packaging factor is defined as the volume of a GTCC LLW disposal container divided by the as-generated or ''unpackaged'' volume of the waste loaded into the disposal container. Packaging factors reflect any processes that reduce or increase an original unpackaged volume of GTCC LLW, the volume inside a waste container not occupied by the waste, and the volume of the waste container itself. Three values are developed that represent (a) the base case or most likely value for a packaging factor, (b) a high case packaging factor that corresponds to the largest anticipated disposal volume of waste, and (c) a low case packaging factor for the smallest volume expected. GTCC LLW is placed in three categories for evaluation in this report: activated metals, sealed sources, and all other waste

  6. Second generation waste package design and storage concept for the Yucca Mountain Repository

    International Nuclear Information System (INIS)

    Armijo, Joseph Sam; Kar, Piyush; Misra, Manoranjan

    2006-01-01

    The reference waste package design and operating mode to be used in the Yucca Mountain Repository is reviewed. An alternate (second generation) operating concept and waste package design is proposed to reduce the risk of localized corrosion of waste packages and to reduce repository costs. The second generation waste package design and storage concept is proposed for implementation after the initial licensing and operation of the reference repository design. Implementation of the second generation concept at Yucca Mountain would follow regulatory processes analogous to those used successfully to extend the design life and uprate the power of commercial light water nuclear reactors in the United States. The second generation concept utilizes the benefits of hot dry storage to minimize the potential for localized corrosion of the waste package by liquid electrolytes. The second generation concept permits major reductions in repository costs by increasing the number of fuel assemblies stored in each waste package, by eliminating the need for titanium drip shields and by fabricating the outer container from corrosion resistant low alloy carbon steel

  7. The packaging and transport of low and intermediate level radioactive wastes

    International Nuclear Information System (INIS)

    Grover, J.R.; Price, M.S.T.

    1985-01-01

    Up to the present time, the majority of the radioactive waste which has been transported in the United Kingdom has been low level waste for disposal in the trenches of the shallow burial site operated by British Nuclear Fuels plc at Drigg and also the packaged waste destined for sea disposal in the annual operation. However, the main bulk of the low and intermediate level wastes which have been generated over the last quarter century remain in store at the various nuclear sites where it originated. Before significant packaging and transport of intermediate level wastes takes place it is desirable to examine the sources and types of wastes, the immobilisation and packaging processes and plants, the transport, and the problems of handling of packages at future land repositories. Optimisation of the packaging and transport must take account of both the upstream and downstream con=straints as well as the implications of complying with both the IAEA Transport Regulations and radiological protection guidelines. Packages for sea disposal must in addition comply with the requirements of the London Dumping Convention and the NEA guidelines. (author)

  8. Compaction and packaging of dry active municipal wastes

    International Nuclear Information System (INIS)

    Chen Zongming; Xi Xinmin

    1994-01-01

    The authors present the feature of a compaction system for active municipal wastes and the radiological monitoring results of workplace and environment. A variety of dry active municipal wastes could be compacted by this system. Volume reduction factor attained to 5 to 7 for soft wastes and 8 to 13 for hard wastes. No evident radiological impact was found on workplace and environment

  9. Integrated Data Base: Status and waste projections

    International Nuclear Information System (INIS)

    Klein, J.A.

    1990-01-01

    The Integrated Data Base (IDB) is the official US Department of Energy (DOE) data base for spent fuel and radioactive waste inventories and projections. DOE low-level waste (LLW) is just one of the many waste types that are documented with the IDB. Summary-level tables and figures are presented illustrating historical and projected volume changes of DOE LLW. This information is readily available through the annual IDB publication. Other presentation formats are also available to the DOE community through a request to the IDB Program. 4 refs., 6 figs., 5 tabs

  10. An analytical one-dimensional model for predicting waste package performance

    International Nuclear Information System (INIS)

    Relyea, J.F.; Wood, M.I.

    1984-01-01

    A method for allocating waste package performance requirements among waste package components with regard to radionuclide isolation has been developed. Modification or change in this approach can be expected as the understanding of radionuclide behavior in the waste package improves. Thus, the performance requirements derived in this document are preliminary and subject to change. However, this kind of analysis is a useful starting point. It has also proved useful for identifying a small group of radionuclides which should be emphasized in a laboratory experimental program designed to characterize the behavior of specific radionuclides in the waste package environment. A simple one-dimensional, two media transport model has been derived and used to calculate radionuclide transport from the waste form-packing material interface of the waste package into the host rock. Cumulative release over 10,000 years, maximum yearly releases and release rates at the packing material-host rock interface were evaluated on a radionuclide-by radionuclide basis. The major parameters controlling radionuclide release were found to be: radionuclide solubility, porosity of the rock, isotopic ratio of the radionuclide and surface area of the waste form-packing material interface. 15 refs., 2 figs., 16 tabs

  11. Project safety studies - nuclear waste management (PSE)

    International Nuclear Information System (INIS)

    1981-10-01

    The project 'Safety Studies-Nuclear Waste Management' (PSE) is a research project performed by order of the Federal Minister for Research and Technology, the general purpose of which is to deepen and ensure the understanding of the safety aspects of the nuclear waste management and to prepare a risk analysis which will have to be established in the future. Owing to this the project is part of a series of projects which serve the further development of the concept of nuclear waste management and its safety, and which are set up in such a way as to accompany the realization of that concept. This report contains the results of the first stage of the project from 1978 to mid-1981. (orig./RW) [de

  12. One project`s waste is another project`s resource

    Energy Technology Data Exchange (ETDEWEB)

    Short, J.

    1997-02-01

    The author describes the efforts being made toward pollution prevention within the DOE complex, as a way to reduce overall project costs, in addition to decreasing the amount of waste to be handled. Pollution prevention is a concept which is trying to be ingrained into project planning. Part of the program involves the concept that ultimately the responsibility for waste comes back to the generator. Parts of the program involve efforts to reuse materials and equipment on new projects, to recycle wastes to generate offsetting revenue, and to increase awareness, accountability and incentives so as to stimulate action on this plan. Summaries of examples are presented in tables.

  13. Options for reducing food waste by ‘Quality Controlled Logistics’ using intelligent packaging along the supply chain

    NARCIS (Netherlands)

    Heising, J.K.; Claassen, G.D.H.; Dekker, M.

    2017-01-01

    Optimizing supply chain management can help to reduce food waste. This article describes how intelligent packaging can be used to reduce food waste when used in supply chain management based on Quality Controlled Logistics (QCL). Intelligent packaging senses compounds in the package that correlate

  14. Waste diminution in Construction projects: Environmental Predicaments

    Science.gov (United States)

    Gharehbaghi, Koorosh; Scott-Young, Christina

    2018-03-01

    Waste diminution in construction projects is not only a behavioural issue, but also an energy consumption and reduction concern. With construction waste equating to the significant amount of exhausted energy together with increased pollution, this contributes to a series of environmental predicaments. The overall goal of construction solid Waste Management is to collect, treat and dispose of solid wastes generated by project activities in an environmentally and socially satisfactory manner, using the most economical means available. As cities expand, their construction activities and consumption patterns further drive up the solid waste quantities. Governments are usually authorized to have responsibility for providing solid Waste Management services, and various administrative laws give them exclusive ownership over the waste produced. In addition, construction waste processing can be further controlled and minimized according to specialized authorities such as Environmental Protection Agencies (EPA) and their relevant acts and regulations. Moreover, a Construction Environmental Management Plan (CEMP) can further control the treatment of waste and therefore, reduce the amount produced. Key elements of a CEMP not only include complying with relevant legislation, standards and guidance from the EPA; however, also to ensuring that there are systems in place to resolve any potential problems associated with site activities. Accordingly, as a part of energy consumption and lessening strategies, this paper will discuss various effective waste reduction methods for construction projects. Finally, this paper will also examine tactics to further improve energy efficiency through innovative construction Waste Management strategies (including desirability rating of most favourable options) to promote the lessening of overall CO2production.

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

    Energy Technology Data Exchange (ETDEWEB)

    Claiborne, H.C.; Croff, A.G.; Griess, J.C.; Smith, F.J.

    1987-09-01

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

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

    International Nuclear Information System (INIS)

    Claiborne, H.C.; Croff, A.G.; Griess, J.C.; Smith, F.J.

    1987-09-01

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

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

    International Nuclear Information System (INIS)

    1983-06-01

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

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

    International Nuclear Information System (INIS)

    PD Meyer; RJ Serne

    1999-01-01

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

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

    NARCIS (Netherlands)

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

    2013-01-01

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

  20. Operating cost estimate low-level radioactive waste volume reduction and packaging options

    International Nuclear Information System (INIS)

    Williams, P.C.; Collins, E.G.

    1983-01-01

    The projected operating cost data examine the differences in compaction and incineration for DAW. Two commonly achieved densities for compaction are presented, while a third case approaching the theoretical density of the cloth and paper materials shows the interesting possibilities of improved compaction. Incineration at two benchmarks for the chlorinated plastic component of typical DAW illustrates the impact waste stream make-up has on economics. Evaporator concentrates are considered as a liquid at four concentrations (6%, 12%, 25% and 50%) and after all water has been removed for the cement and polymer cases. Five separate processing schemes are presented for bead resins. These range from dewatering and packaging in high integrity containers to incineration with solidification

  1. Interpretation of non destructive combined nuclear measurements for the characterization of radioactive wastes and waste packages

    International Nuclear Information System (INIS)

    Raoux, Anne-Cecile

    2000-01-01

    Nuclear industry produces radioactive waste and is faced with the problem of their management, especially for those which have a long radioactive decay time. In view to be able to define the best storage solution, alpha bearing solid waste are identified by different specific parameters (alpha, beta activities,... ). Then, the storage and cost optimizations are essential stakes. The quantification of these parameters can be obtained by the implementation of non destructive nuclear measurement methods generally associated with information from the manufacturing process of the waste. The works presented in this report are dedicated to two complementary aspects of the nuclear waste management issue. On the one hand, an experimental study concerning the possibilities of the prompt and delayed neutron counting with only one measurement result from neutron interrogation is presented. On the other hand, an interpretation method allowing the determination of the waste package specific parameters and their uncertainties has been developed. It is based on random trials which allow to describe the parameters as statistical distributions (Monte Carlo method). It was resulting in the realization of a software called RECITAL (information combination and solving process by random trials). This software was applied to the isotopic quantification of "2"3"5U and "2"3"9Pu from prompt and delayed signals of neutron interrogation. It was also used to demonstrate the complementarity of photofission interrogation with neutron interrogation in view to correct "2"3"8U interference on the delayed fission signal, especially when "2"3"8U contribution is similar to "2"3"5U and "2"3"9Pu ones. (author) [fr

  2. Pilot Project Technology Business Case: Mobile Work Packages

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, Ken [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lawrie, Sean [ScottMadden, Inc., Raleigh, NC (United States); Niedermuller, Josef [ScottMadden, Inc., Raleigh, NC (United States)

    2015-05-01

    Performance advantages of the new pilot project technologies are widely acknowledged, but it has proven difficult for utilities to derive business cases for justifying investment in these new capabilities. Lack of a business case is often cited by utilities as a barrier to pursuing wide-scale application of digital technologies to nuclear plant work activities. The decision to move forward with funding usually hinges on demonstrating actual cost reductions that can be credited to budgets and thereby truly reduce O&M or capital costs. Technology enhancements, while enhancing work methods and making work more efficient, often fail to eliminate workload such that it changes overall staffing and material cost requirements. It is critical to demonstrate cost reductions or impacts on non-cost performance objectives in order for the business case to justify investment by nuclear operators. The Business Case Methodology (BCM) was developed in September of 2015 to frame the benefit side of II&C technologies to address the “benefit” side of the analysis—as opposed to the cost side—and how the organization evaluates discretionary projects (net present value (NPV), accounting effects of taxes, discount rates, etc.). The cost and analysis side is not particularly difficult for the organization and can usually be determined with a fair amount of precision (not withstanding implementation project cost overruns). It is in determining the “benefits” side of the analysis that utilities have more difficulty in technology projects and that is the focus of this methodology. The methodology is presented in the context of the entire process, but the tool provided is limited to determining the organizational benefits only. This report describes a the use of the BCM in building a business case for mobile work packages, which includes computer-based procedures and other automated elements of a work package. Key to those impacts will be identifying where the savings are

  3. Radioactivity evaluation method for pre-packed concrete packages of low-level dry active wastes

    International Nuclear Information System (INIS)

    Sakai, Toshiaki; Funahashi, Tetsuo; Watabe, Kiyomi; Ozawa, Yukitoshi; Kashiwagi, Makoto

    1998-01-01

    Low-level dry active wastes of nuclear power plants are grouted with cement mortal in a container and planned to disposed into the shallow land disposal site. The characteristics of radionuclides contained in dry active wastes are same as homogeneous solidified wastes. In the previous report, we reported the applicability of the radioactivity evaluation methods established for homogeneous solidified wastes to pre-packed concrete packages. This report outlines the developed radioactivity evaluation methods for pre-packed concrete packages based upon recent data. Since the characteristics of dry active wastes depend upon the plant system in which dry active wastes originate and the types of contamination, sampling of wastes and activity measurement were executed to derive scaling factors. The radioactivity measurement methods were also verified. The applicability of non-destructive methods to measure radioactivity concentration of pre-packed concrete packages was examined by computer simulation. It is concluded that those methods are accurate enough to measure actual waste packages. (author)

  4. Quality assurance requirements and methods for high level waste package acceptability

    International Nuclear Information System (INIS)

    1992-12-01

    This document should serve as guidance for assigning the necessary items to control the conditioning process in such a way that waste packages are produced in compliance with the waste acceptance requirements. It is also provided to promote the exchange of information on quality assurance requirements and on the application of quality assurance methods associated with the production of high level waste packages, to ensure that these waste packages comply with the requirements for transportation, interim storage and waste disposal in deep geological formations. The document is intended to assist both the operators of conditioning facilities and repositories as well as national authorities and regulatory bodies, involved in the licensing of the conditioning of high level radioactive wastes or in the development of deep underground disposal systems. The document recommends the quality assurance requirements and methods which are necessary to generate data for these parameters identified in IAEA-TECDOC-560 on qualitative acceptance criteria, and indicates where and when the control methods can be applied, e.g. in the operation or commissioning of a process or in the development of a waste package design. Emphasis is on the control of the process and little reliance is placed on non-destructive or destructive testing. Qualitative criteria, relevant to disposal of high level waste, are repository dependent and are not addressed here. 37 refs, 3 figs, 2 tabs

  5. Waste management project technical baseline description

    International Nuclear Information System (INIS)

    Sederburg, J.P.

    1997-01-01

    A systems engineering approach has been taken to describe the technical baseline under which the Waste Management Project is currently operating. The document contains a mission analysis, function analysis, requirement analysis, interface definitions, alternative analysis, system definition, documentation requirements, implementation definitions, and discussion of uncertainties facing the Project

  6. Preliminary corrosion models for BWIP [Basalt Waste Isolation Project] canister materials

    International Nuclear Information System (INIS)

    Fish, R.L.; Anantatmula, R.P.

    1983-01-01

    Waste package development for the Basalt Waste Isolation Project (BWIP) requires the generation of materials degradation data under repository relevant conditions. These data are used to develop predictive models for the behavior of each component of waste package. The component models are exercised in performance analyses to optimize the waste package design. This document presents all repository relevant canister materials corrosion data that the BWIP and others have developed to date, describes the methodology used to develop preliminary corrosion models and provides the mathematical description of the models for both low carbon steel and Fe9Cr1Mo steel. Example environment/temperature history and model application calculations are presented to aid in understanding the models. The models are preliminary in nature and will be updated as additional corrosion data become available. 6 refs., 5 tabs

  7. Gas generation phenomena in radioactive waste transportation packaging

    International Nuclear Information System (INIS)

    Nigrey, P.J.

    1998-01-01

    The interaction of radiation from radioactive materials with the waste matrix can lead to the deterioration of the waste form resulting in the possible of gaseous species. Depending on the type and characteristics of the radiation source, the generation of hydrogen may predominate. Since the interaction of alpha particles with the waste form results in significant energy transfer, other gases such as carbon oxides, methane, nitrogen oxides, oxygen, water, and helium are possible. The type of gases produced from the waste forms is determined by the mechanisms involved in the waste degradation. For transuranic wastes, the identified degradation mechanisms are reported to be caused by radiolysis, thermal decomposition or dewatering, chemical corrosion, and bacterial action. While all these mechanisms may be responsible for the building of gases during the storage of wastes, radiolysis and thermal decomposition appear to be main contributors during waste transport operations. (authors)

  8. Isotope production potential at Sandia National Laboratories: Product, waste, packaging, and transportation

    International Nuclear Information System (INIS)

    Trennel, A.J.

    1995-01-01

    The U.S. Congress directed the U.S. Department of Energy to establish a domestic source of molybdenum-99, an essential isotope used in nuclear medicine and radiopharmacology. An Environmental Impact Statement for production of 99 Mo at one of four candidate sites is being prepared. As one of the candidate sites, Sandia National Laboratories is developing the Isotope Production Project. Using federally approved processes and procedures now owned by the U.S. Department of Energy, and existing facilities that would be modified to meet the production requirements, the Sandia National Laboratories' Isotope Project would manufacture up to 30 percent of the U.S. market, with the capacity to meet 100 percent of the domestic need if necessary. This paper provides a brief overview of the facility, equipment, and processes required to produce isotopes. Packaging and transportation issues affecting both product and waste are addressed, and the storage and disposal of the four low-level radioactive waste types generated by the production program are considered. Recommendations for future development are provided

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

    International Nuclear Information System (INIS)

    Ferreira, S.; Cabral, M.; Cruz, N.F. da; Simões, P.; Marques, R.C.

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, S.; Cabral, M. [CEG-IST, ULisboa, Av. Rovisco Pais, 1049-001 Lisbon (Portugal); Cruz, N.F. da, E-mail: nunocruz@tecnico.ulisboa.pt [IST, ULisboa, Av. Rovisco Pais, 1049-001 Lisbon (Portugal); Simões, P. [IST, ULisboa, Av. Rovisco Pais, 1049-001 Lisbon (Portugal); Marques, R.C. [CESUR, IST, ULisboa, Av. Rovisco Pais, 1049-001 Lisbon (Portugal)

    2014-09-15

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

  11. Acceptance and tracking of waste packages from nuclear power plants at the Centre de l'Aube

    International Nuclear Information System (INIS)

    Errera, J.; Tison, J.L.

    2001-01-01

    For 30 years, the French National Agency for Radioactive Waste Management (ANDRA) is in charge of the radioactive waste management and acquired a good knowledge relating to the control of low and intermediate level waste produced by nuclear power plants (NPP), the waste characteristics and the waste conditioning. The integrated waste management system for low-level radioactive waste in France implemented by ANDRA covers all stages from waste generation to final disposal at the Centre de I'Aube near surface facility. ANDRA defined a quality assurance program for waste management that specifies the level of quality to be achieved by solidification and packaging processes, defines quality control requirements and defines waste tracking requirements, from waste generation through final disposal. Verification of quality of waste packages is implemented at three levels of the waste management system. The first one consists of inspections of waste packages at the generator's premises and audits of the quality assurance organization of the waste generator. The second level of verification consists of the waste tracking system. It allows identifying and tracking each waste package from the step it is fabricated to its final disposal at the ANDRA site. The third level of verification is obtained by mean of non-destructive and destructive assays of waste packages. These assays allow to verify generator compliance with ANDRA's technical specifications and to investigate the accuracy of physical and radioactive characteristics reported to ANDRA by the generator. (author)

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

    International Nuclear Information System (INIS)

    Richet, C.; Galle, C.; Le Bescop, P.; Peycelon, H.; Bejaoui, S.; Tovena, I.; Pointeau, I.; L'Hostis, V.; Levera, P.

    2004-03-01

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

  13. Specification of safety requirements for waste packages with respect to practicable quality control measures

    International Nuclear Information System (INIS)

    Gruendler, D.; Wurtinger, W.

    1987-01-01

    Waste packages for disposal in a repository in the Federal Republic of Germany have to meet safety requirements derived from site specific safety analyses. The examination of the waste packages with regard to compliance with these requirements is the main objective of quality control measures. With respect to quality control the requirements have to be specified in a way that practicable control measures can be applied. This is dealt with for the quality control of the activity inventory and the quality control of the waste form. The paper discusses the determination of the activity of hard-to-measure radionuclides and the specification of safety related requirements for the waste form and the packaging using typical examples

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

    International Nuclear Information System (INIS)

    Nigrey, P.J.

    1995-01-01

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

  15. Accuracy of hazardous waste project estimates

    International Nuclear Information System (INIS)

    Hackney, J.W.

    1989-01-01

    The HAZRATE system has been developed to appraise the current state of definition of hazardous waste remedial projects. This is shown to have a high degree of correlation to the financial risk of such projects. The method employs a weighted checklist indicating the current degree of definition of some 150 significant project elements. It is based on the author's experience with a similar system for establishing the risk characteristics of process plant projects (Hackney, 1965 and 1989; 1985). In this paper definition ratings for 15 hazardous waste remedial projects have been correlated with the excesses of their actual costs over their base estimates, excluding any allowances for contingencies. Equations are presented, based on this study, for computation of the contingency allowance needed and estimate accuracy possible at a given stage of project development

  16. Technical program plan, Basalt Waste Isolation Project

    International Nuclear Information System (INIS)

    1979-12-01

    The Basalt Waste Isolation Project (BWIP) program as administered by the DOE's Richland Operations Office and Rockwell Hanford Operations is described. The objectives, scope and scientific technologies are discussed. The work breakdown structure of the project includes: project management and support, systems integration, geosciences, hydrology, engineered barriers, test facility design and construction, engineering testing, repository studies, and schedules. The budget of the program including operating and capital cost control is also included

  17. A mechanistic model for leaching from low-level radioactive waste packages

    International Nuclear Information System (INIS)

    Kempf, C.R.

    1988-01-01

    The development of a waste leaching model to predict radionuclide releases from porous wastes in corrodible outer containers in unsaturated conditions and/or conditions of intermittent water flow is summarized in this paper. Three major processes have been conceptualized as necessarily participating in waste leaching: infiltration of water to the waste package; interaction of this water with the waste; and exit of radionuclide-laden water from the waste package. Through the exit point, the main features of the whole leaching process ware held in common. The departure occurs in two main ways: 1) the method of entrance of the radionuclides to leachant (i.e. part of the waste-water interaction phase outlined earlier); and 2) the mode of exit from waste form/waste package (i.e., the exit of radionuclide-laden water phase). The first branching point, which occurs in relation to 1), leads to either readily soluble species directly entering leachant on contact, or to other processes - mainly expected to be diffusion, dissolution or ion exchange, or some combination thereof

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

    International Nuclear Information System (INIS)

    Moak, D.P.

    1986-09-01

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

  19. Industrial Program of Waste Management - Cigeo Project - 13033

    Energy Technology Data Exchange (ETDEWEB)

    Butez, Marc [Agence nationale pour la gestion des dechets radioactifs - Andra, 1-7, rue Jean Monnet 92298 Chatenay-Malabry (France); Bartagnon, Olivier; Gagner, Laurent [AREVA NC Tour AREVA 1 place de la Coupole 92084 Paris La Defense (France); Advocat, Thierry; Sacristan, Pablo [Commissariat a l' energie atomique et aux energies alternatives - CEA, CEA-SACLAY 91191 Gif sur Yvette Cedex (France); Beguin, Stephane [Electricite de France - EDF, Division Combustible Nucleaire, 1, Place Pleyel Site Cap Ampere93282 Saint Denis (France)

    2013-07-01

    The French Planning Act of 28 June 2006 prescribed that a reversible repository in a deep geological formation be chosen as the reference solution for the long-term management of high-level and intermediate-level long-lived radioactive waste. It also entrusted the responsibility of further studies and design of the repository (named Cigeo) upon the French Radioactive Waste Management Agency (Andra), in order for the review of the creation-license application to start in 2015 and, subject to its approval, the commissioning of the repository to take place in 2025. Andra is responsible for siting, designing, implementing, operating the future geological repository, including operational and long term safety and waste acceptance. Nuclear operators (Electricite de France (EDF), AREVA NC, and the French Commission in charge of Atomic Energy and Alternative Energies (CEA) are technically and financially responsible for the waste they generate, with no limit in time. They provide Andra, on one hand, with waste packages related input data, and on the other hand with their long term industrial experiences of high and intermediate-level long-lived radwaste management and nuclear operation. Andra, EDF, AREVA and CEA established a cooperation agreement for strengthening their collaborations in these fields. Within this agreement Andra and the nuclear operators have defined an industrial program for waste management. This program includes the waste inventory to be taken into account for the design of the Cigeo project and the structural hypothesis underlying its phased development. It schedules the delivery of the different categories of waste and defines associated flows. (authors)

  20. Industrial Program of Waste Management - Cigeo Project - 13033

    International Nuclear Information System (INIS)

    Butez, Marc; Bartagnon, Olivier; Gagner, Laurent; Advocat, Thierry; Sacristan, Pablo; Beguin, Stephane

    2013-01-01

    The French Planning Act of 28 June 2006 prescribed that a reversible repository in a deep geological formation be chosen as the reference solution for the long-term management of high-level and intermediate-level long-lived radioactive waste. It also entrusted the responsibility of further studies and design of the repository (named Cigeo) upon the French Radioactive Waste Management Agency (Andra), in order for the review of the creation-license application to start in 2015 and, subject to its approval, the commissioning of the repository to take place in 2025. Andra is responsible for siting, designing, implementing, operating the future geological repository, including operational and long term safety and waste acceptance. Nuclear operators (Electricite de France (EDF), AREVA NC, and the French Commission in charge of Atomic Energy and Alternative Energies (CEA) are technically and financially responsible for the waste they generate, with no limit in time. They provide Andra, on one hand, with waste packages related input data, and on the other hand with their long term industrial experiences of high and intermediate-level long-lived radwaste management and nuclear operation. Andra, EDF, AREVA and CEA established a cooperation agreement for strengthening their collaborations in these fields. Within this agreement Andra and the nuclear operators have defined an industrial program for waste management. This program includes the waste inventory to be taken into account for the design of the Cigeo project and the structural hypothesis underlying its phased development. It schedules the delivery of the different categories of waste and defines associated flows. (authors)

  1. Project No. 4 - Waste incineration facility

    International Nuclear Information System (INIS)

    2000-01-01

    There are currently 12000 m 3 of combustible waste stored at the Ignalina NPP site. It is estimated that by 2005 the volume will have increase to 15000 m 3 (filters, personnel protection, clothing and plastics). As a part of the preparation for the closure of the Ignalina NPP an incineration facility will be required to process combustible wastes to reduce the overall volume of short-lived radioactive wastes stored at the Ignalina NPP site, thus reducing the overall risk to the environment. Project activities includes the design, construction and commissioning of the proposed facility, including all licensing documentation

  2. Hydrothermal carbonization of food waste and associated packaging materials for energy source generation.

    Science.gov (United States)

    Li, Liang; Diederick, Ryan; Flora, Joseph R V; Berge, Nicole D

    2013-11-01

    Hydrothermal carbonization (HTC) is a thermal conversion technique that converts food wastes and associated packaging materials to a valuable, energy-rich resource. Food waste collected from local restaurants was carbonized over time at different temperatures (225, 250 and 275°C) and solids concentrations to determine how process conditions influence carbonization product properties and composition. Experiments were also conducted to determine the influence of packaging material on food waste carbonization. Results indicate the majority of initial carbon remains integrated within the solid-phase at the solids concentrations and reaction temperatures evaluated. Initial solids concentration influences carbon distribution because of increased compound solubilization, while changes in reaction temperature imparted little change on carbon distribution. The presence of packaging materials significantly influences the energy content of the recovered solids. As the proportion of packaging materials increase, the energy content of recovered solids decreases because of the low energetic retention associated with the packaging materials. HTC results in net positive energy balances at all conditions, except at a 5% (dry wt.) solids concentration. Carbonization of food waste and associated packaging materials also results in net positive balances, but energy needs for solids post-processing are significant. Advantages associated with carbonization are not fully realized when only evaluating process energetics. A more detailed life cycle assessment is needed for a more complete comparison of processes. Copyright © 2013 Elsevier Ltd. All rights reserved.

  3. Waste transport and storage: Packaging refused due to failure in fulfilling QC/QA requirements

    International Nuclear Information System (INIS)

    Bruno, N.C.; Brandao, R.O.; Cavalcante, V.L.

    2001-01-01

    The Brazilian Nuclear Programme comprises several nuclear and radioactive facilities including Angra I Nuclear Power Plant, in operation since 1981, and Angra II, scheduled to start its operation by the end of 1999. Among the other ones there are uranium mining and milling facilities, four research reactors and one industrial facility of monazite sands processing. The already existing waste generation and near future ones claim to a solution regarding waste disposal. Although site selection criteria for waste repository in Brazil has already been defined, political and psychosocial aspects have strong impact. Trauma generated by Goiania's radiological accident has led to difficulties when decisions about this matter have to be taken. As a consequence, the waste generated by Angra I is still in a provisional facility at the plant's site. Wastes from the medical sources are stored in research institutes while waste generated from monazite sands is kept in a dam system. In order to overpack non-qualified packages containing waste of Angra I NPP, 70 lost concrete shielding packagings had to be provided. Based on successfully designed and tested prototype, packagings and respective lids specifications were written, approved and released for serial production. As part of packaging certification process, Brazilian Competent Authority performed a regulatory inspection and audit. Various findings, such as weaknesses in quality control and quality assurance records, unacceptable test results as well as failure in modify the concrete composition during a testified packaging manufacturing, led Competent Authority to refuse the packagings as containers until complementary tests could be performed. Further tests and evaluations led the Competent Authority to conclude that the manufacturer failed to both comply with requirements established in packaging specification and fulfill quality control/quality assurance requirements. As responsible by federal law for the reception and

  4. Tank Waste Remediation System Projects Document Control Plan

    International Nuclear Information System (INIS)

    Slater, G.D.; Halverson, T.G.

    1994-01-01

    The purpose of this Tank Waste Remediation System Projects Document Control Plan is to provide requirements and responsibilities for document control for the Hanford Waste Vitrification Plant (HWVP) Project and the Initial Pretreatment Module (IPM) Project

  5. The Reduction of Solid Waste Associated with Military Ration Packaging

    National Research Council Canada - National Science Library

    Ratto, Jo Ann; Lucciarini, Jeanne; Thellen, Christopher; Froio, Danielle; D'Souza, Nandika A

    2006-01-01

    ... decrease the amount of solid waste generated by the military. These nanocomposites formulations were melt processed into films and characterized for barrier, mechanical, thermal, and biodegradation properties...

  6. Project-led education in packaging development and management

    NARCIS (Netherlands)

    Oude Luttikhuis, Ellen; de Lange, Jos; ten Klooster, Roland; Lutters, Diederick

    2014-01-01

    As an implicit subset of industrial design engineering, packaging development and management thereof has long been a changeling, because product and packaging development are usually regarded in a similar manner. At the same time, there is a clear difference between packaging design and product

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-01-18

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

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  9. Study of applicable methods on safety verification of disposal facilities and waste packages

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    Three subjects about safety verification on the disposal of low level radioactive waste were investigated in FY. 2012. For radioactive waste disposal facilities, specs and construction techniques of covering with soil to prevent possible destruction caused by natural events (e.g. earthquake) were studied to consider verification methods for those specs. For waste packages subject to near surface pit disposal, settings of scaling factor and average radioactivity concentration (hereafter referred to as ''SF'') on container-filled and solidified waste packages generated from Kashiwazaki Kariwa Nuclear Power Station Unit 1-5, setting of cesium residual ratio of molten solidified waste generated from Tokai and Tokai No.2 Power Stations, etc. were studied. Those results were finalized in consideration of the opinion from advisory panel, and publicly opened as JNES-EV reports. In FY 2012, five JNES reports were published and these have been used as standards of safety verification on waste packages. The verification method of radioactive wastes subject to near-surface trench disposal and intermediate depth disposal were also studied. For radioactive wastes which will be returned from overseas, determination methods of radioactive concentration, heat rate and hydrogen generation rate of CSD-C were established. Determination methods of radioactive concentration and heat rate of CSD-B were also established. These results will be referred to verification manuals. (author)

  10. Sustainable Steel Carburization by Using Snack Packaging Plastic Waste as Carbon Resources

    Directory of Open Access Journals (Sweden)

    Songyan Yin

    2018-01-01

    Full Text Available In recent years, the research regarding waste conversion to resources technology has attracted growing attention with the continued increase of waste accumulation issues and rapid depletion of natural resources. However, the study, with respect to utilizing plastics waste as carbon resources in the metals industry, is still limited. In this work, an environmentally friendly approach to utilize snack packaging plastic waste as a valuable carbon resources for steel carburization is investigated. At high temperature, plastic waste could be subject to pyrolytic gasification and decompose into small molecular hydrocarbon gaseous products which have the potential to be used as carburization agents for steel. When heating some snack packaging plastic waste and a steel sample together at the carburization temperature, a considerable amount of carbon-rich reducing gases, like methane, could be liberated from the plastic waste and absorbed by the steel sample as a carbon precursor for carburization. The resulting carburization effect on steel was investigated by optical microscopy, scanning electron microscopy, electron probe microanalyzer, and X-ray photoelectron spectrometer techniques. These investigation results all showed that snack packaging plastic waste could work effectively as a valuable carbon resource for steel carburization leading to a significant increase of surface carbon content and the corresponding microstructure evolution in steel.

  11. PACCOM: A nuclear waste packaging facility cost model: Draft technical report

    International Nuclear Information System (INIS)

    Dippold, D.G.; Tzemos, S.; Smith, D.J.

    1985-05-01

    PACCOM is a computerized, parametric model used to estimate the capital, operating, and decommissioning costs of a variety of nuclear waste packaging facility configurations. The model is based upon a modular waste packaging facility concept from which functional components of the overall facility have been identified and their design and costs related to various parameters such as waste type, waste throughput, and the number of operational shifts employed. The model may be used to either estimate the cost of a particular waste packaging facility configuration or to explore the cost tradeoff between plant capital and labor. That is, one may use the model to search for the particular facility sizes and associated cost which when coupled with a particular number of shifts, and thus staffing level, leads to the lowest overall total cost. The functional components which the model considers include hot cells and their supporting facilities, transportation, cask handling facilities, transuranic waste handling facilities, and administrative facilities such as warehouses, security buildings, maintenance buildings, etc. The cost of each of these functional components is related either directly or indirectly to the various independent design parameters. Staffing by shift is reported into direct and indirect support labor. These staffing levels are in turn related to the waste type, waste throughput, etc. 2 refs., 11 figs., 3 tabs

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

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira da Cruz, Nuno, E-mail: nunocruz@ist.utl.pt; Ferreira, Sandra; Cabral, Marta; Simões, Pedro; Marques, Rui Cunha

    2014-02-15

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    Science.gov (United States)

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

    2014-02-01

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

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

    DEFF Research Database (Denmark)

    Götze, Ramona; Astrup, Thomas Fruergaard

    2013-01-01

    criteria of recycling processes. A lack of information in current waste characterisation practise on polymer resin composition, black coloured material content and the influence of surface adherent material on physico-chemical characteristics of plastic packaging waste were identified. These shortcomings...... were addressed by a resin type-based sorting analysis and a washing test for plastic packaging material from Danish household waste. Preliminary results show that, for a quarter of the hand sorted material, no resin type could be identified and that Polypropylene and Polyethylene terephthalate were...... the dominating resin types in plastic packaging. The suggested washing procedure caused a decrease of 70% of the ash content of the plastic material. The analysed metals and nutrients were reduced by up to 24%...

  16. 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

  17. Degradation modes of nickel-base alternate waste package overpack materials

    International Nuclear Information System (INIS)

    Pitman, S.G.

    1988-07-01

    The suitability of Ti Grade 12 for waste package overpacks has been questioned because of its observed susceptibility to crevice corrosion and hydrogen-assisted crack growth. For this reason, materials have been selected for evaluation as alternatives to Ti Grade 12 for use as waste package overpacks. These alternative materials, which are based on the nickel-chromium-molybdenum (Ni-Cr-Mo) alloy system, are Inconel 625, Hastelloy C-276, and Hastelloy C-22. The degradation modes of the Ni-base alternate materials have been examined at Pacific Northwest Laboratory to determine the suitability of these materials for waste package overpack applications in a salt repository. Degradation modes investigated included general corrosion, crevice corrosion, pitting, stress-corrosion cracking, and hydrogen embrittlement

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

    International Nuclear Information System (INIS)

    Schneider, Th.; Schieber, C.; Lavele, S.

    2006-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, Th.; Schieber, C.; Lavele, S.

    2006-12-15

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

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

    International Nuclear Information System (INIS)

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

    1989-09-01

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

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

    International Nuclear Information System (INIS)

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

    1986-04-01

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

  2. Nuclear waste management technical support in the developmnt of nuclear waste form criteria for the NRC. Task 5. National waste package program

    International Nuclear Information System (INIS)

    Davis, M.S.

    1982-02-01

    This report assesses the need for a centrally organized waste package effort and whether the present national program meets those needs. It is the conclusion of the BNL staff that while the DOE has in principle organized a national effort to develop high-integrity waste packages for geologic disposal of high level waste, the effort has not yet produced data to demonstrate that a waste package will comply with NRC's criteria. The BNL staff feels, however, that such a package is achievable either by development of high integrity components which by themselves could comply with 1000-year containment or by the development of new waste package designs that could comply with both the containment and the controlled release criteria in the 10CFR 60 performance objectives. In terms of waste forms, high-integrity components such as pyrolytic carbon coated waste and radioactive glass coated with non-radioactive glass offer higher potential than normal borosilicate waste glass. The existing container research program has yet to produce the data base on which to assess the potential of a container material to contain the waste for 1000 years. However, there may be the potential, based on Swedish calculations and work done on titanium in the DOE program, that Ti or its alloys may satisfy this criterion. Existing data on natural backfills will not be acceptable as the sole source for satisfying containment and the long-term release rate criteria. However, a synthetic zeolite system is an example of a backfill with a potential to satisfy both criteria. In this particular case, it is the BNL staff's opinion that existing technology and data for this system indicate that major development programs may not be required to qualify this material for licensing applications. The most likely means available for satisfying 10 CFR 60 with a single package component is through the performance of a discrete backfill

  3. Recycling potential of post-consumer plastic packaging waste in Finland.

    Science.gov (United States)

    Dahlbo, Helena; Poliakova, Valeria; Mylläri, Ville; Sahimaa, Olli; Anderson, Reetta

    2018-01-01

    Recycling of plastics is urged by the need for closing material loops to maintain our natural resources when striving towards circular economy, but also by the concern raced by observations of plastic scrap in oceans and lakes. Packaging industry is the sector using the largest share of plastics, hence packaging dominates in the plastic waste flow. The aim of this paper was to sum up the recycling potential of post-consumer plastic packaging waste in Finland. This potential was evaluated based on the quantity, composition and mechanical quality of the plastic packaging waste generated by consumers and collected as a source-separated fraction, within the mixed municipal solid waste (MSW) or within energy waste. Based on the assessment 86,000-117,000 tons (18 kg/person/a) of post-consumer plastic packaging waste was generated in Finland in 2014. The majority, 84% of the waste was in the mixed MSW flow in 2014. Due to the launching of new sorting facilities and separate collections for post-consumer plastic packaging in 2016, almost 40% of the post-consumer plastic packaging could become available for recycling. However, a 50% recycling rate for post-consumer plastic packaging (other than PET bottles) would be needed to increase the overall MSW recycling rate from the current 41% by around two percentage points. The share of monotype plastics in the overall MSW plastics fraction was 80%, hence by volume the recycling potential of MSW plastics is high. Polypropylene (PP) and low density polyethylene (LDPE) were the most common plastic types present in mixed MSW, followed by polyethylene terephthalate (PET), polystyrene (PS) and high density polyethylene (HDPE). If all the Finnish plastic packaging waste collected through the three collection types would be available for recycling, then 19,000-25,000 tons of recycled PP and 6000-8000 tons of recycled HDPE would be available on the local market. However, this assessment includes uncertainties due to performing the

  4. Waste package/engineered barrier system design concepts for the direct disposal of spent fuel in the potential United States' repository at Yucca Mountain, Nevada

    International Nuclear Information System (INIS)

    Stahl, D.; Harrison, D.J.

    1993-01-01

    The goal of the US Department of Energy's (DOE) Yucca Mountain Site Characterization Project (YMP) waste package development program is to design a waste package and associated engineered barrier system (EBS) that meets the applicable regulatory requirements for safe disposal of spent nuclear fuel and solidified high-level waste (HLW) in a geologic repository. Attainment of this goal relies on a multi-barrier approach, the unsaturated nature of the Yucca Mountain site, consideration of technical alternatives, and sufficient resolution of technical and regulatory uncertainties. To accomplish this, an iterative system engineering approach will be used. The NWPA of 1982 limits the content of the first US repository to 70,000 metric tons of heavy metal (MTHM). The DOE Mission Plan describes the implementation of the provisions of the NWPA for the waste management system. The Draft 1988 approach will involve selecting candidate designs, evaluating them against performance requirements, and then selecting one or two preferred designs for further detailed evaluation and final design. The reference design of the waste package described in the YMP Site Characterization Plan is a thin-walled, vertical borehole-emplaced waste package with an air gap between the package and the rock wall. The reference design appeared to meet the design requirement. However, the degree of uncertainty was large. This uncertainty led to considering several more-robust design concepts during the Advanced Conceptual Design phase of the program that include small, drift-emplaced packages and higher capacity, drift-emplaced packages, both partially and totally self-shielded. Metallic as well as ceramic materials are being considered

  5. Safety evaluation for packaging (onsite) for concrete-shielded RHTRU waste drum for the 327 postirradiation testing laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Adkins, H.E.

    1996-10-29

    This safety evaluation for packaging authorizes onsite transport of Type B quantities of radioactive material in the Concrete- Shielded Remote-Handled Transuranic Waste (RH TRU) Drum per WHC-CM-2-14, Hazardous Material Packaging and Shipping. The drum will be used for transport of 327 Building legacy waste from the 300 Area to the Transuranic Waste Storage and Assay Facility in the 200 West Area and on to a Solid Waste Storage Facility, also in the 200 Area.

  6. Gas cooled reactor decommissioning. Packaging of waste for disposal in the United Kingdom deep repository

    International Nuclear Information System (INIS)

    Barlow, S.V.; Wisbey, S.J.; Wood, P.

    1998-01-01

    United Kingdom Nirex Limited has been established to develop and operate a deep underground repository for the disposal of the UK's intermediate and certain low level radioactive waste. The UK has a significant Gas Cooled Reactor (GCR) programme, including both Magnox and AGR (Advanced Gas-cooled Reactor) capacity, amounting to 26 Magnox reactors, 15 AGR reactors as well as research and prototype reactor units such as the Windscale AGR and the Windscale Piles. Some of these units are already undergoing decommissioning and Nirex has estimated that some 15,000 m 3 (conditioned volume) will come forward for disposal from GCR decommissioning before 2060. This volume does not include final stage (Stage 3) decommissioning arisings from commercial reactors since the generating utilities in the UK are proposing to adopt a deferred safe store strategy for these units. Intermediate level wastes arising from GCR decommissioning needs to be packaged in a form suitable for on-site interim storage and eventual deep disposal in the planned repository. In the absence of Conditions for Acceptance for a repository in the UK, the dimensions, key features and minimum performance requirements for waste packages are defined in Waste Package Specifications. These form the basis for all assessments of the suitability of wastes for disposal, including GCR wastes. This paper will describe the nature and characteristics of GCR decommissioning wastes which are intended for disposal in a UK repository. The Nirex Waste Package Specifications and the key technical issues, which have been identified when considering GCR decommissioning waste against the performance requirements within the specifications, are discussed. (author)

  7. Mathematical models for diffusive mass transfer from waste package container with multiple perforations

    International Nuclear Information System (INIS)

    Lee, J.H.; Andrews, R.W.; Chambre, P.L.

    1996-01-01

    A robust engineered barrier system (EBS) is employed in the current design concept for the potential high-level nuclear waste repository at Yucca Mountain, Nevada, US. The primary component of the EBS is a multi-barrier waste package container. Simplifying the geometry of the cylindrical waste package container and the underlying invert into the equivalent spherical configuration, mathematical models are developed for steady-state and transient diffusive releases from the failed waste container with multiple perforations (or pit penetrations) at the boundary of the invert. Using the models the steady-state and transient diffusive release behaviors form the failed waste container are studied. The analyses show that the number of perforations, the size of perforation, the container wall thickness, the geometry of the waste container and invert, and the adsorption of radionuclide in the invert are the important parameters that control the diffusive release rate. It is emphasized that the failed (or perforated) waste package container can still perform as a potentially important barrier (or diffusion barrier) to radionuclide release

  8. Site identification presentation: Basalt Waste Isolation Project

    International Nuclear Information System (INIS)

    1979-11-01

    The final step in the site identification process for the Basalt Waste Isolation Project is described. The candidate sites are identified. The site identification methodology is presented. The general objectives which must be met in selecting the final site are listed. Considerations used in the screening process are also listed. Summary tables of the guidelines used are included

  9. Composition and activity variations in bulk gas of drum waste packages of Paks NPP

    International Nuclear Information System (INIS)

    Molnar, M.; Palcsu, L.; Svingor, E.; Szanto, Zs.; Futo, I.; Ormai, P.

    2001-01-01

    To obtain reliable estimates of the quantities and rates of the gas production a series of measurements was carried out in drum waste packages generated and temporarily stored at the site of Paks Nuclear Power Plant (Paks NPP). Ten drum waste packages were equipped with sampling valves for repeated sampling. Nine times between 04/02/2000 and 19/07/2001 qualitative gas component analyses of bulk gases of drums were executed. Gas samples were delivered to the laboratory of the ATOMKI for tritium and radiocarbon content measurements.(author)

  10. 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

  11. WESTINGHOUSE 17X17 MOX PWR ASSEMBLY - WASTE PACKAGE CRITICALITY ANALYSIS (SCPB: N/A)

    International Nuclear Information System (INIS)

    J.W. Davis

    1996-01-01

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to compare the criticality potential of Westinghouse 17 x 17 mixed oxide (MOX) PWR fuel with the Design Basis spent nuclear fuel (SNF) analyzed previously (Ref. 5.1, 5.2). The basis of comparison will be the conceptual design Multi-Purpose Canister (MPC) PWR waste package concepts. The objectives of this evaluation are to show that the criticality potential of the MOX fuel is equal to or lower than the DBF or, if necessary, indicate what additional measures are required to make it so

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-03-27

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

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

    Science.gov (United States)

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

    2014-02-01

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

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

    Science.gov (United States)

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

    2014-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Arantzazu eValdés

    2014-02-01

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

  16. Package

    Directory of Open Access Journals (Sweden)

    Arsić Zoran

    2013-01-01

    Full Text Available It is duty of the seller to pack the goods in a manner which assures their safe arrival and enables their handling in transit and at the place of destination. The problem of packing is relevant in two main respects. First of all the buyer is in certain circumstances entitled to refuse acceptance of the goods if they are not properly packed. Second, the package is relevant to calculation of price and freight based on weight. In the case of export trade, the package should conform to the legislation in the country of destination. The impact of package on environment is regulated by environment protection regulation of Republic if Serbia.

  17. Safety evaluation for packaging (onsite) depleted uranium waste boxes

    Energy Technology Data Exchange (ETDEWEB)

    McCormick, W.A.

    1997-08-27

    This safety evaluation for packaging (SEP) allows the one-time shipment of ten metal boxes and one wooden box containing depleted uranium material from the Fast Flux Test Facility to the burial grounds in the 200 West Area for disposal. This SEP provides the analyses and operational controls necessary to demonstrate that the shipment will be safe for the onsite worker and the public.

  18. Safety evaluation for packaging (onsite) depleted uranium waste boxes

    International Nuclear Information System (INIS)

    McCormick, W.A.

    1997-01-01

    This safety evaluation for packaging (SEP) allows the one-time shipment of ten metal boxes and one wooden box containing depleted uranium material from the Fast Flux Test Facility to the burial grounds in the 200 West Area for disposal. This SEP provides the analyses and operational controls necessary to demonstrate that the shipment will be safe for the onsite worker and the public

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-12-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-04-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-12-15

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

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

    International Nuclear Information System (INIS)

    H.W. Stockman; S. LeStrange

    2000-01-01

    The objective of this calculation is to provide estimates of the amount of fissile material flowing out of the waste package (source term) and the accumulation of fissile elements (U and Pu) in a crushed-tuff invert. These calculations provide input for the analysis of repository impacts of the Pu-ceramic waste forms. In particular, the source term results are used as input to the far-field accumulation calculation reported in Ref. 51, and the in-drift accumulation results are used as inputs for the criticality calculations reported in Ref. 2. The results are also summarized and interpreted in Ref. 52. The scope of this calculation is the waste package (WP) Viability Assessment (VA) design, which consists of an outer corrosion-allowance material (CAM) and an inner corrosion-resistant material (CRM). This design is used in this calculation in order to be consistent with earlier Pu-ceramic degradation calculations (Ref. 15). The impact of the new Enhanced Design Alternative-I1 (EDA-11) design on the results will be addressed in a subsequent report. The design of the invert (a leveling foundation, which creates a level surface of the drift floor and supports the WP mounting structure) is consistent with the EDA-I1 design. The invert will be composed of crushed stone and a steel support structure (Ref. 17). The scope of this calculation is also defined by the nominal degradation scenario, which involves the breach of the WP (Section 10.5.1.2, Ref. 48), followed by the influx of water. Water in the WP may, in time, gradually leach the fissile components and neutron absorbers out of the ceramic waste forms. Thus, the water in the WP may become laden with dissolved actinides (e.g., Pu and U), and may eventually overflow or leak from the WP. Once the water leaves the WP, it may encounter the invert, in which the actinides may reprecipitate. Several factors could induce reprecipitation; these factors include: the high surface area of the crushed stone, and the presence of

  3. In-Drift Accumulation of Fissile Material From Waste Packages Containing Plutonium Disposition Waste Form

    Energy Technology Data Exchange (ETDEWEB)

    H.W> Stockman; S. LeStrange

    2000-09-28

    The objective of this calculation is to provide estimates of the amount of fissile material flowing out of the waste package (source term) and the accumulation of fissile elements (U and Pu) in a crushed-tuff invert. These calculations provide input for the analysis of repository impacts of the Pu-ceramic waste forms. In particular, the source term results are used as input to the far-field accumulation calculation reported in Ref. 51, and the in-drift accumulation results are used as inputs for the criticality calculations reported in Ref. 2. The results are also summarized and interpreted in Ref. 52. The scope of this calculation is the waste package (WP) Viability Assessment (VA) design, which consists of an outer corrosion-allowance material (CAM) and an inner corrosion-resistant material (CRM). This design is used in this calculation in order to be consistent with earlier Pu-ceramic degradation calculations (Ref. 15). The impact of the new Enhanced Design Alternative-I1 (EDA-11) design on the results will be addressed in a subsequent report. The design of the invert (a leveling foundation, which creates a level surface of the drift floor and supports the WP mounting structure) is consistent with the EDA-I1 design. The invert will be composed of crushed stone and a steel support structure (Ref. 17). The scope of this calculation is also defined by the nominal degradation scenario, which involves the breach of the WP (Section 10.5.1.2, Ref. 48), followed by the influx of water. Water in the WP may, in time, gradually leach the fissile components and neutron absorbers out of the ceramic waste forms. Thus, the water in the WP may become laden with dissolved actinides (e.g., Pu and U), and may eventually overflow or leak from the WP. Once the water leaves the WP, it may encounter the invert, in which the actinides may reprecipitate. Several factors could induce reprecipitation; these factors include: the high surface area of the crushed stone, and the presence of

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

    International Nuclear Information System (INIS)

    S. Mastilovic

    1999-01-01

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

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

    International Nuclear Information System (INIS)

    Wurm, K.J.; Miller, N.E.

    1982-11-01

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

  6. 77 FR 17093 - Certain Food Waste Disposers and Components and Packaging Thereof: Notice of Receipt of Complaint...

    Science.gov (United States)

    2012-03-23

    ... INTERNATIONAL TRADE COMMISSION [DN 2886] Certain Food Waste Disposers and Components and Packaging.... International Trade Commission has received a complaint entitled Certain Food Waste Disposers and Components and Packaging Thereof, DN 2886; the Commission is soliciting comments on any public interest issues raised by...

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  8. NWTS program criteria for mined geologic disposal of nuclear waste: functional requirements and performance criteria for waste packages for solidified high-level waste and spent fuel

    International Nuclear Information System (INIS)

    1982-07-01

    The Department of Energy (DOE) has primary federal responsibility for the development and implementation of safe and environmentally acceptable nuclear waste disposal methods. Currently, the principal emphasis in the program is on emplacement of nuclear wastes in mined geologic repositories well beneath the earth's surface. A brief description of the mined geologic disposal system is provided. The National Waste Terminal Storage (NWTS) program was established under DOE's predecessor, the Energy Research and Development Administration, to provide facilities for the mined geologic disposal of radioactive wastes. The NWTS program includes both the development and the implementation of the technology necessary for designing, constructing, licensing, and operating repositories. The program does not include the management of processing radioactive wastes or of transporting the wastes to repositories. The NWTS-33 series, of which this document is a part, provides guidance for the NWTS program in the development and implementation of licensed mined geologic disposal systems for solidified high-level and transuranic (TRU) wastes. This document presents the functional requirements and performance criteria for waste packages for solidified high-level waste and spent fuel. A separate document to be developed, NWTS-33(4b), will present the requirements and criteria for waste packages for TRU wastes. The hierarchy and application of these requirements and criteria are discussed in Section 2.2

  9. Project plans for transuranic waste at small quantity sites in the Department of Energy comples-10522

    International Nuclear Information System (INIS)

    Mctaggart, Jerri Lynne; Lott, Sheila; Gadbury, Casey

    2009-01-01

    Los Alamos National Laboratory, Carlsbad Office (LANL-CO), has been tasked to write Project Plans for all of the Small Quantity Sites (SQS) with defense related Transuranic (TRU) waste in the Department of Energy (DOE) complex. Transuranic Work-Off Plans were precursors to the Project Plans. LANL-CO prepared a Work-Off Plan for each small quantity site. The Work-Off Plan that identified issues, drivers, schedules, and inventory. Eight sites have been chosen to deinventory their legacy TRU waste; Bettis Atomic Power Laboratory, General Electric-Vallecitos Nuclear Center, Lawrence Berkeley National Laboratory, Lawrence Livermore National Laboratory-Area 300, Nevada Test Site, Nuclear Radiation Development, Sandia National Laboratory, and the Separations Process Research Unit. Each plan was written for contact and/or remote handled waste if present at the site. These project plans will assist the small quantity sites to ship legacy TRU waste offsite and de-inventory the site of legacy TRU waste. The DOE is working very diligently to reduce the nuclear foot print in the United States. Each of the eight SQSs will be de-inventoried of legacy TRU waste during a campaign that ends September 2011. The small quantity sites have a fraction of the waste that large quantity sites possess. During this campaign, the small quantity sites will package all of the legacy TRU waste and ship to Idaho or directly to the Waste Isolation Pilot Plant (WIPP) in Carlsbad, New Mexico. The sites will then be removed from the Transuranic Waste Inventory if they are de-inventoried of all waste. Each Project Plan includes the respective site inventory report, schedules, resources, drivers and any issues. These project plans have been written by the difficult waste team and will be approved by each site. Team members have been assigned to each site to write site specific project plans. Once the project plans have been written, the difficult team members will visit the sites to ensure nothing has

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-09-15

    Background. On the basis of an increased focus on emissions of greenhouse gases in general, Waste Norway wanted to prepare a climate accounting for waste management in Norway. Oestfoldforskning was engaged to undertake the project. The aim of the project has been to develop a model for the calculation of net greenhouse gas emissions from different waste types of waste glass containers, metal containers, paper, cardboard, plastic, wet organic waste, wood waste and residual waste. The model is based on life cycle methodology and is used to calculate the net greenhouse gas emissions per kg of waste for the various waste management options and waste types, as well as to calculate the net greenhouse gas emissions for waste management for including waste types and quantities of 2006. There is an emphasis on developing a model so that municipalities / waste companies or regions can develop their own climate accounting for waste management in their region, based on site-specific conditions associated with types and amounts of waste, transport distances, type of treatment, exploitation and use of waste generated energy etc. The model can also be used as the basis for the preparation of useful documentation as the basis for information about waste systems utility in general, and as a basis for strategic reviews for Waste Norway and the waste sector in particular. Conclusions: The main conclusions from the project can be summarized as follows: 1. The results of the study clearly shows that to consider only one environmental indicator is too narrow approach to form the basis for decision making for selection of waste management solutions. 2. Net greenhouse gas emissions for waste management varies greatly, both between the different types of waste and treatment methods which are reviewed. The main results of the ranking of management methods in relation to the net greenhouse effect associated with the waste types and treatment methods are as follows: Recycling of materials

  11. FY-1981 project status for the Transuranic Waste Treatment Facility

    International Nuclear Information System (INIS)

    Benedetti, R.L.; Tait, T.D.

    1981-11-01

    The primary objective of the Transuranic Waste Treatment Facility (TWTF) Project is to provide a facility to process low-level transuranic waste stored at the Idaho National Engineering Laboratory (INEL) into a form acceptable for disposal at the Waste Isolation Pilot Plant. This report provides brief summary descriptions of the project objectives and background, project status through FY-1981, planned activities for FY-1982, and the EG and G TWTF Project office position on processing INEL transuranic waste

  12. Design of a nuclear-waste package for emplacement in tuff

    International Nuclear Information System (INIS)

    O'Neal, W.C.; Rothman, A.J.; Gregg, D.W.; Hockman, J.N.; Revelli, M.A.; Russell, E.W.; Schornhorst, J.R.

    1983-01-01

    Design, modeling, and testing activities are under way at LLNL in the development of high level nuclear waste package designs. We discuss the geological characteristics affecting design, the 10CFR60 design requirements, conceptual designs, metals for containment barriers, economic analysis, thermal modeling, and performance modeling

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

    International Nuclear Information System (INIS)

    Bor Yunchang, Jeffrey; Chien, Y.-L.; Hsu, Esher

    2004-01-01

    This paper presents a new market-incentive (MI) system to recycle waste-packaging containers in Taiwan. Since most used packaging containers have no or insufficient market value, the government imposes a combined product charge and subsidy policy to provide enough economic incentive for recycling various kinds of packaging containers, such as iron, aluminum, paper, glass and plastic. Empirical results show that the new MI approach has stimulated and established the recycling market for waste-packaging containers. The new recycling system has provided 18,356 employment opportunities and generated NT$ 6.97 billion in real-production value and NT$ 3.18 billion in real GDP during the 1998 survey year. Cost-effectiveness analysis constitutes the theoretical foundation of the new scheme, whereas data used to compute empirical product charge are from two sources: marketing surveys of internal conventional costs of solid-waste collection, disposal and recycling in Taiwan, and benefit transfer of external environmental costs in the United States. The new recycling policy designed by the authors provides a reasonable solution for solid-waste management in a country with limited land resources such as Taiwan

  14. 77 FR 50716 - Certain Food Waste Disposers and Components and Packaging Thereof; Notice of Commission...

    Science.gov (United States)

    2012-08-22

    ... INTERNATIONAL TRADE COMMISSION [Investigation No. 337-TA-838] Certain Food Waste Disposers and Components and Packaging Thereof; Notice of Commission Determination Not to Review an Initial Determination Granting Complainant's Motions To Amend the Notice of Investigation and Complaint AGENCY: U.S...

  15. Protecting the Investment: Guidance on the Storage of Packaged Wastes in the UK

    International Nuclear Information System (INIS)

    Naish, Chris; Skelton, Paul; Wisbey, Simon

    2016-01-01

    This presentation will cover: • Introduction to the UK guidance on interim storage; • Waste stores in the UK and the Store Operations Forum; • Example Approach 1 – Operational limits and conditions; • Example Approach 2 – Monitoring the evolution of package performance; • IAEA Independent peer review

  16. Impact of rinsing in pesticide packaging waste management: Economic and environmental benefits

    Directory of Open Access Journals (Sweden)

    Marčeta Una

    2015-01-01

    Full Text Available Pesticides have become dailiness due to inevitable application of these preparations in agricultural activities, with the consequence of generation of large amounts of waste packaging. Impact on the environment and expenses of management of packaging waste can be minimized if the packaging is immediately rinsed after the application of devices and if identified as non-hazardous. Besides, financial losses may be reduced by maximum utilization of the preparation. Considering these two financial aspects this work shows evaluation of quantitative losses of preparations if the triple rising method is not applied. The research was conducted in two phases. Phase I included the examination of the impact of different formulations of the same volume on quantitative and financial losses. Based on the results of the first phase of the research, it was noted that the SC formulation is the most interesting to study because this type of formulation has the highest percentage of residue, as well as the fact that the highest annual consumption is noted percisely in this preparation group. This paper presents the results which indicate the impact of packaging volume of SC formulation (ALVERDE 240 SC, INTERMEZZO and ANTRE PLUS on percentage of preparation residue in packaging if there was no rinsing. The results have shown that the quantitative loss is inversely proportional to the volume of packaging, while financial losses do not only depend on the percentage of residue but also on price and quantity of utilization of preparations.

  17. Role of waste packages in the safety of a high level waste repository in a deep geological formation

    International Nuclear Information System (INIS)

    Bretheau, F.; Lewi, J.

    1990-06-01

    The safety of a radioactive waste disposal facility lays on the three following barriers placed between the radioactive materials and the biosphere: the waste package; the engineered barriers; the geological barrier. The function assigned to each of these barriers in the performance assessment is an option taken by the organization responsible for waste disposal management (ANDRA in France), which must show that: expected performances of each barrier (confinement ability, life-time, etc.) are at least equal to those required to fulfill the assigned function; radiation protection requirements are met in all situations considered as credible, whether they be the normal situation or random event situations. The French waste management strategy is based upon two types of disposal depending on the nature and activity of waste packages: - surface disposal intended for low and medium level wastes having half-lives of about 30 years or less and alpha activity less than 3.7 MBq/kg (0.1 Ci/t), for individual packages and less than 0.37 MBq/kg (0.01 Ci/t) in the average. Deep geological disposal intended for TRU and high level wastes. The conditions of acceptance of packages in a surface disposal site are subject to the two fundamental safety rules no. I.2 and III.2.e. The present paper is only dealing with deep geological disposal. For deep geological repositories, three stages are involved: stage preceding definitive disposal (intermediate storage, transportation, handling, setting up in the disposal cavities); stage subsequent to definitive sealing of the disposal cavities but prior to the end of operation of the repository; stage subsequent to closure of the repository. The role of the geological barrier has been determined as the essential part of long term radioactivity confinement, by a working group, set up by the French safety authorities. Essential technical criteria relating to the choice of a site so defined by this group, are the following: very low permeability

  18. Shielding Calculations on Waste Packages – The Limits and Possibilities of different Calculation Methods by the example of homogeneous and inhomogeneous Waste Packages

    Directory of Open Access Journals (Sweden)

    Adams Mike

    2017-01-01

    Full Text Available For nuclear waste packages the expected dose rates and nuclide inventory are beforehand calculated. Depending on the package of the nuclear waste deterministic programs like MicroShield® provide a range of results for each type of packaging. Stochastic programs like “Monte-Carlo N-Particle Transport Code System” (MCNP® on the other hand provide reliable results for complex geometries. However this type of program requires a fully trained operator and calculations are time consuming. The problem here is to choose an appropriate program for a specific geometry. Therefore we compared the results of deterministic programs like MicroShield® and stochastic programs like MCNP®. These comparisons enable us to make a statement about the applicability of the various programs for chosen types of containers. As a conclusion we found that for thin-walled geometries deterministic programs like MicroShield® are well suited to calculate the dose rate. For cylindrical containers with inner shielding however, deterministic programs hit their limits. Furthermore we investigate the effect of an inhomogeneous material and activity distribution on the results. The calculations are still ongoing. Results will be presented in the final abstract.

  19. Shielding Calculations on Waste Packages - The Limits and Possibilities of different Calculation Methods by the example of homogeneous and inhomogeneous Waste Packages

    Science.gov (United States)

    Adams, Mike; Smalian, Silva

    2017-09-01

    For nuclear waste packages the expected dose rates and nuclide inventory are beforehand calculated. Depending on the package of the nuclear waste deterministic programs like MicroShield® provide a range of results for each type of packaging. Stochastic programs like "Monte-Carlo N-Particle Transport Code System" (MCNP®) on the other hand provide reliable results for complex geometries. However this type of program requires a fully trained operator and calculations are time consuming. The problem here is to choose an appropriate program for a specific geometry. Therefore we compared the results of deterministic programs like MicroShield® and stochastic programs like MCNP®. These comparisons enable us to make a statement about the applicability of the various programs for chosen types of containers. As a conclusion we found that for thin-walled geometries deterministic programs like MicroShield® are well suited to calculate the dose rate. For cylindrical containers with inner shielding however, deterministic programs hit their limits. Furthermore we investigate the effect of an inhomogeneous material and activity distribution on the results. The calculations are still ongoing. Results will be presented in the final abstract.

  20. Tank waste remediation system characterization project quality policies. Revision 1

    International Nuclear Information System (INIS)

    Trimble, D.J.

    1995-01-01

    These Quality Policies (QPs) describe the Quality Management System of the Tank Waste Characterization Project (hereafter referred to as the Characterization Project), Tank Waste Remediation System (TWRS), Westinghouse Hanford Company (WHC). The Quality Policies and quality requirements described herein are binding on all Characterization Project organizations. To achieve quality, the Characterization Project management team shall implement this Characterization Project Quality Management System

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

    International Nuclear Information System (INIS)

    Baekelandt, L.; Libon, H.; Vandorpe, M.; Lafontaine, I.

    1989-01-01

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

  2. An approach for sampling solid heterogeneous waste at the Hanford Site waste receiving and processing and solid waste projects

    International Nuclear Information System (INIS)

    Sexton, R.A.

    1993-03-01

    This paper addresses the problem of obtaining meaningful data from samples of solid heterogeneous waste while maintaining sample rates as low as practical. The Waste Receiving and Processing Facility, Module 1, at the Hanford Site in south-central Washington State will process mostly heterogeneous solid wastes. The presence of hazardous materials is documented for some packages and unknown for others. Waste characterization is needed to segregate the waste, meet waste acceptance and shipping requirements, and meet facility permitting requirements. Sampling and analysis are expensive, and no amount of sampling will produce absolute certainty of waste contents. A sampling strategy is proposed that provides acceptable confidence with achievable sampling rates

  3. STRUCTURAL CALCULATION OF AN EMPLACEMENT PALLET STATICALLY LOADED BY A WASTE PACKAGE

    International Nuclear Information System (INIS)

    S. Mastilovic

    2000-01-01

    The purpose of this calculation is to determine the structural response of the emplacement pallet (EP) subjected to static load from the mounted waste package (WP). The scope of this document is limited to reporting the calculation results in terms of stress intensity magnitudes. This calculation is associated with the waste emplacement systems design; calculations are 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 solutions are performed by using the commercially available ANSYS Version (V) 5.4 finite element code. The results of these calculations are provided in terms of maximum stress intensity magnitudes

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

    International Nuclear Information System (INIS)

    Chesnut, D.A.

    1991-09-01

    Waste packages for a US nuclear waste repository are required to provide reasonable assurance of maintaining substantially complete containment of radionuclides for 300 to 1000 years after closure. The waiting time to failure for complex failure processes affecting engineered or manufactured systems is often found to be an exponentially-distributed random variable. Assuming that this simple distribution can be used to describe the behavior of a hypothetical single barrier waste package, calculations presented in this paper show that the mean time to failure (the only parameter needed to completely specify an exponential distribution) would have to be more than 10 7 years in order to provide reasonable assurance of meeting this requirement. With two independent barriers, each would need to have a mean time to failure of only 10 5 years to provide the same reliability. Other examples illustrate how multiple barriers can provide a strategy for not only achieving but demonstrating regulatory compliance

  5. Cementitious materials for radioactive waste management within IAEA coordinated research project - 59021

    International Nuclear Information System (INIS)

    Drace, Zoran; Ojovan, Michael I.

    2012-01-01

    The IAEA Coordinated Research Project (CRP) on cementitious materials for radioactive waste management was launched in 2007 [1, 2]. The objective of CRP was to investigate the behaviour and performance of cementitious materials used in radioactive waste management system with various purposes and included waste packages, waste-forms and backfills as well as investigation of interactions and interdependencies of these individual elements during long term storage and disposal. The specific research topics considered were: (i) cementitious materials for radioactive waste packaging: including radioactive waste immobilization into a solid waste form, (ii) waste backfilling and containers; (iii) emerging and alternative cementitious systems; (iv) physical-chemical processes occurring during the hydration and ageing of cement matrices and their influence on the cement matrix quality; (v) methods of production of cementitious materials for: immobilization into wasteform, backfills and containers; (vi) conditions envisaged in the disposal environment for packages (physical and chemical conditions, temperature variations, groundwater, radiation fields); (vii) testing and non-destructive monitoring techniques for quality assurance of cementitious materials; (viii) waste acceptance criteria for waste packages, waste forms and backfills; transport, long term storage and disposal requirements;and finally (ix) modelling or simulation of long term behaviours of cementations materials used for packaging, waste immobilization and backfilling, especially in the post-closure phase. The CRP has gathered overall 26 research organizations from 22 Member States aiming to share their research and practices on the use of cementitious materials [2]. The main research outcomes of the CRP were summarized in a summary report currently under preparation to be published by IAEA. The generic topical sections covered by report are: a) conventional cementitious systems; b) novel cementitious

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

    International Nuclear Information System (INIS)

    Davis, D.; Lund, J.S.; Meredith, P.; Walker, P.; Wells, D.A.; Jowett, J.; Kinsella, K.

    1989-01-01

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

  7. Petrologic and geochemical characterization of the Bullfrog Member of the Crater Flat Tuff: outcrop samples used in waste package experiments

    International Nuclear Information System (INIS)

    Knauss, K.G.

    1983-09-01

    In support of the Waste Package Task within the Nevada Nuclear Waste Storage Investigation (NNWSI), experiments on hydrothermal rock/water interaction, corrosion, thermomechanics, and geochemical modeling calculations are being conducted. All of these activities require characterization of the initial bulk composition, mineralogy, and individual phase geochemistry of the potential repository host rock. This report summarizes the characterization done on samples of the Bullfrog Member of the Crater Flat Tuff (Tcfb) used for Waste Package experimental programs. 11 references, 17 figures, 3 tables

  8. Materials for Waste Incinerators and Biomass Plants

    DEFF Research Database (Denmark)

    Rademakers, P.; Grossmann, G.; Karlsson, A.

    1998-01-01

    This paper reviews the projects of the sub-package on waste incineration and biomass firing carried out within COST 501 Round III, Work Package 13.......This paper reviews the projects of the sub-package on waste incineration and biomass firing carried out within COST 501 Round III, Work Package 13....

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

    International Nuclear Information System (INIS)

    DI Kaplan; RJ Serne

    2000-01-01

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

  10. Effect of ionizing radiation on the waste package environment

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-05-01

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

  11. Effect of ionizing radiation on the waste package environment

    International Nuclear Information System (INIS)

    Reed, D.T.; Van Konynenburg, R.A.

    1991-01-01

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

  12. The role of multiple barriers in assuring waste package reliability

    International Nuclear Information System (INIS)

    Bradford, R.M.

    1993-08-01

    Yucca Mountain in southwestern Nevada is being studied as a potential repository site for the permanent storage of high-level nuclear waste. Regulators have set performance standards that the potential repository must meet in order to obtain regulatory approval. Nuclear Regulatory Commission (NRC) regulations state that containment of radioactivity must be ''substantially complete'' for the first 1000 years after closure of the facility. Thereafter, the acceptable annual limit on releases is 1/100,000 of each radionuclide remaining in the inventory after 1000 years. To demonstrate that the potential facility is in compliance with the regulations, it is necessary to obtain some understanding of the probability distribution of the cumulative quantity of releases by certain time points. This paper will discuss the probability distribution of waste container lifetimes and how the understanding of this distribution will play a role in finding the distribution of the release quantities over time. It will be shown that, for reasonable assumptions about the process of barrier failure, the reliability of a multiple-barrier container can be achieved and demonstrated much more readily than a container consisting of a single barrier. The discussion will focus primarily on the requirement of substantially complete containment for the first 1000 years

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

    Science.gov (United States)

    Bacelos, M S; Freire, J T

    2012-01-01

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

  14. Application of geometry correction factors for low-level waste package dose measurements. Revision 1

    International Nuclear Information System (INIS)

    Chandler, M.C.; Parish, B.

    1995-01-01

    Plans are to determine the Cs-137 content of low-level waste packages generated in High-Level Waste by measuring the radiation level at a specified distance from the package with a hand-held radiation instrument. The measurement taken at this specified distance, either 3 or 5 feet, is called the far-field measurement. This report documents a method for adjusting the gamma exposure rate (mR/hr) reading used in dose-to-curie determinations when the far-field measurement equals the background reading. This adjustment is necessary to reduce the conservatism resulting from using a minimum detection limit exposure rate for the dose-to-curie determination for the far-field measurement position. To accomplish this adjustment, the near-field (5 cm) measurement is multiplied by a geometry correction factor to obtain an estimate of the far field exposure rate (which is below instrument sensitivity). This estimate of the far field exposure rate is used to estimate the Cs-137 curie content of the package. This report establishes the geometry correction factors for the dose-to-curie determination when the far-field gamma exposure measurement equals the background reading. This report also provides a means of demonstrating compliance to 1S Manual requirements for exposure rate readings at different locations from waste packages while specifying only two measurement positions. This demonstration of compliance is necessary to minimize the number of locations exposure rate measurements that are required, i.e., ALARA

  15. Options for reducing food waste by quality-controlled logistics using intelligent packaging along the supply chain.

    Science.gov (United States)

    Heising, Jenneke K; Claassen, G D H; Dekker, Matthijs

    2017-10-01

    Optimising supply chain management can help to reduce food waste. This paper describes how intelligent packaging can be used to reduce food waste when used in supply chain management based on quality-controlled logistics (QCL). Intelligent packaging senses compounds in the package that correlate with the critical quality attribute of a food product. The information on the quality of each individual packaged food item that is provided by the intelligent packaging can be used for QCL. In a conceptual approach it is explained that monitoring food quality by intelligent packaging sensors makes it possible to obtain information about the variation in the quality of foods and to use a dynamic expiration date (IP-DED) on a food package. The conceptual approach is supported by quantitative data from simulations on the effect of using the information of intelligent packaging in supply chain management with the goal to reduce food waste. This simulation shows that by using the information on the quality of products that is provided by intelligent packaging, QCL can substantially reduce food waste. When QCL is combined with dynamic pricing based on the predicted expiry dates, a further waste reduction is envisaged.

  16. Nuclear-waste-package program for high-level isolation in Nevada tuff

    International Nuclear Information System (INIS)

    Rothman, A.J.

    1982-01-01

    The objective of the waste package program is to insure that a package is designed suitable for a repository in tuff that meets performance requirements of the NRC. In brief, the current (draft) regulation requires that the radionuclides be contained in the engineered system for 1000 years, and that, thereafter, no more than one part in 10 5 of the nuclides per year leave the boundary of the system. Studies completed as of this writing are thermal modeling of waste packages in a tuff repository and analysis of sodium bentonite as a potential backfill material. Both studies will be presented. Thermal calculations coupled with analysis of the geochemical literature on bentonite indicate that extensive chemical and physical alteration of bentonite would result at the high power densities proposed (ca. 2 kW/package and an area density of 25 W/m 2 ), in part due to compacted bentonite's relatively low thermal conductivity when dehydrated (approx. 0.6 +- 0.2 W/m 0 C). Because our groundwater contains K + , an upper hydrothermal temperature limit appears to be 120 to 150 0 C. At much lower power densities (less than 1 kW per package and an areal density of 12 W/m 2 ), bentonite may be suitable

  17. Life cycle environmental impacts of different construction wood waste and wood packaging waste processing methods

    OpenAIRE

    Manninen, Kaisa; Judl, Jáchym; Myllymaa, Tuuli

    2016-01-01

    This study compared the life cycle environmental impacts of different wood waste processing methods in three impact categories: climate impact, acidification impacts and eutrophication impacts. The wood waste recovery methods examined were the use of wood waste in terrace boards made out of wood composite which replace impregnated terrace boards, incineration of wood waste in a multi-fuel boiler instead of peat and the use of wood waste in the production of particleboard in either Finland or ...

  18. Application of digital radiography for the non-destructive characterization of radioactive waste packages

    International Nuclear Information System (INIS)

    Lierse, C.; Goebel, H.; Kaciniel, E.; Buecherl, T.; Krebs, K.

    1995-01-01

    Digital radiography (DR) using gamma-rays is a powerful tool for the non-destructive determination of various parameters which are relevant within the quality control procedure of radioactive waste packages prior to an interim storage or a final disposal. DR provides information about the waste form and the extent of filling in a typical container. It can identify internal structures and defects, gives their geometric dimensions and helps to detect non-declared inner containers, shielding materials etc. From a digital radiographic image the waste matrix homogeneity may be determined and mean attenuation coefficients as well as density values for selected regions of interest can be calculated. This data provides the basis for an appropriate attenuation correction of gamma emission measurements (gamma scanning) and makes a reliable quantification of gamma emitters in waste containers possible. Information from DR measurements are also used for the selection of interesting height positions of the object which are subsequently inspected in more detail by other non-destructive methods, e. g. by transmission computerized tomography (TCT). The present paper gives important technical specifications of an integrated tomography system (ITS) which is used to perform digital radiography as well as transmission/emission computerized tomography (TCT/ECT) on radioactive waste packages. It describes the DR mode and some of its main applications and shows typical examples of radiographs of real radioactive waste drums

  19. Technical solutions for waste treatment in the Belene project

    International Nuclear Information System (INIS)

    Büttner, K.; Eichhorn, H.

    2011-01-01

    Outline: In June 2010 NUKEM Technologies GmbH was awarded a contract from ATOMSTROYEXPORT JSC to perform the complete work package related to designing and completion of the equipment for treatment of radioactive waste on the turn-key basis for Belene NPP. Technical Solutions: Waste Streams and Technologies at UKC and UKS; Concentration Plant; Thermal Treatment of Resins Sorting Facility; Biological Waste Water Treatment; Conditioning – Cementation • Sorting of Radwaste; Plasma Facility; Grouting; Filter Press; Monitoring and Tracking

  20. Thermal testing of packages for transport of radioactive wastes

    International Nuclear Information System (INIS)

    Koski, J.A.

    1994-01-01

    Shipping containers for radioactive materials must be shown capable of surviving tests specified by regulations such as Title 10, Code of Federal Regulations, Part 71 (called 10CFR71 in this paper) within the United States. Equivalent regulations hold for other countries such as Safety Series 6 issued by the International Atomic Energy Agency. The containers must be shown to be capable of surviving, in order, drop tests, puncture tests, and thermal tests. Immersion testing in water is also required, but must be demonstrated for undamaged packages. The thermal test is intended to simulate a 30 minute exposure to a fully engulfing pool fire that could occur if a transport accident involved the spill of large quantities of hydrocarbon fuels. Various qualification methods ranging from pure analysis to actual pool fire tests have been used to prove regulatory compliance. The purpose of this paper is to consider the alternatives for thermal testing, point out the strengths and weaknesses of each approach, and to provide the designer with the information necessary to make informed decisions on the proper test program for the particular shipping container under consideration. While thermal analysis is an alternative to physical testing, actual testing is often emphasized by regulators, and this report concentrates on these testing alternatives

  1. Development of a working set of waste package performance criteria for deepsea disposal of low-level radioactive waste. Final report

    International Nuclear Information System (INIS)

    Columbo, P.; Fuhrmann, M.; Neilson, R.M. Jr; Sailor, V.L.

    1982-11-01

    The United States ocean dumping regulations developed pursuant to PL92-532, the Marine Protection, Research, and Sanctuaries Act of 1972, as amended, provide for a general policy of isolation and containment of low-level radioactive waste after disposal into the ocean. In order to determine whether any particular waste packaging system is adequate to meet this general requirement, it is necessary to establish a set of performance criteria against which to evaluate a particular packaging system. These performance criteria must present requirements for the behavior of the waste in combination with its immobilization agent and outer container in a deepsea environment. This report presents a working set of waste package performance criteria, and includes a glossary of terms, characteristics of low-level radioactive waste, radioisotopes of importance in low-level radioactive waste, and a summary of domestic and international regulations which control the ocean disposal of these wastes

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

    International Nuclear Information System (INIS)

    Brown, G.S.; Cashwell, J.W.; Apple, M.L.

    1991-01-01

    Shipments of radioactive material (RAM) constitute but a small fraction of the total hazardous materials shipped in the United States each year. Public perception, however, of the potential consequences of a release from a transportation package containing RAM has resulted in significant regulation of transport operations, both to ensure the integrity of a package in accident conditions and to place operational constraints on the shipper. Much of this attention has focused on shipments of spent nuclear fuel and high level wastes which, although comprising a very small number of total shipments, constitute a majority of the total curies transported on an annual basis. This report discusses the shipment of these highly radioactive materials

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

    International Nuclear Information System (INIS)

    Jones, T.E.; Coles, D.G.; Britton, R.C.; Burnell, J.R.

    1986-11-01

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

  4. Application of fluidization to separate packaging waste plastics.

    Science.gov (United States)

    Carvalho, M Teresa; Ferreira, Célia; Portela, Antía; Santos, João Tiago

    2009-03-01

    The objective of the experimental work described in this paper is the study of the separation of PS (polystyrene) from PET (polyethylene terephthalate) and PVC (polyvinyl chloride) from drop-off points using a fluidized bed separator. This is a low-cost process commonly used in the hydro-classification of mineral ores. Firstly, experimental tests were carried out with artificial granulated samples with different grain sizes, types and sources of plastic ("separability tests"). The particle settling velocities were determined under different operating conditions. Then, based on the results, the laboratory tests continued with real mixtures of waste plastics ("separation tests") and the efficiency of the process was evaluated. From a PET-rich mixture, a concentrate of PS with a 75% grade in PS was produced while the underflow was quite clear from PS (grade less than 0.5% in PS).

  5. Candidate container materials for Yucca Mountain waste package designs

    International Nuclear Information System (INIS)

    McCright, R.D.; Halsey, W.G.; Gdowski, G.E.; Clarke, W.L.

    1991-09-01

    Materials considered as candidates for fabricating nuclear waste containers are reviewed in the context of the Conceptual Design phase of a potential repository located at Yucca Mountain. A selection criteria has been written for evaluation of candidate materials for the next phase -- Advanced Conceptual Design. The selection criteria is based on the conceptual design of a thin-walled container fabricated from a single metal or alloy; the criteria consider the performance requirements on the container and the service environment in which the containers will be emplaced. A long list of candidate materials is evaluated against the criteria, and a short list of materials is proposed for advanced characterization in the next design phase

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

    International Nuclear Information System (INIS)

    Odoj, R.

    1991-01-01

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

  7. Solid Waste Projection Model: Database User's Guide

    International Nuclear Information System (INIS)

    Blackburn, C.L.

    1993-10-01

    The Solid Waste Projection Model (SWPM) system is an analytical tool developed by Pacific Northwest Laboratory (PNL) for Westinghouse Hanford Company (WHC) specifically to address Hanford solid waste management issues. This document is one of a set of documents supporting the SWPM system and providing instructions in the use and maintenance of SWPM components. This manual contains instructions for using Version 1.4 of the SWPM database: system requirements and preparation, entering and maintaining data, and performing routine database functions. This document supports only those operations which are specific to SWPM database menus and functions and does not Provide instruction in the use of Paradox, the database management system in which the SWPM database is established

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

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

    International Nuclear Information System (INIS)

    French, David M.; Hayes, Timothy A.; Pope, Howard L.; Enriquez, Alejandro E.; Carson, Peter H.

    2013-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-01-01

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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  12. HISPANIC ENVIRONMENTAL AND WASTE MANAGEMENT OUTREACH PROJECT

    Energy Technology Data Exchange (ETDEWEB)

    Sebastian Puente

    1998-07-25

    The Department of Energy Office of Environmental Management (DOE-EM) in cooperation with the Self Reliance Foundation (SRF) is conducting the Hispanic Environmental and Waste Management Outreach Project (HEWMO) to increase science and environmental literacy, specifically that related to nuclear engineering and waste management in the nuclear industry, among the US Hispanic population. The project will encourage Hispanic youth and young adults to pursue careers through the regular presentation of Spanish-speaking scientists and engineers and other role models, as well as career information on nationally broadcast radio programs reaching youth and parents. This project will encourage making science, mathematics, and technology a conscious part of the everyday life experiences of Hispanic youth and families. The SRF in collaboration with the Hispanic Radio Network (HRN) produces and broadcasts radio programs to address the topics and meet the objectives as outlined in the Environmental Literacy Plan and DOE-EM Communications Plan in this document. The SRF has in place a toll-free ''800'' number Information and Resource Referral (I and RR) service that national radio program listeners can call to obtain information and resource referrals as well as give their reactions to the radio programs that will air. HRN uses this feature to put listeners in touch with local organizations and resources that can provide them with further information and assistance on the related program topics.

  13. HISPANIC ENVIRONMENTAL AND WASTE MANAGEMENT OUTREACH PROJECT

    International Nuclear Information System (INIS)

    Sebastian Puente

    1998-01-01

    The Department of Energy Office of Environmental Management (DOE-EM) in cooperation with the Self Reliance Foundation (SRF) is conducting the Hispanic Environmental and Waste Management Outreach Project (HEWMO) to increase science and environmental literacy, specifically that related to nuclear engineering and waste management in the nuclear industry, among the US Hispanic population. The project will encourage Hispanic youth and young adults to pursue careers through the regular presentation of Spanish-speaking scientists and engineers and other role models, as well as career information on nationally broadcast radio programs reaching youth and parents. This project will encourage making science, mathematics, and technology a conscious part of the everyday life experiences of Hispanic youth and families. The SRF in collaboration with the Hispanic Radio Network (HRN) produces and broadcasts radio programs to address the topics and meet the objectives as outlined in the Environmental Literacy Plan and DOE-EM Communications Plan in this document. The SRF has in place a toll-free ''800'' number Information and Resource Referral (I and RR) service that national radio program listeners can call to obtain information and resource referrals as well as give their reactions to the radio programs that will air. HRN uses this feature to put listeners in touch with local organizations and resources that can provide them with further information and assistance on the related program topics

  14. Gravity packaging final waste recovery based on gravity separation and chemical imaging control.

    Science.gov (United States)

    Bonifazi, Giuseppe; Serranti, Silvia; Potenza, Fabio; Luciani, Valentina; Di Maio, Francesco

    2017-02-01

    Plastic polymers are characterized by a high calorific value. Post-consumer plastic waste can be thus considered, in many cases, as a typical secondary solid fuels according to the European Commission directive on End of Waste (EoW). In Europe the practice of incineration is considered one of the solutions for waste disposal waste, for energy recovery and, as a consequence, for the reduction of waste sent to landfill. A full characterization of these products represents the first step to profitably and correctly utilize them. Several techniques have been investigated in this paper in order to separate and characterize post-consumer plastic packaging waste fulfilling the previous goals, that is: gravity separation (i.e. Reflux Classifier), FT-IR spectroscopy, NIR HyperSpectralImaging (HSI) based techniques and calorimetric test. The study demonstrated as the proposed separation technique and the HyperSpectral NIR Imaging approach allow to separate and recognize the different polymers (i.e. PolyVinyl Chloride (PVC), PolyStyrene (PS), PolyEthylene (PE), PoliEtilene Tereftalato (PET), PolyPropylene (PP)) in order to maximize the removal of the PVC fraction from plastic waste and to perform the full quality control of the resulting products, can be profitably utilized to set up analytical/control strategies finalized to obtain a low content of PVC in the final Solid Recovered Fuel (SRF), thus enhancing SRF quality, increasing its value and reducing the "final waste". Copyright © 2016 Elsevier Ltd. All rights reserved.

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

    International Nuclear Information System (INIS)

    Smith, H.D.; Tracey, E.M.; Darab, J.G.; Smith, P.A.

    1996-03-01

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

  16. Radioactive wastes. New design: the package at the core of investigations

    International Nuclear Information System (INIS)

    Fillet, C.; Cau Dit Coumes, C.; Joussot-Dubien, Ch.; Boen, R.; Ferry, C.; Ribet, I.; Devezeaux de Lavergne, J.G.; Lieven, Th.; Silvy, J.P.; Poinssot, Ch.; Macias, R.M.; Moncouyoux, J.P.

    2005-01-01

    Current waste conditioning processes have demonstrated their effectiveness on an industrial scale. Major advances have nonetheless been achieved, with the development of processes having broader applications, and the ability to achieve further waste volume reductions yet. Meanwhile what will be the long-term behavior of nuclear packages, cornerstone of nuclear wastes management studies? And what about the spent fuels which involve investigation of specific processes, insofar as they may end up in a configuration not originally intended for a long-term storage or disposal facility? Is industrial storage suited for long-term management? How may storage and the long-term be made compatible? This document aims to answer these questions and discusses also on the other countries management of radioactive wastes. (A.L.B.)

  17. Nuclear energy - Waste-packages activity measurement - Part.1: high-resolution gamma spectrometry in integral mode with open geometry

    International Nuclear Information System (INIS)

    2004-01-01

    ISO 14850:2004 describes a procedure for measurements of gamma-emitting radionuclide activity in homogeneous objects such as unconditioned waste (including process waste, dismantling waste, etc.), waste conditioned in various matrices (bitumen, hydraulic binder, thermosetting resins, etc.), notably in the form of 100 L, 200 L, 400 L or 800 L drums, and test specimens or samples, (vitrified waste), and waste packaged in a container, notably technological waste. It also specifies the calibration of the gamma spectrometry chain. The gamma energies used generally range from 0,05 MeV to 3 MeV. (authors)

  18. Prediction of radionuclide invention for low-and intermediate-level radioactive waste by considering concentration limit of waste package

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Kang Il; Kim, Min Seong; Jeong, Noh Gyeon; Park, Jin Beak [Korea Radioactive Waste Agency(KORAD), Daejeon (Korea, Republic of)

    2017-03-15

    The result of a preliminary safety assessment that was completed by applying the radionuclide inventory calculated on the basis of available data from radioactive waste generation agencies suggested that many difficulties are to be expected with regard to disposal safety and operation. Based on the results of the preliminary safety assessment of the entire disposal system, in this paper, a unit package exceeding the safety goal is selected that occupies a large proportion of radionuclides in intermediate-level radioactive waste. We introduce restrictions on the amount of radioactivity in a way that excludes the high surface dose rate of the package. The radioactivity limit for disposal will be used as the baseline data for establishing the acceptance criteria and the disposal criteria for each disposal facility to meet the safety standards. It is necessary to draw up a comprehensive safety development plan for the Gyeongju waste disposal facility that will contribute to the construction of a Safety Case for the safety optimization of radioactive waste disposal facilities.

  19. Development of waste package designs for disposal in a salt repository

    International Nuclear Information System (INIS)

    Balmert, M.E.

    1983-01-01

    Three package design concepts were developed for CHLW and DHLW forms and spent fuel rods: (1) carbon steel overpack, borehole emplacement, (2) titanium clad, carbon steel reinforced overpack, borehole emplacement, and (3) carbon steel (self-shield) overpack, tunnel emplacement. For a DHLW canister with titanium clad overpack, the concept features a 9.5-cm-thick carbon steel overpack reinforcement supporting a 0.25-cm-thick titanium shell. The overall package dimensions are 84 cm diameter x 340 cm long weighing about 8.8 mtons. By contrast, a monolithic DHLW borehole package has a