National Library of Energy BETA

Sample records for waste package materials

  1. Waste Package Materials Performance Peer Review

    Broader source: Energy.gov [DOE]

    A consensus peer review of the current technical basis and the planned experimental and modeling program for the prediction of the long-term performance of waste package materials being considered...

  2. Aqueous Corrosion Rates for Waste Package Materials

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    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.

  4. Materials of Criticality Safety Concern in Waste Packages

    SciTech Connect (OSTI)

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

    2006-07-01

    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)

  5. Waste disposal package

    DOE Patents [OSTI]

    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.

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

    SciTech Connect (OSTI)

    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.

  7. Radioactive waste disposal package

    DOE Patents [OSTI]

    Lampe, Robert F. (Bethel Park, PA)

    1986-01-01

    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.

  8. Radioactive waste disposal package

    DOE Patents [OSTI]

    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.

  9. Waste Package Lifting Calculation

    SciTech Connect (OSTI)

    H. Marr

    2000-05-11

    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. WASTE PACKAGE TRANSPORTER DESIGN

    SciTech Connect (OSTI)

    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''.

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

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    Westerman, R.E.

    1980-10-01

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

  14. A quantitative assessment of microbiological contributions to corrosion of candidate nuclear waste package materials

    SciTech Connect (OSTI)

    Lian, T.; Jones, D.; Martin, S.; Horn, J.

    1999-07-01

    The US Department of Energy is contributing to the design of a potential nuclear waste repository at Yucca Mountain, Nevada. A system to predict the contribution of Yucca Mountain (YM) bacteria to overall corrosion rates of candidate waste package (WP) materials was designed and implemented. DC linear polarization resistance techniques were applied to candidate material coupons that had been inoculated with a mixture of YM-derived bacteria with potentially corrosive activities, or left sterile. Inoculated bacteria caused a 5- to 6-fold increase in corrosion rate of carbon steel C1020 (to approximately 7--8 {micro}m/yr), and an almost 100-fold increase in corrosion rate of Alloy 400 (to approximately 1 {micro}m/yr) was observed due to microbiological activities. Microbiologically Influenced Corrosion (MIC) rates on more resistant materials (CRMs: Alloy 625, Type 304 Stainless Steel, and Alloy C22) were on the order of hundredths of micrometers per year ({micro}m/yr). Bulk chemical and surfacial endpoint analyses of spent media and coupon surfaces showed preferential dissolution of nickel from Alloy 400 coupons and depletion of chromium from CRMs after incubation with YM bacteria. Scanning electron microscopy also showed greater damage to the Alloy 400 surface than that indicated by electrochemical detection methods.

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

    SciTech Connect (OSTI)

    G. Gordon

    2004-10-13

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

  16. Corrosion and environmental-mechanical characterization of iron-base nuclear waste package structural barrier materials. Annual report, FY 1984

    SciTech Connect (OSTI)

    Westerman, R.E.; Haberman, J.H.; Pitman, S.G.; Pulsipher, B.A.; Sigalla, L.A.

    1986-03-01

    Disposal of high-level nuclear waste in deep underground repositories may require the development of waste packages that will keep the radioisotopes contained for up to 1000 y. A number of iron-base materials are being considered for the structural barrier members of waste packages. Their uniform and nonuniform (pitting and intergranular) corrosion behavior and their resistance to stress-corrosion cracking in aqueous environments relevant to salt media are under study at Pacific Northwest Laboratory. The purpose of the work is to provide data for a materials degradation model that can ultimately be used to predict the effective lifetime of a waste package overpack in the actual repository environment. The corrosion behavior of the candidate materials was investigated in simulated intrusion brine (essentially NaCl) in flowing autoclave tests at 150/sup 0/C, and in combinations of intrusion/inclusion (high-Mg) brine environments in moist salt tests, also at 150/sup 0/C. Studies utilizing a /sup 60/Co irradiation facility were performed to determine the corrosion resistance of the candidate materials to products of brine radiolysis at dose rates of 2 x 10/sup 3/ and 1 x 10/sup 5/ rad/h and a temperature of 150/sup 0/C. These irradiation-corrosion tests were ''overtests,'' as the irradiation intensities employed were 10 to 1000 times as high as those expected at the surface of a thick-walled waste package. With the exception of the high general corrosion rates found in the tests using moist salt containing high-Mg brines, the ferrous materials exhibited a degree of corrosion resistance that indicates a potentially satisfactory application to waste package structural barrier members in a salt repository environment.

  17. Feasibility assessment of copper-base waste package container materials in a tuff repository

    SciTech Connect (OSTI)

    Acton, C.F.; McCright, R.D.

    1986-09-30

    This report discussed progress made during the second year of a two-year study on the feasibility of using copper or a copper-base alloy as a container material for a waste package in a potential repository in tuff rock at the Yucca Mountain site in Nevada. Corrosion testing in potentially corrosive irradiated environments received emphasis during the feasibility study. Results of experiments to evaluate the effect of a radiation field on the uniform corrosion rate of the copper-base materials in repository-relevant aqueous environments are given as well as results of an electrochemical study of the copper-base materials in normal and concentrated J-13 water. Results of tests on the irradiation of J-13 water and on the subsequent formation of hydrogen peroxide are given. A theoretical study was initiated to predict the long-term corrosion behavior of copper in the repository. Tests were conducted to determine whether copper would adversely affect release rates of radionuclides to the environment because of degradation of the Zircaloy cladding. A manufacturing survey to determine the feasibility of producing copper containers utilizing existing equipment and processes was completed. The cost and availability of copper was also evaluated and predicted to the year 2000. Results of this feasibility assessment are summarized.

  18. Depleted uranium oxides and silicates as spent nuclear fuel waste package fill materials

    SciTech Connect (OSTI)

    Forsberg, C.W.

    1996-09-10

    A new repository waste package (WP) concept for spent nuclear fuel (SNF) is being investigated that uses depleted uranium (DU) to improve performance and reduce the uncertainties of geological disposal of SNF. The WP would be filled with SNF and then filled with depleted uranium (DU) ({approximately}0.2 wt % {sup 235}U) dioxide (UO{sub 2}) or DU silicate-glass beads. Fission products and actinides can not escape the SNF UO{sub 2} crystals until the UO{sub 2} dissolves or is transformed into other chemical species. After WP failure, the DU fill material slows dissolution by three mechanisms: (1) saturation of AT groundwater with DU and suppression of SNF dissolution, (2) maintenance of chemically reducing conditions in the WP that minimize SNF solubility by sacrificial oxidation of DU from the +4 valence state, and (3) evolution of DU to lower-density hydrated uranium silicates. The fill expansion seals the WP from water flow. The DU also isotopically exchanges with SNF uranium as the SNF degrades to reduce long-term nuclear-criticality concerns.

  19. Influence of lithium-containing waste materials on the melting of packaging glass

    SciTech Connect (OSTI)

    Katkova, K.S.; Balandina, T.I.; Belyaeva, A.G.; Guloyan, Y.A.; Seregina, E.P.

    1986-07-01

    Lithium-containing waste materials from mica enrichment factories are studied. The possibilities of using the wastes for melting of green and semi-white sodium calcium silicate glasses are studied as well. Using physical methods, the authors study the influence of lithium-containing mica waste materials on the process of batch melting, silicate and glass formation, and clarification of molten glass. Tables show melting characteristics with various additions of lithium-containing mica waste, and the influence of added Li/sub 2/O on glass clarification. It is shown that the addition of lithium-containing mica waste materials to sodium calcium silicate glass has a positive effect on silicate and glass formation and on clarification.

  20. Naval Waste Package Design Report

    SciTech Connect (OSTI)

    M.M. Lewis

    2004-03-15

    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.

  1. Tritium waste package

    DOE Patents [OSTI]

    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.

  2. Tritium waste package

    DOE Patents [OSTI]

    Rossmassler, R.; Ciebiera, L.; Tulipano, F.J.; Vinson, S.; Walters, R.T.

    1995-11-07

    A containment and waste package system for processing and shipping tritium oxide waste received from a process gas includes an outer drum and an inner drum containing a disposable molecular sieve bed (DMSB) seated within the 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 and oxygen present in the process gas to recombine to form water vapor, which is then adsorbed onto the DMSB. 1 fig.

  3. FY 1985 status report on feasibility assessment of copper-base waste package container materials in a tuff repository

    SciTech Connect (OSTI)

    McCright, R.D.

    1985-09-30

    This report discusses progress made during the first year of a two-year study on the feasibility of using copper or a copper-base alloy as a container material for a waste package in a potential repository in tuff rock at the Yucca Mountain site in Nevada. The expected corrosion and oxidation performances of oxygen-free copper, aluminum bronze, and 70% copper-30% nickel are presented; a test plan for determining whether copper or one of the alloys can meet the containment requirements is outlined. Some preliminary corrosion test data are presented and discussed. Fabrication and joining techniques for forming waste package containers are descibed. Preliminary test data and analyses indicate that copper and copper-base alloys have several attractive features as waste package container materials, but additional work is needed before definitive conclusions can be made on the feasibility of using copper or a copper-base alloy for containers. Plans for work to be undertaken in the second year are indicated.

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

    SciTech Connect (OSTI)

    Van Konynenburg, R.A.; Halsey, W.G.; McCright, R.D.; Clarke, W.L. Jr.; Gdowski, G.E.

    1993-02-01

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

  5. Nuclear Material Packaging Manual

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2008-03-07

    The manual provides detailed packaging requirements for protecting workers from exposure to nuclear materials stored outside of an approved engineered contamination barrier. Does not cancel/supersede other directives. Certified 11-18-10.

  6. Nuclear Material Packaging

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2008-03-07

    The manual provides detailed packaging requirements for protecting workers from exposure to nuclear materials stored outside of an approved engineered contamination barrier. Supersedes DOE M 441.1-1, dated 3-7-08.

  7. Packaging Materials Properties Data

    SciTech Connect (OSTI)

    Leduc, D.

    1991-10-30

    Several energy absorbing materials are used in nuclear weapons component shipping containers recently designed for the Y-12 Plant Program Management Packaging Group. As a part of the independent review procedure leading to Certificates of Compliance, the U.S. Department of Energy Technical Safety Review Panels requested compression versus deflection . data on these materials. This report is a compilation of that data.

  8. Packaging materials properties data

    SciTech Connect (OSTI)

    Walker, M.S.

    1991-01-01

    Several energy absorbing materials are used in nuclear weapons component shipping containers recently designed for the Y-12 Plant Program Management Packaging Group. As a part of the independent review procedure leading to Certificates of Compliance, the US Department of Energy Technical Safety Review Panels requested compression versus deflection data on these materials. This report is a compilation of that data.

  9. The reduction of packaging waste

    SciTech Connect (OSTI)

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

    1994-04-01

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

  10. EQ6 Calculations for Chemical Degradation of Navy Waste Packages

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    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.

  12. Nuclear waste package fabricated from concrete

    SciTech Connect (OSTI)

    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/sup 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.

  13. Reference waste package environment report

    SciTech Connect (OSTI)

    Glassley, W.E.

    1986-10-01

    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 {similar_to}230{sup 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.

  14. Packaging - Materials review

    SciTech Connect (OSTI)

    Herrmann, Matthias

    2014-06-16

    Nowadays, a large number of different electrochemical energy storage systems are known. In the last two decades the development was strongly driven by a continuously growing market of portable electronic devices (e.g. cellular phones, lap top computers, camcorders, cameras, tools). Current intensive efforts are under way to develop systems for automotive industry within the framework of electrically propelled mobility (e.g. hybrid electric vehicles, plug-in hybrid electric vehicles, full electric vehicles) and also for the energy storage market (e.g. electrical grid stability, renewable energies). Besides the different systems (cell chemistries), electrochemical cells and batteries were developed and are offered in many shapes, sizes and designs, in order to meet performance and design requirements of the widespread applications. Proper packaging is thereby one important technological step for designing optimum, reliable and safe batteries for operation. In this contribution, current packaging approaches of cells and batteries together with the corresponding materials are discussed. The focus is laid on rechargeable systems for industrial applications (i.e. alkaline systems, lithium-ion, lead-acid). In principle, four different cell types (shapes) can be identified - button, cylindrical, prismatic and pouch. Cell size can be either in accordance with international (e.g. International Electrotechnical Commission, IEC) or other standards or can meet application-specific dimensions. Since cell housing or container, terminals and, if necessary, safety installations as inactive (non-reactive) materials reduce energy density of the battery, the development of low-weight packages is a challenging task. In addition to that, other requirements have to be fulfilled: mechanical stability and durability, sealing (e.g. high permeation barrier against humidity for lithium-ion technology), high packing efficiency, possible installation of safety devices (current interrupt device

  15. YUCCA MOUNTAIN WASTE PACKAGE CLOSURE SYSTEM

    SciTech Connect (OSTI)

    G. Housley; C. Shelton-davis; K. Skinner

    2005-08-26

    The method selected for dealing with spent nuclear fuel in the US is to seal the fuel in waste packages and then to place them in an underground repository at the Yucca Mountain Site in Nevada. This article describes the Waste Package Closure System (WPCS) currently being designed for sealing the waste packages.

  16. Depleted uranium as a backfill for nuclear fuel waste package

    DOE Patents [OSTI]

    Forsberg, C.W.

    1998-11-03

    A method is described for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package. 6 figs.

  17. Depleted uranium as a backfill for nuclear fuel waste package

    DOE Patents [OSTI]

    Forsberg, Charles W.

    1998-01-01

    A method for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package.

  18. Waste Package Design Methodology Report

    SciTech Connect (OSTI)

    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.

  19. Background studies in support of a feasibility assessment on the use of copper-base materials for nuclear waste packages in a repository in tuff

    SciTech Connect (OSTI)

    Van Konynenburg, R.A.; Kundig, K.J.A.; Lyman, W.S.; Prager, M.; Meyers, J.R.; Servi, I.S.

    1990-06-01

    This report combines six work units performed in FY`85--86 by the Copper Development Association and the International Copper Research Association under contract with the University of California. The work includes literature surveys and state-of-the-art summaries on several considerations influencing the feasibility of the use of copper-base materials for fabricating high-level nuclear waste packages for the proposed repository in tuff rock at Yucca Mountain, Nevada. The general conclusion from this work was that copper-base materials are viable candidates for inclusion in the materials selection process for this application. 55 refs., 48 figs., 22 tabs.

  20. CH Packaging Operations for High Wattage Waste

    SciTech Connect (OSTI)

    Washington TRU Solutions LLC

    2006-01-06

    This document provides instructions for assembling the following CH packaging payload: Drum payload assembly Standard Waste Box (SWB) assembly Ten-Drum Overpack (TDOP)

  1. Hazardous Materials Packaging and Transportation Safety - DOE...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    60.1D, Hazardous Materials Packaging and Transportation Safety by Ashok Kapoor Functional areas: Hazardous Materials, Packaging and Transportation, Safety and Security, Work...

  2. CRAD, Packaging and Transfer of Hazardous Materials and Materials...

    Office of Environmental Management (EM)

    Packaging and Transfer of Hazardous Materials and Materials of National Security Interest Assessment Plan CRAD, Packaging and Transfer of Hazardous Materials and Materials of...

  3. Thermal Evaluation of the Fort Saint Vrain Codisposal Waste Package

    SciTech Connect (OSTI)

    Adam Scheider; Horia Radulescu

    2001-07-19

    The objective of this calculation is to evaluate the thermal response of the Fort Saint Vrain (FSV) Codisposal Waste Package (WP) design under nominal Monitored Geologic Repository conditions. The objective of the calculation is to provide thermal parameter information to support the FSV waste package design. The information provided by the sketches (Attachment IV) 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. 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. 16). AP-3.124, ''Calculations'' (Ref. 17) is used to perform the calculation and develop the document. The sketches attached to this calculation provide the potential dimensions and materials for the SDHLW (Defense High Level Waste) / DOE (Department of Energy) Long WP.

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

    SciTech Connect (OSTI)

    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]).

  5. CERAMIC WASTE FORM DATA PACKAGE

    SciTech Connect (OSTI)

    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.

  6. Waste Package Component Design Methodology Report

    SciTech Connect (OSTI)

    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 and operational

  7. 21-PWR Waste Package Side and End Impacts

    SciTech Connect (OSTI)

    V. Delabrosse

    2003-02-27

    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.

  8. 21-PWR Waste Package Side and End Impacts

    SciTech Connect (OSTI)

    T. Schmitt

    2005-08-29

    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.

  9. Departmental Materials Transportation and Packaging Management

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2010-11-18

    Establishes requirements and responsibilities for management of Department of Energy (DOE), including National Nuclear Security Administration, materials transportation and packaging and ensures the safe, secure, efficient packaging and transportation of materials, both hazardous and non-hazardous.

  10. Conceptual waste packaging options for deep borehole disposal

    SciTech Connect (OSTI)

    Su, Jiann -Cherng; Hardin, Ernest L.

    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

  11. Swing-Down of 21-PWR Waste Package

    SciTech Connect (OSTI)

    A.K. Scheider

    2001-05-04

    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.

  12. WASTE PACKAGE REMEDIATION SYSTEM DESCRIPTION DOCUMENT

    SciTech Connect (OSTI)

    N.D. Sudan

    2000-06-22

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

  13. The Model 9977 Radioactive Material Packaging Primer

    SciTech Connect (OSTI)

    Abramczyk, G.

    2015-10-09

    The Model 9977 Packaging is a single containment drum style radioactive material (RAM) shipping container designed, tested and analyzed to meet the performance requirements of Title 10 the Code of Federal Regulations Part 71. A radioactive material shipping package, in combination with its contents, must perform three functions (please note that the performance criteria specified in the Code of Federal Regulations have alternate limits for normal operations and after accident conditions): Containment, the package must “contain” the radioactive material within it; Shielding, the packaging must limit its users and the public to radiation doses within specified limits; and Subcriticality, the package must maintain its radioactive material as subcritical

  14. Hanford Site radioactive hazardous materials packaging directory

    SciTech Connect (OSTI)

    McCarthy, T.L.

    1995-12-01

    The Hanford Site Radioactive Hazardous Materials Packaging Directory (RHMPD) provides information concerning packagings owned or routinely leased by Westinghouse Hanford Company (WHC) for offsite shipments or onsite transfers of hazardous materials. Specific information is provided for selected packagings including the following: general description; approval documents/specifications (Certificates of Compliance and Safety Analysis Reports for Packaging); technical information (drawing numbers and dimensions); approved contents; areas of operation; and general information. Packaging Operations & Development (PO&D) maintains the RHMPD and may be contacted for additional information or assistance in obtaining referenced documentation or assistance concerning packaging selection, availability, and usage.

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

    SciTech Connect (OSTI)

    Manaktala, H.K. . Center for Nuclear Waste Regulatory Analyses); Interrante, C.G. . 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. ALTERNATE MATERIALS IN DESIGN OF RADIOACTIVE MATERIAL PACKAGES

    SciTech Connect (OSTI)

    Blanton, P.; Eberl, K.

    2010-07-09

    This paper presents a summary of design and testing of material and composites for use in radioactive material packages. These materials provide thermal protection and provide structural integrity and energy absorption to the package during normal and hypothetical accident condition events as required by Title 10 Part 71 of the Code of Federal Regulations. Testing of packages comprising these materials is summarized.

  17. Drift emplaced waste package thermal response

    SciTech Connect (OSTI)

    Ruffner, D.J.; Johnson, G.L.; Platt, E.A.; Blink, J.A.; Doering, T.W.

    1993-01-01

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

  18. Hazardous Material Packaging for Transport - Administrative Procedures

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1986-09-30

    To establ1sh administrative procedures for the certification and use of radioactive and other hazardous materials packaging by the Department of Energy (DOE).

  19. RECLAMATION OF RADIOACTIVE MATERIAL PACKAGING COMPONENTS

    SciTech Connect (OSTI)

    Abramczyk, G.; Nathan, S.; Loftin, B.; Bellamy, S.

    2011-06-06

    Radioactive material packages are withdrawn from use for various reasons; loss of mission, decertification, damage, replacement, etc. While the packages themselves may be decertified, various components may still be able to perform to their required standards and find useful service. The Packaging Technology and Pressurized Systems group of the Savannah River National Laboratory has been reducing the cost of producing new Type B Packagings by reclaiming, refurbishing, and returning to service the containment vessels from older decertified packagings. The program and its benefits are presented.

  20. Secondary Waste Form Down Selection Data Package – Ceramicrete

    SciTech Connect (OSTI)

    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

  1. Industrial Waste Landfill IV upgrade package

    SciTech Connect (OSTI)

    Not Available

    1994-03-29

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

  2. Departmental Materials Transportation and Packaging Management

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2004-12-22

    The Order establishes requirements and responsibilities for management of Department of Energy (DOE), including National Nuclear Security Administration (NNSA), materials transportation and packaging to ensure the safe, secure, efficient packaging and transportation of materials, both hazardous and nonhazardous. Cancels DOE O 460.2 and DOE O 460.2 Chg 1

  3. Repository Waste Package Transporter Shielding Weight Optimization

    SciTech Connect (OSTI)

    C.E. Sanders; Shiaw-Der Su

    2005-02-02

    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.

  4. 21-PWR WASTE PACKAGE WITH ABSORBER PLATES LOADING CURVE EVALUATION

    SciTech Connect (OSTI)

    J.M. Scaglione

    2004-12-17

    The objective of this calculation is to evaluate the required minimum burnup as a function of initial pressurized water reactor (PWR) assembly enrichment that would permit loading of spent nuclear fuel into the 21 PWR waste package with absorber plates design as provided in Attachment IV. This calculation is an example of the application of the methodology presented in the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2003). The scope of this calculation covers a range of enrichments from 0 through 5.0 weight percent U-235, and a burnup range of 0 through 45 GWd/MTU. Higher burnups were not necessary because 45 GWd/MTU was high enough for the loading curve determination. This activity supports the validation of the use of burnup credit for commercial spent nuclear fuel applications. The intended use of these results will be in establishing PWR waste package configuration loading specifications. Limitations of this evaluation are as follows: (1) The results are based on burnup credit for actinides and selected fission products as proposed in YMP (2003, Table 3-1) and referred to as the ''Principal Isotopes''. Any change to the isotope listing will have a direct impact on the results of this report. (2) The results are based on 1.5 wt% Gd in the Ni-Gd Alloy material and having no tuff inside the waste package. If the Gd loading is reduced or a process to introduce tuff inside the waste package is defined, then this report would need to be reevaluated based on the alternative materials. This calculation is subject to the ''Quality Assurance Requirements and Description'' (QARD) (DOE 2004) because it concerns engineered barriers that are included in the ''Q-List'' (BSC 2004k, Appendix A) as items important to safety and waste isolation.

  5. Departmental Materials Transportation and Packaging Management

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1995-09-27

    Establishes Department of Energy (DOE) policies and requirements to supplement applicable laws, rules, regulations, and other DOE Orders for materials transportation and packaging operations. Cancels DOE 1540.1A, DOE 1540.2, DOE 1540.3A.

  6. Departmental Materials Transportation and Packaging Management

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1995-10-26

    Establishes Department of Energy (DOE) policies and requirements to supplement applicable laws, rules, regulations, and other DOE Orders for materials transportation and packaging operations. Cancels: DOE 1540.1A, DOE 1540.2, and DOE 1540.3A.

  7. Hazardous Materials Packaging and Transportation Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2015-04-20

    The Order establishes safety requirements for the proper packaging and transportation of Department of offsite shipments and onsite transfers of radioactive and other hazardous materials, and for modal transportation.

  8. CH Packaging Operations for High Wattage Waste at LANL

    SciTech Connect (OSTI)

    Washington TRU Solutions LLC

    2005-04-04

    This procedure provides instructions for assembling the following CH packaging payload: Drum payload assembly Standard Waste Box (SWB) assembly Ten-Drum Overpack (TDOP).

  9. CH Packaging Operations for High Wattage Waste at LANL

    SciTech Connect (OSTI)

    Washington TRU Solutions LLC

    2005-04-13

    This procedure provides instructions for assembling the following CH packaging payload: Drum payload assembly Standard Waste Box (SWB) assembly Ten-Drum Overpack (TDOP).

  10. EXTERNAL CRITICALITY CALCULATION FOR DOE SNF CODISPOSAL WASTE PACKAGES

    SciTech Connect (OSTI)

    H. Radulescu

    2002-10-18

    The purpose of this document is to evaluate the potential for criticality for the fissile material that could accumulate in the near-field (invert) and in the far-field (host rock) beneath the U.S. Department of Energy (DOE) spent nuclear fuel (SNF) codisposal waste packages (WPs) as they degrade in the proposed monitored geologic repository at Yucca Mountain. The scope of this calculation is limited to the following DOE SNF types: Shippingport Pressurized Water Reactor (PWR), Enrico Fermi, Fast Flux Test Facility (FFTF), Fort St. Vrain, Melt and Dilute, Shippingport Light Water Breeder Reactor (LWBR), N-Reactor, and Training, Research, Isotope, General Atomics reactor (TRIGA). The results of this calculation are intended to be used for estimating the probability of criticality in the near-field and in the far-field. There are no limitations on use of the results of this calculation. The calculation is associated with the waste package design and was developed in accordance with the technical work plan, ''Technical Work Plan for: Department of Energy Spent Nuclear Fuel and Plutonium Disposition Work Packages'' (Bechtel SAIC Company, LLC [BSC], 2002a). This calculation is subject to the Quality Assurance Requirements and Description (QARD) per the activity evaluation under work package number P6212310Ml in the technical work plan TWP-MGR-MD-0000 10 REV 01 (BSC 2002a).

  11. EM Waste and Materials Disposition & Transportation | Department...

    Office of Environmental Management (EM)

    Waste and Materials Disposition & Transportation EM Waste and Materials Disposition & Transportation DOE's Radioactive Waste Management Priorities: Continue to manage waste ...

  12. Advanced Ceramic Materials and Packaging Technologies for Realizing...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Ceramic Materials and Packaging Technologies for Realizing Sensors for Concentrating Solar Power Systems Advanced Ceramic Materials and Packaging Technologies for Realizing Sensors ...

  13. The Model 9977 Radioactive Material Packaging Primer (Technical...

    Office of Scientific and Technical Information (OSTI)

    Technical Report: The Model 9977 Radioactive Material Packaging Primer Citation Details In-Document Search Title: The Model 9977 Radioactive Material Packaging Primer The Model...

  14. Safety evaluation for packaging (onsite) for concrete-shielded RHTRU waste drum for the 327 postirradiation testing laboratory

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    Blanton, P.; Eberl, K.

    2013-10-10

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

  16. THERMAL EVALUATION OF THE 2-MCO/2-DHLW WASTE PACKAGE

    SciTech Connect (OSTI)

    T. Schmitt

    2000-07-05

    The objective of this calculation was to determine the structural response of multi-canister overpacks (MCO) and the 2-MCO/2-Defense High-Level Waste (DHLW) Waste Package (WP) subjected to tip-over onto an unyielding surface (US). The scope of this calculation was limited to reporting the calculation results in terms of maximum stress intensities. This calculation is associated with the waste package design and was performed by the Waste Package Design Section in accordance with the DOE SNF Analysis Plan for FY 2000.

  17. The radioactive materials packaging handbook: Design, operations, and maintenance

    SciTech Connect (OSTI)

    Shappert, L.B.; Bowman, S.M.; Arnold, E.D.

    1998-08-01

    As part of its required activities in 1994, the US Department of Energy (DOE) made over 500,000 shipments. Of these shipments, approximately 4% were hazardous, and of these, slightly over 1% (over 6,400 shipments) were radioactive. Because of DOE`s cleanup activities, the total quantities and percentages of radioactive material (RAM) that must be moved from one site to another is expected to increase in the coming years, and these materials are likely to be different than those shipped in the past. Irradiated fuel will certainly be part of the mix as will RAM samples and waste. However, in many cases these materials will be of different shape and size and require a transport packaging having different shielding, thermal, and criticality avoidance characteristics than are currently available. This Handbook provides guidance on the design, testing, certification, and operation of packages for these materials.

  18. Contact-Handled and Remote-Handled Transuranic Waste Packaging

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2011-08-09

    Provides specific instructions for packaging and/or repackaging contact-handled transuranic (CH-TRU) and remote-handled transuranic (RH-TRU) waste in a manner consistent with DOE O 435.1, Radioactive Waste Management, DOE M 435.1-1 Chg 1, Radioactive Waste Management Manual, CH-TRU and RH-TRU waste transportation requirements, and Waste Isolation Pilot Plant (WIPP) programmatic requirements. Does not cancel/supersede other directives.

  19. RADIOACTIVE MATERIAL PACKAGING TORQUE REQUIREMENTS COMPLIANCE

    SciTech Connect (OSTI)

    Watkins, R.; Leduc, D.

    2011-03-24

    Shipping containers used to transport radioactive material (RAM) in commerce employ a variety of closure mechanisms. Often, these closure mechanisms require a specific amount of torque be applied to a bolt, nut or other threaded fastener. It is important that the required preload is achieved so that the package testing and analysis is not invalidated for the purpose of protecting the public. Torque compliance is a means of ensuring closure preload, is a major factor in accomplishing the package functions of confinement/containment, sub-criticality, and shielding. This paper will address the importance of applying proper torque to package closures, discuss torque value nomenclature, and present one methodology to ensure torque compliance is achieved.

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

    SciTech Connect (OSTI)

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

    1995-12-01

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

  1. CH Packaging Operations for High Wattage Waste at LANL

    SciTech Connect (OSTI)

    Washington TRU Solutions LLC

    2003-05-06

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

  2. CH Packaging Operations for High Wattage Waste at LANL

    SciTech Connect (OSTI)

    Washington TRU Solutions LLC

    2003-08-28

    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.

  3. CH Packaging Operations for High Wattage Waste at LANL

    SciTech Connect (OSTI)

    Washington TRU Solutions LLC

    2003-03-21

    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.

  4. CH Packaging Operations for High Wattage Waste at LANL

    SciTech Connect (OSTI)

    Washington TRU Solutions LLC

    2002-10-17

    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.

  5. CH Packaging Operations for High Wattage Waste at LANL

    SciTech Connect (OSTI)

    Washington TRU Solutions LLC

    2002-12-18

    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.

  6. DESIGN ANALYSIS FOR THE NAVAL SNF WASTE PACKAGE

    SciTech Connect (OSTI)

    T.L. Mitchell

    2000-05-31

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

  7. Motor Packaging with Consideration of Electromagnetic and Material...

    Broader source: Energy.gov (indexed) [DOE]

    Documents & Publications Motor Packaging with Consideration of Electromagnetic and Material Characteristics Alnico and Ferrite Hybrid Excitation Electric Machines Wireless Charging

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

    SciTech Connect (OSTI)

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

    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.

  9. Measurement of radionuclides in waste packages

    DOE Patents [OSTI]

    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.

  10. Measurement of radionuclides in waste packages

    DOE Patents [OSTI]

    Brodzinski, Ronald L.; Perkins, Richard W.; Rieck, Henry G.; Wogman, Ned A.

    1986-01-01

    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.

  11. Thermal Response of the 44-BWR Waste Package to a Hypothetical Fire Accident

    SciTech Connect (OSTI)

    J.R. Smotrel; H. Marr; M.J. Anderson

    2001-04-05

    The purpose of this calculation is to determine the thermal response of the 44-boiling water reactor (BWR) waste package (WP) to the hypothetical regulatory fire accident. The objective is to calculate the temperature response of the waste package materials to the hypothetical short-term fire defined in 10 CFR 7 1, Section 73(c)(4), Reference 1. The scope of the calculation includes evaluation of the accident with the waste package above ground, at the Yucca Mountain surface facility. The scope 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 is that for the potential design of the type of WP considered in this calculation. In addition to the nominal design configuration thermal load case, the effects of varying the BWR thermal load are determined. The associated activity is the development of engineering evaluations to support the Licensing Application (LA) design activities.

  12. WASTE PACKAGE OPERATIONS FY99 CLOSURE METHODS REPORT

    SciTech Connect (OSTI)

    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.

  13. Cermet Spent Nuclear Fuel Casks and Waste Packages

    SciTech Connect (OSTI)

    Forsberg, Charles W.; Dole, Leslie R.

    2007-07-01

    Multipurpose transport, aging, and disposal casks are needed for the management of spent nuclear fuel (SNF). Self-shielded cermet casks can out-perform current SNF casks because of the superior properties of cermets, which consist of encapsulated hard ceramic particulates dispersed in a continuous ductile metal matrix to produce a strong high-integrity, high-thermal conductivity cask. A multi-year, multinational development and testing program has been developing cermet SNF casks made of steel, depleted uranium dioxide, and other materials. Because cermets are the traditional material of construction for armor, cermet casks can provide superior protection against assault. For disposal, cermet waste packages (WPs) with appropriate metals and ceramics can buffer the local geochemical environment to (1) slow degradation of SNF, (2) reduce water flow though the degraded WP, (3) sorb neptunium and other radionuclides that determine the ultimate radiation dose to the public from the repository, and (4) contribute to long-term nuclear criticality control. Finally, new cermet cask fabrication methods have been partly developed to manufacture the casks with the appropriate properties. The results of this work are summarized with references to the detailed reports. (authors)

  14. Long term nuclear criticality potential in waste packages

    SciTech Connect (OSTI)

    Thomas, D.A.; Doering, T.W.

    1994-12-31

    Title 10 CFR 60.131.(b).(7) requires that the radioactive waste disposed of in the Mined Geologic Disposal System (MGDS) remain subcritical during the period of isolation. The period of waste isolation, approximately 10,000 years, represents a time period greater than any previously examined for criticality control of spent fuel. Change in the criticality potential over long time periods for the Multi-Purpose Canister (MPC) waste package conceptual design has been examined and methods of criticality control over this time have been investigated.

  15. Aging and Phase Stability of Waste Package Outer Barrier

    SciTech Connect (OSTI)

    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 phases (P, {mu}, and {sigma}) are

  16. Base Technology for Radioactive Material Transportation Packaging Systems

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1992-07-08

    To establish Department of Energy (DOE) policies and responsibilities for coordinating and planning base technology for radioactive material transportation packaging systems.

  17. Packaging and Transportation for Offsite Shipment of Materials...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    of Materials of National Security Interests by Matthew Weber Functional areas: Defense Nuclear Facility Safety and Health Requirement, Packaging and Transportation, Security,...

  18. Packaging and Transportation for Offsite Shipment of Materials...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    1.1C Second Draft, Packaging and Transportation for Offsite Shipment of Materials of National Security Interests by Matthew Weber Functional areas: Defense Nuclear Facility Safety...

  19. Motor Packaging with Consideration of Electromagnetic and Material...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Ferrite Hybrid Excitation Electric Machines Motor Packaging with Consideration of Electromagnetic and Material Characteristics Novel Flux Coupling Machine without Permanent Magnets

  20. The Model 9977 Radioactive Material Packaging Primer (Technical...

    Office of Scientific and Technical Information (OSTI)

    Subject: 42 ENGINEERING; RADIOACTIVE MATERIALS; PACKAGING; RADIATION DOSES; PERFORMANCE; CONTAINERS; CRITICALITY; DESIGN; CASKS; SHIELDING Word Cloud More Like This Full Text ...

  1. Safety analysis report for packaging a DOT 7A specification container for tritiated liquid wastes

    SciTech Connect (OSTI)

    Alford, E.

    1980-08-01

    This Safety Analysis Report for Packaging (SARP) was prepared in accordance with ERDA (DOE) Appendix 5201 for DOE/ALO review and approval of packaging of tritiated liquid wastes to be shipped from Sandia National Laboratories, Livermore, (SNLL) California. This report presents information pertinent to the construction of tritiated liquid waste shipping containers. It contains design and development considerations, explains tests and evaluations required to prove the container can withstand normal transportation conditions, and demonstrates that the Sandia container-and-radioactive-material shipment package is in compliance with DOE and Department of Transportation (DOT) safety requirements. An internal review of this SARP has been performed in compliance with the ERDA (DOE) Manual, 5201 Appendix V.

  2. Radioactive waste material melter apparatus

    DOE Patents [OSTI]

    Newman, D.F.; Ross, W.A.

    1990-04-24

    An apparatus for preparing metallic radioactive waste material for storage is disclosed. The radioactive waste material is placed in a radiation shielded enclosure. The waste material is then melted with a plasma torch and cast into a plurality of successive horizontal layers in a mold to form a radioactive ingot in the shape of a spent nuclear fuel rod storage canister. The apparatus comprises a radiation shielded enclosure having an opening adapted for receiving a conventional transfer cask within which radioactive waste material is transferred to the apparatus. A plasma torch is mounted within the enclosure. A mold is also received within the enclosure for receiving the melted waste material and cooling it to form an ingot. The enclosure is preferably constructed in at least two parts to enable easy transport of the apparatus from one nuclear site to another. 8 figs.

  3. Radioactive waste material melter apparatus

    DOE Patents [OSTI]

    Newman, Darrell F.; Ross, Wayne A.

    1990-01-01

    An apparatus for preparing metallic radioactive waste material for storage is disclosed. The radioactive waste material is placed in a radiation shielded enclosure. The waste material is then melted with a plasma torch and cast into a plurality of successive horizontal layers in a mold to form a radioactive ingot in the shape of a spent nuclear fuel rod storage canister. The apparatus comprises a radiation shielded enclosure having an opening adapted for receiving a conventional transfer cask within which radioactive waste material is transferred to the apparatus. A plasma torch is mounted within the enclosure. A mold is also received within the enclosure for receiving the melted waste material and cooling it to form an ingot. The enclosure is preferably constructed in at least two parts to enable easy transport of the apparatus from one nuclear site to another.

  4. Stress Corrosion Cracking Model for High Level Radioactive-Waste Packages

    SciTech Connect (OSTI)

    P. Andresen; G. Gordon; S. Lu

    2004-10-05

    A stress corrosion cracking (SCC) model has been adapted for performance prediction of high level radioactive-waste packages to be emplaced in the proposed Yucca Mountain repository. For waste packages of the proposed Yucca Mountain repository, the outer barrier material is the highly corrosion-resistant Alloy UNS-N06022 (Alloy 22), the environment is represented by aqueous brine films present on the surface of the waste package from dripping or deliquescence of soluble salts present in any surface deposits, and the tensile stress is principally from weld induced residual stress. SCC has historically been separated into ''initiation'' and ''propagation'' phases. Initiation of SCC will not occur on a smooth surface if the surface stress is below a threshold value defined as the threshold stress. Cracks can also initiate at and propagate from flaws (or defects) resulting from manufacturing processes (such as welding); or that develop from corrosion processes such as pitting or dissolution of inclusions. To account for crack propagation, the slip dissolution/film rupture (SDFR) model is adopted to provide mathematical formulae for prediction of the crack growth rate. Once the crack growth rate at an initiated SCC is determined, it can be used by the performance assessment to determine the time to through-wall penetration for the waste package. This paper presents the development of the SDFR crack growth rate model based on technical information in the literature as well as experimentally determined crack growth rates developed specifically for Alloy UNS-N06022 in environments relevant to high level radioactive-waste packages of the proposed Yucca Mountain radioactive-waste repository. In addition, a seismic damage related SCC crack opening area density model is briefly described.

  5. THERMAL ANALYSIS OF GEOLOGIC HIGH-LEVEL RADIOACTIVE WASTE PACKAGES

    SciTech Connect (OSTI)

    Hensel, S.; Lee, S.

    2010-04-20

    The engineering design of disposal of the high level waste (HLW) packages in a geologic repository requires a thermal analysis to provide the temperature history of the packages. Calculated temperatures are used to demonstrate compliance with criteria for waste acceptance into the geologic disposal gallery system and as input to assess the transient thermal characteristics of the vitrified HLW Package. The objective of the work was to evaluate the thermal performance of the supercontainer containing the vitrified HLW in a non-backfilled and unventilated underground disposal gallery. In order to achieve the objective, transient computational models for a geologic vitrified HLW package were developed by using a computational fluid dynamics method, and calculations for the HLW disposal gallery of the current Belgian geological repository reference design were performed. An initial two-dimensional model was used to conduct some parametric sensitivity studies to better understand the geologic system's thermal response. The effect of heat decay, number of co-disposed supercontainers, domain size, humidity, thermal conductivity and thermal emissivity were studied. Later, a more accurate three-dimensional model was developed by considering the conduction-convection cooling mechanism coupled with radiation, and the effect of the number of supercontainers (3, 4 and 8) was studied in more detail, as well as a bounding case with zero heat flux at both ends. The modeling methodology and results of the sensitivity studies will be presented.

  6. Packaging material for thin film lithium batteries

    DOE Patents [OSTI]

    Bates, John B.; Dudney, Nancy J.; Weatherspoon, Kim A.

    1996-01-01

    A thin film battery including components which are capable of reacting upon exposure to air and water vapor incorporates a packaging system which provides a barrier against the penetration of air and water vapor. The packaging system includes a protective sheath overlying and coating the battery components and can be comprised of an overlayer including metal, ceramic, a ceramic-metal combination, a parylene-metal combination, a parylene-ceramic combination or a parylene-metal-ceramic combination.

  7. THERMAL PERFORMANCE OF RADIOACTIVE MATERIAL PACKAGES IN TRANSPORT CONFIGURATION

    SciTech Connect (OSTI)

    Gupta, N.

    2010-03-04

    Drum type packages are routinely used to transport radioactive material (RAM) in the U.S. Department of Energy (DOE) complex. These packages are designed to meet the federal regulations described in 10 CFR Part 71. The packages are transported in specially designed vehicles like Safe Secure Transport (SST) for safety and security. In the transport vehicles, the packages are placed close to each other to maximize the number of units in the vehicle. Since the RAM contents in the packagings produce decay heat, it is important that they are spaced sufficiently apart to prevent overheating of the containment vessel (CV) seals and the impact limiter to ensure the structural integrity of the package. This paper presents a simple methodology to assess thermal performance of a typical 9975 packaging in a transport configuration.

  8. Radioactive waste material disposal

    DOE Patents [OSTI]

    Forsberg, Charles W.; Beahm, Edward C.; Parker, George W.

    1995-01-01

    The invention is a process for direct conversion of solid radioactive waste, particularly spent nuclear fuel and its cladding, if any, into a solidified waste glass. A sacrificial metal oxide, dissolved in a glass bath, is used to oxidize elemental metal and any carbon values present in the waste as they are fed to the bath. Two different modes of operation are possible, depending on the sacrificial metal oxide employed. In the first mode, a regenerable sacrificial oxide, e.g., PbO, is employed, while the second mode features use of disposable oxides such as ferric oxide.

  9. Radioactive waste material disposal

    DOE Patents [OSTI]

    Forsberg, C.W.; Beahm, E.C.; Parker, G.W.

    1995-10-24

    The invention is a process for direct conversion of solid radioactive waste, particularly spent nuclear fuel and its cladding, if any, into a solidified waste glass. A sacrificial metal oxide, dissolved in a glass bath, is used to oxidize elemental metal and any carbon values present in the waste as they are fed to the bath. Two different modes of operation are possible, depending on the sacrificial metal oxide employed. In the first mode, a regenerable sacrificial oxide, e.g., PbO, is employed, while the second mode features use of disposable oxides such as ferric oxide. 3 figs.

  10. Considerations on the performance and fabrication of candidate materials for the Yucca Mountain repository waste packages highly corrosion resistant nickel-base and titanium-base alloys

    SciTech Connect (OSTI)

    Dalder, E; Goldberg, A

    1995-11-30

    Among the metallurgical factors that affect the performance of a material in a given environment are alloy composition, alloy segregation, depletion of alloying elements, non-uniform microstructures, precipitation leading to an increase in susceptibility to corrosion as well as decreases in ductility, residual plastic deformation, and residual stresses. Precipitation often occurs preferentially at grain boundaries, causing depletion of critical elements in regions adjacent to these boundaries. Continuous grain-boundary precipitates can lead to drops in ductility and toughness. The presence of non-metallic inclusions, if excessive and/or segregated, can also cause embrittlement. Segregation of alloying elements can result in localized galvanic action. Depletion of alloying elements as well as segregation can result in reductions in the concentrations of critical elements below those necessary to resist localized corrosion. Segregation and alloy depletion can also facilitate precipitation that could lead to embrittlement.

  11. THERMAL UPGRADING OF 9977 RADIOACTIVE MATERIAL (RAM) TYPE B PACKAGE

    SciTech Connect (OSTI)

    Gupta, N.; Abramczyk, G.

    2012-03-26

    The 9977 package is a radioactive material package that was originally certified to ship Heat Sources and RTG contents up to 19 watts and it is now being reviewed to significantly expand its contents in support of additional DOE missions. Thermal upgrading will be accomplished by employing stacked 3013 containers, a 3013 aluminum spacer and an external aluminum sleeve for enhanced heat transfer. The 7th Addendum to the original 9977 package Safety Basis Report describing these modifications is under review for the DOE certification. The analyses described in this paper show that this well-designed and conservatively analyzed package can be upgraded to carry contents with decay heat up to 38 watts with some simple design modifications. The Model 9977 package has been designed as a replacement for the Department of Transportation (DOT) Fissile Specification 6M package. The 9977 package is a very versatile Type B package which is certified to transport and store a wide spectrum of radioactive materials. The package was analyzed quite conservatively to increase its usefulness and store different payload configurations. Its versatility is evident from several daughter packages such as the 9978 and H1700, and several addendums where the payloads have been modified to suit the Shipper's needs without additional testing.

  12. DEVELOPMENT OF A NEW TYPE A(F)RADIOACTIVE MATERIAL PACKAGING FOR THE DEPARTMENT OF ENERGY

    SciTech Connect (OSTI)

    Blanton, P.; Eberl, K.

    2008-09-14

    In a coordinated effort, the Department of Transportation (DOT) and Nuclear Regulatory Commission (NRC) proposed the elimination of the Specification Packaging from 49 CFR 173.[1] In accordance with the Federal Register, issued on October 1, 2004, new fabrication of Specification Packages would no longer be authorized. In accordance with the NRC final rulemaking published January 26, 2004, Specification Packagings are mandated by law to be removed from service no later than October 1, 2008. This coordinated effort and resulting rulemaking initiated a planned phase out of Specification Type B and Type A fissile (F) material transportation packages within the Department of Energy (DOE) and its subcontractors. One of the Specification Packages affected by this regulatory change is the UN1A2 Specification Package, per DOT 49 CFR 173.417(a)(6). To maintain continuing shipments of DOE materials currently transported in UN1A2 Specification Package after the existing authorization expires, a replacement Type A(F) material packaging design is under development by the Savannah River National Laboratory. This paper presents a summary of the prototype design effort and testing of the new Type A(F) Package development for the DOE. This paper discusses the progress made in the development of a Type A Fissile Packaging to replace the expiring 49 CFR UN1A2 Specification Fissile Package. The Specification Package was mostly a single-use waste disposal container. The design requirements and authorized radioactive material contents of the UN1A2 Specification Package were defined in 49 CFR. A UN1A2 Specification Package was authorized to ship up to 350 grams of U-235 in any enrichment and in any non-pyrophoric form. The design was specified as a 55-gallon 1A2 drum overpack with a body constructed from 18 gauge steel with a 16 gauge drum lid. Drum closure was specified as a standard 12-gauge ring closure. The inner product container size was not specified but was listed as any

  13. BWR ASSEMBLY SOURCE TERMS FOR WASTE PACKAGE DESIGN

    SciTech Connect (OSTI)

    T.L. Lotz

    1997-02-15

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

  14. Extending the utility of a radioactive material package

    SciTech Connect (OSTI)

    Abramczyk, G.; Nathan, S.; Loftin, B.; Bellamy, S.

    2015-06-04

    Once a package has been certified for the transportation of DOT Hazard Class 7 – Radioactive Material in compliance with the requirements of 10 CFR 71, it is often most economical to extend its utility through the addition of content-specific configuration control features or the addition of shielding materials. The SRNL Model 9977 Package’s authorization was expanded from its original single to twenty contents in this manner; and most recently, the 9977 was evaluated for a high-gamma source content. This paper discusses the need for and the proposed shielding modifications to the package for extending the utility of the package for this purpose.

  15. Assessment of Quality Assurance Measures for Radioactive Material Transport Packages not Requiring Competent Authority Design Approval - 13282

    SciTech Connect (OSTI)

    Komann, Steffen; Groeke, Carsten; Droste, Bernhard

    2013-07-01

    The majority of transports of radioactive materials are carried out in packages which don't need a package design approval by a competent authority. Low-active radioactive materials are transported in such packages e.g. in the medical and pharmaceutical industry and in the nuclear industry as well. Decommissioning of NPP's leads to a strong demand for packages to transport low and middle active radioactive waste. According to IAEA regulations the 'non-competent authority approved package types' are the Excepted Packages and the Industrial Packages of Type IP-1, IP-2 and IP-3 and packages of Type A. For these types of packages an assessment by the competent authority is required for the quality assurance measures for the design, manufacture, testing, documentation, use, maintenance and inspection (IAEA SSR 6, Chap. 306). In general a compliance audit of the manufacturer of the packaging is required during this assessment procedure. Their regulatory level in the IAEA regulations is not comparable with the 'regulatory density' for packages requiring competent authority package design approval. Practices in different countries lead to different approaches within the assessment of the quality assurance measures in the management system as well as in the quality assurance program of a special package design. To use the package or packaging in a safe manner and in compliance with the regulations a management system for each phase of the life of the package or packaging is necessary. The relevant IAEA-SSR6 chap. 801 requires documentary verification by the consignor concerning package compliance with the requirements. (authors)

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

    SciTech Connect (OSTI)

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

    2000-10-03

    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.

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

    SciTech Connect (OSTI)

    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.

  18. Elemental characterization of LL-MA radioactive waste packages with the associated particle technique

    SciTech Connect (OSTI)

    Perot, B.; Carasco, C.; Toure, M.; El Kanawati, W.; Eleon, C.

    2011-07-01

    The French Alternative Energies and Atomic Energy Commission (CEA) and National Radioactive Waste Management Agency (ANDRA) are conducting an R and D program to improve the characterization of long-lived and medium activity (LL-MA) radioactive waste packages with analytical methods and with non-destructive nuclear measurements. This paper concerns fast neutron interrogation with the associated particle technique (APT), which brings 3D information about the waste material composition. The characterization of volume elements filled with iron, water, aluminium, and PVC in bituminized and fibre concrete LL-MA waste packages has been investigated with MCNP [1] and MODAR data analysis software [2]. APT provides usable information about major elements presents in the volumes of interest. However, neutron scattering on hydrogen nuclei spreads the tagged neutron beam out of the targeted volume towards surrounding materials, reducing spatial selectivity. Simulation shows that small less than 1 L targets can be characterised up to the half-radius of a 225 L bituminized drum, the matrix of which is very rich in hydrogen. Deeper characterization in concrete is possible but limited by counting statistics due to photon attenuation in this dense matrix and, unless large inspection volumes are considered, by the lack of spatial selectivity of the tagged neutron beam due to neutron scattering. (authors)

  19. Use of depleted uranium silicate glass to minimize release of radionuclides from spent nuclear fuel waste packages

    SciTech Connect (OSTI)

    Forsberg, C.W.

    1996-01-20

    A Depleted Uranium Silicate Container Backfill System (DUSCOBS) is proposed that would use small, isotopically-depleted uranium silicate glass beads as a backfill material inside repository waste packages containing spent nuclear fuel (SNF). The uranium silicate glass beads would fill the void space inside the package including the coolant channels inside SNF assemblies. Based on preliminary analysis, the following benefits have been identified. DUSCOBS improves repository waste package performance by three mechanisms. First, it reduces the radionuclide releases from SNF when water enters the waste package by creating a local uranium silicate saturated groundwater environment that suppresses (a) the dissolution and/or transformation of uranium dioxide fuel pellets and, hence, (b) the release of radionuclides incorporated into the SNF pellets. Second, the potential for long-term nuclear criticality is reduced by isotopic exchange of enriched uranium in SNF with the depleted uranium (DU) in the glass. Third, the backfill reduces radiation interactions between SNF and the local environment (package and local geology) and thus reduces generation of hydrogen, acids, and other chemicals that degrade the waste package system. Finally, DUSCOBS provides a potential method to dispose of significant quantities of excess DU from uranium enrichment plants at potential economic savings. DUSCOBS is a new concept. Consequently, the concept has not been optimized or demonstrated in laboratory experiments.

  20. Packaging and Transfer of Materials of National Security Interest Manual

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2000-09-29

    This Technical Manual establishes requirements for operational safety controls for onsite operations and provides Department of Energy (DOE) technical safety requirements and policy objectives for development of an Onsite Packaging and Transfer Program, pursuant to DOE O 461.1A, Packaging and Transfer or Transportation of Materials of National Security Interest. The DOE contractor must document this program in its Onsite Packaging and Transfer Manual/Procedures. Admin Chg 1, 7-26-05. Certified 2-2-07. Canceled by DOE O 461.2.

  1. Abstraction of Models for Pitting and Crevice Corrosion of Drip Shield and Waste Package Outer Barrier

    SciTech Connect (OSTI)

    K. Mon

    2001-08-29

    This analyses and models report (AMR) was conducted in response to written work direction (CRWMS M and 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). The purpose and scope of this AMR is to review and analyze upstream process-level models (CRWMS M and O 2000a and CRWMS M and O 2000b) and information relevant to pitting and crevice corrosion degradation of waste package outer barrier (Alloy 22) and drip shield (Titanium Grade 7) materials, and to develop abstractions of the important processes in a form that is suitable for input to the WAPDEG analysis for long-term degradation of waste package outer barrier and drip shield in the repository. The abstraction is developed in a manner that ensures consistency with the process-level models and information and captures the essential behavior of the processes represented. Also considered in the model abstraction are the probably range of exposure conditions in emplacement drifts and local exposure conditions on drip shield and waste package surfaces. The approach, method, and assumptions that are employed in the model abstraction are documented and justified.

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

    SciTech Connect (OSTI)

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

    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. Complex-wide representation of material packaged in 3013 containers

    SciTech Connect (OSTI)

    Narlesky, Joshua E.; Peppers, Larry G.; Friday, Gary P.

    2009-06-01

    The DOE sites packaging plutonium oxide materials packaged according to Department of Energy 3013 Standard (DOE-STD-3013) are responsible for ensuring that the materials are represented by one or more samples in the Materials Identification and Surveillance (MIS) program. The sites categorized most of the materials into process groups, and the remaining materials were characterized, based on the prompt gamma analysis results. The sites issued documents to identify the relationships between the materials packaged in 3013 containers and representative materials in the MIS program. These Represented documents were then reviewed and concurred with by the MIS Working Group. However, these documents were developed uniquely at each site and were issued before completion of sample characterization, small-scale experiments, and prompt gamma analysis, which provided more detailed information about the chemical impurities and the behavior of the material in storage. Therefore, based on the most recent data, relationships between the materials packaged in 3013 containers and representative materials in the MIS program been revised. With the prompt gamma analysis completed for Hanford, Rocky Flats, and Savannah River Site 3013 containers, MIS items have been assigned to the 3013 containers for which representation is based on the prompt gamma analysis results. With the revised relationships and the prompt gamma analysis results, a Master Represented table has been compiled to document the linkages between each 3013 container packaged to date and its representative MIS items. This table provides an important link between the Integrated Surveillance Program database, which contains information about each 3013 container to the MIS items database, which contains the characterization, prompt gamma data, and storage behavior data from shelf-life experiments for the representative MIS items.

  4. Thermal Evaluation for the Naval SNF Waste Package

    SciTech Connect (OSTI)

    T.L. Mitchell

    2000-04-25

    The purpose of this calculation is to evaluate the thermal performance of the naval long spent nuclear fuel (SNF) waste package (WP) under multiple disposal conditions in a monitored geologic repository (MGR). The scope of this calculation is limited to determination of thermal temperature profiles upon the surface of, and within, the naval long SNF WP. The objective is to develop a temperature profile history within the WP, at time increments up to 10,000 years of emplacement. The results of this calculation are intended to support the Naval SNF WP Analysis and Model Report (AMR) for Site Recommendation (SR). This calculation was performed to the specifications within its Technical Development Plan (TDP) (Ref. 8.16). This calculation is developed and documented in accordance with the AP-3.12Q/REV. 0IICN. 0 procedure, Calculations.

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

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    Souza, Daiane Cristini B.; Gimenes Tessaro, Ana Paula; Vicente, Roberto

    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)

  8. DRAFT - DOE O 460.1D, Hazardous Materials Packaging and Transportation...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    60.1D, Hazardous Materials Packaging and Transportation Safety by Website Administrator The Order establishes safety requirements for the proper packaging and transportation of...

  9. CRAD, Packaging and Transfer of Hazardous Materials and Materials of National Security Interest Assessment Plan

    Broader source: Energy.gov [DOE]

    Verify that packaging and transportation safety requirements of hazardous materials and materials of national security interest have been established and are in compliance with DOE Orders 461.1 and 460.1B

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

    SciTech Connect (OSTI)

    Calmus, D.B.

    1994-08-25

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

  11. Stailization, Packaging, and Storage of Plutonium-Bearing Materials

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    DOE-STD-3013-2012 MARCH 2012 DOE STANDARD STABILIZATION, PACKAGING, AND STORAGE OF PLUTONIUM-BEARING MATERIALS U.S. Department of Energy AREA PACK Washington, D.C. 20585 DISTRIBUTION STATEMENT A. Approved for public release; distribution is unlimited. TS Available on the Department of Energy Technical Standards Program Web site at http://www.hss.energy.gov/NuclearSafety/ns/techstds/ DOE-STD-3013-2012 iii ABSTRACT This Standard provides guidance for the stabilization, packaging, and safe storage

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

    SciTech Connect (OSTI)

    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.

  13. CALORIMETRY OF TRU WASTE MATERIALS

    SciTech Connect (OSTI)

    C. RUDY; ET AL

    2000-08-01

    Calorimetry has been used for accountability measurements of nuclear material in the US. Its high accuracy, insensitivity to matrix effects, and measurement traceability to National Institute of Standards and Technology have made it the primary accountability assay technique for plutonium (Pu) and tritium in the Department of Energy complex. A measurement of Pu isotopic composition by gamma-ray spectroscopy is required to transform the calorimeter measurement into grams Pu. The favorable calorimetry attributes allow it to be used for verification measurements, for production of secondary standards, for bias correction of other faster nondestructive (NDA) methods, or to resolve anomalous measurement results. Presented in this paper are (1) a brief overview of calorimeter advantages and disadvantages, (2) a description of projected large volume calorimeters suitable for waste measurements, and (3) a new technique, direct measurement of transuranic TRU waste alpha-decay activity through calorimetry alone.

  14. NEW APPROACH TO ADDRESSING GAS GENERATION IN RADIOACTIVE MATERIAL PACKAGING

    SciTech Connect (OSTI)

    Watkins, R; Leduc, D; Askew, N

    2009-06-25

    Safety Analysis Reports for Packaging (SARP) document why the transportation of radioactive material is safe in Type A(F) and Type B shipping containers. The content evaluation of certain actinide materials require that the gas generation characteristics be addressed. Most packages used to transport actinides impose extremely restrictive limits on moisture content and oxide stabilization to control or prevent flammable gas generation. These requirements prevent some users from using a shipping container even though the material to be shipped is fully compliant with the remaining content envelope including isotopic distribution. To avoid these restrictions, gas generation issues have to be addressed on a case by case basis rather than a one size fits all approach. In addition, SARP applicants and review groups may not have the knowledge and experience with actinide chemistry and other factors affecting gas generation, which facility experts in actinide material processing have obtained in the last sixty years. This paper will address a proposal to create a Gas Generation Evaluation Committee to evaluate gas generation issues associated with Safety Analysis Reports for Packaging material contents. The committee charter could include reviews of both SARP approved contents and new contents not previously evaluated in a SARP.

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

    SciTech Connect (OSTI)

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

    2002-02-26

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

  16. Graphite matrix materials for nuclear waste isolation

    SciTech Connect (OSTI)

    Morgan, W.C.

    1981-06-01

    At low temperatures, graphites are chemically inert to all but the strongest oxidizing agents. The raw materials from which artificial graphites are produced are plentiful and inexpensive. Morover, the physical properties of artificial graphites can be varied over a very wide range by the choice of raw materials and manufacturing processes. Manufacturing processes are reviewed herein, with primary emphasis on those processes which might be used to produce a graphite matrix for the waste forms. The approach, recommended herein, involves the low-temperature compaction of a finely ground powder produced from graphitized petroleum coke. The resultant compacts should have fairly good strength, low permeability to both liquids and gases, and anisotropic physical properties. In particular, the anisotropy of the thermal expansion coefficients and the thermal conductivity should be advantageous for this application. With two possible exceptions, the graphite matrix appears to be superior to the metal alloy matrices which have been recommended in prior studies. The two possible exceptions are the requirements on strength and permeability; both requirements will be strongly influenced by the containment design, including the choice of materials and the waste form, of the multibarrier package. Various methods for increasing the strength, and for decreasing the permeability of the matrix, are reviewed and discussed in the sections in Incorporation of Other Materials and Elimination of Porosity. However, it would be premature to recommend a particular process until the overall multi-barrier design is better defined. It is recommended that increased emphasis be placed on further development of the low-temperature compacted graphite matrix concept.

  17. Corrosion of Metal Inclusions In Bulk Vitrification Waste Packages

    SciTech Connect (OSTI)

    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

  18. Treatment of halogen-containing waste and other waste materials

    DOE Patents [OSTI]

    Forsberg, C.W.; Beahm, E.C.; Parker, G.W.

    1997-03-18

    A process is described for treating a halogen-containing waste material. The process provides a bath of molten glass containing a sacrificial metal oxide capable of reacting with a halogen in the waste material. The sacrificial metal oxide is present in the molten glass in at least a stoichiometric amount with respect to the halogen in the waste material. The waste material is introduced into the bath of molten glass to cause a reaction between the halogen in the waste material and the sacrificial metal oxide to yield a metal halide. The metal halide is a gas at the temperature of the molten glass. The gaseous metal halide is separated from the molten glass and contacted with an aqueous scrubber solution of an alkali metal hydroxide to yield a metal hydroxide or metal oxide-containing precipitate and a soluble alkali metal halide. The precipitate is then separated from the aqueous scrubber solution. The molten glass containing the treated waste material is removed from the bath as a waste glass. The process of the invention can be used to treat all types of waste material including radioactive wastes. The process is particularly suited for separating halogens from halogen-containing wastes. 3 figs.

  19. Treatment of halogen-containing waste and other waste materials

    DOE Patents [OSTI]

    Forsberg, Charles W.; Beahm, Edward C.; Parker, George W.

    1997-01-01

    A process for treating a halogen-containing waste material. The process provides a bath of molten glass containing a sacrificial metal oxide capable of reacting with a halogen in the waste material. The sacrificial metal oxide is present in the molten glass in at least a stoichiometric amount with respect to the halogen in the waste material. The waste material is introduced into the bath of molten glass to cause a reaction between the halogen in the waste material and the sacrificial metal oxide to yield a metal halide. The metal halide is a gas at the temperature of the molten glass. The gaseous metal halide is separated from the molten glass and contacted with an aqueous scrubber solution of an alkali metal hydroxide to yield a metal hydroxide or metal oxide-containing precipitate and a soluble alkali metal halide. The precipitate is then separated from the aqueous scrubber solution. The molten glass containing the treated waste material is removed from the bath as a waste glass. The process of the invention can be used to treat all types of waste material including radioactive wastes. The process is particularly suited for separating halogens from halogen-containing wastes.

  20. RECERTIFICATION OF THE MODEL 9977 RADIOACTIVE MATERIAL PACKAGING

    SciTech Connect (OSTI)

    Abramczyk, G.; Bellamy, S.; Loftin, B.; Nathan, S.

    2013-06-05

    The Model 9977 Packaging was initially issued a Certificate of Compliance (CoC) by the Department of Energy’s Office of Environmental Management (DOE-EM) for the transportation of radioactive material (RAM) in the Fall of 2007. This first CoC was for a single radioactive material and two packing configurations. In the five years since that time, seven Addendums have been written to the Safety Analysis Report for Packaging (SARP) and five Letter Amendments have been written that have authorized either new RAM contents or packing configurations, or both. This paper will discuss the process of updating the 9977 SARP to include all the contents and configurations, including the addition of a new content, and its submittal for recertification.

  1. Waste Form Degradation Model Integration for Engineered Materials...

    Office of Environmental Management (EM)

    Waste Form Degradation Model Integration for Engineered Materials Performance Waste Form Degradation Model Integration for Engineered Materials Performance The collaborative ...

  2. Operations to be Performed in the Waste Package Dry Remediation Cell

    SciTech Connect (OSTI)

    Norman E. Cole; Randy K. Elwood

    2003-10-01

    Describes planned and proposed operations for remediating damaged and/or out-of-compliance waste packages, casks, DPCs, overpacks, and containers at the Yucca Mountain Dry Transfer Facility.

  3. Yucca Mountain Waste Package Closure System Robotic Welding and Inspection System

    SciTech Connect (OSTI)

    C. I. Nichol; D. P. Pace; E. D. Larsen; T. R. McJunkin; D. E. Clark; M. L. Clark; K. L. Skinner; A. D. Watkins; H. B. Smartt

    2011-10-01

    The Waste Package Closure System (WPCS), for the closure of radioactive waste in canisters for permanent storage of spent nuclear fuel (SNF) and high-level waste in the Yucca Mountain Repository was designed, fabricated, and successfully demonstrated at the Idaho National Laboratory (INL). This article focuses on the robotic hardware and tools necessary to remotely weld and inspect the closure lid welds. The system was operated remotely and designed for use in a radiation field, due to the SNF contained in the waste packages being closed.

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

    SciTech Connect (OSTI)

    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.

  5. Methane generation from waste materials

    SciTech Connect (OSTI)

    Samani, Zohrab A.; Hanson, Adrian T.; Macias-Corral, Maritza

    2010-03-23

    An organic solid waste digester for producing methane from solid waste, the digester comprising a reactor vessel for holding solid waste, a sprinkler system for distributing water, bacteria, and nutrients over and through the solid waste, and a drainage system for capturing leachate that is then recirculated through the sprinkler system.

  6. In-situ vitrification of waste materials

    DOE Patents [OSTI]

    Powell, James R.; Reich, Morris; Barletta, Robert

    1997-11-14

    A method for the in-situ vitrification of waste materials in a disposable can that includes an inner container and an outer container is disclosed. The method includes the steps of adding frit and waste materials to the inner container, removing any excess water, heating the inner container such that the frit and waste materials melt and vitrify after cooling, while maintaining the outer container at a significantly lower temperature than the inner container. The disposable can is then cooled to ambient temperatures and stored. A device for the in-situ vitrification of waste material in a disposable can is also disclosed.

  7. In-situ vitrification of waste materials

    DOE Patents [OSTI]

    Powell, J.R.; Reich, M.; Barletta, R.

    1997-10-14

    A method for the in-situ vitrification of waste materials in a disposable can that includes an inner container and an outer container is disclosed. The method includes the steps of adding frit and waste materials to the inner container, removing any excess water, heating the inner container such that the frit and waste materials melt and vitrify after cooling, while maintaining the outer container at a significantly lower temperature than the inner container. The disposable can is then cooled to ambient temperatures and stored. A device for the in-situ vitrification of waste material in a disposable can is also disclosed. 7 figs.

  8. EM QA Working Group September 2011 Meeting Materials | Department...

    Energy Savers [EERE]

    Nuclear Materials & Waste Tank Waste and Waste Processing Waste Disposition Packaging and Transportation Site & Facility Restoration Deactivation & Decommissioning (D&D)...

  9. Enterprise Assessments Targeted Assessment of the Waste Treatment and Immobilization Plant High-Level Waste Facility Radioactive Liquid Waste Disposal System Safety Basis Change Package … May 2016

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Assessment of the Waste Treatment and Immobilization Plant High-Level Waste Facility Radioactive Liquid Waste Disposal System Safety Basis Change Package May 2016 Office of Nuclear Safety and Environmental Assessments Office of Environment, Safety and Health Assessments Office of Enterprise Assessments U.S. Department of Energy i Table of Contents Acronyms

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

    SciTech Connect (OSTI)

    PD Meyer; RJ Serne

    1999-12-21

    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.

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

    SciTech Connect (OSTI)

    Not Available

    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.

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

    SciTech Connect (OSTI)

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

    1989-10-01

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

  13. Developing an institutional strategy for transporting defense transuranic waste materials

    SciTech Connect (OSTI)

    Guerrero, J.V.; Kresny, H.S.

    1986-01-01

    In late 1988, the US Department of Energy (DOE) expects to begin emplacing transuranic waste materials in the Waste Isolation Pilot Plant (WIPP), an R and D facility to demonstrate the safe disposal of radioactive wastes resulting from defense program activities. Transuranic wastes are production-related materials, e.g., clothes, rags, tools, and similar items. These materials are contaminated with alpha-emitting transuranium radionuclides with half-lives of > 20 yr and concentrations > 100 nCi/g. Much of the institutional groundwork has been done with local communities and the State of New Mexico on the siting and construction of the facility. A key to the success of the emplacement demonstration, however, will be a qualified transportation system together with institutional acceptance of the proposed shipments. The DOE's Defense Transuranic Waste Program, and its contractors, has lead responsibility for achieving this goal. The Joint Integration Office (JIO) of the DOE, located in Albuquerque, New Mexico, is taking the lead in implementing an integrated strategy for assessing nationwide institutional concerns over transportation of defense transuranic wastes and in developing ways to resolve or mitigate these concerns. Parallel prototype programs are under way to introduce both the new packaging systems and the institutional strategy to interested publics and organizations.

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

    SciTech Connect (OSTI)

    K.D. Draper

    2005-03-31

    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

  15. Analysis and evaluation of a radioactive waste package retrieved from the Farallon Islands 900-meter disposal site

    SciTech Connect (OSTI)

    Colombo, P.; Kendig, M.W.

    1990-09-01

    The Environmental Protection Agency (EPA) was given a Congressional mandate to develop criteria and regulations governing the ocean disposal of all forms of waste. The EPA taken an active role both nationally and within the international nuclear regulatory community to develop the effective controls necessary to protect the health and safety of man and the marine environment. The EPA Office of Radiation Programs (ORP) first initiated feasibility studies to determine whether current technologies could be applied toward determining the fate of radioactive waste disposed of in the past. After successfully locating actual radioactive waste packages in formerly used disposal sites, in the United States, the Office of Radiation Programs developed an intensive program of site characterization studies to examine biological, chemical and physical characteristics including evaluations of the concentration and distribution of radionuclides within these sites, and has conducted a performance evaluation of past packaging techniques and materials. Brookhaven National Laboratory (BNL) has performed container corrosion and matrix analysis studies on the recovered radioactive waste packages. This report presents the final results of laboratory analyses performed. 17 refs., 40 figs., 7 tabs.

  16. Criteria for Packaging and Storing Uranium-233-Bearing Materials

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    3028-2000 July 2000 DOE STANDARD CRITERIA FOR PACKAGING AND STORING URANIUM-233-BEARING MATERIALS U.S. Department of Energy AREA SAFT Washington, D.C. 20585 DISTRIBUTION STATEMENT A. Approved for public release; distribution is unlimited. TS This document has been reproduced from the best available copy. Available to DOE and DOE contractors from ES&H Technical Information Services, U.S. Department of Energy, (800) 473-4375, fax: (301) 903-9823. Available to the public from the U.S.

  17. Stabilization, Packaging, and Storage of Plutonium-Bearing Materials

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    DOE-STD-3013-2000 September 2000 Superseding DOE-STD-3013-99 November 1999 DOE STANDARD STABILIZATION, PACKAGING, AND STORAGE OF PLUTONIUM-BEARING MATERIALS U.S. Department of Energy AREA PACK Washington, D.C. 20585 DISTRIBUTION STATEMENT A. Approved for public release; distribution is unlimited. TS This document has been reproduced from the best available copy. Available to DOE and DOE contractors from ES&H Technical Information Services, U.S. Department of Energy, (800) 473-4375, fax:

  18. Evaluation and compilation of DOE waste package test data: Biannual report, February 1987--July 1987

    SciTech Connect (OSTI)

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

    1988-05-01

    The waste package is a proposed engineering 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 steels, stainless steels, and copper. The current level of understanding of several canister materials is questioned for the candidate repository in tuff. Three issues are addressed, the possibility of the stress-induced failure of Zircaloy, the possible corrosion of copper and copper alloys, and the lack of site-specific characterization data. Discussions are given on problems concerning localized corrosion and environmentally assisted cracking of AISI 1020 steel at elevated temperatures (150{degree}C). For the proposed salt site, the importance of the duration of corrosion tests and some of the conditions that may preclude prompt initiation of needed long-term testing are two issues that are discussed. 31 refs., 5 figs.

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

    SciTech Connect (OSTI)

    Kevin Skinner; Greg Housley; Colleen Shelton-Davis

    2011-11-01

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

  20. Recovery of fissile materials from nuclear wastes

    DOE Patents [OSTI]

    Forsberg, Charles W.

    1999-01-01

    A process for recovering fissile materials such as uranium, and plutonium, and rare earth elements, from complex waste feed material, and converting the remaining wastes into a waste glass suitable for storage or disposal. The waste feed is mixed with a dissolution glass formed of lead oxide and boron oxide resulting in oxidation, dehalogenation, and dissolution of metal oxides. Carbon is added to remove lead oxide, and a boron oxide fusion melt is produced. The fusion melt is essentially devoid of organic materials and halogens, and is easily and rapidly dissolved in nitric acid. After dissolution, uranium, plutonium and rare earth elements are separated from the acid and recovered by processes such as PUREX or ion exchange. The remaining acid waste stream is vitrified to produce a waste glass suitable for storage or disposal. Potential waste feed materials include plutonium scrap and residue, miscellaneous spent nuclear fuel, and uranium fissile wastes. The initial feed materials may contain mixtures of metals, ceramics, amorphous solids, halides, organic material and other carbon-containing material.

  1. PACCOM: A nuclear waste packaging facility cost model: Draft technical report

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    Ferreira da Cruz, Nuno 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

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

    SciTech Connect (OSTI)

    H. Marr

    2006-10-25

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

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

    SciTech Connect (OSTI)

    Ward, Andy L.

    2007-11-26

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

  5. Regulatory and extra-regulatory testing to demonstrate radioactive material packaging safety

    SciTech Connect (OSTI)

    Ammerman, D.J.

    1997-06-01

    Packages for the transportation of radioactive material must meet performance criteria to assure safety and environmental protection. The stringency of the performance criteria is based on the degree of hazard of the material being transported. Type B packages are used for transporting large quantities of radioisotopes (in terms of A{sub 2} quantities). These packages have the most stringent performance criteria. Material with less than an A{sub 2} quantity are transported in Type A packages. These packages have less stringent performance criteria. Transportation of LSA and SCO materials must be in {open_quotes}strong-tight{close_quotes} packages. The performance requirements for the latter packages are even less stringent. All of these package types provide a high level of safety for the material being transported. In this paper, regulatory tests that are used to demonstrate this safety will be described. The responses of various packages to these tests will be shown. In addition, the response of packages to extra-regulatory tests will be discussed. The results of these tests will be used to demonstrate the high level of safety provided to workers, the public, and the environment by packages used for the transportation of radioactive material.

  6. Potential vertical movement of large heat-generating waste packages in salt.

    SciTech Connect (OSTI)

    Clayton, Daniel James; Martinez, Mario J.; Hardin, Ernest L.

    2013-05-01

    With renewed interest in disposal of heat-generating waste in bedded or domal salt formations, scoping analyses were conducted to estimate rates of waste package vertical movement. Vertical movement is found to result from thermal expansion, from upward creep or heave of the near-field salt, and from downward buoyant forces on the waste package. A two-pronged analysis approach was used, with thermal-mechanical creep modeling, and coupled thermal-viscous flow modeling. The thermal-mechanical approach used well-studied salt constitutive models, while the thermal-viscous approach represented the salt as a highly viscous fluid. The Sierra suite of coupled simulation codes was used for both approaches. The waste package in all simulations was a right-circular cylinder with the density of steel, in horizontal orientation. A time-decaying heat generation function was used to represent commercial spent fuel with typical burnup and 50-year age. Results from the thermal-mechanical base case showed approximately 27 cm initial uplift of the package, followed by gradual relaxation closely following the calculated temperature history. A similar displacement history was obtained with the package density set equal to that of salt. The slight difference in these runs is attributable to buoyant displacement (sinking) and is on the order of 1 mm in 2,000 years. Without heat generation the displacement stabilizes at a fraction of millimeter after a few hundred years. Results from thermal-viscous model were similar, except that the rate of sinking was constant after cooldown, at approximately 0.15 mm per 1,000 yr. In summary, all calculations showed vertical movement on the order of 1 mm or less in 2,000 yr, including calculations using well-established constitutive models for temperature-dependent salt deformation. Based on this finding, displacement of waste packages in a salt repository is not a significant repository performance issue.

  7. Maximim Accelerations On The Fuel Assemblies Of a 21-PWR Waste Package During End Impacts 

    SciTech Connect (OSTI)

    V. DeLa Brosse

    2003-03-27

    The objective of this calculation is to determine the acceleration of the fuel assemblies contained in a 21-Pressurized Water Reactor (PWR) spent nuclear fuel waste package impacting an unyielding surface. A range of initial velocities of the waste package is studied. The scope of this calculation is limited to estimating the acceleration of the fuel assemblies during the impact.

  8. Maximim Accelerations On The Fuel Assemblies Of a 21-PWR Waste Package During End Impacts 

    SciTech Connect (OSTI)

    T. Schmitt

    2005-08-17

    The objective of this calculation is to determine the acceleration of the fuel assemblies contained in a 21-Pressurized Water Reactor (PWR) spent nuclear fuel waste package impacting an unyielding surface. A range of initial velocities of the waste package is studied. The scope of this calculation is limited to estimating the acceleration of the fuel assemblies during the impact.

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

    SciTech Connect (OSTI)

    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.

  10. Plasma vitrification of waste materials

    DOE Patents [OSTI]

    McLaughlin, D.F.; Dighe, S.V.; Gass, W.R.

    1997-06-10

    This invention provides a process wherein hazardous or radioactive wastes in the form of liquids, slurries, or finely divided solids are mixed with finely divided glassformers (silica, alumina, soda, etc.) and injected directly into the plume of a non-transferred arc plasma torch. The extremely high temperatures and heat transfer rates makes it possible to convert the waste-glassformer mixture into a fully vitrified molten glass product in a matter of milliseconds. The molten product may then be collected in a crucible for casting into final wasteform geometry, quenching in water, or further holding time to improve homogeneity and eliminate bubbles. 4 figs.

  11. Plasma vitrification of waste materials

    DOE Patents [OSTI]

    McLaughlin, David F.; Dighe, Shyam V.; Gass, William R.

    1997-01-01

    This invention provides a process wherein hazardous or radioactive wastes in the form of liquids, slurries, or finely divided solids are mixed with finely divided glassformers (silica, alumina, soda, etc.) and injected directly into the plume of a non-transferred arc plasma torch. The extremely high temperatures and heat transfer rates makes it possible to convert the waste-glassformer mixture into a fully vitrified molten glass product in a matter of milliseconds. The molten product may then be collected in a crucible for casting into final wasteform geometry, quenching in water, or further holding time to improve homogeneity and eliminate bubbles.

  12. Radiation Effects in Nuclear Waste Materials

    SciTech Connect (OSTI)

    Weber, William J.

    2005-09-30

    The objective of this project is to develop a fundamental understanding of radiation effects in glasses and ceramics, as well as the influence of solid-state radiation effects on aqueous dissolution kinetics, which may impact the performance of nuclear waste forms and stabilized nuclear materials. This work provides the underpinning science to develop improved glass and ceramic waste forms for the immobilization and disposition of high-level tank waste, excess plutonium, plutonium residues and scrap, other actinides, and other nuclear waste streams. Furthermore, this work is developing develop predictive models for the performance of nuclear waste forms and stabilized nuclear materials. Thus, the research performed under this project has significant implications for the immobilization of High-Level Waste (HLW) and Nuclear Materials, two mission areas within the Office of Environmental Management (EM). With regard to the HLW mission, this research will lead to improved understanding of radiation-induced degradation mechanisms and their effects on dissolution kinetics, as well as development of predictive models for waste form performance. In the Nuclear Materials mission, this research will lead to improvements in the understanding of radiation effects on the chemical and structural properties of materials for the stabilization and long-term storage of plutonium, highly-enriched uranium, and other actinides. The research uses plutonium incorporation, ion-beam irradiation, and electron-beam irradiation to simulate the effects of alpha decay and beta decay on relevant glasses and ceramics. The research under this project has the potential to result in improved glass and ceramic materials for the stabilization and immobilization of high-level tank waste, plutonium residues and scraps, surplus weapons plutonium, highly-enriched uranium, other actinides, and other radioactive materials.

  13. Radiation Effects in Nuclear Waste Materials

    SciTech Connect (OSTI)

    Weber, William J.

    2005-06-01

    The objective of this project is to develop a fundamental understanding of radiation effects in glasses and ceramics, as well as the influence of solid-state radiation effects on aqueous dissolution kinetics, which may impact the performance of nuclear waste forms and stabilized nuclear materials. This work provides the underpinning science to develop improved glass and ceramic waste forms for the immobilization and disposition of high-level tank waste, excess plutonium, plutonium residues and scrap, other actinides, and other nuclear waste streams. Furthermore, this work is developing develop predictive models for the performance of nuclear waste forms and stabilized nuclear materials. Thus, the research performed under this project has significant implications for the immobilization of High-Level Waste (HLW) and Nuclear Materials, two mission areas within the Office of Environmental Management (EM). With regard to the HLW mission, this research will lead to improved understanding of radiation-induced degradation mechanisms and their effects on dissolution kinetics, as well as development of predictive models for waste form performance. In the Nuclear Materials mission, this research will lead to improvements in the understanding of radiation effects on the chemical and structural properties of materials for the stabilization and long-term storage of plutonium, highly-enriched uranium, and other actinides. The research uses plutonium incorporation, ion-beam irradiation, and electron-beam irradiation to simulate the effects of alpha decay and beta decay on relevant glasses and ceramics. The research under this project has the potential to result in improved glass and ceramic materials for the stabilization and immobilization of high-level tank waste, plutonium residues and scraps, surplus weapons plutonium, highly-enriched uranium, other actinides, and other radioactive materials.

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

    SciTech Connect (OSTI)

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

    2010-02-01

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

  15. IN-PACKAGE CHEMISTRY ABSTRACTION

    SciTech Connect (OSTI)

    E. Thomas

    2005-07-14

    This report was developed in accordance with the requirements in ''Technical Work Plan for Postclosure Waste Form Modeling'' (BSC 2005 [DIRS 173246]). The purpose of the in-package chemistry model is to predict the bulk chemistry inside of a breached waste package and to provide simplified expressions of that chemistry as a function of time after breach to Total Systems Performance Assessment for the License Application (TSPA-LA). The scope of this report is to describe the development and validation of the in-package chemistry model. The in-package model is a combination of two models, a batch reactor model, which uses the EQ3/6 geochemistry-modeling tool, and a surface complexation model, which is applied to the results of the batch reactor model. The batch reactor model considers chemical interactions of water with the waste package materials, and the waste form for commercial spent nuclear fuel (CSNF) waste packages and codisposed (CDSP) waste packages containing high-level waste glass (HLWG) and DOE spent fuel. The surface complexation model includes the impact of fluid-surface interactions (i.e., surface complexation) on the resulting fluid composition. The model examines two types of water influx: (1) the condensation of water vapor diffusing into the waste package, and (2) seepage water entering the waste package as a liquid from the drift. (1) Vapor-Influx Case: The condensation of vapor onto the waste package internals is simulated as pure H{sub 2}O and enters at a rate determined by the water vapor pressure for representative temperature and relative humidity conditions. (2) Liquid-Influx Case: The water entering a waste package from the drift is simulated as typical groundwater and enters at a rate determined by the amount of seepage available to flow through openings in a breached waste package.

  16. Waste and Materials Disposition Information | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Waste and Materials Disposition Information Waste and Materials Disposition Information Waste and Materials Disposition Information As the Office of Environmental Management (EM) fulfills its mission, waste and materials disposition plays a vital role in the cleanup of radioactive waste and the environmental legacy of nuclear weapons production and nuclear energy research. Disposal of waste frequently falls on the critical path of cleanup projects. Significant planning resources are spent to

  17. Disposal of LLW and ILW in Germany - Characterisation and Documentation of Waste Packages with Respect to the Change of Requirements

    SciTech Connect (OSTI)

    Bandt, G.; Spicher, G.; Steyer, St.; Brennecke, P.

    2008-07-01

    Since the 1998 termination of LLW and ILW emplacement in the Morsleben repository (ERAM), Germany, the treatment, conditioning and documentation of radioactive waste products and packages have been continued on the basis of the waste acceptance requirements as of 1995, prepared for the Konrad repository near Salzgitter in Lower Saxony, Germany. The resulting waste products and packages are stored in interim storage facilities. Due to the Konrad license issued in 2002 the waste acceptance requirements have to be completed by additional requirements imposed by the licensing authority, e. g. for the declaration of chemical waste package constituents. Therefore, documentation of waste products and packages which are checked by independent experts and are in parts approved by the responsible authority (Office for Radiation Protection, BfS) up to now will have to be checked again for fulfilling the final waste acceptance requirements prior to disposal. In order to simplify these additional checks, databases are used to ensure an easy access to all known facts about the waste packages. A short balance of the existing waste products and packages which are already checked and partly approved by BfS as well as an overview on the established databases ensuring a fast access to the known facts about the conditioning processes is presented. (authors)

  18. Nuclear Materials: Reconsidering Wastes and Assets - 13193

    SciTech Connect (OSTI)

    Michalske, T.A.

    2013-07-01

    The nuclear industry, both in the commercial and the government sectors, has generated large quantities of material that span the spectrum of usefulness, from highly valuable ('assets') to worthless ('wastes'). In many cases, the decision parameters are clear. Transuranic waste and high level waste, for example, have no value, and is either in a final disposition path today, or - in the case of high level waste - awaiting a policy decision about final disposition. Other materials, though discardable, have intrinsic scientific or market value that may be hidden by the complexity, hazard, or cost of recovery. An informed decision process should acknowledge the asset value, or lack of value, of the complete inventory of materials, and the structure necessary to implement the range of possible options. It is important that informed decisions are made about the asset value for the variety of nuclear materials available. For example, there is a significant quantity of spent fuel available for recycle (an estimated $4 billion value in the Savannah River Site's (SRS) L area alone); in fact, SRS has already blended down more than 300 metric tons of uranium for commercial reactor use. Over 34 metric tons of surplus plutonium is also on a path to be used as commercial fuel. There are other radiological materials that are routinely handled at the site in large quantities that should be viewed as strategically important and / or commercially viable. In some cases, these materials are irreplaceable domestically, and failure to consider their recovery could jeopardize our technological leadership or national defense. The inventories of nuclear materials at SRS that have been characterized as 'waste' include isotopes of plutonium, uranium, americium, and helium. Although planning has been performed to establish the technical and regulatory bases for their discard and disposal, recovery of these materials is both economically attractive and in the national interest. (authors)

  19. Packaging and Transfer of Materials of National Security Interest Manual

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2000-09-29

    The purpose of this Technical Manual is to establish requirements for operational safety controls for onsite operations. This Technical Manual provides Department of Energy (DOE) technical safety requirements and policy objectives for development of an onsite packaging and transfer program, pursuant to DOE O 461.1; the DOE contractor must document this program in its onsite packaging and transfer manual/procedures. Does not cancel other directives.

  20. Method for recovering materials from waste

    DOE Patents [OSTI]

    Wicks, G.G.; Clark, D.E.; Schulz, R.L.

    1994-01-01

    A method for recovering metals from metals-containing wastes, a vitrifying the remainder of the wastes for disposal. Metals-containing wastes such as circuit boards, cathode ray tubes, vacuum tubes, transistors and so forth, are broken up and placed in a suitable container. The container is heated by microwaves to a first temperature in the range of approximately 300--800{degrees}C to combust organic materials in the waste, then heated further to a second temperature in the range of approximately 1000--1550{degrees}C at which temperature glass formers present in the waste will cause it to melt and vitrify. Low-melting-point metals such as tin and aluminum can be recovered after organics combustion is substantially complete. Metals with higher melting points, such as gold, silver and copper, can be recovered from the solidified product or separated from the waste at their respective melting points. Network former-containing materials can be added at the start of the process to assist vitrification.

  1. Elucidating the role of interfacial materials properties in microfluidic packages.

    SciTech Connect (OSTI)

    Edwards, Thayne L.

    2013-01-01

    The purpose of this work was to discover a method to investigate the properties of interfaces as described by a numerical physical model. The model used was adopted from literature and applied to a commercially available multiphysics software package. By doing this the internal properties of simple structures could be elucidated and then readily applied to more complex structures such as valves and pumps in laminate microfluidic structures. A numerical finite element multi-scale model of a cohesive interface comprised of heterogeneous material properties was used to elucidate irreversible damage from applied strain energy. An unknown internal state variable was applied to characterize the damage process. Using a constrained blister test, this unknown internal state variable could be determined for an adherend/adhesive/adherend body. This is particularly interesting for laminate systems with microfluidic and microstructures contained within the body. A laminate structure was designed and fabricated that could accommodate a variety of binary systems joined using nearly any technique such as adhesive, welding (solvent, laser, ultrasonic, RF, etc.), or thermal. The adhesive method was the most successful and easy to implement but also one of the more difficult to understand, especially over long periods of time. Welding methods are meant to achieve a bond that is similar to bulk properties and so are easier to predict. However, methods of welding often produce defects in the bonds.. Examples of the test structures used to elucidate the internal properties of the model were shown and demonstrated. The real life examples used this research to improve upon current designs and aided in creating complex structures for sensor and other applications.

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

    SciTech Connect (OSTI)

    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 principle{close_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 & 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).

  3. LONG-TERM CORROSION TESTING OF CANDIDATE MATERIALS FOR HIGH-LEVEL RADIOACTIVE WASTE CONTAINMENT

    SciTech Connect (OSTI)

    Estill, J. C.; Doughty, S.; Gdowski, G. E.; Gordon, S.; King, K.; McCright, R. D.; Wang, F.

    1997-10-01

    Preliminary results are presented from the long-term corrosion test program of candidate materials for the high-level radioactive waste packages that would be emplaced in the potential repository at Yucca Mountain, Nevada. The present waste package design is based on a multi-barrier concept having an inner container of a corrosion resistant material and an outer container of a corrosion allowance material. Test specimens have been exposed to simulated bounding environments that may credibly develop in the vicinity of the waste packages. Corrosion rates have been calculated for weight loss and crevice specimens, and U-bend specimens have been examined for evidence of stress corrosion cracking (SCC). Galvanic testing has been started recently and initial results are forthcoming. Pitting characterization of test specimens will be conducted in the coming year. This test program is expected to continue for a minimum of five years so that long-term corrosion data can be determined to support corrosion model development, performance assessment, and waste package design.

  4. DOE Waste Package Project. Quarterly progress report, October 1, 1993--December 31, 1993

    SciTech Connect (OSTI)

    Ladkany, S.G.

    1993-12-31

    The Waste Package Project research activities continued in all research areas. The areas include: Container structural and stress analysis; Nuclear fission criticality studies; Investigation of canister design concepts and corrosion studies; Heat transfer studies; Fluid flow in porous media and radionuclide transport in near field rock; Studies of stresses and stability of the rock formations resulting from the thermal loading of the fuel elements and the multi tunnel concept being analyzed; Characterization of a Faulted Rock Tunnel Model Using Photoelastic and Finite Element Studies; Experiment studies of the dynamic response of a flexible three-link robot using strain gages and Lagrange polynomials; and Robotic manipulation of the nuclear waste container.

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

    SciTech Connect (OSTI)

    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.

  6. COMPACTION OF FIBERBOARD OVERPACK MATERIALS IN A 9975 SHIPPING PACKAGE

    SciTech Connect (OSTI)

    Stefek, T.; Daugherty, W.; Estochen, E.; Murphy, J.

    2010-05-27

    Compaction of lower layers in the 9975 fiberboard overpack has been observed in packages that contain excess moisture. Dynamic loading of the package during transportation may also contribute to compaction of the fiberboard. This condition is being tested and analyzed to better understand these compaction mechanisms and provide a basis from which to evaluate their impact to the safety basis for transportation (Safety Analysis Report for Packaging) and storage (facility Design Safety Analysis) at the Savannah River Site (SRS). A test program has been developed and is being implemented to identify the extent of the compaction as a function of fiberboard moisture and typical transport dynamic loadings. Test conditions will be compared to regulatory requirements for dynamic loading. Characterization of the recovery of short-term compaction following the application of dynamic loading is also being evaluated. Interim results from this test program will be summarized.

  7. Notice of Intent to Revise DOE O 460.2A, Departmental Materials Transportation and Packaging Management

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2010-11-18

    Establishes requirements and responsibilities for management of Department of Energy (DOE), including National Nuclear Security Administration, materials transportation and packaging and ensures the safe, secure, efficient packaging and transportation of materials, both hazardous and non-hazardous.

  8. DOE nuclear material packaging manual: storage container requirements for plutonium oxide materials

    SciTech Connect (OSTI)

    Veirs, D Kirk

    2009-01-01

    Loss of containment of nuclear material stored in containers such as food-pack cans, paint cans, or taped slip lid cans has generated concern about packaging requirements for interim storage of nuclear materials in working facilities such as the plutonium facility at Los Alamos National Laboratory (LANL). In response, DOE has recently issued DOE M 441.1 'Nuclear Material Packaging Manual' with encouragement from the Defense Nuclear Facilities Safety Board. A unique feature compared to transportation containers is the allowance of filters to vent flammable gases during storage. Defining commonly used concepts such as maximum allowable working pressure and He leak rate criteria become problematic when considering vented containers. Los Alamos has developed a set of container requirements that are in compliance with 441.1 based upon the activity of heat-source plutonium (90% Pu-238) oxide, which bounds the requirements for weapons-grade plutonium oxide. The pre and post drop-test He leak rates depend upon container size as well as the material contents. For containers that are routinely handled, ease of handling and weight are a major consideration. Relatively thin-walled containers with flat bottoms are desired yet they cannot be He leak tested at a differential pressure of one atmosphere due to the potential for plastic deformation of the flat bottom during testing. The He leak rates and He leak testing configuration for containers designed for plutonium bearing materials will be presented. The approach to meeting the other manual requirements such as corrosion and thermal degradation resistance will be addressed. The information presented can be used by other sites to evaluate if their conditions are bounded by LANL requirements when considering procurement of 441.1 compliant containers.

  9. Packaging strategies for printed circuit board components. Volume I, materials & thermal stresses.

    SciTech Connect (OSTI)

    Neilsen, Michael K.; Austin, Kevin N.; Adolf, Douglas Brian; Spangler, Scott W.; Neidigk, Matthew Aaron; Chambers, Robert S.

    2011-09-01

    Decisions on material selections for electronics packaging can be quite complicated by the need to balance the criteria to withstand severe impacts yet survive deep thermal cycles intact. Many times, material choices are based on historical precedence perhaps ignorant of whether those initial choices were carefully investigated or whether the requirements on the new component match those of previous units. The goal of this program focuses on developing both increased intuition for generic packaging guidelines and computational methodologies for optimizing packaging in specific components. Initial efforts centered on characterization of classes of materials common to packaging strategies and computational analyses of stresses generated during thermal cycling to identify strengths and weaknesses of various material choices. Future studies will analyze the same example problems incorporating the effects of curing stresses as needed and analyzing dynamic loadings to compare trends with the quasi-static conclusions.

  10. Implementation Guide for Use with DOE O 460.2 Departmental Materials Transportation and Packaging Management

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1996-11-15

    The purpose of this guide is to assist those responsible for transporting and packaging Department materials, and to provide an understanding of Department policies on activities which supplement regulatory requirements. Does not cancel/supersede other directives.

  11. DRAFT - DOE O 460.1D, Hazardous Materials Packaging and Transportation Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    The Order establishes safety requirements for the proper packaging and transportation of Department of offsite shipments and onsite transfers of radioactive and other hazardous materials, and for modal transportation.

  12. LEVERAGING AGING MATERIALS DATA TO SUPPORT EXTENSION OF TRANSPORTATION SHIPPING PACKAGES SERVICE LIFE

    SciTech Connect (OSTI)

    Dunn, K.; Bellamy, S.; Daugherty, W.; Sindelar, R.; Skidmore, E.

    2013-08-18

    Nuclear material inventories are increasingly being transferred to interim storage locations where they may reside for extended periods of time. Use of a shipping package to store nuclear materials after the transfer has become more common for a variety of reasons. Shipping packages are robust and have a qualified pedigree for performance in normal operation and accident conditions but are only certified over an approved transportation window. The continued use of shipping packages to contain nuclear material during interim storage will result in reduced overall costs and reduced exposure to workers. However, the shipping package materials of construction must maintain integrity as specified by the safety basis of the storage facility throughout the storage period, which is typically well beyond the certified transportation window. In many ways, the certification processes required for interim storage of nuclear materials in shipping packages is similar to life extension programs required for dry cask storage systems for commercial nuclear fuels. The storage of spent nuclear fuel in dry cask storage systems is federally-regulated, and over 1500 individual dry casks have been in successful service up to 20 years in the US. The uncertainty in final disposition will likely require extended storage of this fuel well beyond initial license periods and perhaps multiple re-licenses may be needed. Thus, both the shipping packages and the dry cask storage systems require materials integrity assessments and assurance of continued satisfactory materials performance over times not considered in the original evaluation processes. Test programs for the shipping packages have been established to obtain aging data on materials of construction to demonstrate continued system integrity. The collective data may be coupled with similar data for the dry cask storage systems and used to support extending the service life of shipping packages in both transportation and storage.

  13. Certification of the Mound 1 kW package for shipping of plutonium dioxide source material

    SciTech Connect (OSTI)

    Annese, C.E.; Mount, M.K.

    1994-01-01

    The Department of Energy (DOE) has established procedures for obtaining certification of packagings used by DOE and its contractors for the transport of radioactive materials. Specifically, DOE Orders 5480.3 and 1540.2 provide references for other DOE Orders which must be followed when an applicant submits a Safety Analysis Report for Packaging (SARP). From the orders, Department EH of DOE, has internal oversight responsibility for transportation and Packaging safety; package certification falls under EH responsibility; transportation and packaging safety division in EH certifies packages for DOE; and use of DOE certified packages is authorized by DOT. An independent review of the SARP must confirm that the packaging designs and operations meet safety criteria at least equivalent to these standards. This paper will discuss the independent review process of the shielding section of the Mound 1 kW SARP; describe the geometry of the packaging and the load configurations; discuss the analysis of the various neutron and photon source terms that were used for the load configuration under analysis; and provide illustrations of the use of the monte-carlo code, COG{sup 3}, which was utilized to perform the shielding analysis.

  14. Packaging and Transportation | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Packaging and Transportation Packaging and Transportation Packaging and Transportation Radiological shipments are accomplished safely. Annually, about 400 million hazardous materials shipments occur in the United States by rail, air, sea, and land. Of these shipments, about three million are radiological shipments. Since Fiscal Year (FY) 2004, EM has completed over 150,000 shipments of radioactive material/waste. Please click here to see Office of Packaging and Transportation Fiscal Year 2012

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

    SciTech Connect (OSTI)

    Jardine, L J; Borisov, G B

    2004-07-21

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

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

    SciTech Connect (OSTI)

    Roberts, R.S.; Perkins, C.L.

    1982-11-19

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

  17. Approved reference and testing materials for use in Nuclear Waste Management Research and Development Programs

    SciTech Connect (OSTI)

    Mellinger, G.B.; Daniel, J.L.

    1984-12-01

    This document, addressed to members of the waste management research and development community summarizes reference and testing materials available from the Nuclear Waste Materials Characterization Center (MCC). These materials are furnished under the MCC's charter to distribute reference materials essential for quantitative evaluation of nuclear waste package materials under development in the US. Reference materials with known behavior in various standard waste management related tests are needed to ensure that individual testing programs are correctly performing those tests. Approved testing materials are provided to assist the projects in assembling materials data base of defensible accuracy and precision. This is the second issue of this publication. Eight new Approved Testing Materials are listed, and Spent Fuel is included as a separate section of Standard Materials because of its increasing importance as a potential repository storage form. A summary of current characterization information is provided for each material listed. Future issues will provide updates of the characterization status of the materials presented in this issue, and information about new standard materials as they are acquired. 7 references, 1 figure, 19 tables.

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

    SciTech Connect (OSTI)

    T.L. Lotz

    1997-01-29

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

  19. INVESTIGATION OF THE PRESENCE OF DRUGSTORE BEETLES WITHIN CELOTEX ASSEMBLIES IN RADIOACTIVE MATERIAL PACKAGINGS

    SciTech Connect (OSTI)

    Loftin, B; Glenn Abramczyk, G

    2008-06-04

    During normal operations at the Department of Energy's Hanford Site in Hanford, WA, drugstore beetles, (Stegobium paniceum (L.) Coleoptera: Anobiidae), were found within the fiberboard subassemblies of two 9975 Shipping Packages. Initial indications were that the beetles were feeding on the Celotex{trademark} assemblies within the package. Celotex{trademark} fiberboard is used in numerous radioactive material packages serving as both a thermal insulator and an impact absorber for both normal conditions of transport and hypothetical accident conditions. The Department of Energy's Packaging Certification Program (EM-63) directed a thorough investigation to determine if the drugstore beetles were causing damage that would be detrimental to the safety performance of the Celotex{trademark}. The Savannah River National Laboratory is conducting the investigation with entomological expertise provided by Clemson University. The two empty 9975 shipping packages were transferred to the Savannah River National Laboratory in the fall of 2007. This paper will provide details and results of the ongoing investigation.

  20. RELEASE OF DRIED RADIOACTIVE WASTE MATERIALS TECHNICAL BASIS DOCUMENT

    SciTech Connect (OSTI)

    KOZLOWSKI, S.D.

    2007-05-30

    This technical basis document was developed to support RPP-23429, Preliminary Documented Safety Analysis for the Demonstration Bulk Vitrification System (PDSA) and RPP-23479, Preliminary Documented Safety Analysis for the Contact-Handled Transuranic Mixed (CH-TRUM) Waste Facility. The main document describes the risk binning process and the technical basis for assigning risk bins to the representative accidents involving the release of dried radioactive waste materials from the Demonstration Bulk Vitrification System (DBVS) and to the associated represented hazardous conditions. Appendices D through F provide the technical basis for assigning risk bins to the representative dried waste release accident and associated represented hazardous conditions for the Contact-Handled Transuranic Mixed (CH-TRUM) Waste Packaging Unit (WPU). The risk binning process uses an evaluation of the frequency and consequence of a given representative accident or represented hazardous condition to determine the need for safety structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls. A representative accident or a represented hazardous condition is assigned to a risk bin based on the potential radiological and toxicological consequences to the public and the collocated worker. Note that the risk binning process is not applied to facility workers because credible hazardous conditions with the potential for significant facility worker consequences are considered for safety-significant SSCs and/or TSR-level controls regardless of their estimated frequency. The controls for protection of the facility workers are described in RPP-23429 and RPP-23479. Determination of the need for safety-class SSCs was performed in accordance with DOE-STD-3009-94, Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses, as described below.

  1. Regulatory compliance in the design of packages used to transport radioactive materials

    SciTech Connect (OSTI)

    Raske, D.T.

    1993-06-01

    Shipments of radioactive materials within the regulatory jurisdiction of the US Department of Energy (DOE) must meet the package design requirements contained in Title 10 of the Code of Federal Regulations, Part 71, and DOE Order 5480.3. These regulations do not provide design criteria requirements, but only detail the approval standards, structural performance criteria, and package integrity requirements that must be met during transport. The DOE recommended design criterion for high-level Category I radioactive packagings is Section III, Division 1, of the ASME Boiler and Pressure Vessel Code. However, alternative design criteria may be used if all the design requirements are satisfied. The purpose of this paper is to review alternatives to the Code criteria and discuss their applicability to the design of containment vessels in packages for high-level radioactive materials. Issues such as design qualification by physical testing, the use of scale models, and problems encountered using a non-ASME design approach are addressed.

  2. APPLICATION OF POLYURETHANE FOAM FOR IMPACT ABSORPTION AND THERMAL INSULATION FOR GENERAL PURPOSE RADIOACTIVE MATERIALS PACKAGINGS

    SciTech Connect (OSTI)

    Smith, A; Glenn Abramczyk, G; Paul Blanton, P; Steve Bellamy, S; William Daugherty, W; Sharon Williamson, S

    2009-02-18

    Polyurethane foam has been employed in impact limiters for large radioactive materials packagings since the early 1980's. Its consistent crush response, controllable structural properties and excellent thermal insulating characteristics have made it attractive as replacement for the widely used cane fiberboard for smaller, drum size packagings. Accordingly, polyurethane foam was chosen for the overpack material for the 9977 and 9978 packagings. The study reported here was undertaken to provide data to support the analyses performed as part of the development of the 9977 and 9978, and compared property values reported in the literature with published property values and test results for foam specimens taken from a prototype 9977 packaging. The study confirmed that, polyurethane foam behaves in a predictable and consistent manner and fully satisfies the functional requirements for impact absorption and thermal insulation.

  3. Directory of certificiates of compliance for radioactive materials packages: Report of NRC approved packages. Revision 19, Volume 1

    SciTech Connect (OSTI)

    1996-10-01

    This directory provides information on packagings approved by the U.S. Nuclear Regulatory Commission.

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

    SciTech Connect (OSTI)

    T. Wolery

    2005-02-22

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

  5. Evaluation of the corrosivity of dust deposited on waste packages at Yucca Mountain, Nevada.

    SciTech Connect (OSTI)

    Wolery, Thomas J.; Bryan, Charles R.; Barr, Deborah; Jarek, Russell L.; Hardin, Ernest L.; Shields, David R.; Sutton, Mark A.

    2005-08-01

    Potentially corrosive brines can form during post-closure by deliquescence of salt minerals in dust deposited on the surface of waste packages at Yucca Mountain during operations and the pre-closure ventilation period. Although thermodynamic modeling and experimental studies of brine deliquescence indicates that brines are likely to form, they will be nitrate-rich and noncorrosive. Processes that modify the brines following deliquescence are beneficial with respect to inhibition of corrosion. For example, acid degassing (HCl, HNO{sub 3}) could dry out brines, but kinetic limitations are likely to limit the effect to increasing their passivity by raising the pH and increasing the NO{sub 3}/Cl ratio. Predicted dust quantities and maximum brine volumes on the waste package surface are small, and physical isolation of salt minerals in the dust may inhibit formation of eutectic brines and decrease brine volumes. If brines do contact the WP surface, small droplet volumes and layer thicknesses do not support development of diffusive gradients necessary for formation on separate anodic-cathodic zones required for localized corrosion. Finally, should localized corrosion initiate, corrosion product buildup will stifle corrosion, by limiting oxygen access to the metal surface, by capillary retention of brine in corrosion product porosity, or by consumption of brine components (Cl{sup -}).

  6. Packaging and Transfer or Transportation of Materials of National Security Interest

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2000-09-29

    To establish requirements and responsibilities for the Transportation Safeguards System (TSS) packaging and transportation and onsite transfer of nuclear explosives, nuclear components, Naval nuclear fuel elements, Category I and Category II special nuclear materials, special assemblies, and other materials of national security interest. Cancels: DOE 5610.12 and DOE 5610.14.

  7. Molten salt destruction of energetic waste materials

    DOE Patents [OSTI]

    Brummond, William A.; Upadhye, Ravindra S.; Pruneda, Cesar O.

    1995-01-01

    A molten salt destruction process is used to treat and destroy energetic waste materials such as high explosives, propellants, and rocket fuels. The energetic material is pre-blended with a solid or fluid diluent in safe proportions to form a fluid fuel mixture. The fuel mixture is rapidly introduced into a high temperature molten salt bath. A stream of molten salt is removed from the vessel and may be recycled as diluent. Additionally, the molten salt stream may be pumped from the reactor, circulated outside the reactor for further processing, and delivered back into the reactor or cooled and circulated to the feed delivery system to further dilute the fuel mixture entering the reactor.

  8. Molten salt destruction of energetic waste materials

    DOE Patents [OSTI]

    Brummond, W.A.; Upadhye, R.S.; Pruneda, C.O.

    1995-07-18

    A molten salt destruction process is used to treat and destroy energetic waste materials such as high explosives, propellants, and rocket fuels. The energetic material is pre-blended with a solid or fluid diluent in safe proportions to form a fluid fuel mixture. The fuel mixture is rapidly introduced into a high temperature molten salt bath. A stream of molten salt is removed from the vessel and may be recycled as diluent. Additionally, the molten salt stream may be pumped from the reactor, circulated outside the reactor for further processing, and delivered back into the reactor or cooled and circulated to the feed delivery system to further dilute the fuel mixture entering the reactor. 4 figs.

  9. Material Recovery and Waste Form Development FY 2014 Accomplishments Report

    SciTech Connect (OSTI)

    Lori Braase

    2014-11-01

    Develop advanced nuclear fuel cycle separation and waste management technologies that improve current fuel cycle performance and enable a sustainable fuel cycle, with minimal processing, waste generation, and potential for material diversion.

  10. Determination of Fire Enviroment in Stacked Cargo Containers with Radioactive Materials Packages

    SciTech Connect (OSTI)

    Arviso, M.; Bobbe, J.G.; Dukart, R.D.; Koski, J.A.

    1999-05-01

    Results from a Fire Test with a three-by-three stack of standard 6 m long International Standards Organization shipping containers containing combustible fuels and empty radioactive materials packages are reported and discussed. The stack is intended to simulate fire conditions that could occur during on-deck stowage on container cargo ships. The fire is initated by locating the container stack adjacent to a 9.8 x 6 m pool fire. Temperatures of both cargoes (empty and simulated radioactive materials packages) and containers are recorded and reported. Observations on the duration, intensity and spread of the fire are discussed. Based on the results, models for simulation of fire exposure of radioactive materials packages in such fires are suggested.

  11. Terminating Safeguards on Excess Special Nuclear Material: Defense TRU Waste Clean-up and Nonproliferation - 12426

    SciTech Connect (OSTI)

    Hayes, Timothy; Nelson, Roger

    2012-07-01

    The Department of Energy (DOE) and the National Nuclear Security Administration (NNSA) manages defense nuclear material that has been determined to be excess to programmatic needs and declared waste. When these wastes contain plutonium, they almost always meet the definition of defense transuranic (TRU) waste and are thus eligible for disposal at the Waste Isolation Pilot Plant (WIPP). The DOE operates the WIPP in a manner that physical protections for attractiveness level D or higher special nuclear material (SNM) are not the normal operating condition. Therefore, there is currently a requirement to terminate safeguards before disposal of these wastes at the WIPP. Presented are the processes used to terminate safeguards, lessons learned during the termination process, and how these approaches might be useful for future defense TRU waste needing safeguards termination prior to shipment and disposal at the WIPP. Also described is a new criticality control container, which will increase the amount of fissile material that can be loaded per container, and how it will save significant taxpayer dollars. Retrieval, compliant packaging and shipment of retrievably stored legacy TRU waste has dominated disposal operations at WIPP since it began operations 12 years ago. But because most of this legacy waste has successfully been emplaced in WIPP, the TRU waste clean-up focus is turning to newly-generated TRU materials. A major component will be transuranic SNM, currently managed in safeguards-protected vaults around the weapons complex. As DOE and NNSA continue to consolidate and shrink the weapons complex footprint, it is expected that significant quantities of transuranic SNM will be declared surplus to the nation's needs. Safeguards termination of SNM varies due to the wide range of attractiveness level of the potential material that may be directly discarded as waste. To enhance the efficiency of shipping waste with high TRU fissile content to WIPP, DOE designed an over

  12. 9975 SHIPPING PACKAGE PERFORMANCE OF ALTERNATE MATERIALS FOR LONG-TERM STORAGE APPLICATION

    SciTech Connect (OSTI)

    Skidmore, E.; Hoffman, E.; Daugherty, W.

    2010-02-24

    The Model 9975 shipping package specifies the materials of construction for its various components. With the loss of availability of material for two components (cane fiberboard overpack and Viton{reg_sign} GLT O-rings), alternate materials of construction were identified and approved for use for transport (softwood fiberboard and Viton{reg_sign} GLT-S O-rings). As these shipping packages are part of a long-term storage configuration at the Savannah River Site, additional testing is in progress to verify satisfactory long-term performance of the alternate materials under storage conditions. The test results to date can be compared to comparable results on the original materials of construction to draw preliminary conclusions on the performance of the replacement materials.

  13. Application of the ASME code in designing containment vessels for packages used to transport radioactive materials

    SciTech Connect (OSTI)

    Raske, D.T.; Wang, Z.

    1992-07-01

    The primary concern governing the design of shipping packages containing radioactive materials is public safety during transport. When these shipments are within the regulatory jurisdiction of the US Department of Energy, the recommended design criterion for the primary containment vessel is either Section III or Section VIII, Division 1, of the ASME Boiler and Pressure Vessel Code, depending on the activity of the contents. The objective of this paper is to discuss the design of a prototypic containment vessel representative of a packaging for the transport of high-level radioactive material.

  14. Removal of radioactive and other hazardous material from fluid waste

    DOE Patents [OSTI]

    Tranter, Troy J.; Knecht, Dieter A.; Todd, Terry A.; Burchfield, Larry A.; Anshits, Alexander G.; Vereshchagina, Tatiana; Tretyakov, Alexander A.; Aloy, Albert S.; Sapozhnikova, Natalia V.

    2006-10-03

    Hollow glass microspheres obtained from fly ash (cenospheres) are impregnated with extractants/ion-exchangers and used to remove hazardous material from fluid waste. In a preferred embodiment the microsphere material is loaded with ammonium molybdophosphonate (AMP) and used to remove radioactive ions, such as cesium-137, from acidic liquid wastes. In another preferred embodiment, the microsphere material is loaded with octyl(phenyl)-N-N-diisobutyl-carbamoylmethylphosphine oxide (CMPO) and used to remove americium and plutonium from acidic liquid wastes.

  15. DRAFT - DOE O 461.1C, Packaging and Transportation for Offsite Shipment of Materials of National Security Interest

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    The Order establishes requirements and responsibilities for ensuring the safety of packaging and transportation for offsite shipments of Materials of National Security Interest.

  16. In-Package Chemistry Abstraction

    SciTech Connect (OSTI)

    E. Thomas

    2004-11-09

    This report was developed in accordance with the requirements in ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]). The purpose of the in-package chemistry model is to predict the bulk chemistry inside of a breached waste package and to provide simplified expressions of that chemistry as function of time after breach to Total Systems Performance Assessment for the License Application (TSPA-LA). The scope of this report is to describe the development and validation of the in-package chemistry model. The in-package model is a combination of two models, a batch reactor model that uses the EQ3/6 geochemistry-modeling tool, and a surface complexation model that is applied to the results of the batch reactor model. The batch reactor model considers chemical interactions of water with the waste package materials and the waste form for commercial spent nuclear fuel (CSNF) waste packages and codisposed waste packages that contain both high-level waste glass (HLWG) and DOE spent fuel. The surface complexation model includes the impact of fluid-surface interactions (i.e., surface complexation) on the resulting fluid composition. The model examines two types of water influx: (1) the condensation of water vapor that diffuses into the waste package, and (2) seepage water that enters the waste package from the drift as a liquid. (1) Vapor Influx Case: The condensation of vapor onto the waste package internals is simulated as pure H2O and enters at a rate determined by the water vapor pressure for representative temperature and relative humidity conditions. (2) Water Influx Case: The water entering a waste package from the drift is simulated as typical groundwater and enters at a rate determined by the amount of seepage available to flow through openings in a breached waste package. TSPA-LA uses the vapor influx case for the nominal scenario for simulations where the waste package has been

  17. Production of methane by anaerobic fermentation of waste materials

    SciTech Connect (OSTI)

    Hitzman, D.O.

    1989-01-17

    This patent describes an apparatus for producing methane by anaerobic fermentation of waste material, comprising: cavity means in the earth for holding a quantity of the waste material; means for covering a quantity of the waste material in the cavity means and thereby separating the quantity of the waste material from the atmosphere; first conduit means communicating between the waste material in the cavity means and a location remote from the cavity means for conveying gas comprising carbon dioxide and methane from the cavity means to the location; gas separation means communicating with the first conduit means at the location for separating carbon dioxide from methane, the first conduit means including at least one pipe having a plurality of apertures therein and disposed in the cavity means extending into and in fluid flow communication with the waste material for receiving gas liberated by the anaerobic fermentation of the waste material and comprising carbon dioxide and methane, through the apertures therein for conveyance via the first conduit means to the gas separation means; second conduit means communicating between the gas separation means and the waste material in the cavity means for conveying carbon dioxide from the gas separation means to the waste material; and third conduit means communicating with the gas separation means for conveying methane from the gas separation means.

  18. An analysis of the qualification criteria for small radioactive material shipping packages

    SciTech Connect (OSTI)

    McClure, J.D.

    1983-05-01

    The RAM package design certification process has two important elements, testing and acceptance. These terms sound very similar but they have specific meanings. Qualification testing in the context of this study is the imposition of simulated accident test conditions upon the candidate package design. (Normal transportation environments may also be included.) Following qualification testing, the acceptance criteria provide the performance levels which, if demonstrated, indicate the ability of the RAM package to sustain the severity of the qualification testing sequence and yet maintain specified levels of package integrity. This study has used Severities of Transportation Accidents as a data base to examine the regulatory test criteria which are required to be met by small packages containing Type B quantities of radioactive material (RAM). The basic findings indicate that the present regulatory test standards provide significantly higher levels of protection for the surface transportation modes (truck, rail) than for RAM packages shipped by aircraft. It should also be noted that various risk assessment studies have shown that the risk to the public due to severe transport accidents by surface and air transport modes is very low. A key element in this study was the quantification of the severity of the transportation accident environment and the severity of the present qualification test standards (called qualification test standards in this document) so that a direct comparison could be made between them to assess the effectiveness of the existing qualification test standards. The manner in which this was accomplished is described.

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

    SciTech Connect (OSTI)

    J.W. Davis

    2005-12-15

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

  20. Packaging and Transportation for Offsite Shipment of Materials of National Security Interest

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2016-07-20

    The purpose of this Order is to make clear that the packaging and transportation of all offsite shipments of Materials of National Security Interest (MNSI) for DOE must be conducted in accordance with Department of Transportation (DOT) and Nuclear Regulatory Commission (NRC) regulations that would be applicable to comparable commercial shipments, except where an alternative course of action is identified in this Order.

  1. DEVELOPMENT OF BURN TEST SPECIFICATIONS FOR FIRE PROTECTION MATERIALS IN RAM PACKAGES

    SciTech Connect (OSTI)

    Gupta, N.

    2010-03-03

    The regulations in 10 CFR 71 require that the radioactive material (RAM) packages must be able to withstand specific fire conditions given in 10 CFR 71.73 during Hypothetical Accident Conditions (HAC). This requirement is normally satisfied by extensive testing of full scale test specimens under required test conditions. Since fire test planning and execution is expensive and only provides a single snapshot into a package performance, every effort is made to minimize testing and supplement tests with results from computational thermal models. However, the accuracy of such thermal models depends heavily on the thermal properties of the fire insulating materials that are rarely available at the regulatory fire temperatures. To the best of authors knowledge no test standards exist that could be used to test the insulating materials and derive their thermal properties for the RAM package design. This paper presents a review of the existing industry fire testing standards and proposes testing methods that could serve as a standardized specification for testing fire insulating materials for use in RAM packages.

  2. System for chemically digesting low level radioactive, solid waste material

    DOE Patents [OSTI]

    Cowan, Richard G.; Blasewitz, Albert G.

    1982-01-01

    An improved method and system for chemically digesting low level radioactive, solid waste material having a high through-put. The solid waste material is added to an annular vessel (10) substantially filled with concentrated sulfuric acid. Concentrated nitric acid or nitrogen dioxide is added to the sulfuric acid within the annular vessel while the sulfuric acid is reacting with the solid waste. The solid waste is mixed within the sulfuric acid so that the solid waste is substantilly fully immersed during the reaction. The off gas from the reaction and the products slurry residue is removed from the vessel during the reaction.

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

    SciTech Connect (OSTI)

    Marusich, Robert M.

    2012-01-25

    A set of steady state diffusion flow equations, for the hydrogen diffusion from one bag to the next bag (or one plastic waste container to another), within a set of nested waste bags (or nested waste containers), are developed and presented. The input data is then presented and justified. Inputting the data for each volume and solving these equations yields the steady state hydrogen concentration in each volume. The input data (permeability of the bag surface and closure, dimensions and hydrogen generation rate) and equations are analyzed to obtain the hydrogen concentrations in the innermost container for a set of containers which are analyzed for the TRUCON code for the general waste containers and the TRUCON code for the Pencil Tank Overpacks (PTO) in a Standard Waste Box (SWB).

  4. Diffusion and Leaching Behavior of Radionuclides in Category 3 Waste Encasement Concrete and Soil Fill Material Summary Report

    SciTech Connect (OSTI)

    Mattigod, Shas V.; Wellman, Dawn M.; Bovaird, Chase C.; Parker, Kent E.; Clayton, Libby N.; Powers, Laura; Recknagle, Kurtis P.; Wood, Marcus I.

    2011-08-31

    One of the methods being considered for safely disposing of Category 3 low-level radioactive wastes is to encase the waste in concrete. Such concrete encasement would contain and isolate the waste packages from the hydrologic environment and would act as an intrusion barrier. The current plan for waste isolation consists of stacking low-level waste packages on a trench floor, surrounding the stacks with reinforced steel, and encasing these packages in concrete. These concrete-encased waste stacks are expected to vary in size with maximum dimensions of 6.4 m long, 2.7 m wide, and 4 m high. The waste stacks are expected to have a surrounding minimum thickness of 15 cm of concrete encasement. These concrete-encased waste packages are expected to withstand environmental exposure (solar radiation, temperature variations, and precipitation) until an interim soil cover or permanent closure cover is installed, and to remain largely intact thereafter. Any failure of concrete encasement may result in water intrusion and consequent mobilization of radionuclides from the waste packages. The mobilized radionuclides may escape from the encased concrete by mass flow and/or diffusion and move into the surrounding subsurface environment. Therefore, it is necessary to assess the performance of the concrete encasement structure and the ability of the surrounding soil to retard radionuclide migration. The retardation factors for radionuclides contained in the waste packages can be determined from measurements of diffusion coefficients for these contaminants through concrete and fill material. Some of the mobilization scenarios include (1) potential leaching of waste form before permanent closure cover is installed; (2) after the cover installation, long-term diffusion of radionuclides from concrete waste form into surrounding fill material; (3) diffusion of radionuclides from contaminated soils into adjoining concrete encasement and clean fill material. Additionally, the rate of

  5. ENERGY STAR Webinar: Introducing Waste & Materials Tracking in Portfolio

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Manager | Department of Energy Introducing Waste & Materials Tracking in Portfolio Manager ENERGY STAR Webinar: Introducing Waste & Materials Tracking in Portfolio Manager August 18, 2016 2:00PM to 3:30PM EDT Hosted by the U.S. Environmental Protection Agency's (EPA's) ENERGY STAR, this webinar will show how to track your building's waste and materials management. It will introduce the types of waste and disposal methods attendees can track, new metrics and reports, and how you can

  6. THERMAL EVALUATION OF THE USE OF BWR MOX SNF IN THE WASTE PACKAGE DESIGN (SCPB: N/A)

    SciTech Connect (OSTI)

    H. Wang

    1997-01-23

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) as specified in the Waste Package Implementation Plan (pp. 4-8,4-11,4-24, 5-1, and 5-13; Ref. 5.10) and Waste Package Plan (pp. 3-15,3-17, and 3-24; Ref. 5.9). The design data request addressed herein is: (1) Characterize the conceptual 40 BWR and 24 BWR Multi-Purpose Canister (MPC) Waste Package (WP) design to show that the design is feasible for use in the MGDS environment when loaded with BWR MOX SNF. (2) Characterize the conceptual 44 BWR and 24 BWR Uncanistered Fuel (UCF) Waste Package (WP) design to show that the design is feasible for use in the MGDS environment when loaded with BWR MOX SNF. The purpose of this analysis is to respond to a concern that the long-term disposal thermal issues for the WP Design, if used with SNF designed for a MOX fuel cycle, do not preclude WP compatibility with the MGDS. The objective of this analysis is to provide thermal parameter information for the conceptual WP design with disposal container which is loaded with BWR MOX SNF under nominal MGDS repository conditions. The results are intended to show that the design has a reasonable chance to meet the MGDS design requirements for normal MGDS operation, and to provide the required guidance to determining the major design issues for future design efforts, and to show that the BWR MOX SNF loaded WP performance is similar to an WP loaded with commercial BWR SNF.

  7. A COMPARISON OF TWO THERMAL INSULATION AND STRUCTURAL MATERIALS FOR USE IN TYPE B PACKAGINGS

    SciTech Connect (OSTI)

    Blanton, P.; Eberl, K.

    2010-07-16

    This paper presents the summary of design features and test results of two Type B Shipping Package prototype configurations comprising different insulating materials developed by the Savannah River National Laboratory (SRNL) for the Department of Energy. The materials evaluated, a closed-cell polyurethane foam and a vacuformed ceramic fiber material, were selected to provide adequate structural protection to the package containment vessel during Normal Conditions of Transport (NCT) and Hypothetical Accident Condition (HAC) events and to provide thermal protection during the HAC fire. Polyurethane foam has been used in shipping package designs for many years because of the stiffness it provides to the structure and because of the thermal protection it provides during fire scenarios. This comparison describes how ceramic fiber material offers an alternative to the polyurethane foam in a specific overpack design. Because of the high operating temperature ({approx}2,300 F) of the ceramic material, it allows for contents with higher heat loads to be shipped than is possible with polyurethane foam. Methods of manufacturing and design considerations using the two materials will be addressed.

  8. Waste Form Degradation Model Integration for Engineered Materials

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Performance | Department of Energy Waste Form Degradation Model Integration for Engineered Materials Performance Waste Form Degradation Model Integration for Engineered Materials Performance The collaborative approach to the glass and metallic waste form degradation modeling activities includes process model development (including first-principles approaches) and model integration-both internally among developed process models and between developed process models and PA models, and cross

  9. Safety evaluation for packaging 222-S laboratory cargo tank for onetime type B material shipment

    SciTech Connect (OSTI)

    Nguyen, P.M.

    1994-08-19

    The purpose of this Safety Evaluation for Packaging (SEP) is to evaluate and document the safety of the onetime shipment of bulk radioactive liquids in the 222-S Laboratory cargo tank (222-S cargo tank). The 222-S cargo tank is a US Department of Transportation (DOT) MC-312 specification (DOT 1989) cargo tank, vehicle registration number HO-64-04275, approved for low specific activity (LSA) shipments in accordance with the DOT Title 49, Code of Federal Regulations (CFR). In accordance with the US Department of Energy, Richland Operations Office (RL) Order 5480.1A, Chapter III (RL 1988), an equivalent degree of safety shall be provided for onsite shipments as would be afforded by the DOT shipping regulations for a radioactive material package. This document demonstrates that this packaging system meets the onsite transportation safety criteria for a onetime shipment of Type B contents.

  10. Material and energy recovery in integrated waste management systems: Project overview and main results

    SciTech Connect (OSTI)

    Consonni, Stefano; Giugliano, Michele; Massarutto, Antonio; Saccani, Cesare

    2011-09-15

    Highlights: > The source separation level (SSL) of waste management system does not qualify adequately the system. > Separately collecting organic waste gives less advantages than packaging materials. > Recycling packaging materials (metals, glass, plastics, paper) is always attractive. > Composting and anaerobic digestion of organic waste gives questionable outcomes. > The critical threshold of optimal recycling seems to be a SSL of 50%. - Abstract: This paper describes the context, the basic assumptions and the main findings of a joint research project aimed at identifying the optimal breakdown between material recovery and energy recovery from municipal solid waste (MSW) in the framework of integrated waste management systems (IWMS). The project was carried out from 2007 to 2009 by five research groups at Politecnico di Milano, the Universities of Bologna and Trento, and the Bocconi University (Milan), with funding from the Italian Ministry of Education, University and Research (MIUR). Since the optimization of IWMSs by analytical methods is practically impossible, the search for the most attractive strategy was carried out by comparing a number of relevant recovery paths from the point of view of mass and energy flows, technological features, environmental impact and economics. The main focus has been on mature processes applicable to MSW in Italy and Europe. Results show that, contrary to a rather widespread opinion, increasing the source separation level (SSL) has a very marginal effects on energy efficiency. What does generate very significant variations in energy efficiency is scale, i.e. the size of the waste-to-energy (WTE) plant. The mere value of SSL is inadequate to qualify the recovery system. The energy and environmental outcome of recovery depends not only on 'how much' source separation is carried out, but rather on 'how' a given SSL is reached.

  11. Practical Thermal Evaluation Methods For HAC Fire Analysis In Type B Radiaoactive Material (RAM) Packages

    SciTech Connect (OSTI)

    Abramczyk, Glenn; Hensel, Stephen J; Gupta, Narendra K.

    2013-03-28

    Title 10 of the United States Code of Federal Regulations Part 71 for the Nuclear Regulatory Commission (10 CFR Part 71.73) requires that Type B radioactive material (RAM) packages satisfy certain Hypothetical Accident Conditions (HAC) thermal design requirements to ensure package safety during accidental fire conditions. Compliance with thermal design requirements can be met by prototype tests, analyses only or a combination of tests and analyses. Normally, it is impractical to meet all the HAC using tests only and the analytical methods are too complex due to the multi-physics non-linear nature of the fire event. Therefore, a combination of tests and thermal analyses methods using commercial heat transfer software are used to meet the necessary design requirements. The authors, along with his other colleagues at Savannah River National Laboratory in Aiken, SC, USA, have successfully used this 'tests and analyses' approach in the design and certification of several United States' DOE/NNSA certified packages, e.g. 9975, 9977, 9978, 9979, H1700, and Bulk Tritium Shipping Package (BTSP). This paper will describe these methods and it is hoped that the RAM Type B package designers and analysts can use them for their applications.

  12. Greater-than-Class C low-level radioactive waste shipping package/container identification and requirements study. National Low-Level Waste Management Program

    SciTech Connect (OSTI)

    Tyacke, M.

    1993-08-01

    This report identifies a variety of shipping packages (also referred to as casks) and waste containers currently available or being developed that could be used for greater-than-Class C (GTCC) low-level waste (LLW). Since GTCC LLW varies greatly in size, shape, and activity levels, the casks and waste containers that could be used range in size from small, to accommodate a single sealed radiation source, to very large-capacity casks/canisters used to transport or dry-store highly radioactive spent fuel. In some cases, the waste containers may serve directly as shipping packages, while in other cases, the containers would need to be placed in a transport cask. For the purpose of this report, it is assumed that the generator is responsible for transporting the waste to a Department of Energy (DOE) storage, treatment, or disposal facility. Unless DOE establishes specific acceptance criteria, the receiving facility would need the capability to accept any of the casks and waste containers identified in this report. In identifying potential casks and waste containers, no consideration was given to their adequacy relative to handling, storage, treatment, and disposal. Those considerations must be addressed separately as the capabilities of the receiving facility and the handling requirements and operations are better understood.

  13. APPLICATION FO FLOW FORMING FOR USE IN RADIOACTIVE MATERIAL PACKAGING DESIGNS

    SciTech Connect (OSTI)

    Blanton, P.; Eberl, K.; Abramczyk, G.

    2012-07-11

    This paper reports on the development and testing performed to demonstrate the use of flow forming as an alternate method of manufacturing containment vessels for use in radioactive material shipping packaging designs. Additionally, ASME Boiler and Pressure Vessel Code, Section III, Subsection NB compliance along with the benefits compared to typical welding of containment vessels will be discussed. SRNL has completed fabrication development and the testing on flow formed containment vessels to demonstrate the use of flow forming as an alternate method of manufacturing a welded 6-inch diameter containment vessel currently used in the 9975 and 9977 radioactive material shipping packaging. Material testing and nondestructive evaluation of the flow formed parts demonstrate compliance to the minimum material requirements specified in applicable parts of ASME Boiler and Pressure Vessel Code, Section II. Destructive burst testing shows comparable results to that of a welded design. The benefits of flow forming as compared to typical welding of containment vessels are significant: dimensional control is improved due to no weld distortion; less final machining; weld fit-up issues associated with pipes and pipe caps are eliminated; post-weld non-destructive testing (i.e., radiography and die penetrant tests) is not necessary; and less fabrication steps are required. Results presented in this paper indicate some of the benefits in adapting flow forming to design of future radioactive material shipping packages containment vessels.

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

    SciTech Connect (OSTI)

    Vinson, D.W.; Bullen, D.B.

    1995-09-22

    One of the most significant factors impacting the performance of waste package container materials under repository relevant conditions is the thermal environment. This environment will be affected by the areal power density of the repository, which is dictated by facility design, and the dominant heat transfer mechanism at the site. The near-field environment will evolve as radioactive decay decreases the thermal output of each waste package. Recent calculations (Buscheck and Nitao, 1994) have addressed the importance of thermal loading conditions on waste package performance at the Yucca Mountain site. If a relatively low repository thermal loading design is employed, the temperature and relative humidity near the waste package may significantly affect the degradation of corrosion allowance barriers due to moist air oxidation and radiolytically enhanced corrosion. The purpose this report is to present a literature review of the potential degradation modes for moderately corrosion resistant nickel copper and nickel based candidate materials that may be applicable as alternate barriers for the ACD systems in the Yucca Mountain environment. This report presents a review of the corrosion of nickel-copper alloys, summaries of experimental evaluations of oxidation and atmospheric corrosion in nickel-copper alloys, views of experimental studies of aqueous corrosion in nickel copper alloys, a brief review of galvanic corrosion effects and a summary of stress corrosion cracking in these alloys.

  15. Integrating Volume Reduction and Packaging Alternatives to Achieve Cost Savings for Low Level Waste Disposal at the Rocky Flats Environmental Technology Site

    SciTech Connect (OSTI)

    Church, A.; Gordon, J.; Montrose, J. K.

    2002-02-26

    In order to reduce costs and achieve schedules for Closure of the Rocky Flats Environmental Technology Site (RFETS), the Waste Requirements Group has implemented a number of cost saving initiatives aimed at integrating waste volume reduction with the selection of compliant waste packaging methods for the disposal of RFETS low level radioactive waste (LLW). Waste Guidance Inventory and Shipping Forecasts indicate that over 200,000 m3 of low level waste will be shipped offsite between FY2002 and FY2006. Current projections indicate that the majority of this waste will be shipped offsite in an estimated 40,000 55-gallon drums, 10,000 metal and plywood boxes, and 5000 cargo containers. Currently, the projected cost for packaging, shipment, and disposal adds up to $80 million. With these waste volume and cost projections, the need for more efficient and cost effective packaging and transportation options were apparent in order to reduce costs and achieve future Site packaging a nd transportation needs. This paper presents some of the cost saving initiatives being implemented for waste packaging at the Rocky Flats Environmental Technology Site (the Site). There are many options for either volume reduction or alternative packaging. Each building and/or project may indicate different preferences and/or combinations of options.

  16. Some Materials Degradation Issues in the U.S. High-Level Nuclear Waste Repository Study (The Yucca Mountain Project)

    SciTech Connect (OSTI)

    F. Hua; P. Pasupathi; N. Brown; K. Mon

    2005-09-19

    The safe disposal of radioactive waste requires that the waste be isolated from the environment until radioactive decay has reduced its toxicity to innocuous levels for plants, animals, and humans. All of the countries currently studying the options for disposing of high-level nuclear waste (HLW) have selected deep geologic formations to be the primary barrier for accomplishing this isolation. In U.S.A., the Nuclear Waste Policy Act of 1982 (as amended in 1987) designated Yucca Mountain in Nevada as the potential site to be characterized for high-level nuclear waste (HLW) disposal. Long-term containment of waste and subsequent slow release of radionuclides into the geosphere will rely on a system of natural and engineered barriers including a robust waste containment design. The waste package design consists of a highly corrosion resistant Ni-based Alloy 22 cylindrical barrier surrounding a Type 316 stainless steel inner structural vessel. The waste package is covered by a mailbox-shaped drip shield composed primarily of Ti Grade 7 with Ti Grade 24 structural support members. The U.S. Yucca Mountain Project has been studying and modeling the degradation issues of the relevant materials for some 20 years. This paper reviews the state-of-the-art understanding of the degradation processes based on the past 20 years studies on Yucca Mountain Project (YMP) materials degradation issues with focus on interaction between the in-drift environmental conditions and long-term materials degradation of waste packages and drip shields within the repository system during the 10,000 years regulatory period. This paper provides an overview of the current understanding of the likely degradation behavior of the waste package and drip shield in the repository after the permanent closure of the facility. The degradation scenario discussed in this paper include aging and phase instability, dry oxidation, general and localized corrosion, stress corrosion cracking and hydrogen induced

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

    SciTech Connect (OSTI)

    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

  18. MATERIALS COMPATIBILITY OF SNAP FUEL COMPONENTS DURING SHIPMENT IN 9975 PACKAGING

    SciTech Connect (OSTI)

    Vormelker, P

    2006-11-14

    Materials Science and Technology has evaluated materials compatibility for the SNAP (Systems for Nuclear Auxiliary Power) fuel for containment within a 9975 packaging assembly for a shipping period of one year. The evaluation included consideration for potential for water within the convenience can, corrosion from water, galvanic corrosion, tape degradation, and thermal expansion risk. Based on a review of existing literature and assumed conditions, corrosion and/or degradation of the 304 stainless steel (SS) Primary Containment Vessel (PCV) and the 304 stainless steel convenience cans containing the SNAP fuel is not significant to cause failure during the 1 year time shipping period in the 9975 packaging assembly. However, storage beyond the 1 year shipping period has not been validated.

  19. Radioactive material package closures with the use of shape memory alloys

    SciTech Connect (OSTI)

    Koski, J.A.; Bronowski, D.R.

    1997-11-01

    When heated from room temperature to 165 C, some shape memory metal alloys such as titanium-nickel alloys have the ability to return to a previously defined shape or size with dimensional changes up to 7%. In contrast, the thermal expansion of most metals over this temperature range is about 0.1 to 0.2%. The dimension change of shape memory alloys, which occurs during a martensite to austenite phase transition, can generate stresses as high as 700 MPa (100 kspi). These properties can be used to create a closure for radioactive materials packages that provides for easy robotic or manual operations and results in reproducible, tamper-proof seals. This paper describes some proposed closure methods with shape memory alloys for radioactive material packages. Properties of the shape memory alloys are first summarized, then some possible alternative sealing methods discussed, and, finally, results from an initial proof-of-concept experiment described.

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

    SciTech Connect (OSTI)

    Marusich, Robert M.

    2013-08-15

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

  1. Directory of Certificates of Compliance for radioactive materials packages: Report of NRC approved packages. Volume 1, Revision 18

    SciTech Connect (OSTI)

    1995-10-01

    The purpose of this directory is to make available a convenient source of information on packagings which have been approved by the US Nuclear Regulatory Commission. To assist in identifying packaging, an index by Model Number and corresponding Certificate of Compliance Number is included at the front of Volumes 1 and 2. An alphabetical listing by user name is included in the back of Volume 3 of approved QA programs. The reports include a listing of all users of each package design and approved QA programs prior to the publication date.

  2. Research and Development Program for transportation packagings at Sandia National Laboratories

    SciTech Connect (OSTI)

    Hohnstreiter, G.F.; Sorenson, K.B.

    1995-02-01

    This document contains information about the research and development programs dealing with waste transport at Sandia National Laboratories. This paper discusses topics such as: Why new packaging is needed; analytical methodologies and design codes;evaluation of packaging components; materials characterization; creative packaging concepts; packaging engineering and analysis; testing; and certification support.

  3. Packaging and Transportation for Offsite Shipment of Materials of National Security Interest

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2010-12-20

    The purpose of this Order is to make clear that the packaging and transportation of all offsite shipments of materials of national security interest for DOE must be conducted in accordance with DOT and Nuclear Regulatory Commission (NRC) regulations that would be applicable to comparable commercial shipments, except where an alternative course of action is identified in this Order. Supersedes DOE O 461.1A.

  4. Molecular environmental science using synchrotron radiation:Chemistry and physics of waste form materials

    SciTech Connect (OSTI)

    Lindle, Dennis W.; Shuh, David K.

    2005-02-28

    Production of defense-related nuclear materials has generated large volumes of complex chemical wastes containing a mixture of radionuclides. The disposition of these wastes requires conversion of the liquid and solid-phase components into durable, solid forms suitable for long-term immobilization [1]. Specially formulated glass compositions, many of which have been derived from glass developed for commercial purposes, and ceramics such as pyrochlores and apatites, will be the main recipients for these wastes. The performance characteristics of waste-form glasses and ceramics are largely determined by the loading capacity for the waste constituents (radioactive and non-radioactive) and the resultant chemical and radiation resistance of the waste-form package to leaching (durability). There are unique opportunities for the use of near-edge soft-x-ray absorption fine structure (NEXAFS) spectroscopy to investigate speciation of low-Z elements forming the backbone of waste-form glasses and ceramics. Although nuclear magnetic resonance (NMR) is the primary technique employed to obtain speciation information from low-Z elements in waste forms, NMR is incompatible with the metallic impurities contained in real waste and is thus limited to studies of idealized model systems. In contrast, NEXAFS can yield element-specific speciation information from glass constituents without sensitivity to paramagnetic species. Development and use of NEXAFS for eventual studies of real waste glasses has significant implications, especially for the low-Z elements comprising glass matrices [5-7]. The NEXAFS measurements were performed at Beamline 6.3.1, an entrance-slitless bend-magnet beamline operating from 200 eV to 2000 eV with a Hettrick-Underwood varied-line-space (VLS) grating monochromator, of the Advanced Light Source (ALS) at LBNL. Complete characterization and optimization of this beamline was conducted to enable high-performance measurements.

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

    SciTech Connect (OSTI)

    Hargis, Kenneth Marshall

    2015-03-01

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

  6. The biomethanation of waste material; An example in Germany

    SciTech Connect (OSTI)

    Shin, K.C. )

    1991-01-01

    This paper reports that digester gas (biogas) can be generated from anaerobic decomposition of organic waste substances. In the municipal sewage treatment plants in Germany most of the gas production is used for heating and electric power generation. The major portion of solid waste shall be returned to the economical circuit as biogas, compost and recyclable materials.

  7. Definition of Small Gram Quantity Contents for Type B Radioactive Material Transportation Packages: Activity-Based Content Limitations

    SciTech Connect (OSTI)

    Sitaraman, S; Kim, S; Biswas, D; Hafner, R; Anderson, B

    2010-10-27

    Since the 1960's, the Department of Transportation Specification (DOT Spec) 6M packages have been used extensively for transportation of Type B quantities of radioactive materials between Department of Energy (DOE) facilities, laboratories, and productions sites. However, due to the advancement of packaging technology, the aging of the 6M packages, and variability in the quality of the packages, the DOT implemented a phased elimination of the 6M specification packages (and other DOT Spec packages) in favor of packages certified to meet federal performance requirements. DOT issued the final rule in the Federal Register on October 1, 2004 requiring that use of the DOT Specification 6M be discontinued as of October 1, 2008. A main driver for the change was the fact that the 6M specification packagings were not supported by a Safety Analysis Report for Packaging (SARP) that was compliant with Title 10 of the Code of Federal Regulations part 71 (10 CFR 71). Therefore, materials that would have historically been shipped in 6M packages are being identified as contents in Type B (and sometimes Type A fissile) package applications and addenda that are to be certified under the requirements of 10 CFR 71. The requirements in 10 CFR 71 include that the Safety Analysis Report for Packaging (SARP) must identify the maximum radioactivity of radioactive constituents and maximum quantities of fissile constituents (10 CFR 71.33(b)(1) and 10 CFR 71.33(b)(2)), and that the application (i.e., SARP submittal or SARP addendum) demonstrates that the external dose rate (due to the maximum radioactivity of radioactive constituents and maximum quantities of fissile constituents) on the surface of the packaging (i.e., package and contents) not exceed 200 mrem/hr (10 CFR 71.35(a), 10 CFR 71.47(a)). It has been proposed that a 'Small Gram Quantity' of radioactive material be defined, such that, when loaded in a transportation package, the dose rates at external points of an unshielded packaging

  8. PANWAS: A Passive/Active Neutron Waste Assay System for the Radiological Characterization of Waste Packages at the Nucleco Facility at Casaccia

    SciTech Connect (OSTI)

    Alvarez, E.; Wilkins, C.G.; Croft, S.; Villani, M.F.; Ambrifi, A.; Simone, G.

    2006-07-01

    CANBERRA has recently supplied Nucleco SpA with a new Passive/Active Neutron Waste Assay System (PANWAS) for use at their waste management facility at Casaccia in Italy. The system complements two existing CANBERRA high-resolution gamma spectrometry waste assay systems. The three waste assay systems have been integrated into a combined facility for the radiological characterization of the waste managed by Nucleco in order to provide the information required to: - Determine the physical inventory of the nuclear material present for Safeguards purposes, - Segregate the waste into different categories, - Allow transportation to and storage in the final repository for the waste. This paper describes the main characteristics of the PANWAS, how it is used (in conjunction with the two gamma monitoring systems) to determine the radionuclide inventory of the waste and how the system was calibrated and characterized for use in this application. (authors)

  9. Operating Experience and Lessons Learned in the Use of Soft-Sided Packaging for Transportation and Disposal of Low Activity Radioactive Waste

    SciTech Connect (OSTI)

    Kapoor, A.; Gordon, S.; Goldston, W.

    2013-07-08

    This paper describes the operating experience and lessons learned at U.S. Department of Energy (DOE) sites as a result of an evaluation of potential trailer contamination and soft-sided packaging integrity issues related to the disposal of low-level and mixed low-level (LLW/MLLW) radioactive waste shipments. Nearly 4.3 million cubic meters of LLW/MLLW will have been generated and disposed of during fiscal year (FY) 2010 to FY 2015either at commercial disposal sites or disposal sites owned by DOE. The LLW/MLLW is packaged in several different types of regulatory compliant packaging and transported via highway or rail to disposal sites safely and efficiently in accordance with federal, state, and local regulations and DOE orders. In 1999, DOE supported the development of LLW containers that are more volumetrically efficient, more cost effective, and easier to use as compared to metal or wooden containers that existed at that time. The DOE Idaho National Engineering and Environmental Laboratory (INEEL), working in conjunction with the plastic industry, tested several types of soft-sided waste packaging systems that meet U.S. Department of Transportation requirements for transport of low specific activity and surface contaminated objects. Since then, soft-sided packaging of various capacities have been used successfully by the decontamination and decommissioning (D&D) projects to package, transport, and dispose D&D wastes throughout the DOE complex. The joint team of experts assembled by the Energy Facility Contractors Group from DOE waste generating sites, DOE and commercial waste disposal facilities, and soft-sided packaging suppliers conducted the review of soft-sided packaging operations and transportation of these packages to the disposal sites. As a result of this evaluation, the team developed several recommendations and best practices to prevent or minimize the recurrences of equipment contamination issues and proper use of soft-sided packaging for transport

  10. Economical Remediation of Plastic Waste into Advanced Materials with

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Coatings | Argonne National Laboratory Economical Remediation of Plastic Waste into Advanced Materials with Coatings Technology available for licensing: An autogenic pyrolysis process to convert plastic waste into high-value carbon nanotubes (50- to 100-nm outside diameter) and perfectly round carbon spheres (2- to 12-μm outside diameter). The tubes can be used as anode material in advanced batteries such as lithium-ion and eventually, lithium-air batteries. An environmentally-friendly,