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Sample records for canisters

  1. CANISTER TRANSFER SYSTEM DESCRIPTION DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    B. Gorpani

    2000-06-23

    The Canister Transfer System receives transportation casks containing large and small disposable canisters, unloads the canisters from the casks, stores the canisters as required, loads them into disposal containers (DCs), and prepares the empty casks for re-shipment. Cask unloading begins with cask inspection, sampling, and lid bolt removal operations. The cask lids are removed and the canisters are unloaded. Small canisters are loaded directly into a DC, or are stored until enough canisters are available to fill a DC. Large canisters are loaded directly into a DC. Transportation casks and related components are decontaminated as required, and empty casks are prepared for re-shipment. One independent, remotely operated canister transfer line is provided in the Waste Handling Building System. The canister transfer line consists of a Cask Transport System, Cask Preparation System, Canister Handling System, Disposal Container Transport System, an off-normal canister handling cell with a transfer tunnel connecting the two cells, and Control and Tracking System. The Canister Transfer System operating sequence begins with moving transportation casks to the cask preparation area with the Cask Transport System. The Cask Preparation System prepares the cask for unloading and consists of cask preparation manipulator, cask inspection and sampling equipment, and decontamination equipment. The Canister Handling System unloads the canister(s) and places them into a DC. Handling equipment consists of a bridge crane hoist, DC loading manipulator, lifting fixtures, and small canister staging racks. Once the cask has been unloaded, the Cask Preparation System decontaminates the cask exterior and returns it to the Carrier/Cask Handling System via the Cask Transport System. After the DC is fully loaded, the Disposal Container Transport System moves the DC to the Disposal Container Handling System for welding. To handle off-normal canisters, a separate off-normal canister handling

  2. Status report, canister fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Claes-Goeran; Eriksson, Peter; Westman, Marika [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden); Emilsson, Goeran [CSM Materialteknik AB, Linkoeping (Sweden)

    2004-06-01

    The report gives an account of the development of material and fabrication technology for copper canisters with cast inserts during the period from 2000 until the start of 2004. The engineering design of the canister and the choice of materials in the constituent components described in previous status reports have not been significantly changed. In the reference canister, the thickness of the copper shell is 50 mm. Fabrication of individual components with a thinner copper thickness is done for the purpose of gaining experience and evaluating fabrication and inspection methods for such canisters. As a part of the development of cast inserts, computer simulations of the casting processes and techniques used at the foundries have been performed for the purpose of optimizing the material properties. These properties have been evaluated by extensive tensile testing and metallographic inspection of test material taken from discs cut at different points along the length of the inserts. The testing results exhibit a relatively large spread. Low elongation values in certain tensile test specimens are due to the presence of poorly formed graphite, porosities, slag or other casting defects. It is concluded in the report that it will not be possible to avoid some presence of observed defects in castings of this size. In the deep repository, the inserts will be exposed to compressive loading and the observed defects are not critical for strength. An analysis of the strength of the inserts and formulation of relevant material requirements must be based on a statistical approach with probabilistic calculations. This work has been initiated and will be concluded during 2004. An initial verifying compression test of a canister in an isostatic press has indicated considerable overstrength in the structure. Seamless copper tubes are fabricated by means of three methods: extrusion, pierce and draw processing, and forging. It can be concluded that extrusion tests have revealed a

  3. HLW Canister and Can-In-Canister Drop Calculation

    Energy Technology Data Exchange (ETDEWEB)

    H. Marr

    1999-09-15

    The purpose of this calculation is to evaluate the structural response of the standard high-level waste (HLW) canister and the HLW canister containing the cans of immobilized plutonium (''can-in-canister'' throughout this document) to the drop event during the handling operation. The objective of the calculation is to provide the structure parameter information to support the canister design and the waste handling facility design. Finite element solution is performed using the commercially available ANSYS Version (V) 5.4 finite element code. Two-dimensional (2-D) axisymmetric and three-dimensional (3-D) finite element representations for the standard HLW canister and the can-in-canister are developed and analyzed using the dynamic solver.

  4. DISPOSABLE CANISTER WASTE ACCEPTANCE CRITERIA

    Energy Technology Data Exchange (ETDEWEB)

    R.J. Garrett

    2001-07-30

    The purpose of this calculation is to provide the bases for defining the preclosure limits on radioactive material releases from radioactive waste forms to be received in disposable canisters at the Monitored Geologic Repository (MGR) at Yucca Mountain. Specifically, this calculation will provide the basis for criteria to be included in a forthcoming revision of the Waste Acceptance System Requirements Document (WASRD) that limits releases in terms of non-isotope-specific canister release dose-equivalent source terms. These criteria will be developed for the Department of Energy spent nuclear fuel (DSNF) standard canister, the Multicanister Overpack (MCO), the naval spent fuel canister, the High-Level Waste (HLW) canister, the plutonium can-in-canister, and the large Multipurpose Canister (MPC). The shippers of such canisters will be required to demonstrate that they meet these criteria before the canisters are accepted at the MGR. The Quality Assurance program is applicable to this calculation. The work reported in this document is part of the analysis of DSNF and is performed using procedure AP-3.124, Calculations. The work done for this analysis was evaluated according to procedure QAP-2-0, Control of Activities, which has been superseded by AP-2.21Q, Quality Determinations and Planning for Scientific, Engineering, and Regulatory Compliance Activities. This evaluation determined that such activities are subject to the requirements of DOE/RW/0333P, Quality Assurance Requirements and Description (DOE 2000). This work is also prepared in accordance with the development plan titled Design Basis Event Analyses on DOE SNF and Plutonium Can-In-Canister Waste Forms (CRWMS M&O 1999a) and Technical Work Plan For: Department of Energy Spent Nuclear Fuel Work Packages (CRWMS M&O 2000d). This calculation contains no electronic data applicable to any electronic data management system.

  5. Analysis of K west basin canister gas

    Energy Technology Data Exchange (ETDEWEB)

    Trimble, D.J., Fluor Daniel Hanford

    1997-03-06

    Gas and Liquid samples have been collected from a selection of the approximately 3,820 spent fuel storage canisters in the K West Basin. The samples were taken to characterize the contents of the gas and water in the canisters providing source term information for two subprojects of the Spent Nuclear Fuel Project (SNFP) (Fulton 1994): the K Basins Integrated Water Treatment System Subproject (Ball 1996) and the K Basins Fuel Retrieval System Subproject (Waymire 1996). The barrels of ten canisters were sampled for gas and liquid in 1995, and 50 canisters were sampled in a second campaign in 1996. The analysis results from the first campaign have been reported (Trimble 1995a, 1995b, 1996a, 1996b). The analysis results from the second campaign liquid samples have been documented (Trimble and Welsh 1997; Trimble 1997). This report documents the results for the gas samples from the second campaign and evaluates all gas data in terms of expected releases when opening the canisters for SNFP activities. The fuel storage canisters consist of two closed and sealed barrels, each with a gas trap. The barrels are attached at a trunion to make a canister, but are otherwise independent (Figure 1). Each barrel contains up to seven N Reactor fuel element assemblies. A gas space of nitrogen was established in the top 2.2 to 2.5 inches (5.6 to 6.4 cm) of each barrel. Many of the fuel elements were damaged allowing the metallic uranium fuel to be corroded by the canister water. The corrosion releases fission products and generates hydrogen gas. The released gas mixes with the gas-space gas and excess gas passes through the gas trap into the basin water. The canister design does not allow canister water to be exchanged with basin water.

  6. CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    Energy Technology Data Exchange (ETDEWEB)

    C.E. Sanders

    2005-04-07

    This design calculation revises and updates the previous criticality evaluation for the canister handling, transfer and staging operations to be performed in the Canister Handling Facility (CHF) documented in BSC [Bechtel SAIC Company] 2004 [DIRS 167614]. The purpose of the calculation is to demonstrate that the handling operations of canisters performed in the CHF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Canister Handling Facility Description Document'' (BSC 2004 [DIRS 168992], Sections 3.1.1.3.4.13 and 3.2.3). Specific scope of work contained in this activity consists of updating the Category 1 and 2 event sequence evaluations as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004 [DIRS 167268], Section 7). The CHF is limited in throughput capacity to handling sealed U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and high-level radioactive waste (HLW) canisters, defense high-level radioactive waste (DHLW), naval canisters, multicanister overpacks (MCOs), vertical dual-purpose canisters (DPCs), and multipurpose canisters (MPCs) (if and when they become available) (BSC 2004 [DIRS 168992], p. 1-1). It should be noted that the design and safety analyses of the naval canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. In addition, this calculation is valid for

  7. Remote controlled mover for disposal canister transfer

    Energy Technology Data Exchange (ETDEWEB)

    Suikki, M. [Optimik Oy, Turku (Finland)

    2013-10-15

    This working report is an update for an earlier automatic guided vehicle design (Pietikaeinen 2003). The short horizontal transfers of disposal canisters manufactured in the encapsulation process are conducted with remote controlled movers both in the encapsulation plant and in the underground areas at the canister loading station of the disposal facility. The canister mover is a remote controlled transfer vehicle mobile on wheels. The handling of canisters is conducted with the assistance of transport platforms (pallets). The very small automatic guided vehicle of the earlier design was replaced with a commercial type mover. The most important reasons for this being the increased loadbearing requirement and the simpler, proven technology of the vehicle. The larger size of the vehicle induced changes to the plant layouts and in the principles for dealing with fault conditions. The selected mover is a vehicle, which is normally operated from alongside. In this application, the vehicle steering technology must be remote controlled. In addition, the area utilization must be as efficient as possible. This is why the vehicle was downsized in its outer dimensions and supplemented with certain auxiliary equipment and structures. This enables both remote controlled operation and improves the vehicle in terms of its failure tolerance. Operation of the vehicle was subjected to a risk analysis (PFMEA) and to a separate additional calculation conserning possible canister toppling risks. The total cost estimate, without value added tax for manufacturing the system amounts to 730 000 euros. (orig.)

  8. Drop Testing Representative Multi-Canister Overpacks

    Energy Technology Data Exchange (ETDEWEB)

    Snow, Spencer D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Morton, Dana K. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-06-01

    The objective of the work reported herein was to determine the ability of the Multi- Canister Overpack (MCO) canister design to maintain its containment boundary after an accidental drop event. Two test MCO canisters were assembled at Hanford, prepared for testing at the Idaho National Engineering and Environmental Laboratory (INEEL), drop tested at Sandia National Laboratories, and evaluated back at the INEEL. In addition to the actual testing efforts, finite element plastic analysis techniques were used to make both pre-test and post-test predictions of the test MCOs structural deformations. The completed effort has demonstrated that the canister design is capable of maintaining a 50 psig pressure boundary after drop testing. Based on helium leak testing methods, one test MCO was determined to have a leakage rate not greater than 1x10-5 std cc/sec (prior internal helium presence prevented a more rigorous test) and the remaining test MCO had a measured leakage rate less than 1x10-7 std cc/sec (i.e., a leaktight containment) after the drop test. The effort has also demonstrated the capability of finite element methods using plastic analysis techniques to accurately predict the structural deformations of canisters subjected to an accidental drop event.

  9. Grain boundary corrosion of copper canister material

    Energy Technology Data Exchange (ETDEWEB)

    Fennell, P.A.H.; Graham, A.J.; Smart, N.R.; Sofield, C.J. [AEA Technology plc, Harwell (United Kingdom)

    2001-03-01

    The proposed design for a final repository for spent fuel and other long-lived residues in Sweden is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will then be placed in granite bedrock and surrounded by compacted bentonite clay. The canister design is based on a thick cast inner container fitted inside a corrosion-resistant copper canister. During fabrication of the outer copper canisters there will be some unavoidable grain growth in the welded areas. As grains grow they will tend to concentrate impurities within the copper at the new grain boundaries. The work described in this report was undertaken to determine whether there is any possibility of enhanced corrosion at grain boundaries within the copper canister. The potential for grain boundary corrosion was investigated by exposing copper specimens, which had undergone different heat treatments and hence had different grain sizes, to aerated artificial bentonite-equilibrated groundwater with two concentrations of chloride, for increasing periods of time. The degree of grain boundary corrosion was determined by atomic force microscopy (AFM) and optical microscopy. AFM showed no increase in grain boundary 'ditching' for low chloride groundwater. In high chloride groundwater the surface was covered uniformly with a fine-grained oxide. No increases in oxide thickness were observed. No significant grain boundary attack was observed using optical microscopy either. The work suggests that in aerated artificial groundwaters containing chloride ions, grain boundary corrosion of copper is unlikely to adversely affect SKB's copper canisters.

  10. Drop Calculations of HLW Canister and Pu Can-in-Canister

    Energy Technology Data Exchange (ETDEWEB)

    Sreten Mastilovic

    2001-07-31

    The objective of this calculation is to determine the structural response of the standard high-level waste (HLW) canister and the canister containing the cans of immobilized plutonium (Pu) (''can-in-canister'' [CIC] throughout this document) subjected to drop DBEs (design basis events) during the handling operation. The evaluated DBE in the former case is 7-m (23-ft) vertical (flat-bottom) drop. In the latter case, two 2-ft (0.61-m) corner (oblique) drops are evaluated in addition to the 7-m vertical drop. These Pu CIC calculations are performed at three different temperatures: room temperature (RT) (20 C ), T = 200 F = 93.3 C , and T = 400 F = 204 C ; in addition to these the calculation characterized by the highest maximum stress intensity is performed at T = 750 F = 399 C as well. The scope of the HLW canister calculation is limited to reporting the calculation results in terms of: stress intensity and effective plastic strain in the canister, directional residual strains at the canister outer surface, and change of canister dimensions. The scope of Pu CIC calculation is limited to reporting the calculation results in terms of stress intensity, and effective plastic strain in the canister. The information provided by the sketches from Reference 26 (Attachments 5.3,5.5,5.8, and 5.9) is that of the potential CIC design considered in this calculation, and all obtained results are valid for this design only. This calculation is associated with the Plutonium Immobilization Project and is performed by the Waste Package Design Section in accordance with Reference 24. It should be noted that the 9-m vertical drop DBE, included in Reference 24, is not included in the objective of this calculation since it did not become a waste acceptance requirement. AP-3.124, ''Calculations'', is used to perform the calculation and develop the document.

  11. Design report of the disposal canister for twelve fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Raiko, H. [VTT Energy, Espoo (Finland); Salo, J.P. [Posiva Oy, Helsinki (Finland)

    1999-05-01

    The report provides a summary of the design of the canister for final disposal of spent nuclear fuel. The canister structure consists of a cylindrical massive nodular graphite cast iron insert covered by a 50 mm thick copper overlay. The capacity of the canister is 12 assemblies of BWR or VVER 440 fuel. The canister shall be tight with a high probability for about 100 000 years. The good and long lasting tightness requires: (1) The good initial tightness that is achieved by high quality requirements and extensive quality control, (2) The good corrosion resistance, which is obtained by the overpack of oxygen free copper, and (3) Mechanical strength of the canister, that is ensured by analyses (the following loads are considered: hydrostatic pressure, even and uneven swelling pressure of bentonite, thermal effects, and elevated hydrostatic pressure during glaciation. The allowed stresses and strains are set in such a way that reasonable engineering safety factors are obtained in all assessed design base loading cases). The canister shall limit the radiation dose rate outside the canister to minimise the radiolysis of the water in the vicinity of the canister. The canister insert shall keep the fuel assemblies in a subcritical configuration even if the void in the canister is filled with water due to postulated leakage. The design basis of the canister is set, the performed analyses are summarised and the results are assessed and discussed in the report. (orig.) 35 refs.

  12. Groundwork for Universal Canister System Development

    Energy Technology Data Exchange (ETDEWEB)

    Price, Laura L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gross, Mike [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Prouty, Jeralyn L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Rigali, Mark J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Craig, Brian [Argonne National Lab. (ANL), Argonne, IL (United States); Han, Zenghu [Argonne National Lab. (ANL), Argonne, IL (United States); Lee, John Hok [Argonne National Lab. (ANL), Argonne, IL (United States); Liu, Yung [Argonne National Lab. (ANL), Argonne, IL (United States); Pope, Ron [Argonne National Lab. (ANL), Argonne, IL (United States); Connolly, Kevin [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Feldman, Matt [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jarrell, Josh [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Radulescu, Georgeta [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wells, Alan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-01

    The mission of the United States Department of Energy's Office of Environmental Management is to complete the safe cleanup of the environmental legacy brought about from five decades of nuclear weapons development and go vernment - sponsored nuclear energy re search. S ome of the waste s that that must be managed have be en identified as good candidates for disposal in a deep borehole in crystalline rock (SNL 2014 a). In particular, wastes that can be disposed of in a small package are good candidates for this disposal concept. A canister - based system that can be used for handling these wastes during the disposition process (i.e., storage, transfers, transportation, and disposal) could facilitate the eventual disposal of these wastes. This report provides information for a program plan for developing specifications regarding a canister - based system that facilitates small waste form packaging and disposal and that is integrated with the overall efforts of the DOE's Office of Nuclear Energy Used Fuel Dis position Camp aign's Deep Borehole Field Test . Groundwork for Universal Ca nister System Development September 2015 ii W astes to be considered as candidates for the universal canister system include capsules containing cesium and strontium currently stored in pools at the Hanford Site, cesium to be processed using elutable or nonelutable resins at the Hanford Site, and calcine waste from Idaho National Laboratory. The initial emphasis will be on disposal of the cesium and strontium capsules in a deep borehole that has been drilled into crystalline rock. Specifications for a universal canister system are derived from operational, performance, and regulatory requirements for storage, transfers, transportation, and disposal of radioactive waste. Agreements between the Department of Energy and the States of Washington and Idaho, as well as the Deep Borehole Field Test plan provide schedule requirements for development of the universal canister system

  13. Canister storage building hazard analysis report

    Energy Technology Data Exchange (ETDEWEB)

    Krahn, D.E.; Garvin, L.J.

    1997-07-01

    This report describes the methodology used in conducting the Canister Storage Building (CSB) hazard analysis to support the final CSB safety analysis report (SAR) and documents the results. The hazard analysis was performed in accordance with DOE-STD-3009-94, Preparation Guide for US Department of Energy Nonreactor Nuclear Facility Safety Analysis Report, and implements the requirements of DOE Order 5480.23, Nuclear Safety Analysis Report.

  14. Telescoping Sample Canister Capture Mechanism (TSCCM)

    Science.gov (United States)

    Kong, Kin Yuen; Gorevan, Stephen; Mukherjee, Suparna; Wilson, Jack

    2003-11-01

    Sample return from solar system bodies including planets, moons, comets and asteroids is of high importance within the space science community. A returned sample will allow much more elaborate and detailed analysis not feasible through remote robotic analysis. For this reason, Honeybee Robotics has developed a low-cost reusable, automated on-orbit sample canister capture mechanism. The purpose of the mechanism is to capture a full sample canister and transfer it to a storage cache, sample return spacecraft, or on-orbit laboratory for further scientific study. The current design allows for reliable misalignment-compensated capture for various sample container geometries in any initial orientation. After capture, the sample canister is aligned and presented for transfer. Honeybee has demonstrated the concept through tests of two- and three-dimensional telescopic capture mechanism breadboards. The telescopic capture mechanism design is scalable, minimizes volume and can be made of lightweight material to minmize mass, all of which are critical aspects of spacecraft design.

  15. Stress corrosion cracking of copper canisters

    Energy Technology Data Exchange (ETDEWEB)

    King, Fraser (Integrity Corrosion Consulting Limited (Canada)); Newman, Roger (Univ. of Toronto (Canada))

    2010-12-15

    A critical review is presented of the possibility of stress corrosion cracking (SCC) of copper canisters in a deep geological repository in the Fennoscandian Shield. Each of the four main mechanisms proposed for the SCC of pure copper are reviewed and the required conditions for cracking compared with the expected environmental and mechanical loading conditions within the repository. Other possible mechanisms are also considered, as are recent studies specifically directed towards the SCC of copper canisters. The aim of the review is to determine if and when during the evolution of the repository environment copper canisters might be susceptible to SCC. Mechanisms that require a degree of oxidation or dissolution are only possible whilst oxidant is present in the repository and then only if other environmental and mechanical loading conditions are satisfied. These constraints are found to limit the period during which the canisters could be susceptible to cracking via film rupture (slip dissolution) or tarnish rupture mechanisms to the first few years after deposition of the canisters, at which time there will be insufficient SCC agent (ammonia, acetate, or nitrite) to support cracking. During the anaerobic phase, the supply of sulphide ions to the free surface will be transport limited by diffusion through the highly compacted bentonite. Therefore, no HS. will enter the crack and cracking by either of these mechanisms during the long term anaerobic phase is not feasible. Cracking via the film-induced cleavage mechanism requires a surface film of specific properties, most often associated with a nano porous structure. Slow rates of dissolution characteristic of processes in the repository will tend to coarsen any nano porous layer. Under some circumstances, a cuprous oxide film could support film-induced cleavage, but there is no evidence that this mechanism would operate in the presence of sulphide during the long-term anaerobic period because copper sulphide

  16. Thermal Predictions of the Cooling of Waste Glass Canisters

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen

    2014-11-01

    Radioactive liquid waste from five decades of weapons production is slated for vitrification at the Hanford site. The waste will be mixed with glass forming additives and heated to a high temperature, then poured into canisters within a pour cave where the glass will cool and solidify into a stable waste form for disposal. Computer simulations were performed to predict the heat rejected from the canisters and the temperatures within the glass during cooling. Four different waste glass compositions with different thermophysical properties were evaluated. Canister centerline temperatures and the total amount of heat transfer from the canisters to the surrounding air are reported.

  17. CANISTER HANDLING FACILITY WORKER DOSE ASSESSMENT

    Energy Technology Data Exchange (ETDEWEB)

    D.T. Dexheimer

    2004-02-27

    The purpose of this calculation is to estimate radiation doses received by personnel working in the Canister Handling Facility (CHF) performing operations to receive transportation casks, transfer wastes, prepare waste packages, perform associated equipment maintenance. The specific scope of work contained in this calculation covers individual worker group doses on an annual basis, and includes the contributions due to external and internal radiation. The results of this calculation will be used to support the design of the CHF and provide occupational dose estimates for the License Application.

  18. Canister storage building hazard analysis report

    Energy Technology Data Exchange (ETDEWEB)

    POWERS, T.B.

    1999-05-11

    This report describes the methodology used in conducting the Canister Storage Building (CSB) hazard analysis to support the CSB final safety analysis report (FSAR) and documents the results. The hazard analysis was performed in accordance with the DOE-STD-3009-94, ''Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports'', and meets the intent of HNF-PRO-704, ''Hazard and Accident Analysis Process''. This hazard analysis implements the requirements of DOE Order 5480.23, ''Nuclear Safety Analysis Reports''.

  19. EVALUATION OF REQUIREMENTS FOR THE DWPF HIGHER CAPACITY CANISTER

    Energy Technology Data Exchange (ETDEWEB)

    Miller, D.; Estochen, E.; Jordan, J.; Kesterson, M.; Mckeel, C.

    2014-08-05

    The Defense Waste Processing Facility (DWPF) is considering the option to increase canister glass capacity by reducing the wall thickness of the current production canister. This design has been designated as the DWPF Higher Capacity Canister (HCC). A significant decrease in the number of canisters processed during the life of the facility would be achieved if the HCC were implemented leading to a reduced overall reduction in life cycle costs. Prior to implementation of the change, Savannah River National Laboratory (SRNL) was requested to conduct an evaluation of the potential impacts. The specific areas of interest included loading and deformation of the canister during the filling process. Additionally, the effect of the reduced wall thickness on corrosion and material compatibility needed to be addressed. Finally the integrity of the canister during decontamination and other handling steps needed to be determined. The initial request regarding canister fabrication was later addressed in an alternate study. A preliminary review of canister requirements and previous testing was conducted prior to determining the testing approach. Thermal and stress models were developed to predict the forces on the canister during the pouring and cooling process. The thermal model shows the HCC increasing and decreasing in temperature at a slightly faster rate than the original. The HCC is shown to have a 3°F ΔT between the internal and outer surfaces versus a 5°F ΔT for the original design. The stress model indicates strain values ranging from 1.9% to 2.9% for the standard canister and 2.5% to 3.1% for the HCC. These values are dependent on the glass level relative to the thickness transition between the top head and the canister wall. This information, along with field readings, was used to set up environmental test conditions for corrosion studies. Small 304-L canisters were filled with glass and subjected to accelerated environmental testing for 3 months. No evidence of

  20. Shippingport Spent Fuel Canister (SSFC) Design Report Project W-518

    Energy Technology Data Exchange (ETDEWEB)

    JOHNSON, D.M.

    2000-01-27

    The SSFC Design Report Describes A spent fuel canister for Shippingport Core 2 blanket fuel assemblies. The design of the SSFC is a minor modification of the MCO. The modification is limited to the Shield Plug which remains unchanged with regard to interfaces with the canister shell. The performance characteristics remain those for the MCO, which bounds the payload of the SSFC.

  1. Canister storage building design basis accident analysis documentation

    Energy Technology Data Exchange (ETDEWEB)

    KOPELIC, S.D.

    1999-02-25

    This document provides the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report.

  2. Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

    Energy Technology Data Exchange (ETDEWEB)

    CROWE, R.D.; PIEPHO, M.G.

    2000-03-23

    This document provided the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report''. All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report.

  3. Description of DWPF reference waste form and canister

    Energy Technology Data Exchange (ETDEWEB)

    1981-06-01

    This document describes the reference waste form and canister for the Defense Waste Processing Facility (DWPF). The facility is planned for location at the Savannah River Plant in Aiken, SC, and is scheduled for construction authorization during FY-1983. The reference canister is fabricated of 24-in.-OD 304L stainless steel pipe with a dished bottom, domed head, and lifting and welding flanges on the head neck. The overall canister length is 9 ft 10 in., with a wall thickness of 3/8-in. (schedule 20 pipe). The canister length was selected to reduce equipment cell height in the DWPF to a practical size. The canister diameter was selected to ensure that a filled canister with its shipping cask could be accommodated on a legal-weight truck. The overall dimensions and weight appear to be generally compatible with preliminary assessments of repository requirements. The reference waste form is borosilicate glass containing approximately 28 wt % sludge oxides with the balance glass frit. Borosilicate glass was chosen because of its high resistance to leaching by water, its relatively high solubility for nuclides found in the sludge, and its reasonably low melting temperature. The glass frit contains approximately 58% SiO/sub 2/ and 15% B/sub 2/O/sub 3/. This composition results in a low average leachability in the waste form of approximately 5 x 10/sup -9/ g/cm/sup 2/-day based on /sup 137/Cs over 365 days in 25/sup 0/C water. The canister is filled with 3260 lb of glass which occupies about 85% of the free canister volume. The filled canister will generate approximately 425 watts when filled with oxides from 5-year-old sludge and 15-year-old supernate from the Stage 1 and Stage 2 processes. The radionuclide content of the canister is about 150,000 curies, with a radiation level of 2 x 10/sup 4/ rem/hour at 1 cm.

  4. Assessment of a spent fuel disposal canister. Assessment studies for a copper canister with cast steel inner component

    Energy Technology Data Exchange (ETDEWEB)

    Bond, A.E.; Hoch, A.R.; Jones, G.D.; Tomczyk, A.J.; Wiggin, R.M.; Worraker, W.J. [AEA Technology, Harwell (United Kingdom)

    1997-05-01

    The proposed design for a final repository for spent fuel and other long-lived residues in Sweden, is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will then be placed in vertical storage holes drilled in a series of caverns excavated from the granite bedrock at a depth of about 500 m. Each canister will be surrounded by compacted bentonite clay. In this report, a simple model of the behaviour of the canister subsequent to a first breach in its copper overpack is developed. This model is used to predict: -the ingress of water to the canister (as a function of the size and the shape of the initial defect, the buffer conductivity, the corrosion rate and the pressure inside the canister); -the build-up of corrosion products in the canister (as a function of the available water in the canister, the corrosion rate and the properties of the corrosion products); -the effect of corrosion on the structural integrity of the canister. A number of different scenarios for the location of the breach in the copper overpack are considered.

  5. COMSOL Multiphysics Model for HLW Canister Filling

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-11

    The U.S. Department of Energy (DOE) is building a Tank Waste Treatment and Immobilization Plant (WTP) at the Hanford Site in Washington to remediate 55 million gallons of radioactive waste that is being temporarily stored in 177 underground tanks. Efforts are being made to increase the loading of Hanford tank wastes in glass while meeting melter lifetime expectancies and process, regulatory, and product quality requirements. Wastes containing high concentrations of Al2O3 and Na2O can contribute to nepheline (generally NaAlSiO4) crystallization, which can sharply reduce the chemical durability of high level waste (HLW) glass. Nepheline crystallization can occur during slow cooling of the glass within the stainless steel canister. The purpose of this work was to develop a model that can be used to predict temperatures of the glass in a WTP HLW canister during filling and cooling. The intent of the model is to support scoping work in the laboratory. It is not intended to provide precise predictions of temperature profiles, but rather to provide a simplified representation of glass cooling profiles within a full scale, WTP HLW canister under various glass pouring rates. These data will be used to support laboratory studies for an improved understanding of the mechanisms of nepheline crystallization. The model was created using COMSOL Multiphysics, a commercially available software. The model results were compared to available experimental data, TRR-PLT-080, and were found to yield sufficient results for the scoping nature of the study. The simulated temperatures were within 60 ºC for the centerline, 0.0762m (3 inch) from centerline, and 0.2286m (9 inch) from centerline thermocouples once the thermocouples were covered with glass. The temperature difference between the experimental and simulated values reduced to 40 ºC, 4 hours after the thermocouple was covered, and down to 20 ºC, 6 hours after the thermocouple was covered

  6. Structural Sensitivity of Dry Storage Canisters

    Energy Technology Data Exchange (ETDEWEB)

    Klymyshyn, Nicholas A.; Karri, Naveen K.; Adkins, Harold E.; Hanson, Brady D.

    2013-09-27

    This LS-DYNA modeling study evaluated a generic used nuclear fuel vertical dry storage cask system under tip-over, handling drop, and seismic load cases to determine the sensitivity of the canister containment boundary to these loads. The goal was to quantify the expected failure margins to gain insight into what material changes over the extended long-term storage lifetime could have the most influence on the security of the containment boundary. It was determined that the tip-over case offers a strong challenge to the containment boundary, and identifies one significant material knowledge gap, the behavior of welded stainless steel joints under high-strain-rate conditions. High strain rates are expected to increase the material’s effective yield strength and ultimate strength, and may decrease its ductility. Determining and accounting for this behavior could potentially reverse the model prediction of a containment boundary failure at the canister lid weld. It must be emphasized that this predicted containment failure is an artifact of the generic system modeled. Vendor specific designs analyze for cask tip-over and these analyses are reviewed and approved by the Nuclear Regulatory Commission. Another location of sensitivity of the containment boundary is the weld between the base plate and the canister shell. Peak stresses at this location predict plastic strains through the whole thickness of the welded material. This makes the base plate weld an important location for material study. This location is also susceptible to high strain rates, and accurately accounting for the material behavior under these conditions could have a significant effect on the predicted performance of the containment boundary. The handling drop case was largely benign to the containment boundary, with just localized plastic strains predicted on the outer surfaces of wall sections. It would take unusual changes in the handling drop scenario to harm the containment boundary, such as

  7. Radon measurements with charcoal canisters temperature and humidity considerations

    Directory of Open Access Journals (Sweden)

    Živanović Miloš Z.

    2016-01-01

    Full Text Available Radon testing by using open-faced charcoal canisters is a cheap and fast screening method. Many laboratories perform the sampling and measurements according to the United States Environmental Protection Agency method - EPA 520. According to this method, no corrections for temperature are applied and corrections for humidity are based on canister mass gain. The EPA method is practiced in the Vinča Institute of Nuclear Sciences with recycled canisters. In the course of measurements, it was established that the mass gain of the recycled canisters differs from mass gain measured by Environmental Protection Agency in an active atmosphere. In order to quantify and correct these discrepancies, in the laboratory, canisters were exposed for periods of 3 and 4 days between February 2015 and December 2015. Temperature and humidity were monitored continuously and mass gain measured. No significant correlation between mass gain and temperature was found. Based on Environmental Protection Agency calibration data, functional dependence of mass gain on humidity was determined, yielding Environmental Protection Agency mass gain curves. The results of mass gain measurements of recycled canisters were plotted against these curves and a discrepancy confirmed. After correcting the independent variable in the curve equation and calculating the corrected mass gain for recycled canisters, the agreement between measured mass gain and Environmental Protection Agency mass gain curves was attained. [Projekat Ministarstva nauke Republike Srbije, br. III43009: New Technologies for Monitoring and Protection of Environment from Harmful Chemical Substances and Radiation Impact

  8. Radiolysis Model Sensitivity Analysis for a Used Fuel Storage Canister

    Energy Technology Data Exchange (ETDEWEB)

    Wittman, Richard S.

    2013-09-20

    This report fulfills the M3 milestone (M3FT-13PN0810027) to report on a radiolysis computer model analysis that estimates the generation of radiolytic products for a storage canister. The analysis considers radiolysis outside storage canister walls and within the canister fill gas over a possible 300-year lifetime. Previous work relied on estimates based directly on a water radiolysis G-value. This work also includes that effect with the addition of coupled kinetics for 111 reactions for 40 gas species to account for radiolytic-induced chemistry, which includes water recombination and reactions with air.

  9. Design analysis report for the canister

    Energy Technology Data Exchange (ETDEWEB)

    Raiko, Heikki (VTT (Finland)); Sandstroem, Rolf (Materials Science and Engineering, Royal Inst. of Technology, Stockholm (Sweden)); Ryden, Haakan; Johansson, Magnus (Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden))

    2010-04-15

    The mechanical strength of the canister (BWR and PWR types) has been studied. The loading processes are taken from the design premises report and some of them, especially the uneven bentonite swelling cases, are further developed in this study and in its references. The canister geometry is described in detail including the manufacturing tolerances of the dimensions. The canister material properties are summarised and the wide material testing programmes and model developments are referenced. The combination of various load cases are rationalised and the conservative combinations are defined. Also the probabilities of various load cases and combinations are assessed for setting reasonable safety margins. The safety margins are used according to ASME Code principles for safety class 1 components. The governing load cases are analysed with 2D- or global 3D-finite-element models including large deformation and non-linear material modelling and, in some cases, also creep. The integrity assessments are partly made from the stress and strain results using global models and partly from fracture resistance analyses using the sub-modelling technique. The sub-model analyses utilize the deformations from the global analyses as constraints on the sub-model boundaries and more detailed finite-element meshes are defined with defects included in the models together with elastic-plastic material models. The J-integral is used as the fracture parameter for the postulated defects. The allowable defect sizes are determined using the measured fracture resistance curves of the insert iron as a reference with respective safety factors according to the ASME Pressure Vessel Code requirements. Based on the BWR canister analyses, the following conclusions can be drawn. The 45 MPa isostatic pressure load case shows very robust and distinct results in that the risk for local collapse is vanishingly small. The probabilistic analysis of plastic collapse only considers the initial local collapse

  10. Deflection measurements of LABAN canister sections in horizontal attitude

    Energy Technology Data Exchange (ETDEWEB)

    Wakeman, W.

    1985-01-08

    Deflection measurements made on the LABAN canister sections indicate that the apparent stiffness of its frames, with all the diagnostics experiments installed, is not significantly different from the stiffness of the bare frames.

  11. Spent nuclear fuel canister storage building conceptual design report

    Energy Technology Data Exchange (ETDEWEB)

    Swenson, C.E. [Westinghouse Hanford Co., Richland, WA (United States)

    1996-01-01

    This Conceptual Design Report provides the technical basis for the Spent Nuclear Fuels Project, Canister Storage Building, and as amended by letter (correspondence number 9555700, M.E. Witherspoon to E.B. Sellers, ``Technical Baseline and Updated Cost Estimate for the Canister Storage Building``, dated October 24, 1995), includes the project cost baseline and Criteria to be used as the basis for starting detailed design in fiscal year 1995.

  12. Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

    Energy Technology Data Exchange (ETDEWEB)

    CROWE, R.D.

    1999-09-09

    This document provides the detailed accident analysis to support ''HNF-3553, Spent Nuclear Fuel Project Final Safety, Analysis Report, Annex A,'' ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report.

  13. Remote Welding, NDE and Repair of DOE Standardized Canisters

    Energy Technology Data Exchange (ETDEWEB)

    Eric Larsen; Art Watkins; Timothy R. McJunkin; Dave Pace; Rodney Bitsoi

    2006-05-01

    The U.S. Department of Energy (DOE) created the National Spent Nuclear Fuel Program (NSNFP) to manage DOE’s spent nuclear fuel (SNF). One of the NSNFP’s tasks is to prepare spent nuclear fuel for storage, transportation, and disposal at the national repository. As part of this effort, the NSNFP developed a standardized canister for interim storage and transportation of SNF. These canisters will be built and sealed to American Society of Mechanical Engineers (ASME) Section III, Division 3 requirements. Packaging SNF usually is a three-step process: canister loading, closure welding, and closure weld verification. After loading SNF into the canisters, the canisters must be seal welded and the welds verified using a combination of visual, surface eddy current, and ultrasonic inspection or examination techniques. If unacceptable defects in the weld are detected, the defective sections of weld must be removed, re-welded, and re-inspected. Due to the high contamination and/or radiation fields involved with this process, all of these functions must be performed remotely in a hot cell. The prototype apparatus to perform these functions is a floor-mounted carousel that encircles the loaded canister; three stations perform the functions of welding, inspecting, and repairing the seal welds. A welding operator monitors and controls these functions remotely via a workstation located outside the hot cell. The discussion describes the hardware and software that have been developed and the results of testing that has been done to date.

  14. Data quality objectives for K West canister sludge sampling

    Energy Technology Data Exchange (ETDEWEB)

    Makenas, B.J., Westinghouse Hanford

    1996-12-11

    Data Quality Objectives have been developed for a limited campaign of sampling K Basin canister sludge. Specifically, samples will be taken from the sealed K West Basin fuel canisters. Characterization of the sludge in these canisters will address the needs of fuel retrieval which are to collect and transport sludge which is currently in the canisters. Data will be gathered on physical properties (such as viscosity, particle size, density, etc.) as well as on chemical and radionuclide constituents and radiation levels of sludge. The primary emphasis will be on determining radionuclide concentrations to be deposited on Ion Exchange Modules (IXMS) during canister opening and fuel retrieval. The data will also be useful in determining whether K West Basin sludge meets the waste acceptance criteria for Hanford waste tanks as a backup disposal concept and these data will also supply information on the properties of sludge material which will1403 accompany fuel elements in the Multi-Canister Overpacks (MCOS) as envisioned in the Integrated Process Strategy (IPS).

  15. Impact analysis of stainless steel spent fuel canisters

    Energy Technology Data Exchange (ETDEWEB)

    Aramayo, G.A. [Oak Ridge National Lab., TN (United States); Turner, D.W. [Lockheed Martin Energy Systems, Oak Ridge, TN (United States). Waste Management Organization

    1998-04-01

    This paper presents the results of the numerical analysis performed to asses the structural integrity of spent nuclear fuel (SNF) stainless steel canisters when subjected to impact loads associated with free gravity drops from heights not exceeding 20 ft. The SNF canisters are to be used for the Shipment of radioactive material from the Oak Ridge National Laboratory (ORNL) Site to the Idaho National Engineering and Environmental Laboratory (INEEL) for storage. The Idaho chemical Processing Plant Fuel Receipt Criteria Questionnaire requires that the vertical drop accidents from two heights be analyze. These heights are those that are considered to be critical at the time of unloading the canisters from the shipping cask. The configurations analyzed include a maximum payload of 90 lbs dropping from heights of 20 and 3 ft. The nominal weight of the canister is 23.3 lbs. The analysis has been performed using finite element methods. Innovative analysis techniques are used to capture the effects of failure and separation of canister components. The structural integrity is evaluated in terms of physical deformation and separation of the canister components that may result from failure of components at selected interfaces.

  16. Analysis for Eccentric Multi Canister Overpack (MCO) Drops at the Canister Storage Building (CSB) (CSB-S-0073)

    Energy Technology Data Exchange (ETDEWEB)

    TU, K.C.

    1999-10-08

    Multi-Canister Overpacks (MCOs) containing spent nuclear fuel (SNF) will be routinely handled at the Canister Storage Building (CSB) during fuel movement operations in the SNF Project. This analysis was performed to investigate the potential for damage from an eccentric accidental drop onto the standard storage tube, overpack tube, service station, or sample/weld station. Appendix D was added to the FDNW document to include the peer Review Comment Record & transmittal record.

  17. GIBNE canister: a comprehensive analytical and experimental evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Gerhard, M.A.

    1983-01-17

    The finite-element computer program GEMINI was used to efficiently and accurately characterize the GIBNE 86 in canister. GEMINI accurately calculated the GIBNE bare frame deflections for all four load cases. The center bulkhead of the 45 ft long cnaister deflected .323 in. when the canister was horizontally supported by its end bulkheads. Several large lead weights were used to simulate (but not accurately characterize) the addition of internal hardware to the canister. The devlection increased to .512 in. with the addition of 8000 lbs to bulkhead 5. With the 8000 lb load moved to bulkhead 4 and 8260 lbs added to bulkhead 6, the deflection increased to .678 in. Deflections calculated by GEMINI were conservative by 3 to 5%. GEMINI correctly predicted the stress distribution in the bare frame cable trays. The GIBNE tests and analyses accurately characterized the GIBNE bare frame. However, the experimental results did not separate individual effects of the lines-of-sight and end fixtures. As a result, the numerical model can not be validated for a canister including lines-of-sight. During calendar year 1983, the LABAN test will characterize the individual effects of the lines-of-sight and the end fixtures. At that time the numerical model will be fine-tuned to match the experimental results. We will then be able to analytically predict canister alignment changes under a wide variety of loading conditions.

  18. Preliminary Transportation, Aging and Disposal Canister System Performance Specification

    Energy Technology Data Exchange (ETDEWEB)

    C.A Kouts

    2006-11-22

    This document provides specifications for selected system components of the Transportation, Aging and Disposal (TAD) canister-based system. A list of system specified components and ancillary components are included in Section 1.2. The TAD canister, in conjunction with specialized overpacks will accomplish a number of functions in the management and disposal of spent nuclear fuel. Some of these functions will be accomplished at purchaser sites where commercial spent nuclear fuel (CSNF) is stored, and some will be performed within the Office of Civilian Radioactive Waste Management (OCRWM) transportation and disposal system. This document contains only those requirements unique to applications within Department of Energy's (DOE's) system. DOE recognizes that TAD canisters may have to perform similar functions at purchaser sites. Requirements to meet reactor functions, such as on-site dry storage, handling, and loading for transportation, are expected to be similar to commercially available canister-based systems. This document is intended to be referenced in the license application for the Monitored Geologic Repository (MGR). As such, the requirements cited herein are needed for TAD system use in OCRWM's disposal system. This document contains specifications for the TAD canister, transportation overpack and aging overpack. The remaining components and equipment that are unique to the OCRWM system or for similar purchaser applications will be supplied by others.

  19. Design, production and initial state of the canister

    Energy Technology Data Exchange (ETDEWEB)

    Cederqvist, Lars; Johansson, Magnus; Leskinen, Nina; Ronneteg, Ulf

    2010-12-15

    The report is included in a set of Production reports, presenting how the KBS-3 repository is designed, produced and inspected. The set of reports is included in the safety report for the KBS-3 repository and repository facility.The report provides input on the initial state of the canisters to the assessment of the long-term safety, SR-Site. The initial state refers to the properties of the engineered barriers once they have been finally placed in the KBS-3 repository and will not be further handled within the repository facility. In addition, the report provides input to the operational safety report, SR-Operation, on how the canisters shall be handled and disposed. The report presents the design premises and reference design of the canister and verifies the conformity of the reference design to the design premises. The production methods and the ability to produce canisters according to the reference design are described. Finally, the initial state of the canisters and their conformity to the reference design and design premises are presented

  20. SNF Interim Storage Canister Corrosion and Surface Environment Investigations

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Enos, David G. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

    2015-09-01

    This progress report describes work being done at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of spent nuclear fuel (SNF). Of particular concern is stress corrosion cracking (SCC), by which a through-wall crack could potentially form in a canister outer wall over time intervals that are shorter than possible dry storage times. In order for SCC to occur, three criteria must be met. A corrosive environment must be present on the canister surface, the metal must susceptible to SCC, and sufficient tensile stress to support SCC must be present through the entire thickness of the canister wall. SNL is currently evaluating the potential for each of these criteria to be met.

  1. Criticality safety evaluation report for the multi-canister overpack

    Energy Technology Data Exchange (ETDEWEB)

    KESSLER, S.F.

    1999-05-21

    This criticality evaluation is for Spent N Reactor fuel unloaded from the existing canisters in both KE and KW Basins, and loaded into multiple canister overpack (MCO) containers with specially built baskets containing a maximum of either 54 Mark 1V or 48 Mark IA fuel assemblies. The criticality evaluations include loading baskets into the cask-MCO, operations at the Cold Vacuum Drying Facility, and storage in the Canister Storage Building. Many conservatisms have been built into this analysis, the primary one being the selection of the k{sub eff} = 0.95 criticality safety limit. Additional analyses in this revision include partial fuel basket loadings, loading 26.1 inch Mark IA fuel assemblies into Mark IV fuel baskets, and the revised fuel and scrap basket designs. The MCO MCNP model was revised to include the shield plug assembly.

  2. Debris Removal Project K West Canister Cleaning System Performance Specification

    Energy Technology Data Exchange (ETDEWEB)

    FARWICK, C.C.

    1999-12-09

    Approximately 2,300 metric tons Spent Nuclear Fuel (SNF) are currently stored within two water filled pools, the 105 K East (KE) fuel storage basin and the 105 K West (KW) fuel storage basin, at the U.S. Department of Energy, Richland Operations Office (RL). The SNF Project is responsible for operation of the K Basins and for the materials within them. A subproject to the SNF Project is the Debris Removal Subproject, which is responsible for removal of empty canisters and lids from the basins. Design criteria for a Canister Cleaning System to be installed in the KW Basin. This documents the requirements for design and installation of the system.

  3. Materials for Consideration in Standardized Canister Design Activities.

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R.; Ilgen, Anastasia Gennadyevna; Enos, David George; Teich-McGoldrick, Stephanie; Hardin, Ernest

    2014-10-01

    This document identifies materials and material mitigation processes that might be used in new designs for standardized canisters for storage, transportation, and disposal of spent nuclear fuel. It also addresses potential corrosion issues with existing dual-purpose canisters (DPCs) that could be addressed in new canister designs. The major potential corrosion risk during storage is stress corrosion cracking of the weld regions on the 304 SS/316 SS canister shell due to deliquescence of chloride salts on the surface. Two approaches are proposed to alleviate this potential risk. First, the existing canister materials (304 and 316 SS) could be used, but the welds mitigated to relieve residual stresses and/or sensitization. Alternatively, more corrosion-resistant steels such as super-austenitic or duplex stainless steels, could be used. Experimental testing is needed to verify that these alternatives would successfully reduce the risk of stress corrosion cracking during fuel storage. For disposal in a geologic repository, the canister will be enclosed in a corrosion-resistant or corrosion-allowance overpack that will provide barrier capability and mechanical strength. The canister shell will no longer have a barrier function and its containment integrity can be ignored. The basket and neutron absorbers within the canister have the important role of limiting the possibility of post-closure criticality. The time period for corrosion is much longer in the post-closure period, and one major unanswered question is whether the basket materials will corrode slowly enough to maintain structural integrity for at least 10,000 years. Whereas there is extensive literature on stainless steels, this evaluation recommends testing of 304 and 316 SS, and more corrosion-resistant steels such as super-austenitic, duplex, and super-duplex stainless steels, at repository-relevant physical and chemical conditions. Both general and localized corrosion testing methods would be used to

  4. Evaluation of the Frequencies for Canister Inspections for SCC

    Energy Technology Data Exchange (ETDEWEB)

    Stockman, Christine [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-02-02

    This report fulfills the M3 milestone M3FT-15SN0802042, “Evaluate the Frequencies for Canister Inspections for SCC” under Work Package FT-15SN080204, “ST Field Demonstration Support – SNL”. It reviews the current state of knowledge on the potential for stress corrosion cracking (SCC) of dry storage canisters and evaluates the implications of this state of knowledge on the establishment of an SCC inspection frequency. Models for the prediction of SCC by the Japanese Central Research Institute of Electric Power Industry (CRIEPI), the United States (U.S.) Electric Power Research Institute (EPRI), and Sandia National Laboratories (SNL) are summarized, and their limitations discussed.

  5. Analysis for Eccentric Multi Canister Overpack (MCO) Drops at the Canister Storage Building (CSB) (CSB-S-0073)

    Energy Technology Data Exchange (ETDEWEB)

    HOLLENBECK, R.G.

    2000-05-08

    The Spent Nuclear Fuel (SNF) Canister Storage Building (CSB) is the interim storage facility for the K-Basin SNF at the US. Department of Energy (DOE) Hanford Site. The SNF is packaged in multi-canister overpacks (MCOs). The MCOs are placed inside transport casks, then delivered to the service station inside the CSB. At the service station, the MCO handling machine (MHM) moves the MCO from the cask to a storage tube or one of two sample/weld stations. There are 220 standard storage tubes and six overpack storage tubes in a below grade reinforced concrete vault. Each storage tube can hold two MCOs.

  6. Estimation of CANDU spent fuel disposal canister lifetime

    Energy Technology Data Exchange (ETDEWEB)

    Kook, Dong Hak; Lee, Min Soo; Hwang, Yong Soo; Choi, Heui Joo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-05-15

    Active nuclear energy utilization causes significant spent fuel accumulation problem. The cumulative amount of spent fuel is about 10,083 ton as of Dec. 2008, and is expected to increase up to 19,000 ton by 2020. Of those, CANDU spent fuels account for more than 60% of the total amounts. CANDU spent fuels had been stored in dry concrete silos since 1991 and during the past 15 years, 300 silos were constructed and {approx}3,200 ton of spent fuels are stored now. Another dry storage facility MACSTOR /KN-400 will store new-coming CANDU spent fuels from 2009. But, after intermediate storage ends, all CANDU spent fuels have to be disposed within multi-layer metallic canister which is composed of cast iron inside and copper outside. Canister lifetime estimation, therefore, is very important for the final disposal safety analysis. The most significant factor of lifetime is copper corrosion, and Y. S. Hwang developed a corrosion model in order to predict the general corrosion effect on copper canister lifetime during the final disposal period. This research applied his model to KURT1 where many disposal researches are being performed actively and the results shows safe margin of the copper canister for the very long-term disposal.

  7. OCRWM Bulletin: Westinghouse begins designing multi-purpose canister

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    This publication consists of two parts: OCRWM (Office of Civilian Radioactive Waste Management) Bulletin; and Of Mountains & Science which has articles on the Yucca Mountain project. The OCRWM provides information about OCRWM activities and in this issue has articles on multi-purpose canister design, and transportation cask trailer.

  8. Evaluation of Accident Frequencies at the Canister Storage Bldg (CSB)

    Energy Technology Data Exchange (ETDEWEB)

    POWERS, T.B.

    2000-03-20

    By using simple frequency calculations and fault tree logic, an evaluation of the design basis accident frequencies at the Canister Storage Building has been performed. The following are the design basis accidents: Mechanical damage of MCO; Gaseous release from the MCO; MCO internal hydrogen deflagration; MCO external hydrogen deflagration; Thermal runaway reactions inside the MCO; and Violation of design temperature criteria.

  9. Storage and disposal of radioactive waste as glass in canisters

    Energy Technology Data Exchange (ETDEWEB)

    Mendel, J.E.

    1978-12-01

    A review of the use of waste glass for the immobilization of high-level radioactive waste glass is presented. Typical properties of the canisters used to contain the glass, and the waste glass, are described. Those properties are used to project the stability of canisterized waste glass through interim storage, transportation, and geologic disposal.

  10. Canister Cleaning System Final Design Report Project A-2A

    Energy Technology Data Exchange (ETDEWEB)

    FARWICK, C.C.

    2000-06-15

    Approximately 2,300 metric tons Spent Nuclear Fuel (SNF) are currently stored within two water filled pools, the 105 K East (KE) fuel storage basin and the 105 K West (KW) fuel storage basin, at the U.S. Department of Energy, Richland Operations Office (RL). The SNF Project is responsible for operation of the K Basins and for the materials within them. A subproject to the SNF Project is the Debris Removal Subproject, which is responsible for removal of empty canisters and lids from the basins. The Canister Cleaning System (CCS) is part of the Debris Removal Project. The CCS will be installed in the KW Basin and operated during the fuel removal activity. The KW Basin has approximately 3600 canisters that require removal from the basin. The CCS is being designed to ''clean'' empty fuel canisters and lids and package them for disposal to the Environmental Restoration Disposal Facility complex. The system will interface with the KW Basin and be located in the Dummy Elevator Pit.

  11. The development of a Martian atmospheric Sample collection canister

    Science.gov (United States)

    Kulczycki, E.; Galey, C.; Kennedy, B.; Budney, C.; Bame, D.; Van Schilfgaarde, R.; Aisen, N.; Townsend, J.; Younse, P.; Piacentine, J.

    The collection of an atmospheric sample from Mars would provide significant insight to the understanding of the elemental composition and sub-surface out-gassing rates of noble gases. A team of engineers at the Jet Propulsion Laboratory (JPL), California Institute of Technology have developed an atmospheric sample collection canister for Martian application. The engineering strategy has two basic elements: first, to collect two separately sealed 50 cubic centimeter unpressurized atmospheric samples with minimal sensing and actuation in a self contained pressure vessel; and second, to package this atmospheric sample canister in such a way that it can be easily integrated into the orbiting sample capsule for collection and return to Earth. Sample collection and integrity are demonstrated by emulating the atmospheric collection portion of the Mars Sample Return mission on a compressed timeline. The test results achieved by varying the pressure inside of a thermal vacuum chamber while opening and closing the valve on the sample canister at Mars ambient pressure. A commercial off-the-shelf medical grade micro-valve is utilized in the first iteration of this design to enable rapid testing of the system. The valve has been independently leak tested at JPL to quantify and separate the leak rates associated with the canister. The results are factored in to an overall system design that quantifies mass, power, and sensing requirements for a Martian atmospheric Sample Collection (MASC) canister as outlined in the Mars Sample Return mission profile. Qualitative results include the selection of materials to minimize sample contamination, preliminary science requirements, priorities in sample composition, flight valve selection criteria, a storyboard from sample collection to loading in the orbiting sample capsule, and contributions to maintaining “ Earth” clean exterior surfaces on the orbiting sample capsule.

  12. Defects which might occur in the copper-iron canister classified according to their likely effect on canister integrity

    Energy Technology Data Exchange (ETDEWEB)

    Bowyer, W.H. [Meadow End Farm, Farnham (United Kingdom)

    2000-06-15

    Earlier studies identified the material and manufacturing defects that might occur in serially produced canisters to the SKB reference design. This study has considered the defects, which were identified in the earlier works and classified them in terms of their importance to the durability of the canister in service. It has depended on, observations made by the writer over a seven-year involvement with SKI, literature studies and consultation with experts. For ease of reference each section of the report contains a table which includes information on defects taken from the earlier work plus the classification arising from this work. A study has been conducted to identify the material and manufacturing defects that might occur in serially produced canisters to the SKB reference design. The study has depended on cooperation of contractors engaged by SKB to participate in the development program, SKB staff, observations made by the writer over a five-year involvement with SKI, literature studies and consultation with experts. The candidate manufacturing procedures have been described inasmuch as it has been necessary to do so to make the points related to defects. Where possible, the cause of defects, their likely effects on manufacturing procedures or on durability of the canister and the methods available for their detection are given. For ease of reference each section of the report contains a table which summarises the information in it and, in the final section of the report, all the tables are presented en-bloc.

  13. Water soluble decontamination coating for Defense Waste Processing Facility (DWPF) canisters

    Energy Technology Data Exchange (ETDEWEB)

    Selby, C.L.

    1986-12-17

    Water soluble sodium borate glass coating was successfully codeveloped by Clemson University (Dr. H.G. Lefort) and Du Pont as an alternative decontamination process to frit slurry blasting of Defense Waste Processing Facility (DWPF) canisters. Slurry blasting requires transport of abrasive slurries, might cause galling by entrapped frit particles, and could result in frit slurry freezeup in pumps and retention basins. Contamination can be removed from precoated canisters with a gentle hot water rinse. Glass waste spilled on a coated canister will spall off spontaneously during canister cooling. A glass coating appears to prevent transfer of contamination to the Canister Decontamination Cell (CDC) guides and cradle. 1 ref., 5 tabs.

  14. Biological Research in Canisters (BRIC) - Light Emitting Diode (LED)

    Science.gov (United States)

    Levine, Howard G.; Caron, Allison

    2016-01-01

    The Biological Research in Canisters - LED (BRIC-LED) is a biological research system that is being designed to complement the capabilities of the existing BRIC-Petri Dish Fixation Unit (PDFU) for the Space Life and Physical Sciences (SLPS) Program. A diverse range of organisms can be supported, including plant seedlings, callus cultures, Caenorhabditis elegans, microbes, and others. In the event of a launch scrub, the entire assembly can be replaced with an identical back-up unit containing freshly loaded specimens.

  15. Thermal assessment of Shippingport pressurized water reactor blanket fuel assemblies within a multi-canister overpack within the canister storage building

    Energy Technology Data Exchange (ETDEWEB)

    HEARD, F.J.

    1999-04-09

    A series of analyses were performed to assess the thermal performance characteristics of the Shippingport Pressurized Water Reactor Core 2 Blanket Fuel Assemblies as loaded within a Multi-Canister Overpack within the Canister Storage Building. A two-dimensional finite element was developed, with enough detail to model the individual fuel plates: including the fuel wafers, cladding, and flow channels.

  16. Final Report: Characterization of Canister Mockup Weld Residual Stresses

    Energy Technology Data Exchange (ETDEWEB)

    Enos, David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-12-01

    Stress corrosion cracking (SCC) of interim storage containers has been indicated as a high priority data gap by the Department of Energy (DOE) (Hanson et al., 2012), the Electric Power Research Institute (EPRI, 2011), the Nuclear Waste Technical Review Board (NWTRB, 2010a), and the Nuclear Regulatory Commission (NRC, 2012a, 2012b). Uncertainties exist in terms of the environmental conditions that prevail on the surface of the storage containers, the stress state within the container walls associated both with weldments as well as within the base metal itself, and the electrochemical properties of the storage containers themselves. The goal of the work described in this document is to determine the stress states that exists at various locations within a typical storage canister by evaluating the properties of a full-diameter cylindrical mockup of an interim storage canister. This mockup has been produced using the same manufacturing procedures as the majority of the fielded spent nuclear fuel interim storage canisters. This document describes the design and procurement of the mockup and the characterization of the stress state associated with various portions of the container. It also describes the cutting of the mockup into sections for further analyses, and a discussion of the potential impact of the results from the stress characterization effort.

  17. PAUT inspection of copper canister: Structural attenuation and POD formulation

    Science.gov (United States)

    Gianneo, A.; Carboni, M.; Mueller, C.; Ronneteg, U.

    2016-02-01

    For inspection of thick-walled (50mm) copper canisters for final disposal of spent nuclear fuel in Sweden, ultrasonic inspection using phased array technique (PAUT) is applied. Because thick-walled copper is not commonly used as a structural material, previous experience on Phased Array Ultrasonic Testing for this type of application is limited. The paper presents the progress in understanding the amplitudes and attenuation changes acting on the Phased Array Ultrasonic Testing inspection of copper canisters. Previous studies showed the existence of a low pass filtering effect and a heterogeneous grain size distribution along the depth, thus affecting both the detectability of defects and their "Probability of Detection" determination. Consequently, the difference between the first and second back wall echoes were not sufficient to determine the local attenuation (within the inspection range), which affects the signal response for each individual defect. Experimental evaluation of structural attenuation was carried out onto step-wedge samples cut from full-size, extruded and pierced & drawn, copper canisters. Effective attenuation values has been implemented in numerical simulations to achieve a Multi Parameter Probability of Detection and to formulate a Model Assisted Probability of Detection through a Monte-Carlo extraction model.

  18. NDE to Manage Atmospheric SCC in Canisters for Dry Storage of Spent Fuel: An Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, Ryan M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pardini, Allan F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Cuta, Judith M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Adkins, Harold E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Casella, Andrew M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Qiao, Hong [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Larche, Michael R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Diaz, Aaron A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Doctor, Steven R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-09-01

    This report documents efforts to assess representative horizontal (Transuclear NUHOMS®) and vertical (Holtec HI-STORM) storage systems for the implementation of non-destructive examination (NDE) methods or techniques to manage atmospheric stress corrosion cracking (SCC) in canisters for dry storage of used nuclear fuel. The assessment is conducted by assessing accessibility and deployment, environmental compatibility, and applicability of NDE methods. A recommendation of this assessment is to focus on bulk ultrasonic and eddy current techniques for direct canister monitoring of atmospheric SCC. This assessment also highlights canister regions that may be most vulnerable to atmospheric SCC to guide the use of bulk ultrasonic and eddy current examinations. An assessment of accessibility also identifies canister regions that are easiest and more difficult to access through the ventilation paths of the concrete shielding modules. A conceivable sampling strategy for canister inspections is to sample only the easiest to access portions of vulnerable regions. There are aspects to performing an NDE inspection of dry canister storage system (DCSS) canisters for atmospheric SCC that have not been addressed in previous performance studies. These aspects provide the basis for recommendations of future efforts to determine the capability and performance of eddy current and bulk ultrasonic examinations for atmospheric SCC in DCSS canisters. Finally, other important areas of investigation are identified including the development of instrumented surveillance specimens to identify when conditions are conducive for atmospheric SCC, characterization of atmospheric SCC morphology, and an assessment of air flow patterns over canister surfaces and their influence on chloride deposition.

  19. LABAN emplacement pipe load-release test and stemming/canister alignment study

    Energy Technology Data Exchange (ETDEWEB)

    Howard, D.L.

    1983-12-02

    An Emplacement Pipe Load-Release Test and a study of downhole alignment during stemming were performed on the LABAN event. The purpose of these experiments was to determine canister and line of sight (LOS) distortion induced by downhole stemming and load-release procedures. The load-release test was aborted at approximately 40% completion due to excessive canister distortions. This report summarizes test results in terms of emplacement pipe loads vs vertical canister motions, canister and LOS lateral displacements, and the changes in LOS alignment that resulted from the downhole stemming and load-release processes.

  20. The Unity connecting module is moved to payload canister

    Science.gov (United States)

    1998-01-01

    In the Space Station Processing Facility, an overhead crane moves the Unity connecting module to the payload canister for transfer to the launch pad. Part of the International Space Station (ISS), Unity is scheduled for launch aboard Space Shuttle Endeavour on Mission STS-88 in December. The Unity is a connecting passageway to the living and working areas of ISS. While on orbit, the flight crew will deploy Unity from the payload bay and attach Unity to the Russian-built Zarya control module which will be in orbit at that time.

  1. The Unity connecting module is moved to payload canister

    Science.gov (United States)

    1998-01-01

    In the Space Station Processing Facility, workers attach the overhead crane that will lift the Unity connecting module from its workstand to move the module to the payload canister. Part of the International Space Station (ISS), Unity is scheduled for launch aboard Space Shuttle Endeavour on Mission STS-88 in December. The Unity is a connecting passageway to the living and working areas of ISS. While on orbit, the flight crew will deploy Unity from the payload bay and attach Unity to the Russian-built Zarya control module which will be in orbit at that time.

  2. Spent Nuclear Fuel Project Canister Storage Building Functions and Requirements

    Energy Technology Data Exchange (ETDEWEB)

    KLEM, M.J.

    2000-10-18

    In 1998, a major change in the technical strategy for managing Multi Canister Overpacks (MCO) while stored within the Canister Storage Building (CSB) occurred. The technical strategy is documented in Baseline Change Request (BCR) No. SNF-98-006, Simplified SNF Project Baseline (MCO Sealing) (FDH 1998). This BCR deleted the hot conditioning process initially adopted for the Spent Nuclear Fuel Project (SNF Project) as documented in WHC-SD-SNF-SP-005, Integrated Process Strategy for K Basins Spent Nuclear Fuel (WHC 199.5). In summary, MCOs containing Spent Nuclear Fuel (SNF) from K Basins would be placed in interim storage following processing through the Cold Vacuum Drying (CVD) facility. With this change, the needs for the Hot Conditioning System (HCS) and inerting/pressure retaining capabilities of the CSB storage tubes and the MCO Handling Machine (MHM) were eliminated. Mechanical seals will be used on the MCOs prior to transport to the CSB. Covers will be welded on the MCOs for the final seal at the CSB. Approval of BCR No. SNF-98-006, imposed the need to review and update the CSB functions and requirements baseline documented herein including changing the document title to ''Spent Nuclear Fuel Project Canister Storage Building Functions and Requirements.'' This revision aligns the functions and requirements baseline with the CSB Simplified SNF Project Baseline (MCO Sealing). This document represents the Canister Storage Building (CSB) Subproject technical baseline. It establishes the functions and requirements baseline for the implementation of the CSB Subproject. The document is organized in eight sections. Sections 1.0 Introduction and 2.0 Overview provide brief introductions to the document and the CSB Subproject. Sections 3.0 Functions, 4.0 Requirements, 5.0 Architecture, and 6.0 Interfaces provide the data described by their titles. Section 7.0 Glossary lists the acronyms and defines the terms used in this document. Section 8

  3. Multi Canister Overpack (MCO) Topical Report [SEC 1 THRU 3

    Energy Technology Data Exchange (ETDEWEB)

    LORENZ, B.D.

    2000-05-11

    In February 1995, the US Department of Energy (DOE) approved the Spent Nuclear Fuel (SNF) Project's ''Path Forward'' recommendation for resolution of the safety and environmental concerns associated with the deteriorating SNF stored in the Hanford Site's K Basins (Hansen 1995). The recommendation included an aggressive series of projects to design, construct, and operate systems and facilitates to permit the safe retrieval, packaging, transport, conditions, and interim storage of the K Basins' SNF. The facilities are the Cold VAcuum Drying Facility (CVDF) in the 100 K Area of the Hanford Site and the Canister Storage building (CSB) in the 200 East Area. The K Basins' SNF is to be cleaned, repackaged in multi-canister overpacks (MCOs), removed from the K Basins, and transported to the CVDF for initial drying. The MCOs would then be moved to the CSB and weld sealed (Loscoe 1996) for interim storage (about 40 years). One of the major tasks associated with the initial Path Forward activities is the development and maintenance of the safety documentation. In addition to meeting the construction needs for new structures, the safety documentation for each must be generated.

  4. Microwave Temperature Profiler Mounted in a Standard Airborne Research Canister

    Science.gov (United States)

    Mahoney, Michael J.; Denning, Richard F.; Fox, Jack

    2009-01-01

    Many atmospheric research aircraft use a standard canister design to mount instruments, as this significantly facilitates their electrical and mechanical integration and thereby reduces cost. Based on more than 30 years of airborne science experience with the Microwave Temperature Profiler (MTP), the MTP has been repackaged with state-of-the-art electronics and other design improvements to fly in one of these standard canisters. All of the controlling electronics are integrated on a single 4 5-in. (.10 13- cm) multi-layer PCB (printed circuit board) with surface-mount hardware. Improved circuit design, including a self-calibrating RTD (resistive temperature detector) multiplexer, was implemented in order to reduce the size and mass of the electronics while providing increased capability. A new microcontroller-based temperature controller board was designed, providing better control with fewer components. Five such boards are used to provide local control of the temperature in various areas of the instrument, improving radiometric performance. The new stepper motor has an embedded controller eliminating the need for a separate controller board. The reference target is heated to avoid possible emissivity (and hence calibration) changes due to moisture contamination in humid environments, as well as avoiding issues with ambient targets during ascent and descent. The radiometer is a double-sideband heterodyne receiver tuned sequentially to individual oxygen emission lines near 60 GHz, with the line selection and intermediate frequency bandwidths chosen to accommodate the altitude range of the aircraft and mission.

  5. Measurements of Fundamental Fluid Physics of SNF Storage Canisters

    Energy Technology Data Exchange (ETDEWEB)

    Condie, Keith Glenn; Mc Creery, Glenn Ernest; McEligot, Donald Marinus

    2001-09-01

    With the University of Idaho, Ohio State University and Clarksean Associates, this research program has the long-term goal to develop reliable predictive techniques for the energy, mass and momentum transfer plus chemical reactions in drying / passivation (surface oxidation) operations in the transfer and storage of spent nuclear fuel (SNF) from wet to dry storage. Such techniques are needed to assist in design of future transfer and storage systems, prediction of the performance of existing and proposed systems and safety (re)evaluation of systems as necessary at later dates. Many fuel element geometries and configurations are accommodated in the storage of spent nuclear fuel. Consequently, there is no one generic fuel element / assembly, storage basket or canister and, therefore, no single generic fuel storage configuration. One can, however, identify generic flow phenomena or processes which may be present during drying or passivation in SNF canisters. The objective of the INEEL tasks was to obtain fundamental measurements of these flow processes in appropriate parameter ranges.

  6. Thermal-hydraulic assessment of concrete storage cubicle with horizontal 3013 canisters

    Energy Technology Data Exchange (ETDEWEB)

    HEARD, F.J.

    1999-04-08

    The FIDAP computer code was used to perform a series of analyses to assess the thermal-hydraulic performance characteristics of the concrete plutonium storage cubicles, as modified for the horizontal placement of 3013 canisters. Four separate models were developed ranging from a full height model of the storage cubicle to a very detailed standalone model of a horizontal 3013 canister.

  7. Two-dimensional model of a Space Station Freedom thermal energy storage canister

    Science.gov (United States)

    Kerslake, Thomas W.; Ibrahim, Mounir B.

    1990-01-01

    The Solar Dynamic Power Module being developed for Space Station Freedom uses a eutectic mixture of LiF-CaF2 phase change salt contained in toroidal canisters for thermal energy storage. Results are presented from heat transfer analyses of the phase change salt containment canister. A 2-D, axisymmetric finite difference computer program which models the canister walls, salt, void, and heat engine working fluid coolant was developed. Analyses included effects of conduction in canister walls and solid salt, conduction and free convection in liquid salt, conduction and radiation across salt vapor filled void regions and forced convection in the heat engine working fluid. Void shape, location, growth or shrinkage (due to density difference between the solid and liquid salt phases) were prescribed based on engineering judgement. The salt phase change process was modeled using the enthalpy method. Discussion of results focuses on the role of free-convection in the liquid salt on canister heat transfer performance. This role is shown to be important for interpreting the relationship between ground based canister performance (in l-g) and expected on-orbit performance (in micro-g). Attention is also focused on the influence of void heat transfer on canister wall temperature distributions. The large thermal resistance of void regions is shown to accentuate canister hot spots and temperature gradients.

  8. Commercial radioactive waste management system feasibility with the universal canister concept. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Morissette, R.P.; Schneringer, P.E.; Lane, R.K.; Moore, R.L.; Young, K.A.

    1986-01-01

    A Program Research and Development Announcement (PRDA) was initiated by DOE to solicit from industry new and novel ideas for improvements in the nuclear waste management system. GA Technologies Inc. was contracted to study a system utilizing a universal canister which could be loaded at the reactor and used throughout the waste management system. The proposed canister was developed with the objective of meeting the mission requirements with maximum flexibility and at minimum cost. Canister criteria were selected from a thorough analysis of the spent fuel inventory, and canister concepts were evaluated along with the shipping and storage casks to determine the maximum payload. Engineering analyses were performed on various cask/canister combinations. One important criterion was the interchangeability of the canisters between truck and rail cask systems. A canister was selected which could hold three PWR intact fuel elements or up to eight consolidated PWR fuel elements. One canister could be shipped in an overweight truck cask or six in a rail cask. Economic analysis showed a cost savings of the reference system under consideration at that time.

  9. Aespoe Hard Rock Laboratory Canister Retrieval Test. Microorganisms in buffer from the Canister Retrieval Test - numbers and metabolic diversity

    Energy Technology Data Exchange (ETDEWEB)

    Lydmark, Sara; Pedersen, Karsten (Microbial Analytics Sweden AB (Sweden))

    2011-03-15

    'Canister Retrieval Test' (CRT) is an experiment that started at Aespoe Hard Rock Laboratory (HRL) 2000. CRT is a part of the investigations which evaluate a possible KBS-3 storage of nuclear waste. The primary aim was to see whether it is possible or not to retrieve a copper canister after storage under authentic KBS-3 conditions. However, CRT also provided a unique opportunity to investigate if bacteria survived in the bentonite buffer during storage. Therefore, in connection to the retrieval of the canister microbiological samples were extracted from the bentonite buffer and the bacterial composition was studied. In this report, microbiological analyses of a total of 66 samples at the C2, R10, R9 and R6 levels in the bentonite from CRT are presented and discussed. By culturing bacteria from the bentonite in specific media the following bacterial parameters were investigated: The total amount of culturable heterotrophic aerobic bacteria, sulphate-reducing bacteria, and bacteria that produce the organic compound acetate (acetogens). The biovolume in the bentonite was determined by detection of the ATP content. In addition, bacteria from the bentonite were cultured in different sulphate-reducing media. In these cultures, the presence of the biotic compounds sulphide and acetate was investigated, since these have potentially negative effect on the copper canister in a KBS-3 repository. The results were to some extent compared to density, water content, and temperature data provided by Clay Technology AB. The results showed that 100-102 viable sulphate-reducing and acetogenic bacteria and 102-104 heterotrophic aerobic bacteria g-1 bentonite were present after five years of storage in the rock. Bacteria with several morphologies could be found in the cultures with bentonite. The most bacteria were detected in the bentonite buffer close to the rock but in a few samples also in bentonite close to the copper canister. When the presence of bacteria in the

  10. Reliability in sealing of canister for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ronneteg, Ulf [Bodycote Materials Testing AB, Nykoeping (Sweden); Cederqvist, Lars; Ryden, Haakan [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden); Oeberg, Tomas [Tomas Oeberg Konsult AB, Karlskrona (Sweden); Mueller, Christina [Federal Inst. for Materials Research and Testing, Berlin (Germany)

    2006-06-15

    The reliability of the system for sealing the canister and inspecting the weld that has been developed for the Encapsulation plant was investigated. In the investigation the occurrence of discontinuities that can be formed in the welds was determined both qualitatively and quantitatively. The probability that these discontinuities can be detected by nondestructive testing (NDT) was also studied. The friction stir welding (FSW) process was verified in several steps. The variables in the welding process that determine weld quality were identified during the development work. In order to establish the limits within which they can be allowed to vary, a screening experiment was performed where the different process settings were tested according to a given design. In the next step the optimal process setting was determined by means of a response surface experiment, whereby the sensitivity of the process to different variable changes was studied. Based on the optimal process setting, the process window was defined, i.e. the limits within which the welding variables must lie in order for the process to produce the desired result. Finally, the process was evaluated during a demonstration series of 20 sealing welds which were carried out under production-like conditions. Conditions for the formation of discontinuities in welding were investigated. The investigations show that the occurrence of discontinuities is dependent on the welding variables. Discontinuities that can arise were classified and described with respect to characteristics, occurrence, cause and preventive measures. To ensure that testing of the welds has been done with sufficient reliability, the probability of detection (POD) of discontinuities by NDT and the accuracy of size determination by NDT were determined. In the evaluation of the demonstration series, which comprised 20 welds, a statistical method based on the generalized extreme value distribution was fitted to the size estimate of the indications

  11. Study of the consequences of secondary water radiolysis within and surrounding a defective canister

    Energy Technology Data Exchange (ETDEWEB)

    Jinsong Liu; Neretnieks, I. [Royal Inst. of Tech., Stockholm (Sweden). Dept. of Chemical Engineering and Technology; Stroemberg, Bo [Swedish Nuclear Power Inspectorate, Stockholm (Sweden)

    2000-11-01

    Consequences of secondary water radiolysis, caused by dispersed radionuclides released from spent nuclear fuel, both inside a defective canister and in the bentonite buffer surrounding the canister have been studied. The dissolution rate of the spent fuel is assumed to be controlled by chemical kinetics. Several cases have been addressed. First a simple mass balance model is presented. Some very conservative assumptions like complete failure of the canister one thousand years after its deposition in the repository and instantaneous oxidation rate of the spent fuel are deliberately made, to explore the upper bound limit of the effect of the secondary water radiolysis on the spent fuel dissolution. The model results show that the spent fuel could possibly be oxidised in an ever-increasing rate with these very simplified assumptions. More realistic and less conservative cases are then considered. In these cases, the canister is assumed to be initially defective with a hole of a few millimeters on its wall. The small hole will considerably restrict the transport of oxidants through the canister wall and the release of radionuclides to the outside of the canister. The spent fuel dissolution is assumed to be controlled by chemical kinetics at rates extrapolated from experimental studies. The cases are modelled with progressive complication. In the first case the effect of the secondary radiolysis inside fuel canister is neglected. It is also assumed that secondary phases of radionuclides do not precipitate inside the canister. The model results show that a relatively large domain of the near-field can be oxidised by the oxidants of secondary radiolysis. In the second case it is assumed that the radionuclide concentration within the canister is controlled by its respective solubility limit. The amount of radionuclides released out of the canister will then be limited by the solubility of the secondary phases. The effect of the secondary radiolysis will be quite limited in

  12. Analysis of sludge from Hanford K East Basin canisters

    Energy Technology Data Exchange (ETDEWEB)

    Makenas, B.J. [ed.] [comp.] [DE and S Hanford, Inc., Richland, WA (United States); Welsh, T.L. [B and W Protec, Inc. (United States); Baker, R.B. [DE and S Hanford, Inc., Richland, WA (United States); Hoppe, E.W.; Schmidt, A.J.; Abrefah, J.; Tingey, J.M.; Bredt, P.R.; Golcar, G.R. [Pacific Northwest National Lab., Richland, WA (United States)

    1997-09-12

    Sludge samples from the canisters in the Hanford K East Basin fuel storage pool have been retrieved and analyzed. Both chemical and physical properties have been determined. The results are to be used to determine the disposition of the bulk of the sludge and to assess the impact of residual sludge on dry storage of the associated intact metallic uranium fuel elements. This report is a summary and review of the data provided by various laboratories. Although raw chemistry data were originally reported on various bases (compositions for as-settled, centrifuged, or dry sludge) this report places all of the data on a common comparable basis. Data were evaluated for internal consistency and consistency with respect to the governing sample analysis plan. Conclusions applicable to sludge disposition and spent fuel storage are drawn where possible.

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

    Energy Technology Data Exchange (ETDEWEB)

    Colleen Shelton-Davis

    2005-11-01

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

  14. Molecular Contamination on Anodized Aluminum Components of the Genesis Science Canister

    Science.gov (United States)

    Burnett, D. S.; McNamara, K. M.; Jurewicz, A.; Woolum, D.

    2005-01-01

    Inspection of the interior of the Genesis science canister after recovery in Utah, and subsequently at JSC, revealed a darkening on the aluminum canister shield and other canister components. There has been no such observation of film contamination on the collector surfaces, and preliminary spectroscopic ellipsometry measurements support the theory that the films observed on the anodized aluminum components do not appear on the collectors to any significant extent. The Genesis Science Team has made an effort to characterize the thickness and composition of the brown stain and to determine if it is associated with molecular outgassing.Detailed examination of the surfaces within the Genesis science canister reveals that the brown contamination is observed to varying degrees, but only on surfaces exposed in space to the Sun and solar wind hydrogen. In addition, the materials affected are primarily composed of anodized aluminum. A sharp line separating the sun and shaded portion of the thermal closeout panel is shown. This piece was removed from a location near the gold foil collector within the canister. Future plans include a reassembly of the canister components to look for large-scale patterns of contamination within the canister to aid in revealing the root cause.

  15. Analysis for Eccentric Multi Canister Overpack (MCO) Drops at the Canister Storage Building (CSB) (CSB-S-0073)

    Energy Technology Data Exchange (ETDEWEB)

    HOLLENBECK, R.G.

    2000-06-07

    The purpose of this report is to investigate the potential for damage to the multi-canister overpack (MCO) during impact from an eccentric accidental drop onto the standard storage tube, overpack storage tube, service station or sampling/weld station. Damage to the storage tube and sample/weld station is beyond the scope of this report. The results of this analysis are required to show the following: (1) If a breach resulting in unacceptable release of contamination could occur in the MCO. (2) If the dropped MCO could become stuck in the storage tube after the drop. (3) Maximum deceleration of the spent nuclear fuel baskets. The model appropriate for the standard storage tubes with the smaller gap is the basis for the analysis and results reported herein in this SNF-5204, revision 2 report. Revision 1 of this report is based on a model that includes the larger gap appropriate for the overpack tubes.

  16. Evaluation of the potential for gas pressurization and free liquid accumulation in a WVDP canister

    Energy Technology Data Exchange (ETDEWEB)

    Hazelton, R.F.; Thornhill, C.K.

    1993-12-01

    A full-scale canister provided by the West Valley Demonstration Project, filled during the SF-11 vitrification qualification test, was tested to determine its potential for gas generation (non-radiolitic only) and liquid accumulation. The canister was sealed and held at a temperature of about 500{degrees}C for eight weeks. Gas samples obtained during the test were analyzed using mass spectroscopy to determine the composition of gases within the canister. At the end of the eight weeks the canister gases were evacuated through a desiccant to capture any water that had been released by the glass during the test. In addition, an analysis of the glass using fourier transform infrared spectroscopy was performed to determine the water content in the glass both before and after the temperature exposure.

  17. Multi Canister Overpack (MCO) Design Report [SEC 1 Thru 3

    Energy Technology Data Exchange (ETDEWEB)

    GOLDMANN, L.H.

    2000-02-29

    The MCO is designed to facilitate the removal, processing and storage of the spent nuclear fuel currently stored in the East and West K-Basins. The MCO is a stainless steel canister approximately 24 inches in diameter and 166 inches long with cover cap installed. The shell and the collar which is welded to the shell are fabricated from 304/304L dual certified stainless steel for the shell and F304/F304L dual certified for the collar. The shell has a nominal thickness of 1/2 inch. The top closure consists of a shield plug with four processing ports and a locking ring with jacking bolts to pre-load a metal seal under the shield plug. The fuel is placed in one of four types of baskets, excluding the SPR fuel baskets, in the fuel retention basin. Each basket is then loaded into the MCO which is inside the transfer cask. Once all of the baskets are loaded into the MCO, the shield plug with a process tube is placed into the open end of the MCO. This shield plug provides shielding for workers when the transfer cask, containing the MCO, is lifted from the pool. After being removed from the pool, the locking ring is installed and the jacking bolts are tightened to pre-load the metal main closure seal. The cask is then sealed and the MCO taken to the Cold Vacuum Drying (CVD) facility for bulk water removal and vacuum drying through the process ports. Covers for the process ports may be installed or removed as needed per operating procedures. The MCO is then transferred to the Canister Storage Building (CSB), in the closed transfer cask. At the CSB, the MCO is then removed from the cask and becomes one of two MCOs stacked in a storage tube. MCOs will have a cover cap welded over the shield plug providing a complete welded closure. A number of MCOs may be stored with just the mechanical seal to allow monitoring of the MCO pressure, temperature, and gas composition.

  18. Assessment study of the stresses induced by corrosion in the Advanced Cold Process Canister

    Energy Technology Data Exchange (ETDEWEB)

    Hoch, A.R.; Sharland, S.M. [Chemical Studies Department, Radwaste Disposal Division, AEA Decommissioning and Radwaste, Harwell Laboratory, Oxfordshire (United Kingdom)

    1993-10-01

    The Advanced Cold Process Canister (ACPC) is a concept for the encapsulation of spent nuclear fuel for geological disposal. The basic design of the ACPC consists of an outer oxygen free copper overpack covering a carbon steel inner container. In this report the stresses exerted on the copper overpack as a result of an early failure of the canister and the subsequent corrosion of the steel are calculated. 4 figs, 8 refs, 2 tabs.

  19. Draft report: Results of stainless steel canister corrosion studies and environmental sample investigations

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R. [Sandia National Laboratories, Albuquerque, NM (United States); Enos, David [Sandia National Laboratories, Albuquerque, NM (United States)

    2014-09-01

    This progress report describes work being done at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of used nuclear fuel. The work involves both characterization of the potential physical and chemical environment on the surface of the storage canisters and how it might evolve through time, and testing to evaluate performance of the canister materials under anticipated storage conditions.

  20. Physical properties of encapsulate spent fuel in canisters; Comportamiento fisico de las capsulas de almacenamiento

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-07-01

    Spent fuel and high-level wastes will be permanently stored in a deep geological repository (AGP). Prior to this, they will be encapsulated in canisters. The present report is dedicated to the study of such canisters under the different physical demands that they may undergo, be those in operating or accident conditions. The physical demands of interest include mechanical demands, both static and dynamic, and thermal demands. Consideration is given to the complete file of the canister, from the time when it is empty and without lid to the final conditions expected in the repository. Thermal analyses of canisters containing spent fuel are often carried out in two dimensions, some times with hypotheses of axial symmetry and some times using a plane transverse section through the centre of the canister. The results obtained in both types of analyses are compared here to those of complete three-dimensional analyses. The latter generate more reliable information about the temperatures that may be experienced by the canister and its contents; they also allow calibrating the errors embodied in the two-dimensional calculations. (Author)

  1. Design basis for the copper canister. Stage one

    Energy Technology Data Exchange (ETDEWEB)

    Bowyer, W. H. [ERA Technology Limited, Leatherhead, Surrey (United Kingdom)

    1995-02-01

    The copper/iron canister which has been proposed for containment of high level waste in the Swedish Nuclear Waste Disposal Programme has been studied from the points of view of choice of materials, manufacturing technology and quality assurance. The choice of High Strength Low Alloy steel for the load bearing element appears to be a good choice but it is necessary to understand the effect of laser welding on the structure of the chosen alloy and to ensure that the very rapid cooling rates which attend laser welding of thick material do not lead to the development of untempered martensite. The choice of an almost pure copper for the corrosion barrier is based on the very good corrosion resistance claimed for it under repository conditions. Production trials are in progress using this material and serious difficulties are expected both in manufacture and in quality assurance. The trials may or may not produce a satisfactory prototype but they will give pointers towards modifications in choice of material and processing technology. This study concludes that the chosen material is particularly difficult to process and to test, and that the claimed good corrosion resistance in in doubt. 54 refs.

  2. Inorganic analyses of volatilized and condensed species within prototypic Defense Waste Processing Facility (DWPF) canistered waste

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C.M.

    1992-06-30

    The high-level radioactive waste currently stored in carbon steel tanks at the Savannah River Site (SRS) will be immobilized in a borosilicate glass in the Defense Waste Processing Facility (DWPF). The canistered waste will be sent to a geologic repository for final disposal. The Waste Acceptance Preliminary Specifications (WAPS) require the identification of any inorganic phases that may be present in the canister that may lead to internal corrosion of the canister or that could potentially adversely affect normal canister handling. During vitrification, volatilization of mixed (Na, K, Cs)Cl, (Na, K, Cs){sub 2}SO{sub 4}, (Na, K, Cs)BF{sub 4}, (Na, K){sub 2}B{sub 4}O{sub 7} and (Na,K)CrO{sub 4} species from glass melt condensed in the melter off-gas and in the cyclone separator in the canister pour spout vacuum line. A full-scale DWPF prototypic canister filled during Campaign 10 of the SRS Scale Glass Melter was sectioned and examined. Mixed (NaK)CI, (NaK){sub 2}SO{sub 4}, (NaK) borates, and a (Na,K) fluoride phase (either NaF or Na{sub 2}BF{sub 4}) were identified on the interior canister walls, neck, and shoulder above the melt pour surface. Similar deposits were found on the glass melt surface and on glass fracture surfaces. Chromates were not found. Spinel crystals were found associated with the glass pour surface. Reference amounts of the halides and sulfates were found retained in the glass and the glass chemistry, including the distribution of the halides and sulfates, was homogeneous. In all cases where rust was observed, heavy metals (Zn, Ti, Sn) from the cutting blade/fluid were present indicating that the rust was a reaction product of the cutting fluid with glass and heat sensitized canister or with carbon-steel contamination on canister interior. Only minimal water vapor is present so that internal corrosion of the canister, will not occur.

  3. Description of Defense Waste Processing Facility reference waste form and canister. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Baxter, R.G.

    1983-08-01

    The Defense Waste Processing Facility (DWPF) will be located at the Savannah River Plant in Aiken, SC, and is scheduled for construction authorization during FY-1984. The reference waste form is borosilicate glass containing approx. 28 wt % sludge oxides, with the balance glass frit. Borosilicate glass was chosen because of its high resistance to leaching by water, its relatively high solubility for nuclides found in the sludge, and its reasonably low melting temperature. The glass frit contains about 58% SiO/sub 2/ and 15% B/sub 2/O/sub 3/. Leachabilities of SRP waste glasses are expected to approach 10/sup -8/ g/m/sup 2/-day based upon 1000-day tests using glasses containing SRP radioactive waste. Tests were performed under a wide variety of conditions simulating repository environments. The canister is filled with 3260 lb of glass which occupies about 85% of the free canister volume. The filled canister will generate approx. 470 watts when filled with oxides from 5-year-old sludge and 15-year-old supernate from the sludge and supernate processes. The radionuclide content of the canister is about 177,000 ci, with a radiation level of 5500 rem/h at canister surface contact. The reference canister is fabricated of standard 24-in.-OD, Schedule 20, 304L stainless steel pipe with a dished bottom, domed head, and a combined lifting and welding flange on the head neck. The overall canister length is 9 ft 10 in. with a 3/8-in. wall thickness. The 3-m canister length was selected to reduce equipment cell height in the DWPF to a practical size. The canister diameter was selected as an optimum size from glass quality considerations, a logical size for repository handling and to ensure that a filled canister with its double containment shipping cask could be accommodated on a legal-weight truck. The overall dimensions and weight appear to be compatible with preliminary assessments of repository requirements. 10 references.

  4. System-Level Logistics for Dual Purpose Canister Disposal

    Energy Technology Data Exchange (ETDEWEB)

    Kalinina, Elena A.

    2014-06-03

    The analysis presented in this report investigated how the direct disposal of dual purpose canisters (DPCs) may be affected by the use of standard transportation aging and disposal canisters (STADs), early or late start of the repository, and the repository emplacement thermal power limits. The impacts were evaluated with regard to the availability of the DPCs for emplacement, achievable repository acceptance rates, additional storage required at an interim storage facility (ISF) and additional emplacement time compared to the corresponding repackaging scenarios, and fuel age at emplacement. The result of this analysis demonstrated that the biggest difference in the availability of UNF for emplacement between the DPC-only loading scenario and the DPCs and STADs loading scenario is for a repository start date of 2036 with a 6 kW thermal power limit. The differences are also seen in the availability of UNF for emplacement between the DPC-only loading scenario and the DPCs and STADs loading scenario for the alternative with a 6 kW thermal limit and a 2048 start date, and for the alternatives with a 10 kW thermal limit and 2036 and 2048 start dates. The alternatives with disposal of UNF in both DPCs and STADs did not require additional storage, regardless of the repository acceptance rate, as compared to the reference repackaging case. In comparison to the reference repackaging case, alternatives with the 18 kW emplacement thermal limit required little to no additional emplacement time, regardless of the repository start time, the fuel loading scenario, or the repository acceptance rate. Alternatives with the 10 kW emplacement thermal limit and the DPCs and STADs fuel loading scenario required some additional emplacement time. The most significant decrease in additional emplacement time occurred in the alternative with the 6 kW thermal limit and the 2036 repository starting date. The average fuel age at emplacement ranges from 46 to 88 years. The maximum fuel age at

  5. Tests for manufacturing technology of disposal canisters for nuclear spent fuel; Loppusijoituskapselin valmistustekniset kokeet

    Energy Technology Data Exchange (ETDEWEB)

    Raiko, H. [VTT Energy (Finland); Salonen, T. [Outokumpu Poricopper Oy (Finland); Meuronen, I. [Suomen Teknohaus Oy (Finland); Lehto, K. [Valmet Oyj Rautpohja Foundry (Finland)

    1999-06-01

    The summary and status of the results of the manufacturing technology programmes concerning the disposal canister for spent nuclear fuel conducted by Posiva Oy are given in this report. Posiva has maintained a draft plan for a disposal canister design and an assessment of potential manufacturing technologies for about ten years in Finland. Now, during the year 1999, the first full scale demonstration canister is manufactured in Finland. The technology used for manufacturing of this prototype is developed by Posiva Oy mainly in co-operation with domestic industry. The main partner in developing the manufacturing technology for the copper shell has been Outokumpu Poricopper Oy, Pori, Finland, and the main partner in developing the technology for the iron insert of the canister has been Valmet Oyj Rautpohja Foundry, Jyvaeskylae, Finland. In both areas many subcontractors have been used, predominantly domestic engineering workshops, but also some foreign subcontractors, e.g. for EB-welding, who have had large enough welding equipment. This report describes the developing programmes for canister manufacturing, evaluates the results and presents some alternative methods, and tries to evaluate the pros and contras of them. In addition, the adequacy of the achieved technological know-how is assessed in respect of the required quality of the disposal canister. The following manufacturing technologies have been the concrete topics of the development programme: Electron beam welding technology development for thick-walled copper, Casting of massive copper billets, Hot rolling of thick-walled copper plates, Hot pressing and forging in lid manufacture, Extrusion and drawing of copper tubes, Bending of copper plates by roller or press, Machining of copper, Residual stress removal by heat treatment, Non-destructive testing, Long-term strength of EB-welds, Casting and machining of the iron insert of the canister The specialists from all the main developing partner companies have

  6. Progress in the understanding of the long-term corrosion behaviour of copper canisters

    Science.gov (United States)

    King, Fraser; Lilja, Christina; Vähänen, Marjut

    2013-07-01

    Copper has been proposed as a canister material for the disposal of spent nuclear fuel in a deep geologic repository in a number of countries worldwide. Since it was first proposed for this purpose in 1978, a significant number of studies have been performed to assess the corrosion performance of copper under repository conditions. These studies are critically reviewed and the suitability of copper as a canister material for nuclear waste is re-assessed. Over the past 30-35 years there has been considerable progress in our understanding of the expected corrosion behaviour of copper canisters. Crucial to this progress has been the improvement in the understanding of the nature of the repository environment and how it will evolve over time. With this improved understanding, it has been possible to predict the evolution of the corrosion behaviour from the initial period of warm, aerobic conditions in the repository to the long-term phase of cool, anoxic conditions dominated by the presence of sulphide. An historical review of the treatment of the corrosion behaviour of copper canisters is presented, from the initial corrosion assessment in 1978, through a major review of the corrosion behaviour in 2001, through to the current level of understanding based on the results of on-going studies. Compared with the initial corrosion assessment, there has been considerable progress in the treatment of localised corrosion, stress corrosion cracking, and microbiologically influenced corrosion of the canisters. Progress in the mechanistic modelling of the evolution of the corrosion behaviour of the canister is also reviewed, as is the continuing debate about the thermodynamic stability of copper in pure water. The overall conclusion of this critical review is that copper is a suitable material for the disposal of spent nuclear fuel and offers the prospect of containment of the waste for an extended period of time. The corrosion behaviour is influenced by the presence of the

  7. Development of a Universal Canister for Disposal of High-Level Waste in Deep Boreholes.

    Energy Technology Data Exchange (ETDEWEB)

    Price, Laura L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gomberg, Steve [USDOE, Washington, DC (United States)

    2015-11-01

    The mission of the United States Department of Energy’s Office of Environmental Management is to complete the safe cleanup of the environmental legacy brought about from five decades of nuclear weapons development and government-sponsored nuclear energy research. Some of the wastes that must be managed have been identified as good candidates for disposal in a deep borehole in crystalline rock. In particular, wastes that can be disposed of in a small package are good candidates for this disposal concept. A canister-based system that can be used for handling these wastes during the disposition process (i.e., storage, transfer, transportation, and disposal) could facilitate the eventual disposal of these wastes. Development of specifications for the universal canister system will consider the regulatory requirements that apply to storage, transportation, and disposal of the capsules, as well as operational requirements and limits that could affect the design of the canister (e.g., deep borehole diameter). In addition, there are risks and technical challenges that need to be recognized and addressed as Universal Canister system specifications are developed. This paper provides an approach to developing specifications for such a canister system that is integrated with the overall efforts of the DOE’s Used Fuel Disposition Campaign's Deep Borehole Field Test and compatible with planned storage of potential borehole-candidate wastes.

  8. A New Frangible Composite Canister Cover with the Function of Specified Direction Separation

    Science.gov (United States)

    Zhou, Guangming; Cai, Deng'an; Qian, Yuan; Deng, Jian; Wang, Xiaopei

    2016-08-01

    A lightweight and auto-separated canister cover is required for quick launching in some specific missile launchers. In this paper, a new frangible composite canister cover with the function of specified direction separation is proposed and studied via both experimental and numerical approaches. The frangible canister cover with local non-split weak zone structure, which is manufactured by traditional hand lay-up process with vacuum assisted resin infusion (VARI) method, is designed to fail and separate in a predetermined and specified direction in comparison with the cover with full split weak zone structure. This design is innovative and also necessary for reduction of potential risk to peripheral equipment around the missile launcher. The failure pressure of the cover is determined on the basis of the failure criteria used in finite element (FE) model. In experimental pressurized testing, a number of frangible canister covers subjected to pressure loadings in six cases are studied. Close agreements between the experimental and numerical results have been examined. The frangible canister covers with local non-split weak zone structure which have been studied can be separated and fly out to the specified direction.

  9. Deep penetrating eddy current for copper canister inspection. Main results

    Energy Technology Data Exchange (ETDEWEB)

    Tadeusz Stepinski [TSonic, Uppsala (Sweden)

    2004-02-01

    The aim of this project was to optimize the detection and characterization of deep flaws (voids) in copper plates. Two types of voids were investigated and compared: circular and rectangular. The circular voids had the form of cylindrical cavities while the rectangular ones were cavities with a rectangular cross section. All probes were of the same type, transmit-receive transducers with four pick-ups connected in a double differential configuration. Comparison of the EC responses to circular and rectangular voids obtained using the MDF12 probe has shown that both types of voids can be characterized using phase and amplitude of their responses in the complex impedance plane. Phase of the response in the impedance plane appeared to be a reliable measure of void depth. Phase dependence on the void depth is linear (which agrees with the theory) and its slope is approx -37 deg/mm. Magnitude of the EC response contains information on the void size provided that the void depth is known. It has been shown that magnitude of the EC responses is correlated to the lengths of the rectangular voids and hole diameter, this is, similar lengths and diameters result in similar response magnitudes. It should be noted, however, that multi-differential MDF probes generate responses with different shapes for circular and rectangular voids. First, shapes of the MDF probe responses in the impedance plane depend on the probe's orientation with respect to scanning direction. Second, they also depend upon the direction of scanning with respect to the void orientation. The measurements presented in this report were performed for the probe axis aligned along with the scanning direction and, in case of rectangular voids, for scanning direction along the void lengths. Comparison of the responses obtained from flat bottom holes in copper material taken from different canister parts has not shown any essential differences between the material samples. Conductivity measurement performed using

  10. Corrosion of the copper canister in the repository environment

    Energy Technology Data Exchange (ETDEWEB)

    Hermansson, H.P.; Eriksson, Sture [Studsvik Material AB, Nykoeping (Sweden)

    1999-12-01

    The present report accounts for studies on copper corrosion performed at Studsvik Material AB during 1997-1999 on commission by SKI. The work has been focused on localised corrosion and electrochemistry of copper in the repository environment. The current theory of localised copper corrosion is not consistent with recent practical experiences. It is therefore desired to complete and develop the theory based on knowledge about the repository environment and evaluations of previous as well as recent experimental and field results. The work has therefore comprised a thorough compilation and up-date of literature on copper corrosion and on the repository environment. A selection of a 'working environment', defining the chemical parameters and their ranges of variation has been made and is used as a fundament for the experimental part of the work. Experiments have then been performed on the long-range electrochemical behaviour of copper in selected environments simulating the repository. Another part of the work has been to further develop knowledge about the thermodynamic limits for corrosion in the repository environment. Some of the thermodynamic work is integrated here. Especially thermodynamics for the system Cu-Cl-H-O up to 150 deg C and high chloride concentrations are outlined. However, there is also a rough overview of the whole system Cu-Fe-Cl-S-C-H-O as a fundament for the discussion. Data are normally accounted as Pourbaix diagrams. Some of the conclusions are that general corrosion on copper will probably not be of significant importance in the repository as far as transportation rates are low. However, if such rates were high, general corrosion could be disastrous, as there is no passivation of copper in the highly saline environment. The claim on knowledge of different kinds of localised corrosion and pitting is high, as pitting damages can shorten the lifetime of a canister dramatically. Normal pitting can happen in oxidising environment, but

  11. Three-Dimensional Heat Transfer Analysis for A Thermal Energy Storage Canister

    Institute of Scientific and Technical Information of China (English)

    Hou Xinbin; Xin Yuming; Yang Chunxin; Yuan Xiugan; Dong Keyong

    2001-01-01

    High temperature latent thermal storage is one of the critical techniques for a solar dynamic power system. This paper presents results from heat transfer analysis of a phase change salt containment canister. A three dimensional analysis program is developed to model heat transfer in a PCM canister. Analysis include effects of asymmetric circumference heat flux, conduction in canister walls, liquid PCM and solid PCM, void volume change and void location, and conduction and radiation across PCM vapor void. The PCM phase change process is modeled using the enthalpy method and the simulation results are compared with those of other two dimensional investigations. It's shown that there are large difference with two-dimensional analysis, therefore the three-dimensional model is necessary for system design of high temperature latent thermal storage.

  12. Uncertainty quantification methodologies development for stress corrosion cracking of canister welds

    Energy Technology Data Exchange (ETDEWEB)

    Dingreville, Remi Philippe Michel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-09-30

    This letter report presents a probabilistic performance assessment model to evaluate the probability of canister failure (through-wall penetration) by SCC. The model first assesses whether environmental conditions for SCC – the presence of an aqueous film – are present at canister weld locations (where tensile stresses are likely to occur) on the canister surface. Geometry-specific storage system thermal models and weather data sets representative of U.S. spent nuclear fuel (SNF) storage sites are implemented to evaluate location-specific canister surface temperature and relative humidity (RH). As the canister cools and aqueous conditions become possible, the occurrence of corrosion is evaluated. Corrosion is modeled as a two-step process: first, pitting is initiated, and the extent and depth of pitting is a function of the chloride surface load and the environmental conditions (temperature and RH). Second, as corrosion penetration increases, the pit eventually transitions to a SCC crack, with crack initiation becoming more likely with increasing pit depth. Once pits convert to cracks, a crack growth model is implemented. The SCC growth model includes rate dependencies on both temperature and crack tip stress intensity factor, and crack growth only occurs in time steps when aqueous conditions are predicted. The model suggests that SCC is likely to occur over potential SNF interim storage intervals; however, this result is based on many modeling assumptions. Sensitivity analyses provide information on the model assumptions and parameter values that have the greatest impact on predicted storage canister performance, and provide guidance for further research to reduce uncertainties.

  13. Radon-222 activity flux measurement using activated charcoal canisters: revisiting the methodology.

    Science.gov (United States)

    Alharbi, Sami H; Akber, Riaz A

    2014-03-01

    The measurement of radon ((222)Rn) activity flux using activated charcoal canisters was examined to investigate the distribution of the adsorbed (222)Rn in the charcoal bed and the relationship between (222)Rn activity flux and exposure time. The activity flux of (222)Rn from five sources of varying strengths was measured for exposure times of one, two, three, five, seven, 10, and 14 days. The distribution of the adsorbed (222)Rn in the charcoal bed was obtained by dividing the bed into six layers and counting each layer separately after the exposure. (222)Rn activity decreased in the layers that were away from the exposed surface. Nevertheless, the results demonstrated that only a small correction might be required in the actual application of charcoal canisters for activity flux measurement, where calibration standards were often prepared by the uniform mixing of radium ((226)Ra) in the matrix. This was because the diffusion of (222)Rn in the charcoal bed and the detection efficiency as a function of the charcoal depth tended to counterbalance each other. The influence of exposure time on the measured (222)Rn activity flux was observed in two situations of the canister exposure layout: (a) canister sealed to an open bed of the material and (b) canister sealed over a jar containing the material. The measured (222)Rn activity flux decreased as the exposure time increased. The change in the former situation was significant with an exponential decrease as the exposure time increased. In the latter case, lesser reduction was noticed in the observed activity flux with respect to exposure time. This reduction might have been related to certain factors, such as absorption site saturation or the back diffusion of (222)Rn gas occurring at the canister-soil interface.

  14. Coupled Transport/Reaction Modelling of Copper Canister Corrosion Aided by Microbial Processes

    Energy Technology Data Exchange (ETDEWEB)

    Jinsong Liu [Royal Institute of Technology, Stockholm (Sweden). Dept. of Chemical Engineering and Technology

    2006-04-15

    Copper canister corrosion is an important issue in the concept of a nuclear fuel repository. Previous studies indicate that the oxygen-free copper canister could hold its integrity for more than 100,000 years in the repository environment. Microbial processes may reduce sulphate to sulphide and considerably increase the amount of sulphides available for corrosion. In this paper, a coupled transport/reaction model is developed to account for the transport of chemical species produced by microbial processes. The corroding agents like sulphide would come not only from the groundwater flowing in a fracture that intersects the canister, but also from the reduction of sulphate near the canister. The reaction of sulphate-reducing bacteria and the transport of sulphide in the bentonite buffer are included in the model. The depth of copper canister corrosion is calculated by the model. With representative 'central values' of the concentrations of sulphate and methane at repository depth at different sites in Fennoscandian Shield the corrosion depth predicted by the model is a few millimetres during 10{sup 5} years. As the concentrations of sulphate and methane are extremely site-specific and future climate changes may significantly influence the groundwater compositions at potential repository sites, sensitivity analyses have been conducted. With a broad perspective of the measured concentrations at different sites in Sweden and in Finland, and some possible mechanisms (like the glacial meltwater intrusion and interglacial seawater intrusion) that may introduce more sulphate into the groundwater at intermediate depths during future climate changes, higher concentrations of either/both sulphate and methane than what is used as the representative 'central' values would be possible. In worst cases. locally, half of the canister thickness could possibly be corroded within 10{sup 5} years.

  15. System Configuration Management Implementation Procedure for the Canister Storage Building (CSB)

    Energy Technology Data Exchange (ETDEWEB)

    GARRISON, R.C.

    2000-11-28

    This document provides configuration management for the Distributed Control System (DCS), the Gaseous Effluent Monitoring System (GEMS-100) System, the Heating Ventilation and Air Conditioning (HVAC) Programmable Logic Controller (PLC), the Canister Receiving Crane (CRC) CRN-001 PLC, and both North and South vestibule door interlock system PLCs at the Canister Storage Building (CSB). This procedure identifies and defines software configuration items in the CSB control and monitoring systems, and defines configuration control throughout the system life cycle. Components of this control include: configuration status accounting; physical protection and control; and verification of the completeness and correctness of these items.

  16. Transportation considerations related to waste forms and canisters for Defense TRU wastes

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, K.J.; Andrews, W.B.; Schreiber, A.M.; Rosenthal, L.J.; Odle, C.J.

    1981-09-01

    This report identifies and discusses the considerations imposed by transportation on waste forms and canisters for contact-handled, solid transuranic wastes from the US Department of Energy (DOE) activities. The report reviews (1) the existing raw waste forms and potential immobilized waste forms, (2) the existing and potential future DOE waste canisters and shipping containers, (3) regulations and regulatory trends for transporting commercial transuranic wastes on the ISA, (4) truck and rail carrier requirements and preferences for transporting the wastes, and (5) current and proposed Type B external packagings for transporting wastes.

  17. Friction stir welding - an alternative method for sealing nuclear waste storage canisters

    Energy Technology Data Exchange (ETDEWEB)

    Andrews, R.E. [TWI Ltd, Cambridge (United Kingdom)

    2004-12-01

    When welding 50 mm thick copper a very high heat input is required to combat the high thermal diffusivity and only the Electron Beam Welding (EBW) process had this capability when this copper canister concept was conceived. Despite the encouraging results achieved using EBW with thick section copper, SKB felt that it would be prudent to assess other joining methods. This assessment concluded that friction welding, could also provide very high quality welds to satisfy the service life requirements of the SKB canister design. A friction welding variant called Friction Stir Welding (FSW) was shown to have the capability of welding 3 mm thick copper sheet with excellent integrity and reproducibility. This later provided sufficient encouragement for SKB to consider the potential of FSW as a method for joining thick section copper, using relatively simple machine tool based technology. It was thought that FSW might provide an alternative or complementary method for welding lids, or bases to canisters. In 1997 an FSW development programme started at TWI, focussed on the feasibility of welding 10 mm thick copper plate. Once this task was successfully completed, work continued to demonstrate that progressively thicker plate, up to 50 mm thick, could be joined. At this stage, with process viability established, a full size experimental FSW canister machine was designed and built. Work with this machine finished in January 2003, when it had been shown that FSW could definitely be used to weld lids to full size canisters. This report summarises the TWI development of FSW for SKB from 1997 to January 2003. It also highlights the important aspects of the process and the project milestones that will help to ensure that SKB has a welding technology that can be used with confidence for production fabrication of copper waste storage canisters in the future. The overall conclusion to this FSW development is that there is no doubt that the FSW process could be used to produce full

  18. Results of stainless steel canister corrosion studies and environmental sample investigations

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R. [Sandia National Laboratories, Albuquerque, NM (United States); Enos, David [Sandia National Laboratories, Albuquerque, NM (United States)

    2014-12-01

    This progress report describes work being done at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of used nuclear fuel. The work involves both characterization of the potential physical and chemical environment on the surface of the storage canisters and how it might evolve through time, and testing to evaluate performance of the canister materials under anticipated storage conditions. To evaluate the potential environment on the surface of the canisters, SNL is working with the Electric Power Research Institute (EPRI) to collect and analyze dust samples from the surface of in-service SNF storage canisters. In FY 13, SNL analyzed samples from the Calvert Cliffs Independent Spent Fuel Storage Installation (ISFSI); here, results are presented for samples collected from two additional near-marine ISFSI sites, Hope Creek NJ, and Diablo Canyon CA. The Hope Creek site is located on the shores of the Delaware River within the tidal zone; the water is brackish and wave action is normally minor. The Diablo Canyon site is located on a rocky Pacific Ocean shoreline with breaking waves. Two types of samples were collected: SaltSmart™ samples, which leach the soluble salts from a known surface area of the canister, and dry pad samples, which collected a surface salt and dust using a swipe method with a mildly abrasive ScotchBrite™ pad. The dry samples were used to characterize the mineralogy and texture of the soluble and insoluble components in the dust via microanalytical techniques, including mapping X-ray Fluorescence spectroscopy and Scanning Electron Microscopy. For both Hope Creek and Diablo Canyon canisters, dust loadings were much higher on the flat upper surfaces of the canisters than on the vertical sides. Maximum dust sizes collected at both sites were slightly larger than 20 μm, but Phragmites grass seeds ~1 mm in size, were observed on the tops of the Hope Creek canisters

  19. Multi-dimensional modeling of a thermal energy storage canister. M.S. Thesis - Cleveland State Univ., Dec. 1990

    Science.gov (United States)

    Kerslake, Thomas W.

    1991-01-01

    The Solar Dynamic Power Module being developed for Space Station Freedom uses a eutectic mixture of LiF-CaF2 phase change material (PCM) contained in toroidal canisters for thermal energy storage. Presented are the results from heat transfer analyses of a PCM containment canister. One and two dimensional finite difference computer models are developed to analyze heat transfer in the canister walls, PCM, void, and heat engine working fluid coolant. The modes of heat transfer considered include conduction in canister walls and solid PCM, conduction and pseudo-free convection in liquid PCM, conduction and radiation across PCM vapor filled void regions, and forced convection in the heat engine working fluid. Void shape, location, growth or shrinkage (due to density difference between the solid and liquid PCM phases) are prescribed based on engineering judgment. The PCM phase change process is analyzed using the enthalpy method. The discussion of the results focuses on how canister thermal performance is affected by free convection in the liquid PCM and void heat transfer. Characterizing these effects is important for interpreting the relationship between ground-based canister performance (in 1-g) and expected on-orbit performance (in micro-g). Void regions accentuate canister hot spots and temperature gradients due to their large thermal resistance. Free convection reduces the extent of PCM superheating and lowers canister temperatures during a portion of the PCM thermal charge period. Surprisingly small differences in canister thermal performance result from operation on the ground and operation on-orbit. This lack of a strong gravity dependency is attributed to the large contribution of container walls in overall canister energy redistribution by conduction.

  20. Examining the role of canister cooling conditions on the formation of nepheline from nuclear waste glasses

    Energy Technology Data Exchange (ETDEWEB)

    Christian, J. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-09-01

    Nepheline (NaAlSiO₄) crystals can form during slow cooling of high-level waste (HLW) glass after it has been poured into a waste canister. Formation of these crystals can adversely affect the chemical durability of the glass. The tendency for nepheline crystallization to form in a HLW glass increases with increasing concentrations of Al₂O₃ and Na₂O.

  1. Instrumentation. Nondestructive Examination for Verification of Canister and Cladding Integrity - FY2013 Status Update

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, Ryan M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jones, Anthony M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pardini, Allan F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Denslow, Kayte M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Crawford, Susan L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Larche, Michael R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-09-30

    This report documents FY13 efforts for two instrumentation subtasks under storage and transportation. These instrumentation tasks relate to developing effective nondestructive evaluation (NDE) methods and techniques to (1) verify the integrity of metal canisters for the storage of used nuclear fuel (UNF) and to (2) characterize hydrogen effects in UNF cladding to facilitate safe storage and retrieval.

  2. Instrumentation: Nondestructive Examination for Verification of Canister and Cladding Integrity. FY2014 Status Update

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, Ryan M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Suter, Jonathan D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jones, Anthony M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2014-09-12

    This report documents FY14 efforts for two instrumentation subtasks under storage and transportation. These instrumentation tasks relate to developing effective nondestructive evaluation (NDE) methods and techniques to (1) verify the integrity of metal canisters for the storage of used nuclear fuel (UNF) and to (2) verify the integrity of dry storage cask internals.

  3. Fuel and canister process report for the safety assessment SR-Site

    Energy Technology Data Exchange (ETDEWEB)

    Werme, Lars; Lilja, Christina (eds.)

    2010-12-15

    This report documents fuel and canister processes identified as relevant to the long-term safety of a KBS-3 repository. It forms an important part of the reporting of the safety assessment SR-Site. The detailed assessment methodology, including the role of the process reports in the assessment, is described in the SR-Site Main report /SKB 2011/

  4. Miniature Canister (MiniCan) Corrosion experiment progress report 4 for 2008-2011

    Energy Technology Data Exchange (ETDEWEB)

    Smart, Nick; Reddy, Bharti; Rance, Andy [Serco, Hook (United Kingdom)

    2012-06-15

    To ensure the safe encapsulation of spent nuclear fuel rods for geological disposal, SKB of Sweden are considering using the Copper-Iron Canister, which consists of an outer copper canister and a cast iron insert. Over the years a programme of laboratory work has been carried out to investigate a range of corrosion issues associated with the canister, including the possibility of expansion of the outer copper canister as a result of the anaerobic corrosion of the cast iron insert. Previous experimental work using stacks of test specimens has not shown any evidence of corrosion-induced expansion. However, as a further step in developing an understanding of the likely performance of the canister in a repository environment, Serco has set up a series of experiments in SKB's Aespoe Hard Rock Laboratory (HRL) using inactive model canisters, in which leaks were deliberately introduced into the outer copper canister while surrounded by bentonite, with the aim of obtaining information about the internal corrosion evolution of the internal environment. The experiments use five small scale model canisters (300 mm long x 150 mm diameter) that simulate the main features of the SKB canister design (hence the project name, 'MiniCan'). The main aim of the work is to examine how corrosion of the cast iron insert will evolve if a leak is present in the outer copper canister. This report describes the progress on the five experiments running at the Aespoe Hard Rock Laboratory and the data obtained from the start of the experiments in late 2006 up to Winter 2011. The full details of the design and installation of the experiments are given in a previous report and this report concentrates on summarising and interpreting the data obtained to date. This report follows the earlier progress reports presenting results up to December 2010. The current document (progress report 4) describes work up to December 2011. The current report presents the results of the water analyses

  5. Miniature Canister (MiniCan) Corrosion Experiment Progress Report 3 for 2008-2010

    Energy Technology Data Exchange (ETDEWEB)

    Smart, N.R.; Reddy, B.; Rance, A.P. (Serco (United Kingdom))

    2011-08-15

    To ensure the safe encapsulation of spent nuclear fuel rods for geological disposal, SKB of Sweden are considering using the Copper-Iron Canister, which consists of an outer copper canister and a cast iron insert. Over the years a programme of laboratory work has been carried out to investigate a range of corrosion issues associated with the canister, including the possibility of expansion of the outer copper canister as a result of the anaerobic corrosion of the cast iron insert. Previous experimental work using stacks of test specimens has not shown any evidence of corrosion-induced expansion. However, as a further step in developing an understanding of the likely performance of the canister in a repository environment, Serco has set up a series of experiments in SKB's Aespoe Hard Rock Laboratory (HRL) using inactive model canisters, in which leaks were deliberately introduced into the outer copper canister while surrounded by bentonite, with the aim of obtaining information about the internal corrosion evolution of the internal environment. The experiments use five small-scale model canisters (300 mm long x 150 mm diameter) that simulate the main features of the SKB canister design (hence the project name, 'MiniCan'). The main aim of the work is to examine how corrosion of the cast iron insert will evolve if a leak is present in the outer copper canister. This report describes the progress on the five experiments running at the Aespoe Hard Rock Laboratory and the data obtained from the start of the experiments in late 2006 up to Winter 2010. The full details of the design and installation of the experiments are given in a previous report and this report concentrates on summarising and interpreting the data obtained to date. This report follows two earlier progress reports presenting results up to December 2009. The current document (progress report 3) describes work up to December 2010. The current report presents the results of the water analyses

  6. Clean Assembly of Genesis Collector Canister for Flight: Lessons for Planetary Sample Return

    Science.gov (United States)

    Allton, J. H.; Stansbery, E. K.; Allen, C. C.; Warren, J. L.; Schwartz, C. M.

    2007-01-01

    Measurement of solar composition in the Genesis collectors requires not only high sensitivity but very low blanks; thus, very strict collector contamination minimization was required beginning with mission planning and continuing through hardware design, fabrication, assembly and testing. Genesis started with clean collectors and kept them clean inside of a canister. The mounting hardware and container for the clean collectors were designed to be cleanable, with access to all surfaces for cleaning. Major structural components were made of aluminum and cleaned with megasonically energized ultrapure water (UPW). The UPW purity was >18 M resistivity. Although aluminum is relatively difficult to clean, the Genesis protocol achieved level 25 and level 50 cleanliness on large structural parts; however, the experience suggests that surface treatments may be helpful on future missions. All cleaning was performed in an ISO Class 4 (Class 10) cleanroom immediately adjacent to an ISO Class 4 assembly room; thus, no plastic packaging was required for transport. Persons assembling the canister were totally enclosed in cleanroom suits with face shield and HEPA filter exhaust from suit. Interior canister materials, including fasteners, were installed, untouched by gloves, using tweezers and other stainless steel tools. Sealants/lubricants were not exposed inside the canister, but vented to the exterior and applied in extremely small amounts using special tools. The canister was closed in ISO Class 4, not to be opened until on station at Earth-Sun L1. Throughout the cleaning and assembly, coupons of reference materials that were cleaned at the same time as the flight hardware were archived for future reference and blanks. Likewise reference collectors were archived. Post-mission analysis of collectors has made use of these archived reference materials.

  7. Evaluation of DUSTRAN Software System for Modeling Chloride Deposition on Steel Canisters

    Energy Technology Data Exchange (ETDEWEB)

    Tran, Tracy T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jensen, Philip J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Fritz, Brad G. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rutz, Frederick C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Devanathan, Ram [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-07-29

    The degradation of steel by stress corrosion cracking (SCC) when exposed to atmospheric conditions for decades is a significant challenge in the fossil fuel and nuclear industries. SCC can occur when corrosive contaminants such as chlorides are deposited on a susceptible material in a tensile stress state. The Nuclear Regulatory Commission has identified chloride-induced SCC as a potential cause for concern in stainless steel used nuclear fuel (UNF) canisters in dry storage. The modeling of contaminant deposition is the first step in predictive multiscale modeling of SCC that is essential to develop mitigation strategies, prioritize inspection, and ensure the integrity and performance of canisters, pipelines, and structural materials. A multiscale simulation approach can be developed to determine the likelihood that a canister would undergo SCC in a certain period of time. This study investigates the potential of DUSTRAN, a dust dispersion modeling system developed by Pacific Northwest National Laboratory, to model the deposition of chloride contaminants from sea salt aerosols on a steel canister. Results from DUSTRAN simulations run with historical meteorological data were compared against measured chloride data at a coastal site in Maine. DUSTRAN’s CALPUFF model tended to simulate concentrations higher than those measured; however, the closest estimations were within the same order of magnitude as the measured values. The decrease in discrepancies between measured and simulated values as the level of abstraction in wind speed decreased suggest that the model is very sensitive to wind speed. However, the influence of other parameters such as the distinction between open-ocean and surf-zone sources needs to be explored further. Deposition values predicted by the DUSTRAN system were not in agreement with concentration values and suggest that the deposition calculations may not fully represent physical processes. Overall, results indicate that with parameter

  8. SLUDGE TREATMENT PROJECT COST COMPARISON BETWEEN HYDRAULIC LOADING AND SMALL CANISTER LOADING CONCEPTS

    Energy Technology Data Exchange (ETDEWEB)

    GEUTHER J; CONRAD EA; RHOADARMER D

    2009-08-24

    The Sludge Treatment Project (STP) is considering two different concepts for the retrieval, loading, transport and interim storage of the K Basin sludge. The two design concepts under consideration are: (1) Hydraulic Loading Concept - In the hydraulic loading concept, the sludge is retrieved from the Engineered Containers directly into the Sludge Transport and Storage Container (STSC) while located in the STS cask in the modified KW Basin Annex. The sludge is loaded via a series of transfer, settle, decant, and filtration return steps until the STSC sludge transportation limits are met. The STSC is then transported to T Plant and placed in storage arrays in the T Plant canyon cells for interim storage. (2) Small Canister Concept - In the small canister concept, the sludge is transferred from the Engineered Containers (ECs) into a settling vessel. After settling and decanting, the sludge is loaded underwater into small canisters. The small canisters are then transferred to the existing Fuel Transport System (FTS) where they are loaded underwater into the FTS Shielded Transfer Cask (STC). The STC is raised from the basin and placed into the Cask Transfer Overpack (CTO), loaded onto the trailer in the KW Basin Annex for transport to T Plant. At T Plant, the CTO is removed from the transport trailer and placed on the canyon deck. The CTO and STC are opened and the small canisters are removed using the canyon crane and placed into an STSC. The STSC is closed, and placed in storage arrays in the T Plant canyon cells for interim storage. The purpose of the cost estimate is to provide a comparison of the two concepts described.

  9. Results of stainless steel canister corrosion studies and environmental sample investigations

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R. [Sandia National Laboratories, Albuquerque, NM (United States); Enos, David [Sandia National Laboratories, Albuquerque, NM (United States)

    2014-12-01

    This progress report describes work being done at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of used nuclear fuel. The work involves both characterization of the potential physical and chemical environment on the surface of the storage canisters and how it might evolve through time, and testing to evaluate performance of the canister materials under anticipated storage conditions. To evaluate the potential environment on the surface of the canisters, SNL is working with the Electric Power Research Institute (EPRI) to collect and analyze dust samples from the surface of in-service SNF storage canisters. In FY 13, SNL analyzed samples from the Calvert Cliffs Independent Spent Fuel Storage Installation (ISFSI); here, results are presented for samples collected from two additional near-marine ISFSI sites, Hope Creek NJ, and Diablo Canyon CA. The Hope Creek site is located on the shores of the Delaware River within the tidal zone; the water is brackish and wave action is normally minor. The Diablo Canyon site is located on a rocky Pacific Ocean shoreline with breaking waves. Two types of samples were collected: SaltSmart™ samples, which leach the soluble salts from a known surface area of the canister, and dry pad samples, which collected a surface salt and dust using a swipe method with a mildly abrasive ScotchBrite™ pad. The dry samples were used to characterize the mineralogy and texture of the soluble and insoluble components in the dust via microanalytical techniques, including mapping X-ray Fluorescence spectroscopy and Scanning Electron Microscopy. For both Hope Creek and Diablo Canyon canisters, dust loadings were much higher on the flat upper surfaces of the canisters than on the vertical sides. Maximum dust sizes collected at both sites were slightly larger than 20 μm, but Phragmites grass seeds ~1 mm in size, were observed on the tops of the Hope Creek canisters

  10. Removal plan for Shippingport pressurized water reactor core 2 blanket fuel assemblies form T plant to the canister storage building

    Energy Technology Data Exchange (ETDEWEB)

    Lata

    1996-09-26

    This document presents the current strategy and path forward for removal of the Shippingport Pressurized Water Reactor Core 2 blanket fuel assemblies from their existing storage configuration (wet storage within the T Plant canyon) and transport to the Canister Storage Building (designed and managed by the Spent Nuclear Fuel. Division). The removal plan identifies all processes, equipment, facility interfaces, and documentation (safety, permitting, procedures, etc.) required to facilitate the PWR Core 2 assembly removal (from T Plant), transport (to the Canister storage Building), and storage to the Canister Storage Building. The plan also provides schedules, associated milestones, and cost estimates for all handling activities.

  11. The effect of discontinuities on the corrosion behaviour of copper canisters

    Energy Technology Data Exchange (ETDEWEB)

    King, F. [Integrity Corrosion Consulting Ltd, Calgary, AL (Canada)

    2004-03-01

    Discontinuities may remain in the weld region of copper canisters following the final closure welding and inspection procedures. Although the shell of the copper canister is expected to exhibit excellent corrosion properties in the repository environment, the question remains what impact these discontinuities might have on the long-term performance and service life of the canister. A review of the relevant corrosion literature has been carried out and an expert opinion of the impact of these discontinuities on the canister lifetime has been developed. Since the amount of oxidant in the repository is limited and the maximum wall penetration is expected to be < 2 mm, discontinuities will only be significant if they impact the localised corrosion or stress corrosion cracking (SCC) behaviour of the canister. Not all of the discontinuities will impact the corrosion behaviour of the canister. Only surface-breaking discontinuities and those discontinuities within 2 mm of the surface will affect the corrosion behaviour. Defects located further away from the finished surface will have no impact. The relevant literature on the initiation and propagation of localised corrosion and SCC has been reviewed. Initiation of localised corrosion occurs at the microscopic scale at grain boundaries, and will not be affected by the presence of macroscopic discontinuities. The localised breakdown of a passive Cu{sub 2}O/Cu(OH){sub 2} film at a critical electrochemical potential determines where and when pits initiate, not the presence of pit-shaped surface discontinuities. The factors controlling pit growth and death are well understood. There is evidence for a maximum pit radius for copper in chloride solutions, above which the small anodic: cathodic surface area ratio required for the formation of deep pits cannot be sustained. This maximum pit radius is of the order of 0.1-0.5 mm. Surface discontinuities larger than this size are unlikely to propagate as pits, and pits generated from

  12. A study of defects which might arise in the copper steel canister

    Energy Technology Data Exchange (ETDEWEB)

    Bowyer, W.H. [Meadow End Farm, Farnham (United Kingdom)

    1999-05-15

    A study has been conducted to identify the material and manufacturing defects that might occur in serially produced canisters to the SKB reference design. The study has depended on cooperation of contractors engaged by SKB to participate in the development program, SKB staff, observations made by the writer over a five-year involvement with SKI, literature studies and consultation with experts. The candidate manufacturing procedures have been described inasmuch as it has been necessary to do so to make the points related to defects. Where possible, the cause of defects, their likely effects on manufacturing procedures or on durability of the canister and the methods available for their detection are given. For ease of reference each section of the report contains a table which summarizes the information in it and, in the final section of the report, all the tables are presented en-bloc.

  13. Oxidative Dissolution of Spent Fuel and Release of Nuclides from a Copper/Iron Canister : Model Developments and Applications

    OpenAIRE

    Liu, Longcheng

    2001-01-01

    Three models have been developed and applied in the performance assessment of a final repository. They are based on accepted theories and experimental results for known and possible mechanisms that may dominate in the oxidative dissolution of spent fuel and the release of nuclides from a canister. Assuming that the canister is breached at an early stage after disposal, the three models describe three sub-systems in the near field of the repository, in which the governing processes and mechani...

  14. NDT Reliability - Final Report. Reliability in non-destructive testing (NDT) of the canister components

    Energy Technology Data Exchange (ETDEWEB)

    Pavlovic, Mato; Takahashi, Kazunori; Mueller, Christina; Boehm, Rainer (BAM, Federal Inst. for Materials Research and Testing, Berlin (Germany)); Ronneteg, Ulf (Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden))

    2008-12-15

    This report describes the methodology of the reliability investigation performed on the ultrasonic phased array NDT system, developed by SKB in collaboration with Posiva, for inspection of the canisters for permanent storage of nuclear spent fuel. The canister is composed of a cast iron insert surrounded by a copper shell. The shell is composed of the tube and the lid/base which are welded to the tube after the fuel has been place, in the tube. The manufacturing process of the canister parts and the welding process are described. Possible defects, which might arise in the canister components during the manufacturing or in the weld during the welding, are identified. The number of real defects in manufactured components have been limited. Therefore the reliability of the NDT system has been determined using a number of test objects with artificial defects. The reliability analysis is based on the signal response analysis. The conventional signal response analysis is adopted and further developed before applied on the modern ultrasonic phased-array NDT system. The concept of multi-parameter a, where the response of the NDT system is dependent on more than just one parameter, is introduced. The weakness of use of the peak signal response in the analysis is demonstrated and integration of the amplitudes in the C-scan is proposed as an alternative. The calculation of the volume POD, when the part is inspected with more configurations, is also presented. The reliability analysis is supported by the ultrasonic simulation based on the point source synthesis method

  15. ALPHN: A computer program for calculating ([alpha], n) neutron production in canisters of high-level waste

    Energy Technology Data Exchange (ETDEWEB)

    Salmon, R.; Hermann, O.W.

    1992-10-01

    The rate of neutron production from ([alpha], n) reactions in canisters of immobilized high-level waste containing borosilicate glass or glass-ceramic compositions is significant and must be considered when estimating neutron shielding requirements. The personal computer program ALPHA calculates the ([alpha], n) neutron production rate of a canister of vitrified high-level waste. The user supplies the chemical composition of the glass or glass-ceramic and the curies of the alpha-emitting actinides present. The output of the program gives the ([alpha], n) neutron production of each actinide in neutrons per second and the total for the canister. The ([alpha], n) neutron production rates are source terms only; that is, they are production rates within the glass and do not take into account the shielding effect of the glass. For a given glass composition, the user can calculate up to eight cases simultaneously; these cases are based on the same glass composition but contain different quantities of actinides per canister. In a typical application, these cases might represent the same canister of vitrified high-level waste at eight different decay times. Run time for a typical problem containing 20 chemical species, 24 actinides, and 8 decay times was 35 s on an IBM AT personal computer. Results of an example based on an expected canister composition at the Defense Waste Processing Facility are shown.

  16. Topical safety analysis report for the transportation of the NUHOMS{reg_sign} dry shielded canister. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    None

    1993-08-01

    This Topical Safety Analysis Report (SAR) describes the design and the generic transportation licensing basis for utilizing the NUTECH HORIZONTAL MODULAR STORAGE (NUHOMS{reg_sign}) system dry shielded canister (DSC) containing twenty-four pressurized water reactor (PWR) spent fuel assemblies (SFA) in conjunction with a conceptually designed Transportation Cask. This SAR documents the design qualification of the NUHOMS{reg_sign} DSC as an integral part of a 10CFR71 Fissile Material Class III, Type B(M) Transportation Package. The package consists of the canister and a conceptual transportation cask (NUHOMS{reg_sign} Transportation Cask) with impact limiters. Engineering analysis is performed for the canister to confirm that the existing canister design complies with 10CFR71 transportation requirements. Evaluations and/or analyses is performed for criticality safety, shielding, structural, and thermal performance. Detailed engineering analysis for the transportation cask will be submitted in a future SAR requesting 10CFR71 certification of the complete waste package. Transportation operational considerations describe various operational aspects of the canister/transportation cask system. operational sequences are developed for canister transfer from storage to the transportation cask and interfaces with the cask auxiliary equipment for on- and off-site transport.

  17. Criticality Analysis for Proposed Maximum Fuel Loading in a Standardized SNF Canister with Type 1a Baskets

    Energy Technology Data Exchange (ETDEWEB)

    Chad Pope; Larry L. Taylor; Soon Sam Kim

    2007-02-01

    This document represents a summary version of the criticality analysis done to support loading SNF in a Type 1a basket/standard canister combination. Specifically, this engineering design file (EDF) captures the information pertinent to the intact condition of four fuel types with different fissile loads and their calculated reactivities. These fuels are then degraded into various configurations inside a canister without the presence of significant moderation. The important aspect of this study is the portrayal of the fuel degradation and its effect on the reactivity of a single canister given the supposition there will be continued moderation exclusion from the canister. Subsequent analyses also investigate the most reactive ‘dry’ canister in a nine canister array inside a hypothetical transport cask, both dry and partial to complete flooding inside the transport cask. The analyses also includes a comparison of the most reactive configuration to other benchmarked fuels using a software package called TSUNAMI, which is part of the SCALE 5.0 suite of software.

  18. ALPHN: A computer program for calculating ({alpha}, n) neutron production in canisters of high-level waste

    Energy Technology Data Exchange (ETDEWEB)

    Salmon, R.; Hermann, O.W.

    1992-10-01

    The rate of neutron production from ({alpha}, n) reactions in canisters of immobilized high-level waste containing borosilicate glass or glass-ceramic compositions is significant and must be considered when estimating neutron shielding requirements. The personal computer program ALPHA calculates the ({alpha}, n) neutron production rate of a canister of vitrified high-level waste. The user supplies the chemical composition of the glass or glass-ceramic and the curies of the alpha-emitting actinides present. The output of the program gives the ({alpha}, n) neutron production of each actinide in neutrons per second and the total for the canister. The ({alpha}, n) neutron production rates are source terms only; that is, they are production rates within the glass and do not take into account the shielding effect of the glass. For a given glass composition, the user can calculate up to eight cases simultaneously; these cases are based on the same glass composition but contain different quantities of actinides per canister. In a typical application, these cases might represent the same canister of vitrified high-level waste at eight different decay times. Run time for a typical problem containing 20 chemical species, 24 actinides, and 8 decay times was 35 s on an IBM AT personal computer. Results of an example based on an expected canister composition at the Defense Waste Processing Facility are shown.

  19. Estimates of durability of TMI-2 core debris canisters and cask liners

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, A.B. Jr.; Lund, A.L.; Pednekar, S.P.

    1994-04-01

    Core debris from the Three Mile Island-2 (TMI-2) reactor is currently stored in stainless steel canisters. The need to maintain the integrity of the TMI-2 core debris containers through the period of extended storage and possibly into disposal prompted this assessment. In the assessment, corrosion-induced degradation was estimated for two materials: type 304L stainless steel (SS) canisters that contain the core debris, and type 1020 carbon steel (CS) liners in the concrete casks planned for containing the canisters from 2000 AD until the TMI-2 core debris is placed in a repository. Three environments were considered: air-saturated water (with 2 ppM Cl{sup {minus}}) at 20{degree}C, and air at 20{degree}C with two relative humidities (RHs), 10 and 40%. Corrosion mechanisms assessed included general corrosion (failure criterion: 50% loss of wall thickness) and localized attack (failure criterion: through-wall pinhole penetration). Estimation of carbon steel corrosion after 50 y also was requested.

  20. Development of flaw acceptance criteria for aging management of spent nuclear fuel multi-purpose canisters

    Energy Technology Data Exchange (ETDEWEB)

    Lam, Poh -Sang [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL). Materials Science and Technology; Sindelar, Robert L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL). Materials Science and Technology

    2015-03-09

    A typical multipurpose canister (MPC) is made of austenitic stainless steel and is loaded with spent nuclear fuel assemblies. The canister may be subject to service-induced degradation when it is exposed to aggressive atmospheric environments during a possibly long-term storage period if the permanent repository is yet to be identified and readied. Because heat treatment for stress relief is not required for the construction of an MPC, stress corrosion cracking may be initiated on the canister surface in the welds or in the heat affected zone. An acceptance criteria methodology is being developed for flaw disposition should the crack-like defects be detected by periodic in-service Inspection. The first-order instability flaw sizes has been determined with bounding flaw configurations, that is, through-wall axial or circumferential cracks, and part-through-wall long axial flaw or 360° circumferential crack. The procedure recommended by the American Petroleum Institute (API) 579 Fitness-for-Service code (Second Edition) is used to estimate the instability crack length or depth by implementing the failure assessment diagram (FAD) methodology. The welding residual stresses are mostly unknown and are therefore estimated with the API 579 procedure. It is demonstrated in this paper that the residual stress has significant impact on the instability length or depth of the crack. The findings will limit the applicability of the flaw tolerance obtained from limit load approach where residual stress is ignored and only ligament yielding is considered.

  1. Development of flaw acceptance criteria for aging management of spent nuclear fuel multiple-purpose canisters

    Energy Technology Data Exchange (ETDEWEB)

    Lam, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL). Materials Science and Technology; Sindelar, R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL). Materials Science and Technology

    2015-03-09

    A typical multipurpose canister (MPC) is made of austenitic stainless steel and is loaded with spent nuclear fuel assemblies. The canister may be subject to service-induced degradation when it is exposed to aggressive atmospheric environments during a possibly long-term storage period if the permanent repository is yet to be identified and readied. Because heat treatment for stress relief is not required for the construction of an MPC, stress corrosion cracking may be initiated on the canister surface in the welds or in the heat affected zone. An acceptance criteria methodology is being developed for flaw disposition should the crack-like defects be detected by periodic In-service Inspection. The first-order instability flaw sizes has been determined with bounding flaw configurations, that is, through-wall axial or circumferential cracks, and part-through-wall long axial flaw or 360° circumferential crack. The procedure recommended by the American Petroleum Institute (API) 579 Fitness-for-Service code (Second Edition) is used to estimate the instability crack length or depth by implementing the failure assessment diagram (FAD) methodology. The welding residual stresses are mostly unknown and are therefore estimated with the API 579 procedure. It is demonstrated in this paper that the residual stress has significant impact on the instability length or depth of the crack. The findings will limit the applicability of the flaw tolerance obtained from limit load approach where residual stress is ignored and only ligament yielding is considered.

  2. Calculation of displacements on fractures intersecting canisters induced by earthquakes: Aberg, Beberg and Ceberg examples

    Energy Technology Data Exchange (ETDEWEB)

    LaPointe, P.R.; Cladouhos, T. [Golder Associates Inc. (Sweden); Follin, S. [Golder Grundteknik KB (Sweden)

    1999-01-01

    This study shows how the method developed in La Pointe and others can be applied to assess the safety of canisters due to secondary slippage of fractures intersecting those canisters in the event of an earthquake. The method is applied to the three generic sites Aberg, Beberg and Ceberg. Estimation of secondary slippage or displacement is a four-stage process. The first stage is the analysis of lineament trace data in order to quantify the scaling properties of the fractures. This is necessary to insure that all scales of fracturing are properly represented in the numerical simulations. The second stage consists of creating stochastic discrete fracture network (DFN) models for jointing and small faulting at each of the generic sites. The third stage is to combine the stochastic DFN model with mapped lineament data at larger scales into data sets for the displacement calculations. The final stage is to carry out the displacement calculations for all of the earthquakes that might occur during the next 100,000 years. Large earthquakes are located along any lineaments in the vicinity of the site that are of sufficient size to accommodate an earthquake of the specified magnitude. These lineaments are assumed to represent vertical faults. Smaller earthquakes are located at random. The magnitude of the earthquake that any fault could generate is based upon the mapped surface trace length of the lineaments, and is calculated from regression relations. Recurrence rates for a given magnitude of earthquake are based upon published studies for Sweden. A major assumption in this study is that future earthquakes will be similar in magnitude, location and orientation as earthquakes in the geological and historical records of Sweden. Another important assumption is that the displacement calculations based upon linear elasticity and linear elastic fracture mechanics provides a conservative (over-)estimate of possible displacements. A third assumption is that the world

  3. NDE of copper canisters for long-term storage of spent nuclear fuel from the Swedish nuclear power plants

    Science.gov (United States)

    Stepinski, Tadeusz

    2003-07-01

    Sweden has been intensively developing methods for long term storage of spent fuel from the nuclear power plants for twenty-five years. A dedicated research program has been initiated and conducted by the Swedish company SKB (Swedish Nuclear Fuels and Waste Management Co.). After the interim storage SKB plans to encapsulate spent nuclear fuel in copper canisters that will be placed at a deep repository located in bedrock. The canisters filled with fuel rods will be sealed by an electron beam weld. This paper presents three complementary NDE techniques used for assessing the sealing weld in copper canisters, radiography, ultrasound, and eddy current. A powerful X-ray source and a digital detector are used for the radiography. An ultrasonic array system consisting of a phased ultrasonic array and a multi-channel electronics is used for the ultrasonic examination. The array system enables electronic focusing and rapid electronic scanning eliminating the use of a complicated mechanical scanner. A specially designed eddy current probe capable of detecting small voids at the depth up to 4 mm in copper is used for the eddy current inspection. Presently, all the NDE techniques are verified in SKB's Canister Laboratory where full scale canisters are welded and examined.

  4. Design Verification Report Spent Nuclear Fuel (SNF) Project Canister Storage Building (CSB)

    Energy Technology Data Exchange (ETDEWEB)

    PICKETT, W.W.

    2000-09-22

    The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. Because this sub-project is still in the construction/start-up phase, all verification activities have not yet been performed (e.g., canister cover cap and welding fixture system verification, MCO Internal Gas Sampling equipment verification, and As-built verification.). The verification activities identified in this report that still are to be performed will be added to the start-up punchlist and tracked to closure.

  5. Filter Measurement System for Nuclear Material Storage Canisters. End of Year Report FY 2013

    Energy Technology Data Exchange (ETDEWEB)

    Moore, Murray E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Reeves, Kirk P. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-02-03

    A test system has been developed at Los Alamos National Laboratory to measure the aerosol collection efficiency of filters in the lids of storage canisters for special nuclear materials. Two FTS (filter test system) devices have been constructed; one will be used in the LANL TA-55 facility with lids from canisters that have stored nuclear material. The other FTS device will be used in TA-3 at the Radiation Protection Division’s Aerosol Engineering Facility. The TA-3 system will have an expanded analytical capability, compared to the TA-55 system that will be used for operational performance testing. The LANL FTS is intended to be automatic in operation, with independent instrument checks for each system component. The FTS has been described in a complete P&ID (piping and instrumentation diagram) sketch, included in this report. The TA-3 FTS system is currently in a proof-of-concept status, and TA-55 FTS is a production-quality prototype. The LANL specification for (Hagan and SAVY) storage canisters requires the filter shall “capture greater than 99.97% of 0.45-micron mean diameter dioctyl phthalate (DOP) aerosol at the rated flow with a DOP concentration of 65±15 micrograms per liter”. The percent penetration (PEN%) and pressure drop (DP) of fifteen (15) Hagan canister lids were measured by NFT Inc. (Golden, CO) over a period of time, starting in the year 2002. The Los Alamos FTS measured these quantities on June 21, 2013 and on Oct. 30, 2013. The LANL(6-21-2013) results did not statistically match the NFT Inc. data, and the LANL FTS system was re-evaluated, and the aerosol generator was replaced and the air flow measurement method was corrected. The subsequent LANL(10-30-2013) tests indicate that the PEN% results are statistically identical to the NFT Inc. results. The LANL(10-30-2013) pressure drop measurements are closer to the NFT Inc. data, but future work will be investigated. An operating procedure for the FTS (filter test system) was written, and

  6. Very deep borehole. Deutag's opinion on boring, canister emplacement and retrievability

    Energy Technology Data Exchange (ETDEWEB)

    Harrison, Tim [Well Engineering Partners BV, The Hague (Netherlands)

    2000-05-01

    An engineering feasibility study has been carried out to determine whether or not it is possible to drill the proposed Very Deep Borehole concept wells required by SKB for nuclear waste disposal. A conceptual well design has been proposed. All aspects of well design have been considered, including drilling tools, rig design, drilling fluids, casing design and annulus isolation. The proposed well design is for 1168.4 mm hole to be drilled to 500 m. A 1066.8 mm outer diameter (OD) casing will be run and cemented. A 1016 mm hole will be drilled to approximately 2000 m, where 914.4 mm OD casing will be run. This annulus will be sealed with bentonite slurry apart from the bottom 100 m which will be cemented. 838.2 mm hole will be drilled to a final depth of 4000 m, where 762 mm OD slotted casing will be run. All the hole sections will be drilled using a downhole hammer with foam as the drilling fluid medium. Prior to running each casing string, the hole will be displaced to mud to assist with casing running and cementing. The waste canisters will be run on a simple J-slot tool, with integral backup system in case the J-slot fails. The canisters will all be centralised. Canisters can be retrieved using the same tool as used to run them. Procedures are given for both running and retrieving. Logging and testing is recommended only in the exploratory wells, in a maximum hole size of 311.1 mm. This will require the drilling of pilot holes to enable logging and testing to take place. It is estimated that each well will take approximately 137 days to drill and case, at an estimated cost of 4.65 Meuro per well. This time and cost estimate does not include any logging, testing, pilot hole drilling or time taken to run the canisters. New technology developments to enhance the drilling process are required in recyclable foam systems, in hammer bit technology, and in the development of robust under-reamers. It is the authors conclusion that it is possible to drill the well with

  7. Summary of canister overheating incident at the Carbon Tetrachloride Expedited Response Action site

    Energy Technology Data Exchange (ETDEWEB)

    Driggers, S.A.

    1994-03-10

    The granular activated carbon (GAC)-filled canister that overheated was being used to adsorb carbon tetrachloride vapors drawn from a well near the 216-Z-9 Trench, a subsurface disposal site in the 200 West Area of the Hanford Site. The overheating incident resulted in a band of discolored paint on the exterior surface of the canister. Although there was no other known damage to equipment, no injuries to operating personnel, and no releases of hazardous materials, the incident is of concern because it was not anticipated. It also poses the possibility of release of carbon tetrachloride and other hazardous vapors if the incident were to recur. All soil vapor extraction system (VES) operations were halted until a better understanding of the cause of the incident could be determined and controls implemented to reduce the possibility of a recurrence. The focus of this report and the intent of all the activities associated with understanding the overheating incident has been to provide information that will allow safe restart of the VES operations, develop operational limits and controls to prevent recurrence of an overheating incident, and safely optimize recovery of carbon tetrachloride from the ground.

  8. Yucca Mountain project canister material corrosion studies as applied to the electrometallurgical treatment metallic waste form

    Energy Technology Data Exchange (ETDEWEB)

    Keiser, D.D.

    1996-11-01

    Yucca Mountain, Nevada is currently being evaluated as a potential site for a geologic repository. As part of the repository assessment activities, candidate materials are being tested for possible use as construction materials for waste package containers. A large portion of this testing effort is focused on determining the long range corrosion properties, in a Yucca Mountain environment, for those materials being considered. Along similar lines, Argonne National Laboratory is testing a metallic alloy waste form that also is scheduled for disposal in a geologic repository, like Yucca Mountain. Due to the fact that Argonne`s waste form will require performance testing for an environment similar to what Yucca Mountain canister materials will require, this report was constructed to focus on the types of tests that have been conducted on candidate Yucca Mountain canister materials along with some of the results from these tests. Additionally, this report will discuss testing of Argonne`s metal waste form in light of the Yucca Mountain activities.

  9. Genesis Solar Wind Science Canister Components Curated as Potential Solar Wind Collectors and Reference Contamination Sources

    Science.gov (United States)

    Allton, J. H.; Gonzalez, C. P.; Allums, K. K.

    2016-01-01

    The Genesis mission collected solar wind for 27 months at Earth-Sun L1 on both passive and active collectors carried inside of a Science Canister, which was cleaned and assembled in an ISO Class 4 cleanroom prior to launch. The primary passive collectors, 271 individual hexagons and 30 half-hexagons of semiconductor materials, are described in. Since the hard landing reduced the 301 passive collectors to many thousand smaller fragments, characterization and posting in the online catalog remains a work in progress, with about 19% of the total area characterized to date. Other passive collectors, surfaces of opportunity, have been added to the online catalog. For species needing to be concentrated for precise measurement (e.g. oxygen and nitrogen isotopes) an energy-independent parabolic ion mirror focused ions onto a 6.2 cm diameter target. The target materials, as recovered after landing, are described in. The online catalog of these solar wind collectors, a work in progress, can be found at: http://curator.jsc.nasa.gov/gencatalog/index.cfm This paper describes the next step, the cataloging of pieces of the Science Canister, which were surfaces exposed to the solar wind or component materials adjacent to solar wind collectors which may have contributed contamination.

  10. Rates and mechanisms of radioactive release and retention inside a waste disposal canister - in Can Processes

    Energy Technology Data Exchange (ETDEWEB)

    Oversby, V.M. (ed.) [and others

    2003-10-01

    Sweden and Finland are planning to dispose of spent nuclear fuel in a deep underground repository constructed in granitic rock. Each country is investigating candidate sites and developing the scientific and technical basis for assessing the safety of an eventual repository. An essential part of the safety assessment involves understanding the behaviour of the spent fuel after it is placed in the geologic environment. The fuel will be sealed inside a copper canister that contains a cast iron insert. The copper functions as a corrosion resistant barrier, while the cast iron insert fills much of the internal void space, adding strength to the canister and reducing the space available for water to accumulate inside the canister after the corrosion barrier is breached. The canisters will be surrounded by compressed bentonite, which will limit the access of water and dissolved species to the canister. Oxygen that is initially present when the disposal environment is sealed will be rapidly consumed by pyrite in the bentonite, bacterial species in the rock, and reduced inorganic materials in the rock. The copper canister will prevent access of water to the iron until it is corroded through, a process that is expected to take millions of years. After water contacts the iron, anaerobic corrosion of the insert will generate hydrogen gas and introduce Fe(II) ions into the water. The long-term environment for the fuel, therefore, is a highly reducing environment. The only possible source of oxidising agents is radiolysis of the water by radiation from the fuel. In the long-term, the radioactivity in the fuel is due to isotopes that decay by alpha decay; most of the activity from beta and gamma radiation will have decayed away. Spent fuel that is available for testing contains high levels of beta and gamma activity. Even when testing is done in the presence of hydrogen or actively corroding iron, the radiolysis due to beta and gamma radiation can introduce oxidising agents into

  11. ASME Code requirements for multi-canister overpack design and fabrication

    Energy Technology Data Exchange (ETDEWEB)

    SMITH, K.E.

    1998-11-03

    The baseline requirements for the design and fabrication of the MCO include the application of the technical requirements of the ASME Code, Section III, Subsection NB for containment and Section III, Subsection NG for criticality control. ASME Code administrative requirements, which have not historically been applied at the Hanford site and which have not been required by the US Nuclear Regulatory Commission (NRC) for licensed spent fuel casks/canisters, were not invoked for the MCO. As a result of recommendations made from an ASME Code consultant in response to DNFSB staff concerns regarding ASME Code application, the SNF Project will be making the following modifications: issue an ASME Code Design Specification and Design Report, certified by a Registered Professional Engineer; Require the MCO fabricator to hold ASME Section III or Section VIII, Division 2 accreditation; and Use ASME Authorized Inspectors for MCO fabrication. Incorporation of these modifications will ensure that the MCO is designed and fabricated in accordance with the ASME Code. Code Stamping has not been a requirement at the Hanford site, nor for NRC licensed spent fuel casks/canisters, but will be considered if determined to be economically justified.

  12. ALARA Analysis for Shippingport Pressurized Water Reactor Core 2 Fuel Storage in the Canister Storage Building (CSB)

    CERN Document Server

    Lewis, M E

    2000-01-01

    The addition of Shippingport Pressurized Water Reactor (PWR) Core 2 Blanket Fuel Assembly storage in the Canister Storage Building (CSB) will increase the total cumulative CSB personnel exposure from receipt and handling activities. The loaded Shippingport Spent Fuel Canisters (SSFCs) used for the Shippingport fuel have a higher external dose rate. Assuming an MCO handling rate of 170 per year (K East and K West concurrent operation), 24-hr CSB operation, and nominal SSFC loading, all work crew personnel will have a cumulative annual exposure of less than the 1,000 mrem limit.

  13. Multi Canister Overpack (MCO) Handling Machine Independent Review of Seismic Structural Analysis

    Energy Technology Data Exchange (ETDEWEB)

    SWENSON, C.E.

    2000-09-22

    The following separate reports and correspondence pertains to the independent review of the seismic analysis. The original analysis was performed by GEC-Alsthom Engineering Systems Limited (GEC-ESL) under subcontract to Foster-Wheeler Environmental Corporation (FWEC) who was the prime integration contractor to the Spent Nuclear Fuel Project for the Multi-Canister Overpack (MCO) Handling Machine (MHM). The original analysis was performed to the Design Basis Earthquake (DBE) response spectra using 5% damping as required in specification, HNF-S-0468 for the 90% Design Report in June 1997. The independent review was performed by Fluor-Daniel (Irvine) under a separate task from their scope as Architect-Engineer of the Canister Storage Building (CSB) in 1997. The comments were issued in April 1998. Later in 1997, the response spectra of the Canister Storage Building (CSB) was revised according to a new soil-structure interaction analysis and accordingly revised the response spectra for the MHM and utilized 7% damping in accordance with American Society of Mechanical Engineers (ASME) NOG-1, ''Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder).'' The analysis was re-performed to check critical areas but because manufacturing was underway, designs were not altered unless necessary. FWEC responded to SNF Project correspondence on the review comments in two separate letters enclosed. The dispositions were reviewed and accepted. Attached are supplier source surveillance reports on the procedures and process by the engineering group performing the analysis and structural design. All calculation and analysis results are contained in the MHM Final Design Report which is part of the Vendor Information File 50100. Subsequent to the MHM supplier engineering analysis, there was a separate analyses for nuclear safety accident concerns that used the electronic input data files provided by FWEC/GEC-ESL and are contained in

  14. Corrosion of high-level radioactive waste iron-canisters in contact with bentonite

    Energy Technology Data Exchange (ETDEWEB)

    Kaufhold, Stephan, E-mail: s.kaufhold@bgr.de [BGR, Bundesanstalt für Geowissenschaften und Rohstoffe, Stilleweg 2, D-30655 Hannover (Germany); Hassel, Achim Walter [Max-Planck-Institut für Eisenforschung GmbH, Max-Planck-Straße 1, D-40237 Düsseldorf (Germany); Institute for Chemical Technology of Inorganic Materials, Johannes Kepler University Linz, Altenberger Straße 69, 4040 Linz (Austria); Sanders, Daniel [Max-Planck-Institut für Eisenforschung GmbH, Max-Planck-Straße 1, D-40237 Düsseldorf (Germany); Dohrmann, Reiner [BGR, Bundesanstalt für Geowissenschaften und Rohstoffe, Stilleweg 2, D-30655 Hannover (Germany); LBEG, Landesamt für Bergbau, Energie und Geologie, Stilleweg 2, D-30655 Hannover (Germany)

    2015-03-21

    Graphical abstract: Corrosion at the bentonite iron interface proceeds unaerobically with formation of an 1:1 Fe silicate mineral. A series of exposure tests with different types of bentonites showed that Na–bentonites are slightly less corrosive than Ca–bentonites and highly charges smectites are less corrosive compared to low charged ones. The formation of a patina was observed in some cases and has to be investigated further. - Highlights: • At the iron bentonite interface a 1:1 Fe layer silicate forms upon corrosion. • A series of iron–bentonite corrosion products showed slightly less corrosion for Na-rich and high-charged bentonites. • In some tests the formation of a patina was observed consisting of Fe–silicate, which has to be investigated further. - Abstract: Several countries favor the encapsulation of high-level radioactive waste (HLRW) in iron or steel canisters surrounded by highly compacted bentonite. In the present study the corrosion of iron in contact with different bentonites was investigated. The corrosion product was a 1:1 Fe layer silicate already described in literature (sometimes referred to as berthierine). Seven exposition test series (60 °C, 5 months) showed slightly less corrosion for the Na–bentonites compared to the Ca–bentonites. Two independent exposition tests with iron pellets and 38 different bentonites clearly proved the role of the layer charge density of the swelling clay minerals (smectites). Bentonites with high charged smectites are less corrosive than bentonites dominated by low charged ones. The type of counterion is additionally important because it determines the density of the gel and hence the solid/liquid ratio at the contact to the canister. The present study proves that the integrity of the multibarrier-system is seriously affected by the choice of the bentonite buffer encasing the metal canisters in most of the concepts. In some tests the formation of a patina was observed consisting of Fe

  15. Design Verification Report Spent Nuclear Fuel (SNF) Project Canister Storage Building (CSB)

    Energy Technology Data Exchange (ETDEWEB)

    BAZINET, G.D.

    2000-11-03

    The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. The original version of this document was prepared by Vista Engineering for the SNF Project. The purpose of this revision is to document completion of verification actions that were pending at the time the initial report was prepared. Verification activities for the installed and operational SSCs have been completed. Verification of future additions to the CSB related to the canister cover cap and welding fixture system and MCO Internal Gas Sampling equipment will be completed as appropriate for those components. The open items related to verification of those

  16. Biogeochemistry of Redox at Repository Depth and Implications for the Canister

    Energy Technology Data Exchange (ETDEWEB)

    Bath, Adrian; Hermansson, Hans-Peter

    2009-08-15

    The present groundwater chemical conditions at the candidate sites for a spent nuclear fuel repository in Sweden (the Forsmark and Laxemar sites) and processes affecting its future evolution comprise essential conditions for the evaluation of barrier performance and long-term safety. This report reviews available chemical sampling information from the site investigations at the candidate sites, with a particular emphasis on redox active groundwater components and microbial populations that influence redox affecting components. Corrosion of copper canister material is the main barrier performance influence of redox conditions that is elaborated in the report. One section addresses native copper as a reasonable analogue for canister materials and another addresses the feasibility of methane hydrate ice accumulation during permafrost conditions. Such an accumulation could increase organic carbon availability in scenarios involving microbial sulphate reduction. The purpose of the project is to evaluate and describe the available knowledge and data for interpretation of geochemistry, microbiology and corrosion in safety assessment. A conclusive assessment of the sufficiency of information can, however, only be done in the future context of a full safety assessment. The authors conclude that SKB's data and models for chemical and microbial processes are adequate and reasonably coherent. The redox conditions in the repository horizon are predominantly established through the SO{sub 4}2-/HS- and Fe3+/Fe2+ redox couples. The former may exhibit a more significant buffering effect as suggested by measured Eh values, while the latter is associated with a lager capacity due to abundant Fe(II) minerals in the bedrock. Among a large numbers of groundwater features considered in geochemical equilibrium modelling, Eh, pH, temperature and concentration of dissolved sulphide comprise the most essential canister corrosion influences. Groundwater sulphide may originate from

  17. Galvanic and stress corrosion of copper canisters in repository environment. A short review

    Energy Technology Data Exchange (ETDEWEB)

    Hermansson, H.P.; Koenig, M. [Studsvik Nuclear AB, Nykoeping (Sweden)

    2001-02-01

    The Swedish Nuclear Power Inspectorate, SKI, has studied different aspects of canister and copper corrosion as part of the general improvement of the knowledge base within the area. General and local corrosion has earlier been treated by experiments as well as by thermodynamic calculations. For completeness also galvanic and stress corrosion should be treated. The present work is a short review, intended to indicate areas needing further focus. The work consists of two parts, the first of which contains a judgement of statements concerning risk of galvanic corrosion of copper in the repository. The second part concerns threshold values for the stress intensity factor of stress corrosion in copper. A suggestion is given on how such values possibly could be measured for copper at repository conditions. In early investigations by SKB, galvanic corrosion is not mentioned or at least not treated. In later works it is treated but often in a theoretical way without indications of any further treatment or investigation. Several pieces of work indicate that further investigations are required to ensure that different types of corrosion, like galvanic, cannot occur in the repository environment. There are for example effects of grain size, grain boundary conditions, impurities and other factors that could influence the appearance of galvanic corrosion that are not treated. Those factors have to be considered to be completely sure that galvanic corrosion and related effects does not occur for the actual canister in the specific environment of the repository. The circumstances are so specific, that a rather general discussion indicating that galvanic corrosion is not probable just is not enough. Experiments should also be performed for verification. It is concluded that the following specific areas, amongst others, could benefit from further consideration. Galvanic corrosion of unbreached copper by inhomogeneities in the environment and in the copper metal should be addressed

  18. The Effect of Flow Rate and Canister Geometry on the Effectiveness of Removing Carbon Dioxide with Soda Lime.

    Science.gov (United States)

    1980-09-01

    pressure was measured using a Meriam type W,0 - 30 inches of mercury manometer . The gas was then piped to the water bath. At the water bath, the gas was...bypass pressure as indicatedon the 30-inch mercury manometer was recorded. The canister was then allowed to remain in the water bath for forty-five

  19. Canister storage building (CSB) safety analysis report phase 3: Safety analysis documentation supporting CSB construction

    Energy Technology Data Exchange (ETDEWEB)

    Garvin, L.J.

    1997-04-28

    The Canister Storage Building (CSB) will be constructed in the 200 East Area of the U.S. Department of Energy (DOE) Hanford Site. The CSB will be used to stage and store spent nuclear fuel (SNF) removed from the Hanford Site K Basins. The objective of this chapter is to describe the characteristics of the site on which the CSB will be located. This description will support the hazard analysis and accident analyses in Chapter 3.0. The purpose of this report is to provide an evaluation of the CSB design criteria, the design's compliance with the applicable criteria, and the basis for authorization to proceed with construction of the CSB.

  20. Site-to-canister scale flow and transport in Haestholmen, Kivetty, Olkiluoto and Romuvaara

    Energy Technology Data Exchange (ETDEWEB)

    Poteri, A.; Laitinen, M. [VTT Energy, Espoo (Finland)

    1999-05-01

    Radioactive waste is originating from production of electricity in nuclear power plants. Most of the waste has only low or intermediate levels of radioactivity. However, the spent nuclear fuel is highly radioactive and it has to be isolated from the biosphere. The current nuclear waste management plan in Finland is based on direct disposal of the spent nuclear fuel deep underground. The only feasible mechanism for the radionuclides to escape from an underground repository is to be carried by the groundwater flow after the failure of waste containers. The scope of this study is to examine the groundwater flow situation and transport properties in the vicinity of the disposal canister and along the potential release paths from the repository into the biosphere. The results of this study are further applied in the site specific safety analysis of a spent fuel repository. Synthesis is made of the porous medium estimates of the groundwater flow in the regional and site scales and the detailed fracture network analysis of the flow in the canister scale. This synthesis includes estimation of the transport properties from the canister into the biosphere and flow rates around the deposition holes of the waste canisters. The modelling has been carried out for four different sites: Hastholmen, Kivetty, Olkiluoto and Romavaara. According to the simulations groundwater flow rate around the deposition holes is less than about 1 litre/a for about 75 % of the deposition holes. For about 5 % of the deposition holes the flow rates are a few litres per year or higher. The highest flow rates resulted at Hastholmen, in fresh water conditions 10 000 years after present, and at Kivetty. The transport resistances were calculated for the `worst` flow paths that might have impact on the safety of the repository. The total transport resistances from the repository into the biosphere along those flow paths varied between about 40 000 a/m and 5-10{sup 6} a/m. Most of the total transport

  1. FEMA and RAM Analysis for the Multi Canister Overpack (MCO) Handling Machine

    Energy Technology Data Exchange (ETDEWEB)

    SWENSON, C.E.

    2000-06-01

    The Failure Modes and Effects Analysis and the Reliability, Availability, and Maintainability Analysis performed for the Multi-Canister Overpack Handling Machine (MHM) has shown that the current design provides for a safe system, but the reliability of the system (primarily due to the complexity of the interlocks and permissive controls) is relatively low. No specific failure modes were identified where significant consequences to the public occurred, or where significant impact to nearby workers should be expected. The overall reliability calculation for the MHM shows a 98.1 percent probability of operating for eight hours without failure, and an availability of the MHM of 90 percent. The majority of the reliability issues are found in the interlocks and controls. The availability of appropriate spare parts and maintenance personnel, coupled with well written operating procedures, will play a more important role in successful mission completion for the MHM than other less complicated systems.

  2. Fuel and canister process report for the safety assessment SR-Can

    Energy Technology Data Exchange (ETDEWEB)

    Werme, Lars (ed.)

    2006-10-15

    This report documents fuel and canister processes identified as relevant to the long-term safety of a KBS-3 repository. It forms an important part of the reporting of the safety assessment SR-Can. The detailed assessment methodology, including the role of the process report in the assessment, is described in the SR-Can Main report. The report is written by, and for, experts in the relevant scientific fields. It should though be possible for a generalist in the area of long-term safety assessments of geologic nuclear waste repositories to comprehend the contents of the report. The report is an important part of the documentation of the SR-Can project and an essential reference within the project, providing a scientifically motivated plan for the handling of geosphere processes. It is, furthermore, foreseen that the report will be essential for reviewers scrutinising the handling of geosphere issues in the SR-Can assessment. Several types of fuel will be emplaced in the repository. For the reference case with 40 years of reactor operation, the fuel quantity from boiling water reactors, BWR fuel, is estimated at 7,000 tonnes, while the quantity from pressurized water reactors, PWR fuel, is estimated at about 2,300 tonnes. In addition, 23 tonnes of mixed-oxide fuel (MOX) fuel of German origin from BWR and PWR reactors and 20 tonnes of fuel from the decommissioned heavy water reactor in Aagesta will be disposed of. To allow for future changes in the Swedish nuclear programme, the safety assessment assumes a total of 6,000 canister corresponding to 12,000 tonnes of fuel.

  3. Creep of the Copper Canister. A Critical Review of the Literature

    Energy Technology Data Exchange (ETDEWEB)

    Bowyer, William H. [Meadow End Farm, Farnham (United Kingdom)

    2003-04-01

    Literature relevant to creep of the copper shell of the copper-iron canister has been reviewed. Two classes of copper have been examined, Oxygen Free High Conductivity (OFHC), which is referred to in the relevant literature and this report as OF material, and OF material with 50 ppm of phosphorus added. The second material is referred to as OFP. Creep processes occurring in copper are briefly described and a deformation diagram, after Frost and Ashby is provided. It is concluded that the diagram adequately describes the processes observed for the two materials of interest without necessarily being in exact agreement at a quantitative level. There are two regimes of time, temperature and stress which are important when creep of the copper shell is considered. The first is a holding period between welding of the lid to the canister and placing the canister in the repository and the second is the storage period in the repository. In the holding period, residual stresses arising from the manufacturing processes are important and in the second period stresses arising from repository pressures are important as well as the residual pressures arising from manufacture. The holding period may extend up to one year and the temperature of the copper shell may decline from the immediate post welding temperature to 100 deg C in this interval. Initial peak localised stresses may give rise to strains of up to 14 %. Dynamic recovery immediately after welding reduces the stresses associated with these strains to levels which correspond to stresses for approximately 0.1 % strain at the ruling temperature. This is 75 MPa at 100 deg C and 50 MPa for 150 deg C. A further stress relaxation of up to 30 % occurs in the first 20 days after welding. Localised stresses are therefore unlikely to exceed 50 MPa when the canister is placed into storage. No negative effects have been observed in connection with this stress relaxation process. In the storage period, which is indefinite, the

  4. Final Report: Part 1. In-Place Filter Testing Instrument for Nuclear Material Containers. Part 2. Canister Filter Test Standards for Aerosol Capture Rates.

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Austin Douglas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Runnels, Joel T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Moore, Murray E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Reeves, Kirk Patrick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-11-02

    A portable instrument has been developed to assess the functionality of filter sand o-rings on nuclear material storage canisters, without requiring removal of the canister lid. Additionally, a set of fifteen filter standards were procured for verifying aerosol leakage and pressure drop measurements in the Los Alamos Filter Test System. The US Department of Energy uses several thousand canisters for storing nuclear material in different chemical and physical forms. Specialized filters are installed into canister lids to allow gases to escape, and to maintain an internal ambient pressure while containing radioactive contaminants. Diagnosing the condition of container filters and canister integrity is important to ensure worker and public safety and for determining the handling requirements of legacy apparatus. This report describes the In-Place-Filter-Tester, the Instrument Development Plan and the Instrument Operating Method that were developed at the Los Alamos National Laboratory to determine the “as found” condition of unopened storage canisters. The Instrument Operating Method provides instructions for future evaluations of as-found canisters packaged with nuclear material. Customized stainless steel canister interfaces were developed for pressure-port access and to apply a suction clamping force for the interface. These are compatible with selected Hagan-style and SAVY-4000 storage canisters that were purchased from NFT (Nuclear Filter Technology, Golden, CO). Two instruments were developed for this effort: an initial Los Alamos POC (Proof-of-Concept) unit and the final Los Alamos IPFT system. The Los Alamos POC was used to create the Instrument Development Plan: (1) to determine the air flow and pressure characteristics associated with canister filter clogging, and (2) to test simulated configurations that mimicked canister leakage paths. The canister leakage scenarios included quantifying: (A) air leakage due to foreign material (i.e. dust and hair

  5. Evaluation of the conservativeness of the methodology for estimating earthquake-induced movements of fractures intersecting canisters

    Energy Technology Data Exchange (ETDEWEB)

    La Pointe, Paul R.; Cladouhos, Trenton T. [Golder Associates Inc., Las Vegas, NV (United States); Outters, Nils; Follin, Sven [Golder Grundteknik KB, Stockholm (Sweden)

    2000-04-01

    This study evaluates the parameter sensitivity and the conservativeness of the methodology outlined in TR 99-03. Sensitivity analysis focuses on understanding how variability in input parameter values impacts the calculated fracture displacements. These studies clarify what parameters play the greatest role in fracture movements, and help define critical values of these parameters in terms of canister failures. The thresholds or intervals of values that lead to a certain level of canister failure calculated in this study could be useful for evaluating future candidate sites. Key parameters include: 1. magnitude/frequency of earthquakes; 2. the distance of the earthquake from the canisters; 3. the size and aspect ratio of fractures intersecting canisters; and 4. the orientation of the fractures. The results of this study show that distance and earthquake magnitude are the most important factors, followed by fracture size. Fracture orientation is much less important. Regression relations were developed to predict induced fracture slip as a function of distance and either earthquake magnitude or slip on the earthquake fault. These regression relations were validated by using them to estimate the number of canister failures due to single damaging earthquakes at Aberg, and comparing these estimates with those presented in TR 99-03. The methodology described in TR 99-03 employs several conservative simplifications in order to devise a numerically feasible method to estimate fracture movements due to earthquakes outside of the repository over the next 100,000 years. These simplifications include: 1. fractures are assumed to be frictionless and cohesionless; 2. all energy transmitted to the fracture by the earthquake is assumed to produce elastic deformation of the fracture; no energy is diverted into fracture propagation; and 3. shielding effects of other fractures between the earthquake and the fracture are neglected. The numerical modeling effectively assumes that the

  6. Modeling of molecular and particulate transport in dry spent nuclear fuel canisters

    Science.gov (United States)

    Casella, Andrew M.

    2007-09-01

    The transportation and storage of spent nuclear fuel is one of the prominent issues facing the commercial nuclear industry today, as there is still no general consensus regarding the near- and long-term strategy for managing the back-end of the nuclear fuel cycle. The debate continues over whether the fuel cycle should remain open, in which case spent fuel will be stored at on-site reactor facilities, interim facilities, or a geologic repository; or if the fuel cycle should be closed, in which case spent fuel will be recycled. Currently, commercial spent nuclear fuel is stored at on-site reactor facilities either in pools or in dry storage containers. Increasingly, spent fuel is being moved to dry storage containers due to decreased costs relative to pools. As the number of dry spent fuel containers increases and the roles they play in the nuclear fuel cycle increase, more regulations will be enacted to ensure that they function properly. Accordingly, they will have to be carefully analyzed for normal conditions, as well as any off-normal conditions of concern. This thesis addresses the phenomena associated with one such concern; the formation of a microscopic through-wall breach in a dry storage container. Particular emphasis is placed on the depressurization of the canister, release of radioactivity, and plugging of the breach due to deposition of suspended particulates. The depressurization of a dry storage container upon the formation of a breach depends on the temperature and quantity of the fill gas, the pressure differential across the breach, and the size of the breach. The first model constructed in this thesis is capable of determining the depressurization time for a breached container as long as the associated parameters just identified allow for laminar flow through the breach. The parameters can be manipulated to quantitatively determine their effect on depressurization. This model is expanded to account for the presence of suspended particles. If

  7. Criticality Safety Evaluation Report CSER-96-019 for Spent Nuclear Fuel (SNF) Processing and Storage Facilities Multi Canister Overpack (MCO)

    Energy Technology Data Exchange (ETDEWEB)

    KESSLER, S.F.

    1999-10-20

    This criticality evaluation is for Spent N Reactor fuel unloaded from the existing canisters in both KE and KW Basins, and loaded into multiple canister overpack (MCO) containers with specially built baskets containing a maximum of either 54 Mark IV or 48 Mark IA fuel assemblies. The criticality evaluations include loading baskets into the cask-MCO, operation at the Cold Vacuum Drying Facility,a nd storage in the Canister Storage Building. Many conservatisms have been built into this analysis, the primary one being the selection of the K{sub eff} = 0.95 criticality safety limit. This revision incorporates the analyses for the sampling/weld station in the Canister Storage Building and additional analysis of the MCO during the draining at CVDF. Additional discussion of the scrap basket model was added to show why the addition of copper divider plates was not included in the models.

  8. End of FY2014 Report - Filter Measurement System for Nuclear Material Storage Canisters (Including Altitude Correction for Filter Pressure Drop)

    Energy Technology Data Exchange (ETDEWEB)

    Moore, Murray E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Reeves, Kirk Patrick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-02-24

    Two LANL FTS (Filter Test System ) devices for nuclear material storage canisters are fully operational. One is located in PF-4 ( i.e. the TA-55 FTS) while the other is located at the Radiation Protection Division’s Aerosol Engineering Facility ( i.e. the TA-3 FTS). The systems are functionally equivalent , with the TA-3 FTS being the test-bed for new additions and for resolving any issues found in the TA-55 FTS. There is currently one unresolved issue regarding the TA-55 FTS device. The canister lid clamp does not give a leak tight seal when testing the 1 QT (quart) or 2 QT SAVY lids. An adapter plate is being developed that will ensure a correct test configuration when the 1 or 2 QT SAVY lid s are being tested .

  9. Report on hydro-mechanical and chemical-mineralogical analyses of the bentonite buffer in Canister Retrieval Test

    Energy Technology Data Exchange (ETDEWEB)

    Dueck, Ann; Johannesson, Lars-Erik; Kristensson, Ola; Olsson, Siv [Clay Technology AB (Sweden)

    2011-12-15

    The effect of five years of exposure to repository-like conditions on compacted Wyoming bentonite was determined by comparing the hydraulic, mechanical, and mineralogical properties of samples from the bentonite buffer of the Canister Retrieval Test (CRT) with those of reference material. The CRT, located at the Swedish Aspo Hard Rock Laboratory (HRL), was a full-scale field experiment simulating conditions relevant for the Swedish KBS-3 concept for disposal of high-level radioactive waste in crystalline host rock. The compacted bentonite, surrounding a copper canister equipped with heaters, had been subjected to heating at temperatures up to 95 deg C and hydration by natural Na-Ca-Cl type groundwater for almost five years at the time of retrieval. Under the thermal and hydration gradients that prevailed during the test, sulfate in the bentonite was redistributed and accumulated as anhydrite close to the canister. The major change in the exchangeable cation pool was a loss in Mg in the outer parts of the blocks, suggesting replacement of Mg mainly by Ca along with the hydration with groundwater. Close to the copper canister, small amounts of Cu were incorporated in the bentonite. A reduction of strain at failure was observed in the innermost part of the bentonite buffer, but no influence was seen on the shear strength. No change of the swelling pressure was observed, while a modest decrease in hydraulic conductivity was found for the samples with the highest densities. No coupling was found between these changes in the hydro-mechanical properties and the montmorillonite . the X-ray diffraction characteristics, the cation exchange properties, and the average crystal chemistry of the Na-converted < 1 {mu}m fractions provided no evidence of any chemical/structural changes in the montmorillonite after the 5-year hydrothermal test.

  10. Spent Nuclear Fuel Project (SNFP) gas generation from N-Fuel in multi-canister overpacks

    Energy Technology Data Exchange (ETDEWEB)

    Cooper, T.D.

    1996-08-01

    During the conversion from wet pool storage for spent nuclear fuel at Hanford, gases will be generated from both radiolysis and chemical reactions. The gas generation phenomenon needs to be understood as it applies to safety and design issues,specifically over pressurization of sealed storage containers,and detonation/deflagration of flammable gases. This study provides an initial basis to predict the implications of gas generation on the proposed functional processes for spent nuclear fuel conversion from wet to dry storage. These projections are based upon examination of the history of fuel manufacture at Hanford, irradiation in the reactors, corrosion during wet pool storage, available fuel characterization data and available information from literature. Gas generation via radiolysis and metal corrosion are addressed. The study examines gas generation, the boundary conditions for low medium and high levels of sludge in SNF storage/processing containers. The functional areas examined include: flooded and drained Multi-Canister Overpacks, cold vacuum drying, shipping and staging and long term storage.

  11. Multi Canister Overpack (MCO) Combustible Gas Management Leak Test Acceptance Criteria (OCRWM)

    Energy Technology Data Exchange (ETDEWEB)

    SHERRELL, D.L.

    2000-10-10

    The purpose of this document is to support the Spent Nuclear Fuel Project's combustible gas management strategy while avoiding the need to impose any requirements for oxygen free atmospheres within storage tubes that contain multi-canister overpacks (MCO). In order to avoid inerting requirements it is necessary to establish and confirm leak test acceptance criteria for mechanically sealed and weld sealed MCOs that are adequte to ensure that, in the unlikely event the leak test results for any MCO were to approach either of those criteria, it could still be handled and stored in stagnant air without compromising the SNF Project's overall strategy to prevent accumulation of combustible gas mixtures within MCOs or within their surroundings. To support that strategy, this document: (1) establishes combustible gas management functions and minimum functional requirements for the MCO's mechanical seals and closure weld(s); (2) establishes a maximum practical value for the minimum required initial MCO inert backfill gas pressure; and (3) based on items 1 and 2, establishes and confirms leak test acceptance criteria for the MCO's mechanical seal and final closure weld(s).

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

    Energy Technology Data Exchange (ETDEWEB)

    J.R. Massari

    1995-10-06

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

  13. Human Factors Engineering and Ergonomics Analysis for the Canister Storage Building (CSB) Results and Findings

    Energy Technology Data Exchange (ETDEWEB)

    GARVIN, L.J.

    1999-09-20

    The purpose for this supplemental report is to follow-up and update the information in SNF-3907, Human Factors Engineering (HFE) Analysis: Results and Findings. This supplemental report responds to applicable U.S. Department of Energy Safety Analysis Report review team comments and questions. This Human Factors Engineering and Ergonomics (HFE/Erg) analysis was conducted from April 1999 to July 1999; SNF-3907 was based on analyses accomplished in October 1998. The HFE/Erg findings presented in this report and SNF-3907, along with the results of HNF-3553, Spent Nuclear Fuel Project, Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report,'' Chapter A3.0, ''Hazards and Accidents Analyses,'' provide the technical basis for preparing or updating HNF-3553. Annex A, Chaptex A13.0, ''Human Factors Engineering.'' The findings presented in this report allow the HNF-3553 Chapter 13.0, ''Human Factors,'' to respond fully to the HFE requirements established in DOE Order 5480.23, Nuclear Safety Analysis Reports.

  14. Inspection of copper canisters for spent nuclear fuel by means of ultrasound. Algorithms for ultrasonic imaging

    Energy Technology Data Exchange (ETDEWEB)

    Stepinski, Tadeusz (ed.); Engholm, Marcus; Olofsson, Tomas (Uppsala Univ., Signals and Systems, Dept. of Technical Sciences (Sweden))

    2011-07-15

    This report contains research results concerning the use of advanced ultrasound for the inspection of copper canisters for spent nuclear fuel obtained at Signals and Systems, Uppsala Univ. in 2009 and 2010. The first part of the report deals with ultrasonic imaging of damage in planar structures using Lamb waves. We present results of the first successful attempt to apply an adaptive beamformer for Lamb waves. Our algorithm is an extension of the adaptive beamformer based on minimum variance distortion less response (MVDR) approach to dispersive, multimodal Lamb waves. We present simulation and experimental results illustrating the performance of the MVDR applied to imaging artificial damage in an aluminum plate. In the second part of the report we present two extensions of the previously proposed 2D phase shift migration algorithms for enhancing resolution in ultrasonic imaging of solid objects. The first extension enables processing 3D data in order to fully utilize the resolution enhancement potential of the technique. The second extension, consists in generalizing the technique to allow for the processing of data acquired using an array instead of a previously concerned single transducer. Robustness issue related to objects having front surfaces that are slightly tilted relative to the scanning axis is also considered

  15. A methodology to estimate earthquake effects on fractures intersecting canister holes

    Energy Technology Data Exchange (ETDEWEB)

    La Pointe, P.; Wallmann, P.; Thomas, A.; Follin, S. [Golder Assocites Inc. (Sweden)

    1997-03-01

    A literature review and a preliminary numerical modeling study were carried out to develop and demonstrate a method for estimating displacements on fractures near to or intersecting canister emplacement holes. The method can be applied during preliminary evaluation of candidate sites prior to any detailed drilling or underground excavation, utilizing lineament maps and published regression relations between surface rupture trace length and earthquake magnitude, rupture area and displacements. The calculated displacements can be applied to lineament traces which are assumed to be faults and may be the sites for future earthquakes. Next, a discrete fracture model is created for secondary faulting and jointing in the vicinity of the repository. These secondary fractures may displace due to the earthquake on the primary faults. The three-dimensional numerical model assumes linear elasticity and linear elastic fracture mechanics which provides a conservative displacement estimate, while still preserving realistic fracture patterns. Two series of numerical studies were undertaken to demonstrate how the methodology could be implemented and how results could be applied to questions regarding site selection and performance assessment. The first series illustrates how earthquake damage to a hypothetical repository for a specified location (Aespoe) could be estimated. A second series examined the displacements induced by earthquakes varying in magnitude from 6.0 to 8.2 as a function of how close the earthquake was in relation to the repository. 143 refs, 25 figs, 7 tabs.

  16. 金属氢化物储氢装置研究%Study on Metal Hydride Canister

    Institute of Scientific and Technical Information of China (English)

    刘晓鹏; 蒋利军; 陈立新

    2009-01-01

    The temperature field of the inner cylindrical canister was simulated by using finite difference method and 2D heat transfer model during the hydrogenation process. It is showed that a temperature gradient is distributed obviously in the metal hydride bed, and the ceutric place of the canister has the highest temperature. Therefore, heat assembled in the centric place must be intensively transferred to improve the hydrogen storage properties of the metal hydride canister. In order to improve the hydrogen absorption/desorption cycle performance of the canister, the cycle life of as-cast and melt-spinning TiV0.41 Fe0.09Mn1.5 alloy was comparatively studied. It is indicated that the cycle life of the melt -spinning alloy is considerably longer than that of the as-cast one. The canister prepared by using melt-spinning TiV0.41Fe0.09Mn1.5 alloy has 94% of the hydrogen storage capacity after 3600 cycles.%用有限差分法和二维导热模型计算了圆柱形金属氢化物储氢装置内部储氢过程的温度场分布,结果表明空气换热型储氢装置内部的合金反应床存在明显的温度梯度场,吸氢时储氢装置中心部位的温度最高,需要强化其芯部换热条件,以提高储氢装置的储放氢性能.对比研究了铸态以及甩带快淬工艺制备 TiV0.41 Fe0.09Mn1.5合金吸放氢循环寿命,表明甩带快淬工艺可以显著提高储氢合金的吸放氢循环性能.以甩带快淬工艺制备的TiV0.41Fe0.09Mn1.5合金为工质的储氢装置,经过3 600次吸放氢循环后的容量保持率达到94%以上.

  17. EB-welding of the copper canister for the nuclear waste disposal. Final report of the development programme 1994-1997

    Energy Technology Data Exchange (ETDEWEB)

    Aalto, H. [Outokumpu Oy Poricopper, Pori (Finland)

    1998-10-01

    During 1994-1997 Posiva Oy and Outokumpu Poricopper Oy had a joint project Development of EB-welding method for massive copper canister manufacturing. The project was part of the national technology program `Weld 2000` and it was supported financially by Technology Development Centre (TEKES). The spent fuel from Finnish nuclear reactors is planned to be encapsulated in thick-walled copper canisters and placed deep into the bedrock. The thick copper layer of the canister provides a long time corrosion resistance and prevents deposited nuclear fuel from contact with water. The quality requirements of the copper components are high because of the designed long lifetime of the canister. The EB-welding technology has proved to be applicable method for the production of the copper canisters and the EB-welding technique is needed at least when the lids of the copper canister will be closed. There are a number of parameters in EB-welding which affect weldability. However, the effect of the welding parameters and their optimization has not been extensively studied in welding of thick copper sections using conventional high vacuum EB-welding. One aim of this development work was to extensively study effect of welding parameters on weld quality. The final objective was to minimise welding defects in the main weld and optimize slope out procedure in thick copper EB-welding. Welding of 50 mm thick copper sections was optimized using vertical and horizontal EB-welding techniques. As a result two full scale copper lids were welded to a short cylinder successfully. The resulting weld quality with optimised welding parameters was reasonable good. The optimised welding parameters for horizontal and vertical beam can be applied to the longitudinal body welds of the canister. The optimal slope out procedure for the lid closure needs some additional development work. In addition of extensive EB-welding program ultrasonic inspection and creep strength of the weld were studied. According

  18. Analysis of the effect of vibrations on the bentonite buffer in the canister hole

    Energy Technology Data Exchange (ETDEWEB)

    Jonsson, Martin (AaF- Berg och Maetteknik, Stockholm (Sweden)); Hakami, Hossein; Ekneligoda, Thushan (Itasca Geomekanik AB, Solna (Sweden))

    2009-09-15

    During the construction of a final repository for spent nuclear fuel in crystalline rock, blasting activities in certain deposition tunnels will occur at the same time as the deposition of canisters containing the waste is going on in another adjacent access tunnel. In fact, the deposition consists of several stages after the drilling of the deposition hole. The most vulnerable stage from a vibration point of view is when the bentonite buffer is placed in the deposition hole but the canister has not been placed yet. During this stage, a hollow column of bentonite blocks remains free to vibrate inside the deposition hole. The goal of this study was to investigate the displacement of the bentonite blocks when exposed to the highest vibration level that can be expected during the drill and blast operations. In order to investigate this, a three dimensional model in 3DEC, capable of capturing the dynamic behaviour of the bentonite buffer was set up. To define the vibration levels, which serve as input data for the 3DEC model, an extensive analysis of the recorded vibrations from the TASQ - tunnel was carried out. For this purpose, an upper expected vibration limit was defined. This was done outgoing from the fact that the planned charging for the construction of the geological repository will lie in the interval 2 to 4 kg. Furthermore, at the first stage for this study, it was decided that the vibration should be conservatively evaluated for 30 m distance. Using these data, it was concluded that the maximum vibration level that can be expected will be approximately 60 mm/s. After simplifying the vibration signal, a sinusoidal wave with the amplitude 60 mm/s was applied at the bottom of the column and it was assumed that the vibrations only affect the bentonite buffer in one direction (horizontal direction). From this simulation, it was concluded that hardly any displacements occurred. However, when applying the same sinusoidal wave both in the horizontal and the

  19. Strategy for verification and demonstration of the sealing process for canisters for spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, Christina [Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin (Germany); Oeberg, Tomas [Tomas Oeberg Konsult AB, Lyckeby (Sweden)

    2004-08-01

    Electron beam welding and friction stir welding are the two processes now being considered for sealing copper canisters with Sweden's radioactive waste. This report outlines a strategy for verification and demonstration of the encapsulation process which here is considered to consist of the sealing of the canister by welding followed by quality control of the weld by non-destructive testing. Statistical methodology provides a firm basis for modern quality technology and design of experiments has been successful part of it. Factorial and fractional factorial designs can be used to evaluate main process factors and their interactions. Response surface methodology with multilevel designs enables further optimisation. Empirical polynomial models can through Taylor series expansions approximate the true underlying relationships sufficiently well. The fitting of response measurements is based on ordinary least squares regression or generalised linear methods. Unusual events, like failures in the lid welds, are best described with extreme value statistics and the extreme value paradigm give a rationale for extrapolation. Models based on block maxima (the generalised extreme value distribution) and peaks over threshold (the generalised Pareto distribution) are considered. Experiences from other fields of the materials sciences suggest that both of these approaches are useful. The initial verification experiments of the two welding technologies considered are suggested to proceed by experimental plans that can be accomplished with only four complete lid welds each. Similar experimental arrangements can be used to evaluate process 'robustness' and optimisation of the process window. Two series of twenty demonstration trials each, mimicking assembly-line production, are suggested as a final evaluation before the selection of welding technology. This demonstration is also expected to provide a data base suitable for a baseline estimate of future performance

  20. Cleaning Genesis Sample Return Canister for Flight: Lessons for Planetary Sample Return

    Science.gov (United States)

    Allton, J. H.; Hittle, J. D.; Mickelson, E. T.; Stansbery, Eileen K.

    2016-01-01

    Sample return missions require chemical contamination to be minimized and potential sources of contamination to be documented and preserved for future use. Genesis focused on and successfully accomplished the following: - Early involvement provided input to mission design: a) cleanable materials and cleanable design; b) mission operation parameters to minimize contamination during flight. - Established contamination control authority at a high level and developed knowledge and respect for contamination control across all institutions at the working level. - Provided state-of-the-art spacecraft assembly cleanroom facilities for science canister assembly and function testing. Both particulate and airborne molecular contamination was minimized. - Using ultrapure water, cleaned spacecraft components to a very high level. Stainless steel components were cleaned to carbon monolayer levels (10 (sup 15) carbon atoms per square centimeter). - Established long-term curation facility Lessons learned and areas for improvement, include: - Bare aluminum is not a cleanable surface and should not be used for components requiring extreme levels of cleanliness. The problem is formation of oxides during rigorous cleaning. - Representative coupons of relevant spacecraft components (cut from the same block at the same time with identical surface finish and cleaning history) should be acquired, documented and preserved. Genesis experience suggests that creation of these coupons would be facilitated by specification on the engineering component drawings. - Component handling history is critical for interpretation of analytical results on returned samples. This set of relevant documents is not the same as typical documentation for one-way missions and does include data from several institutions, which need to be unified. Dedicated resources need to be provided for acquiring and archiving appropriate documents in one location with easy access for decades. - Dedicated, knowledgeable

  1. Deep geological disposal system development; mechanical structural stability analysis of spent nuclear fuel disposal canister under the internal/external pressure variation

    Energy Technology Data Exchange (ETDEWEB)

    Kwen, Y. J.; Kang, S. W.; Ha, Z. Y. [Hongik University, Seoul (Korea)

    2001-04-01

    This work constitutes a summary of the research and development work made for the design and dimensioning of the canister for nuclear fuel disposal. Since the spent nuclear fuel disposal emits high temperature heats and much radiation, its careful treatment is required. For that, a long term(usually 10,000 years) safe repository for spent fuel disposal should be securred. Usually this repository is expected to locate at a depth of 500m underground. The canister construction type introduced here is a solid structure with a cast iron insert and a corrosion resistant overpack, which is designed for spent nuclear fuel disposal in a deep repository in the crystalline bedrock, which entails an evenly distributed load of hydrostatic pressure from undergroundwater and high pressure from swelling of bentonite buffer. Hence, the canister must be designed to withstand these high pressure loads. Many design variables may affect the structural strength of the canister. In this study, among those variables array type of inner baskets and thicknesses of outer shell and lid and bottom are tried to be determined through the mechanical linear structural analysis, thicknesses of outer shell is determined through the nonlinear structural analysis, and the bentonite buffer analysis for the rock movement is conducted through the of nonlinear structural analysis Also the thermal stress effect is computed for the cast iron insert. The canister types studied here are one for PWR fuel and another for CANDU fuel. 23 refs., 60 figs., 23 tabs. (Author)

  2. Inspection of copper canisters for spent nuclear fuel by means of ultrasound. Ultrasonic imaging, FSW monitoring with acoustic emission

    Energy Technology Data Exchange (ETDEWEB)

    Stepinski, Tadeusz (ed.); Olofsson, Tomas; Wennerstroem, Erik [Uppsala Univ., Dept. of Technical Sciences (Sweden). Signals and Systems

    2006-12-15

    This report contains the research results concerning advanced ultrasound for the inspection of copper canisters for spent nuclear fuel obtained at Signals and Systems, Uppsala University in years 2005/2006. In the first part of the report we propose a concept of monitoring of the friction stir welding (FSW) process by means of acoustic emission (AE) technique. First, we introduce the AE technique and then we present the principle of the system for monitoring the FSW process in cylindrical symmetry specific for the SKB canisters. We propose an omnidirectional circular array of ultrasonic transducers for receiving the AE signals generated by the FSW tool and the releases of the residual stress at canister's circumference. Finally, we review the theory of uniform circular arrays. The second part of the report is concerned with synthetic aperture focusing technique (SAFT) characterized by enhanced spatial resolution. We evaluate three different approaches to perform imaging with less computational cost than that of the extended SAFT (ESAFT) method proposed in our previous reports. First, a sparse version of ESAFT is presented, which solves the reconstruction problem only for a small set of the most probable scatterers in the image. A frequency domain the {omega}-k SAFT algorithm, which relies on the far-field approximation is presented in the second part. Finally, a detailed analysis of the most computationally intense step in the ESAFT and the sparse 2D deconvolution is presented. In the final part of the report we introduce basics of the 3D ultrasonic imaging that has a great potential in the inspection of the FSW welds. We discuss in some detail the three interrelated steps involved in the 3D ultrasonic imaging: data acquisition, 3D reconstruction, and 3D visualization.

  3. Fire simulation of the canister transfer and installation vehicle; Kapselin siirto- ja asennusajoneuvon palosimulointi

    Energy Technology Data Exchange (ETDEWEB)

    Peltokorpi, L. [Fortum Power and Heat Oy, Espoo (Finland)

    2012-12-15

    A pyrolysis model of the canister transfer and installation vehicle was developed and vehicle fires in the final disposal tunnel and in the central tunnel were simulated using the fire simulation program FDS (Fire Dynamics Simulator). For comparison, same vehicle fire was also simulated at conditions in which the fire remained as a fuel controlled during the whole simulation. The purpose of the fire simulations was to simulate the fire behaviour realistically taking into account for example the limitations coming from the lack of oxygen. The material parameters for the rubber were defined and the simulation models for the tyres developed by simulating the fire test of a front wheel loader rubber tyre done by SP Technical Research Institute of Sweden. In these simulations the most important phenomena were successfully brought out but the timing of the phenomena was difficult. The final values for the rubber material parameters were chosen so that the simulated fire behaviour was at least as intense as the measured one. In the vehicle fire simulations a hydraulic oil or diesel leak causing a pool fire size of 2 MW and 2 m{sup 2} was assumed. The pool fire was assumed to be located under the tyres of the SPMT (Self Propelled Modular Transporters) transporter. In each of the vehicle fire simulations only the tyres of the SPMT transporter were observed to be burning whereas the tyres of the trailer remained untouched. In the fuel controlled fire the maximum power was slightly under 10 MW which was reached in about 18 minutes. In the final disposal tunnel the growth of the fire was limited due to the lack of oxygen and the relatively fast air flows existing in the tunnel. Fast air flows caused the flame spreading to be limited to the certain directions. In the final disposal tunnel fire the maximum power was slightly over 7 MW which was reached about 8 minutes after the ignition. In the central tunnel there was no shortage of oxygen but the spread of the fire was limited

  4. Development of a constitutive model for the plastic deformation and creep of copper and its use in the estimate of the creep life of the copper canister

    Energy Technology Data Exchange (ETDEWEB)

    Pettersson, Kjell [Matsafe AB, Stockholm (Sweden)

    2006-12-15

    A previously developed model for the plastic deformation and creep of copper (included as an Appendix to the present report) has been used as the basis for a discussion on the possibility of brittle creep fracture of the copper canister during long term storage of nuclear waste. Reported creep tests on oxygen free (OF) copper have demonstrated that copper can have an extremely low creep ductility. However with the addition of about 50 ppm phosphorus to the copper it appears as if the creep brittleness problem is avoided and that type of copper (OFP) has consequently been chosen as the canister material. It is shown in the report that the experiments performed on OFP copper does not exclude the possibility of creep brittleness of OFP copper in the very long term. The plasticity and creep model has been used to estimate creep life under conditions of intergranular creep cracking according to a model formulated by Cocks and Ashby. The estimated life times widely exceed the design life of the canister. However the observations of creep brittleness in OF copper indicate that the Cocks-Ashby model probably does not apply to the OF copper. Thus additional calculations have been done with the plasticity and creep model in order to estimate stress as a function of time for the probably most severe loading case of the canister with regard to creep failure, an earth quake shear. Despite the fact that the stress in the canister will remain at the 100 MPa level for thousands of years after an earth quake the low temperature, about 50 deg C or less, will make the solid state diffusion process assumed to control the brittle cracking process, too slow to lead to any significant brittle creep cracking in the canister.

  5. Inspection of copper canister for spent nuclear fuel by means of ultrasound. Copper characterization, FSW monitoring with acoustic emission and ultrasonic imaging

    Energy Technology Data Exchange (ETDEWEB)

    Stepinski, Tadeusz (ed.); Engholm, Marcus; Olofsson, Tomas (Uppsala Univ., Signals and Systems, Dept. of Technical Sciences, Uppsala (Sweden))

    2009-08-15

    This report contains the research results concerning advanced ultrasound for the inspection of copper canisters for spent nuclear fuel obtained at Signals and Systems, Uppsala University in 2008. The first part of the report is concerned with aspects related to ultrasonic attenuation of copper material used for canisters. We present results of attenuation measurement performed for a number of samples taken from a real canister; two from the lid and four from different parts of canister wall. Ultrasonic attenuation of the material originating from canister lid is relatively low (less that 50 dB/m) and essentially frequency independent in the frequency range up to 5 MHz. However, for the material originating from the extruded canister part considerable variations of the attenuation are observed, which can reach even 200 dB/m at 3.5 MHz. In the second part of the report we present further development of the concept of the friction stir welding process monitoring by means of multiple sensors formed into a uniform circular array (UCA). After a brief introduction into modeling Lamb waves and UCA we focus on array processing techniques that enable estimating direction of arrival of multimodal Lamb waves. We consider two new techniques, the Capon beamformer and the broadband multiple signal classification technique (MUSIC). We present simulation results illustrating their performance. In the final part we present the phase shift migration algorithm for ultrasonic imaging of layered media using synthetic aperture concept. We start from explaining theory of the phase migration concept, which is followed by the results of experiments performed on copper blocks with drilled holes. We show that the proposed algorithm performs well for immersion inspection of metal objects and yields both improved spatial resolution and suppressed grain noise

  6. Oxidative dissolution of spent fuel and release of nuclides from a copper/iron canister. Model developments and applications

    Energy Technology Data Exchange (ETDEWEB)

    Longcheng Liu

    2001-12-01

    Three models have been developed and applied in the performance assessment of a final repository. They are based on accepted theories and experimental results for known and possible mechanisms that may dominate in the oxidative dissolution of spent fuel and the release of nuclides from a canister. Assuming that the canister is breached at an early stage after disposal, the three models describe three sub-systems in the near field of the repository, in which the governing processes and mechanisms are quite different. In the model for the oxidative dissolution of the fuel matrix, a set of kinetic descriptions is provided that describes the oxidative dissolution of the fuel matrix and the release of the embedded nuclides. In particular, the effect of autocatalytic reduction of hexavalent uranium by dissolved H{sub 2}, using UO{sub 2} (s) on the fuel pellets as a catalyst, is taken into account. The simulation results suggest that most of the radiolytic oxidants will be consumed by the oxidation of the fuel matrix, and that much less will be depleted by dissolved ferrous iron. Most of the radiolytically produced hexavalent uranium will be reduced by the autocatalytic reaction with H{sub 2} on the fuel surface. It will reprecipitate as UO{sub 2} (s) on the fuel surface, and thus very little net oxidation of the fuel will take place. In the reactive transport model, the interactions of multiple processes within a defective canister are described, in which numerous redox reactions take place as multiple species diffuse. The effect of corrosion of the cast iron insert of the canister and the reduction of dissolved hexavalent uranium by ferrous iron sorbed onto iron corrosion products and by dissolved H{sub 2} are particularly included. Scoping calculations suggest that corrosion of the iron insert will occur primarily under anaerobic conditions. The escaping oxidants from the fuel rods will migrate toward the iron insert. Much of these oxidants will, however, be consumed

  7. Horizontal deposition of canisters for spent nuclear fuel. Summary of the KBS-3H Project 2004-2007

    Energy Technology Data Exchange (ETDEWEB)

    2008-12-15

    SKB and Posiva both selected the KBS-3 method for the geologic disposal of spent nuclear fuel. The KBS-3 method relies on stable and favourable conditions of the bedrock, long-lived canisters containing the spent fuel and the buffer functions of clay surrounding the canister. The reference design is the KBS-3V, in which the canisters with spent nuclear fuel are emplaced vertically in individual deposition holes. For a number of years SKB and Posiva have also jointly studied a design in which the canisters are instead serially emplaced in long horizontal drifts (KBS-3H). The drivers behind the development of the KBS-3H concept are that both cost and environmental impact could be reduced without compromising long-term safety. There are many similarities between KBS-3H and KBS-3V as both designs are based on the KBS-3 method. The main objectives of KBS-3H Project 2004-2007 were to demonstrate that the deposition alternative is technically feasible and that it fulfils the same long-term safety requirements as KBS-3V. These main objectives have only been partially met owing to the restrictions imposed before the start of the project and during its execution. More work is needed for the full demonstration of the engineering feasibility with due consideration to anticipated, site-specific conditions. In KBS-3H Project 2004-2007, it was demonstrated that it was possible to excavate horizontal drifts that would fulfil most of the stringent requirements on geometry dictated by the use of current standard technology. It was further demonstrated that it is possible to emplace a 46-tonne supercontainer in a deposition drift using water-cushion technology. A critical1 issue for the robustness of the KBS-3H during emplacement and saturation is that the groundwater seepage into the deposition drift is low (< 0.1 l/min over the entire length of the supercontainer section) as higher inflow may cause piping/erosion of the buffer during the saturation period. A Mega-Packer was

  8. A FRAMEWORK TO DEVELOP FLAW ACCEPTANCE CRITERIA FOR STRUCTURAL INTEGRITY ASSESSMENT OF MULTIPURPOSE CANISTERS FOR EXTENDED STORAGE OF USED NUCLEAR FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Lam, P.; Sindelar, R.; Duncan, A.; Adams, T.

    2014-04-07

    A multipurpose canister (MPC) made of austenitic stainless steel is loaded with used nuclear fuel assemblies and is part of the transfer cask system to move the fuel from the spent fuel pool to prepare for storage, and is part of the storage cask system for on-site dry storage. This weld-sealed canister is also expected to be part of the transportation package following storage. The canister may be subject to service-induced degradation especially if exposed to aggressive environments during possible very long-term storage period if the permanent repository is yet to be identified and readied. Stress corrosion cracking may be initiated on the canister surface in the welds or in the heat affected zone because the construction of MPC does not require heat treatment for stress relief. An acceptance criteria methodology is being developed for flaw disposition should the crack-like defects be detected by periodic Inservice Inspection. The external loading cases include thermal accident scenarios and cask drop conditions with the contribution from the welding residual stresses. The determination of acceptable flaw size is based on the procedure to evaluate flaw stability provided by American Petroleum Institute (API) 579 Fitness-for-Service (Second Edition). The material mechanical and fracture properties for base and weld metals and the stress analysis results are obtained from the open literature such as NUREG-1864. Subcritical crack growth from stress corrosion cracking (SCC), and its impact on inspection intervals and acceptance criteria, is not addressed.

  9. 新型舰载同心筒发射过程流场研究%Launching Process of the New Type Shipborne Concentric Canister Launcher

    Institute of Scientific and Technical Information of China (English)

    邵立武; 姜毅; 马艳丽; 王伟臣

    2011-01-01

    The internal-external canister space should be enough to exhaust jet flow when the concentric canister launcher is used. The common-frame launch is used for the shipborne weapon which would cause the small sizes of the internal canister and ballistic missile. To solve the problem, the external canister is designed to be rectangular. The internal-external canister space increases with the same common frame and the missile temperature decreases. The three-dimensional dynamic meshes were used to study the launching process of the concentric canister launcher. The zone moving and dynamic methods were used to update the meshes. The results showed that the location curve accord well with experiment. The missile surface temperature decreased greatly with the new type concentric canister launcher.%采用同心筒垂直发射装置,内外筒要保证足够的间隙尺寸用来排导燃气,目前舰载发射均采用通垂方式,势必使得内筒尺寸较小,从而导弹的直径也就较小,不利于发挥弹道导弹的优势.在此基础上提出将传统的外筒设计为一方形结构,在与相同尺寸的通垂架相连接的前提下,增加了内外筒之间燃气排导空间,降低发射过程中导弹表面的温度.计算中使用三维动网格更新方法对同心筒发射过程进行了计算研究,网格更新方法采用域动分层法.结果表明,导弹运动位移曲线与试验符合较好,新型同心筒方案使得整个发射过程中导弹表面的温度均明显降低.

  10. State of the art of the welding method for sealing spent nuclear fuel canister made of copper. Part 1 - FSW

    Energy Technology Data Exchange (ETDEWEB)

    Purhonen, T.

    2014-05-15

    The purpose of this report is to gather together comprehensive information concerning FSW as an optional welding method for welding the nuclear waste copper canister at the disposal facility. This report discusses the current situation, knowledge of the process and information concerning results of the development and research work related to welding thick copper and the special needs of the disposal environment. Most of the research work and development work has been done by Posiva's Swedish partner SKB, Swedish Nuclear Fuel and Waste Management Co. SKB chose FSW as their reference welding method in 2005. FSW (friction stir welding) is a solid-state welding method, invented in 1991, in which frictional heat is generated between the tool and the weld metal, causing the metal to soften, normally without reaching the melting point, and allowing the tool to traverse the joint line. Friction stir welding can be used for joining many types of materials and material combinations, if the tool materials and designs can be found which operate at the forging temperature of the workpiece. The general requirements for the copper canister weld and base material are presented in Posiva's VAHA-system, which sets the most critical values or demands concerning the short- and long-term properties or other needs. The sections in this report are set out in a similar way as in the VAHA-system. Concerning the results from the research and development work, it can be said that FS weld material fulfils the values set by VAHA. The quality of the welds fulfils the set demands for intact weld material and the welding process is robust using an automatic control system. There still remains work concerning the acceptance procedure for the welding process and other open issues which are described in this report. (orig.)

  11. Evaluation of Multi Canister Overpack (MCO) Handling Machine Uplift Restraint for a Seismic Event During Repositioning Operations

    Energy Technology Data Exchange (ETDEWEB)

    SWENSON, C.E.

    2000-05-15

    Insertion of the Multi-Canister Overpack (MCO) assemblies into the Canister Storage Building (CSB) storage tubes involves the use of the MCO Handling Machine (MHM). During MCO storage tube insertion operations, inadvertent movement of the MHM is prevented by engaging seismic restraints (''active restraints'') located adjacent to both the bridge and trolley wheels. During MHM repositioning operations, the active restraints are not engaged. When the active seismic restraints are not engaged, the only functioning seismic restraints are non-engageable (''passive'') wheel uplift restraints which function only if the wheel uplift is sufficient to close the nominal 0.5-inch gap at the uplift restraint interface. The MHM was designed and analyzed in accordance with ASME NOG-1-1995. The ALSTHOM seismic analysis reported seismic loads on the MHM uplift restraints and EDERER performed corresponding structural calculations to demonstrate structural adequacy of the seismic uplift restraint hardware. The ALSTHOM and EDERER calculations were performed for a parked MHM with the active seismic restraints engaged, resulting in uplift restraint loading only in the vertical direction. In support of development of the CSB Safety Analysis Report (SAR), an evaluation of the MHM seismic response was requested for the case where the active seismic restraints are not engaged. If a seismic event occurs during MHM repositioning operations, a moving contact at a seismic uplift restraint would introduce a friction load on the restraint in the direction of the movement. These potential horizontal friction loads on the uplift restraints were not included in the existing restraint hardware design calculations. One of the purposes of the current evaluation is to address the structural adequacy of the MHM seismic uplift restraints with the addition of the horizontal friction associated with MHM repositioning movements.

  12. X-38: Parachute Canister Fired from Plywood Mockup during Flight Termination System Test

    Science.gov (United States)

    1996-01-01

    The canister containing a seven-foot-diameter X-38 Flight Termination System (FTS) parachute is launched safely away from a plywood mockup of the X-38 by a pyrotechnic firing system on December 19, 1996, at NASA Dryden Flight Research Center, Edwards, California. The test was economically accomplished by mounting the mockup of the X-38's aft end, minus vertical stabilizers, on a truck prior to installation in the X-38. The X-38 Crew Return Vehicle (CRV) research project is designed to develop the technology for a prototype emergency crew return vehicle, or lifeboat, for the International Space Station. The project is also intended to develop a crew return vehicle design that could be modified for other uses, such as a joint U.S. and international human spacecraft that could be launched on the French Ariane-5 Booster. The X-38 project is using available technology and off-the-shelf equipment to significantly decrease development costs. Original estimates to develop a capsule-type crew return vehicle were estimated at more than $2 billion. X-38 project officials have estimated that development costs for the X-38 concept will be approximately one quarter of the original estimate. Off-the-shelf technology is not necessarily 'old' technology. Many of the technologies being used in the X-38 project have never before been applied to a human-flight spacecraft. For example, the X-38 flight computer is commercial equipment currently used in aircraft and the flight software operating system is a commercial system already in use in many aerospace applications. The video equipment for the X-38 is existing equipment, some of which has already flown on the space shuttle for previous NASA experiments. The X-38's primary navigational equipment, the Inertial Navigation System/Global Positioning System, is a unit already in use on Navy fighters. The X-38 electromechanical actuators come from previous joint NASA, U.S. Air Force, and U.S. Navy research and development projects. Finally

  13. 湿式独立自排导垂直发射技术研究%Research on the Wet-type Concentric Canister Launcher

    Institute of Scientific and Technical Information of China (English)

    马艳丽; 姜毅; 王伟臣; 刘伯伟; 颜凤

    2011-01-01

    To study the overhigh pressure in the canister during the wet-type canister missile vertical launching, the wet-type concentric canister launchers of various structure parameters are calculated numerically. The launching process of the concentric canister with 17mm space between the inside and outside and single-canister are analyzed numerically with the dynamic update method. Considering the effect of water vaporization, the flow field of vapor and liquid is resolved by Mixture model of dual-phase flow. The zone moving and dynamic laying method was adopted in the meshes update. The result indicates that the smaller space between the inside and outside or the distance between nozzle and rear cover, the bigger pressure in the canister, and the guiding cone in the bottom can obviously decrease the pressure on the rear cover. Therefore, the calculation result accords with the experiment well.%湿式独立自排导垂直发射技术在发射过程中可能会存在发射筒内压力过高的问题,就不同结构参数的湿式独立自排导垂直发射装置进行数值计算,并采用动网格对内外筒间距为17 mm及单筒发射的导弹发射过程进行数值研究和分析.计算中考虑水的汽化效应,采用Mixture两相流计算模型求解气液两相流场,网格更新方法采用域动分层法.结果表明,内外筒间距越小,燃气的排导越受限制,筒内的压力越大;喷管出口与后盖部的距离越小,筒内压力越大;底部加导流锥对于降低后盖上的压力作用明显,计算结果与试验符合较好.

  14. B类滤毒罐防护磷化氢性能评价探讨%Evaluation of B-Type Canister Protective Performance against Phosphine

    Institute of Scientific and Technical Information of China (English)

    汪东旺; 李泽; 赵鑫华; 尹维东; 李志坚; 元以栋

    2012-01-01

    磷化氢是粮食仓储企业使用效果最好的杀虫剂,是一种剧毒的气体熏蒸剂。所以对于从业人员来讲,安全防护就显得格外重要。长期以来粮食仓储企业一直使用自吸过滤式防毒面具(配套选用B类滤毒罐)为首选器材。近期业内对B类滤毒罐防护磷化氢的有效性提出质疑,为此,本文通过性能评价试验,说明B类滤毒罐防护磷化氢是有效的,并从理论上说明B类滤毒罐防护磷化氢是有科学依据的。%Phosphine is the best pesticides of the grain storage enterprise, and is the toxic gaseous fumigant. For users, it is important for the security. For a long time, the grain storage enterprise uses the serf-absorption filtering gas mask (selecting the B-type canister) . Recently, the protective performance against phosphine of B-type canister is doubted, through the test, this paper will explain the effective and scientific theory of B-type canister.

  15. Inspection of copper canisters for spent nuclear fuel by means of ultrasound. Phased arrays, ultrasonic imaging and nonlinear acoustics

    Energy Technology Data Exchange (ETDEWEB)

    Stepinski, Tadeusz (ed.); Ping Wu; Wennerstroem, Erik [Uppsala Univ. (Sweden). Signals and Systems

    2004-09-01

    This report contains the research results concerning advanced ultrasound for the inspection of copper canisters for spent nuclear fuel obtained at Signals and Systems, Uppsala University in years 2003/2004. After a short introduction a review of beam forming fundamentals required for proper understanding phased array operation is included. The factors that determine lateral resolution during ultrasonic imaging of flaws in solids are analyzed and results of simulations modelling contact inspection of copper are presented. In the second chapter an improved synthetic aperture imaging (SAI) technique is introduced. The proposed SAI technique is characterized by an enhanced lateral resolution compared with the previously proposed extended synthetic aperture focusing technique (ESAFT). The enhancement of imaging performance is achieved due to more realistic assumption concerning the probability density function of scatterers in the region of interest. The proposed technique takes the form of a two-step algorithm using the result obtained in the first step as a prior for the second step. Final chapter contains summary of our recent experimental and theoretical research on nonlinear ultrasonics of unbounded interfaces. A new theoretical model for rough interfaces is developed, and the experimental results from the copper specimens that mimic contact cracks of different types are presented. Derivation of the theory and selected measurement results are given in appendix.

  16. Inspection of copper canisters for spent nuclear fuel by means of ultrasound. Nonlinear acoustics, synthetic aperture imaging

    Energy Technology Data Exchange (ETDEWEB)

    Lingvall, Fredrik; Ping Wu; Stepinski, Tadeusz [Uppsala Univ., (Sweden). Dept. of Materials Science

    2003-03-01

    This report contains results concerning inspection of copper canisters for spent nuclear fuel by means of ultrasound obtained at Signals and Systems, Uppsala University in year 2001/2002. The first chapter presents results of an investigation of a new method for synthetic aperture imaging. The new method presented here takes the form of a 2D filter based on minimum mean squared error (MMSE) criteria. The filter, which varies with the target position in two dimensions includes information about spatial impulse response (SIR) of the imaging system. Spatial resolution of the MMSE method is investigated and compared experimentally to that of the classical SAFT and phased array imaging. It is shown that the resolution of the MMSE algorithm, evaluated for imaging immersed copper specimen is superior to that observed for the two above-mentioned methods. Extended experimental and theoretical research concerning the potential of nonlinear waves and material harmonic imaging is presented in the second chapter. An experimental work is presented that was conducted using the RITEC RAM-5000 ultrasonic system capable of providing a high power tone-burst output. A new method for simulation of nonlinear acoustic waves that is a combination of the angular spectrum approach and the Burger's equation is also presented. This method was used for simulating nonlinear elastic waves radiated by the annular transducer that was used in the experiments.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-01

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

  18. Volatile Profiles of Emissions from Different Activities Analyzed Using Canister Samplers and Gas Chromatography-Mass Spectrometry (GC/MS) Analysis: A Case Study

    Science.gov (United States)

    Orecchio, Santino; Fiore, Michele; Barreca, Salvatore; Vara, Gabriele

    2017-01-01

    The objective of present study was to identify volatile organic compounds (VOCs) emitted from several sources (fuels, traffic, landfills, coffee roasting, a street-food laboratory, building work, indoor use of incense and candles, a dental laboratory, etc.) located in Palermo (Italy) by using canister autosamplers and gas chromatography-mass spectrometry (GC-MS) technique. In this study, 181 VOCs were monitored. In the atmosphere of Palermo city, propane, butane, isopentane, methyl pentane, hexane, benzene, toluene, meta- and para-xylene, 1,2,4 trimethyl benzene, 1,3,5 trimethyl benzene, ethylbenzene, 4 ethyl toluene and heptane were identified and quantified in all sampling sites. PMID:28212294

  19. HYDRA-I: a three-dimensional finite difference code for calculating the thermohydraulic performance of a fuel assembly contained within a canister

    Energy Technology Data Exchange (ETDEWEB)

    McCann, R.A.

    1980-12-01

    A finite difference computer code, named HYDRA-I, has been developed to simulate the three-dimensional performance of a spent fuel assembly contained within a cylindrical canister. The code accounts for the coupled heat transfer modes of conduction, convection, and radiation and permits spatially varying boundary conditions, thermophysical properties, and power generation rates. This document is intended as a manual for potential users of HYDRA-I. A brief discussion of the governing equations, the solution technique, and a detailed description of how to set up and execute a problem are presented. HYDRA-I is designed for operation on a CDC 7600 computer. An appendix is included that summarizes approximately two dozen different cases that have been examined. The cases encompass variations in fuel assembly and canister configurations, power generation rates, filler materials, and gases. The results presented show maximum and various local temperatures and heat fluxes illustrating the changing importance of the three heat transfer modes. Finally, the need for comparison with experimental data is emphasized as an aid in code verification although the limited data available indicate excellent agreement.

  20. Whole air canister sampling coupled with preconcentration GC/MS analysis of part-per-trillion levels of trimethylsilanol in semiconductor cleanroom air.

    Science.gov (United States)

    Herrington, Jason S

    2013-08-20

    The costly damage airborne trimethylsilanol (TMS) exacts on optics in the semiconductor industry has resulted in the demand for accurate and reliable methods for measuring TMS at trace levels (i.e., parts per trillion, volume per volume of air [ppt(v)] [~ng/m(3)]). In this study I developed a whole air canister-based approach for field sampling trimethylsilanol in air, as well as a preconcentration gas chromatography/mass spectrometry laboratory method for analysis. The results demonstrate clean canister blanks (0.06 ppt(v) [0.24 ng/m(3)], which is below the detection limit), excellent linearity (a calibration relative response factor relative standard deviation [RSD] of 9.8%) over a wide dynamic mass range (1-100 ppt(v)), recovery/accuracy of 93%, a low selected ion monitoring method detection limit of 0.12 ppt(v) (0.48 ng/m(3)), replicate precision of 6.8% RSD, and stability (84% recovery) out to four days of storage at room temperature. Samples collected at two silicon wafer fabrication facilities ranged from 10.0 to 9120 ppt(v) TMS and appear to be associated with the use of hexamethyldisilazane priming agent. This method will enable semiconductor cleanroom managers to monitor and control for trace levels of trimethylsilanol.

  1. Engineered Barrier System - Assessment of the Corrosion Properties of Copper Canisters. Report from a Workshop. Synthesis and extended abstract

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, Peter (ed.) [Quintessa Ltd., Henley-on-Thames (GB)] (and others)

    2006-03-15

    assumption turns out not to be valid at some stage during the repository evolution. Workshop participants suggested a need for SKI to review SKB's canister corrosion model in more detail as part of future safety assessment reviews (calculations, assumptions and data). Additional experimental work might be needed for the assessment of copper corrosion in high chloride environments and with simultaneous presence of chloride and sulphide. It is essential that altogether consistent facts, understanding and models are used when developing an argument. Any inconsistency regarding these three aspects (facts, understanding, models) needs to be identified. An example would be if thermodynamic data and theoretical calculations suggest that corrosion will not happen, while kinetic data (experimental results) suggest a significant corrosion rate. For future safety assessments, SKB is recommended to use a consistent template for the handling of different corrosion mechanisms even if their final treatment will be quite different. This may include e.g. an extended application of the exclusion principle and/or application of the decision tree approach (as applied for stress corrosion cracking in the Canadian programme). However, it should be noted that the reliability of the exclusion principle depends on the quantity and quality of information on which it is based, and that more explicit criteria might be needed to support the decision tree approach. There is also a need for a well structured approach to handling uncertainties. Examples include those that can be characterised as variability (welding defects, sulphide content of groundwater and bentonite) and as lack of knowledge (e.g. microbial viability, the existence of an unidentified groundwater component affecting corrosion or an unknown corrosion mechanism). A suitable combination of a probabilistic application of the main copper corrosion model, well supported calculation cases with mechanistic models and possibly a selection

  2. Engineered Barrier System - Assessment of the Corrosion Properties of Copper Canisters. Report from a Workshop. Synthesis and extended abstract

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, Peter (ed.) [Quintessa Ltd., Henley-on-Thames (GB)] (and others)

    2006-03-15

    assumption turns out not to be valid at some stage during the repository evolution. Workshop participants suggested a need for SKI to review SKB's canister corrosion model in more detail as part of future safety assessment reviews (calculations, assumptions and data). Additional experimental work might be needed for the assessment of copper corrosion in high chloride environments and with simultaneous presence of chloride and sulphide. It is essential that altogether consistent facts, understanding and models are used when developing an argument. Any inconsistency regarding these three aspects (facts, understanding, models) needs to be identified. An example would be if thermodynamic data and theoretical calculations suggest that corrosion will not happen, while kinetic data (experimental results) suggest a significant corrosion rate. For future safety assessments, SKB is recommended to use a consistent template for the handling of different corrosion mechanisms even if their final treatment will be quite different. This may include e.g. an extended application of the exclusion principle and/or application of the decision tree approach (as applied for stress corrosion cracking in the Canadian programme). However, it should be noted that the reliability of the exclusion principle depends on the quantity and quality of information on which it is based, and that more explicit criteria might be needed to support the decision tree approach. There is also a need for a well structured approach to handling uncertainties. Examples include those that can be characterised as variability (welding defects, sulphide content of groundwater and bentonite) and as lack of knowledge (e.g. microbial viability, the existence of an unidentified groundwater component affecting corrosion or an unknown corrosion mechanism). A suitable combination of a probabilistic application of the main copper corrosion model, well supported calculation cases with mechanistic models and possibly a selection

  3. Evaluating the use of PAO (4 cSt polyalphaoelfin) oil instead of DOP (di-octyl phthalate) oil for measuring the aerosol capture of nuclear canister filters

    Energy Technology Data Exchange (ETDEWEB)

    Moore, Murray E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-07-18

    This document details the distinction between using PAO (4 cSt polyalphaoelfin) oil instead of DOP (di-octyl phthalate) oil for measuring the aerosol capture of filters. This document is developed to justify the use of PAO rather than DOP for evaluating the performance of filters in the SAVY 4000 and Hagan containers. The design criteria (Anderson et al, 2012) for purchasing SAVY 4000 containers and the Safety Analysis Report for the SAVY 4000 Container Series specified that the filter must “capture greater than 99.97% of 0.45 μm mean diameter dioctyl phthalate (DOP) aerosol at the rated flow with a DOP concentration of 65±15 micrograms per liter.”This corresponds to a leakage percent of 0.03% (3.0x10-2). The density of DOP oil is 985 kg/m3 and the density of PAO oil is 819 kg/m3. ATI Test Inc measured the mass mean diameter of aerosol distributions produced by a single Laskin type III-A nozzle operating at a 20 psig air pressure as 0.563 μm for DOP oil and 0.549 μm for PAO oil. (See Appendix A.) For both types of oil in this document, the single fiber method calculated the leakage percent to be 4.4x10-5 for DOP oil and 4.7x10-5 for PAO oil. Although the percent error between these two quantities is 7.7%, these calculated leakage percent values are more than two orders of magnitude less than the criterion specified in the SAVY canister SAR. As a point of reference, the photometer used to measure the SAVY canister filter performance cannot resolve values for the leakage percent below 1.0x10-5. Additionally, over a range of particle sizes from 0.01 μm to 3.0 μm, there was less than 4.0x10-5 error between the calculated filter efficiency for the two types of oil at any particular particle size diameter. In conclusion, the difference between using DOP and PAO for testing SAVY canister filters is of inconsequential concern.

  4. Stack Flow Rate Changes and the ANSI/N13.1-1999 Qualification Criteria: Application to the Hanford Canister Storage Building Stack

    Energy Technology Data Exchange (ETDEWEB)

    Flaherty, Julia E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Glissmeyer, John A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-02-29

    The Canister Storage Building (CSB), located in the 200-East Area of the Hanford Site, is a 42,000 square foot facility used to store spent nuclear fuel from past activities at the Hanford Site. Because the facility has the potential to emit radionuclides into the environment, its ventilation exhaust stack has been equipped with an air monitoring system. Subpart H of the National Emissions Standards for Hazardous Air Pollutants requires that a sampling probe be located in the exhaust stack in accordance with criteria established by the American National Standards Institute/Health Physics Society Standard N13.1-1999, Sampling and Monitoring Releases of Airborne Radioactive Substances from the Stack and Ducts of Nuclear Facilities.

  5. Geological Disposal of Nuclear Waste: Investigating the Thermo-Hygro-Mechanical-Chemical (THMC) Coupled Processes at the Waste Canister- Bentonite Barrier Interface

    Science.gov (United States)

    Davies, C. W.; Davie, D. C.; Charles, D. A.

    2015-12-01

    Geological disposal of nuclear waste is being increasingly considered to deal with the growing volume of waste resulting from the nuclear legacy of numerous nations. Within the UK there is 650,000 cubic meters of waste safely stored and managed in near-surface interim facilities but with no conclusive permanent disposal route. A Geological Disposal Facility with incorporated Engineered Barrier Systems are currently being considered as a permanent waste management solution (Fig.1). This research focuses on the EBS bentonite buffer/waste canister interface, and experimentally replicates key environmental phases that would occur after canister emplacement. This progresses understanding of the temporal evolution of the EBS and the associated impact on its engineering, mineralogical and physicochemical state and considers any consequences for the EBS safety functions of containment and isolation. Correlation of engineering properties to the physicochemical state is the focus of this research. Changes to geotechnical properties such as Atterberg limits, swelling pressure and swelling kinetics are measured after laboratory exposure to THMC variables from interface and batch experiments. Factors affecting the barrier, post closure, include corrosion product interaction, precipitation of silica, near-field chemical environment, groundwater salinity and temperature. Results show that increasing groundwater salinity has a direct impact on the buffer, reducing swelling capacity and plasticity index by up to 80%. Similarly, thermal loading reduces swelling capacity by 23% and plasticity index by 5%. Bentonite/steel interaction studies show corrosion precipitates diffusing into compacted bentonite up to 3mm from the interface over a 4 month exposure (increasing with temperature), with reduction in swelling capacity in the affected zone, probably due to the development of poorly crystalline iron oxides. These results indicate that groundwater conditions, temperature and corrosion

  6. Inspection of copper canisters for spent nuclear fuel by means of ultrasonic array system. Modelling, defect detection and grain noise estimation

    Energy Technology Data Exchange (ETDEWEB)

    Wu Ping; Stepinski, T. [Uppsala Univ., (Sweden). Dept. of Material Science

    1998-07-01

    The work presented in the report has been split into three overlapping tasks which have the following objectives: (1) development of beam-forming tools, and verification of modeling tools; (2) investigation of detection and resolution limits; (3) evaluation of attenuation, estimation and suppression of grain noise. For beam-forming tools, a method of designing steered and/or focused beams in immersed solids is presented based on geometrical acoustics. Presently, the beam designs are only related to delays but not to apodization. These focused, steered beams are intended to be used for sizing defects and inspecting the regions close to canisters outer walls. The modeling tool developed previously for simulating elastic fields radiated by planar arrays into immersed solids has been verified by comparing with the results obtained from PASS, a software developed by Dr. Didier Cassereau, France. The results from our modeling tool are in excellent agreement with those from PASS. Since the array coming with the ALLIN ultrasonic array system is not planar, but cylindrically curved in elevation, and it works not in transmission mode, but in pulse echo mode, the above modeling tool for the planar arrays cannot be applied directly. Therefore, the modeling tool has been upgraded for the ALLIN array. The theory underlying this modeling tool is the extended angular spectrum approach (ASA) which was developed based on the conventional ASA that only applies to planar sources. Experimental verification of the modeling tool has shown that the results from the tool agree very well with the measurements. To quantify the fields from the ALLIN array and to facilitate the comparison of simulated results with the measured ones, the ALLIN array system has been calibrated based on the existing functionality, and an analytical model has been proposed for simulating measured acoustic echo pulses. To investigate the detection and resolution limits, we have carried out a series of experiments

  7. Interior Ballistic Modeling and Simulation of Underwater Launched Missile Using Concentric Canister Launcher%同心筒水下发射内弹道建模与仿真研究

    Institute of Scientific and Technical Information of China (English)

    袁绪龙; 王亚东; 刘维

    2013-01-01

    To construct a fast calculation method of interior ballistics of underwater launched missile using Concentric Canister Launcher(CCL),a simulation model of CCL was established according to the first law of thermodynamics,and power characteristics and underwater environment were considered.The empirical parameters in this model were decided using CFD solutions,and they were verified by more calculations.The influences of design parameters on the interior ballistics were studied by the verified model.The simulation shows that the sizes of canister top and bottom restricting parts can be used to adjust launching velocity of missile;the size of canister bottom restricting part can be adapted to modify the acceleration of missile.It can increase the adjusting range of velocity to increase initial volumes of inner and outer canister.The simulation method and results offer reference for engineers.%为构建一种快速的同心筒水下发射内弹道算法,采用热力学第一定律,结合导弹动力装置特性及水下环境需求,建立了同心筒水下发射内弹道计算模型,用CFD结果辨识并校验模型经验参数.应用校验的模型研究了发射装置设计参数对内弹道参数的影响规律.结果表明:筒口、筒底限流尺寸均可用于调速,筒底限流尺寸可用于调节过载,内、外筒初始容积增大可增大调速范围.仿真方法和结果可供工程设计人员参考.

  8. 罐采样与GC-MS联用测定空气中三甲胺%Determining Trimethylamine in Air by Canister Sampling and GC-MS

    Institute of Scientific and Technical Information of China (English)

    黄旭锋

    2015-01-01

    采用真空罐采样,三级冷阱预浓缩仪处理样品,气相色谱-质谱法测定空气中的三甲胺.结果表明,检出限为0.005 mg/m3,相对标准偏差为4.1%,加标回收率为92.0%~108%,能达到国家标准要求.该方法操作简单,测定准确可靠,可用于环境空气中三甲胺的测定.%With Sampling by vacuum canister and processing samples by three stage cold trap,we determined trimethylamine in the air samples by gas chromatography⁃mass spectrometry. The results show that the detection limit was 0.005 mg/m3,the relative standard deviation is 4.1%,and the recovery rate was 92.0%~108%,which can meet the requirements of Emission Standards for Odor Pollutants. This method is simple,accurate and reliable,which can be used for determination of trimethylamine in ambient air.

  9. Inspection of copper canisters for spent nuclear fuel by means of ultrasound. NDE of friction stir welds, nonlinear acoustics, ultrasonic imaging

    Energy Technology Data Exchange (ETDEWEB)

    Stepinski, Tadeusz (ed.); Lingvall, Fredrik; Wennerstroem, Erik; Ping Wu [Uppsala Univ., Dept. of Materials Science (Sweden). Signals and Systems

    2004-01-01

    This report contains results concerning advanced ultrasound for the inspection of copper canisters for spent nuclear fuel obtained at Signals and Systems, Uppsala University in years 2002/2003. After a short introduction a review of the NDE techniques that have been applied to the assessment of friction stir welds (FSW) is presented. The review is based on the results reported by the specialists from the USA, mostly from the aerospace industry. A separate chapter is devoted to the extended experimental and theoretical research concerning potential of nonlinear waves in NDE applications. Further studies concerning nonlinear propagation of acoustic and elastic waves (classical nonlinearity) are reported. Also a preliminary investigation of the nonlinear ultrasonic detection of contacts and interfaces (non-classical nonlinearity) is included. Report on the continuation of previous work concerning computer simulation of nonlinear propagations of ultrasonic beams in water and in immersed solids is also presented. Finally, results of an investigation concerning a new method of synthetic aperture imaging (SAI) and its comparison to the traditional phased array (PA) imaging and to the synthetic aperture focusing technique (SAFT) are presented. A new spatial-temporal filtering method is presented that is a generalization of the previously proposed filter. Spatial resolution of the proposed method is investigated and compared experimentally to that of classical SAFT and PA imaging. Performance of the proposed method for flat targets is also investigated.

  10. Review of NDE Methods for Detection and Monitoring of Atmospheric SCC in Welded Canisters for the Storage of Used Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, Ryan M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pardini, Allan F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hanson, Brady D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Sorenson, Ken B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-01-14

    Dry cask storage systems (DCSSs) for used nuclear fuel (UNF) were originally envisioned for storage periods of short duration (~ a few decades). However, uncertainty challenges the opening of a permanent repository for UNF implying that UNF will need to remain in dry storage for much longer durations than originally envisioned (possibly for centuries). Thus, aging degradation of DCSSs becomes an issue that may not have been sufficiently considered in the design phase and that can challenge the efficacy of very long-term storage of UNF. A particular aging degradation concern is atmospheric stress corrosion cracking (SCC) of DCSSs located in marine environments. In this report, several nondestructive (NDE) methods are evaluated with respect to their potential for effective monitoring of atmospheric SCC in welded canisters of DCSSs. Several of the methods are selected for evaluation based on their usage for in-service inspection applications in the nuclear power industry. The technologies considered include bulk ultrasonic techniques, acoustic emission, visual techniques, eddy current, and guided ultrasonic waves.

  11. Inspection of copper canisters for spent nuclear fuel by means of ultrasound. Electron beam evaluation, harmonic imaging, materials characterization, and ultrasonic modelling

    Energy Technology Data Exchange (ETDEWEB)

    Wu Ping; Lingvall, Fredrik; Stepinski, Tadeusz [Uppsala Univ. (Sweden). Dept. of Materials Science

    2000-12-01

    This report presents the research in the sixth phase that is concerned with ultrasonic techniques for assessing electron beam (EB) welds in copper canisters. The research has been carried out in three main aspects: (1) comparative inspections of EB welds, (2) EB weld evaluation, and (3) quantitative evaluation of attenuation in copper. Comparative inspections of EB welds in two copper canister blocks have been made by means of ultrasound and radiography. Comparison of the inspected results demonstrate that both techniques complement each other very well. The radiographic technique on the whole gives relatively better spatial resolution but low contrast in radiographs. It can reliably detect voids in EB, but cannot provide information about material structure in the EB weld. Ultrasonic technique provides information about flaw locations and shapes similar to the radiographs. Moreover, it can easily distinguish welded and non-welded zones and be used to study weld's macro- and microstructure. The defects in ultrasonic images often show higher contrast, and some flaw indications may be seen in ultrasonic inspection but not in radiographs. But small flaws are hard to distinguish from grain noise. For EB weld evaluation, first, scattering from EB weld has been investigated using three broadband transducers with different center frequencies. The investigation has shown that more information on scattering and attenuation can be exploited in this case so that the EB welds can be better characterized, and that the best frequency range for characterizing welds is 2 - 5 MHz. Secondly, harmonic imaging (HI) of EB welds have been studied using two different sources of harmonics: (i) transducer harmonics, originating from the high-order resonant modes of transmitters excited by a broadband pulse, and (ii) material harmonics, stemming from the nonlinear distortion of waves propagating in materials. The transducer HI exploits additional information due to transducer harmonics

  12. Inspection of copper canisters for spent nuclear fuel by means of Ultrasonic Array System. Electron beam evaluation, modeling and materials characterization

    Energy Technology Data Exchange (ETDEWEB)

    Ping Wu; Lingvall, F.; Stepinski, T. [Uppsala Univ. (Sweden). Dept. of Material Science

    1999-12-01

    Research conducted in the fifth phase of the SKB's study aimed at developing ultrasonic techniques for assessing EB welds copper canisters is reported here. This report covers three main tasks: evaluation of electron beam (EB) welds, modeling of ultrasonic fields and characterization of copper material. A systematic analysis of ultrasonic interaction and imaging of an EB weld has been performed. From the analysis of histograms of the weld ultrasonic image, it appeared that the porosity tended to be concentrated towards the upper side of a HV weld, and a guideline on how to select the gates for creating C-scans has been proposed. The spatial diversity method (SDM) has shown a limited ability to suppress grain noise both in the parent material (copper) and in the weld so that the ultrasonic image of the weld could be improved. The suppression was achieved at the price of reduced spatial resolution. The ability of wavelet filters to enhance flaw responses has been studied. An FIR (finite impulse response) filter, based on Sombrero mother wavelet, has yield encouraging results concerning clutter suppression. However, the physical explanation for the results is still missing and needs further research. For modeling of ultrasonic fields of the ALLIN array, an approach to computing the SIR (spatial impulse response) of a cylindrically curved, rectangular aperture has been developed. The aperture is split into very narrow strips in the cylindrically curved direction and SIR of the whole aperture by superposing the individual impulse responses of those strips. Using this approach, the SIR of the ALLIN array with a cylindrically curved surface has been calculated. The pulse excitation of normal velocity on the surface of the array, that is required for simulating actual ultrasonic fields, has been determined by measurement in combination with a deconvolution technique. Using the SIR and the pulse excitation obtained, the pulsed-echo fields from the array have been

  13. Development of a method for the study of H{sub 2} gas emission in sealed compartments containing canister copper immersed in O{sub 2}-free water

    Energy Technology Data Exchange (ETDEWEB)

    Bengtsson, Andreas; Chukharkina, Alexandra; Eriksson, Lena; Hallbeck, Bjoern; Hallbeck, Lotta; Johansson, Jessica; Johansson, Linda; Pedersen, Karsten [Microbial Analytics Sweden AB, Moelnlycke (Sweden)

    2013-06-15

    Current models of copper corrosion indicate that copper is not subject to corrosion by water in itself, but that additional components, such as O{sub 2}, chloride or sulphide are needed to initiate a corrosive process. Of late however, a number of reports have suggested that copper may be susceptible to water-induced corrosion in the absence of external constituents affecting the process. The process has been proposed to rely the auto-ionization driven presence of the hydroxide ions in pure water, and to result in the development of atomic hydrogen (H), with subsequent release of H{sub 2} gas. A suggested equilibrium is reached at a partial pressure of H{sub 2} of about 1 mbar (0.1 kPa) in 73 deg C, and the corrosion reaction is proposed to be rate-limited by the supply of hydroxide ions from the water, a process being slower than proposed formation of water from a H{sub 2}-O{sub 2} reaction. In consequence, the presence of O{sub 2} in the system would result in no detectable release of H{sub 2} until all O{sub 2} was consumed, while the absence of O{sub 2} would lead to water-driven corrosion of copper proceeding until the H{sub 2} equilibrium is reached, at a partial H{sub 2} pressure of about 1 mbar. The proposed mechanism presents a novel aspect on copper corrosion processes. By extension, the suggested corrosion process may have implications for proposed strategies for long-term storage of spent nuclear fuel waste (SNF), which in part rely on the long-term (>105 years) integrity of copper canisters stored in anoxic water inundated environments (SKB 2010)

  14. 过滤罐微生物气溶胶过滤效率及其评价方法的研究%A methodological study on testing and evaluating of filtration efficiency of canister against microbial aerosol

    Institute of Scientific and Technical Information of China (English)

    温占波; 赵建军; 李劲松; 王洁; 鹿建春; 李娜

    2009-01-01

    目的 建立防护面具高效过滤罐微生物气溶胶测试评价方法,对过滤罐的实际防护效果进行测试评价.方法 Serratia marcescens作为模式细菌繁殖体气溶胶、Bacillus subtilis var niger芽孢作为模式芽孢气溶胶、噬菌体f2作为模式病毒气溶胶,使用实验室建立的微生物气溶胶检测技术平台,人工发生模式微生物气溶胶,分别在过滤罐过滤前后使用空气微生物采样器进行定量采样,根据过滤前后模式微生物气溶胶的浓度分别计算细菌、芽孢、病毒气溶胶过滤效率.1-1、1-2、1-3、1-44个只含有高效过滤材料的过滤罐分别测试了Serratia marcescens、Bacillus subtilis var niger、噬菌体f2气溶胶的过滤效率.543、544 2个装有活性炭的高效过滤罐测试了对Scrratia marcescens气溶胶的过滤效率.结果 1-1、1-2、1-3 3个高效过滤罐对Serratia marcescens、Bacillus subtilis var niger芽孢、噬菌体f2的气溶胶的过滤效率为100.000%,1-4高效过滤罐对Bacillus subtilis var niger芽孢气溶胶的过滤效率为990997%、Serratia marcescerts和噬菌体f2气溶胶的过滤效率均为100.000%.加入活性炭后543、544 2个过滤罐对Serratia marcescens气溶胶的过滤效率均为100.000%.结论 建立的检测方法可以用于高效过滤罐微生物气溶胶防护效果的评价,高效过滤罐(包括装有活性炭者)微生物气溶胶防护效果均佳.%Objective To establish a testing and evaluating method for filtration efficiency of the eanister against microbial aerosol. Methods Serratia marcescens aerosol served as model of bacterial aerosol, Bacillus subtilis var niger aerosol as model of spores aerosol, bacteriophage f2 aerosol as model of viral aerosol. Employing the microbial aerosol testing platform was established in lab, models of microbial aerosol generated artificially were sampled quantitatively by air samplers before and after filtrating by canisters, respectively. Filtration

  15. 在高湿度环境下用活性炭盒测量氡浓度的研究%STUDY ON RADON CONCENTRATION MONITORING USING ACTIVATED CHARCOAL CANISTERS IN HIGH HUMIDITY ENVIRONMENTS

    Institute of Scientific and Technical Information of China (English)

    王月兴; 王海军; 杨翊方; 秦思昌; 王震涛; 张振江

    2009-01-01

    The effects of humidity on the sensitivity using activated charcoal canisters for measuring radon concentrations in high humidity environments were studied.Every canister filled with 80 g of activated charcoal,and they were exposed to 48 h or 72 h in the relative humidity of 68%,80%,88% and 96% (28℃),respectively.The amount of radon absorbed in the canisters was determined by counting the gamma rays from 214 Pb and 214 Bi (radon progeny).The results showed that counts decreased with the increase of relative humidity.There was a negative linear relationship between count and humidity.In the relative humidity range of 68%-96%,the sensitivity of radon absorption decreased about 2.4% for every 1% (degree) rise in humidity.The results also showed that the exposure time of the activated charcoal canisters should be less than 3 days.%本文研究了在高湿度环境中使用活性炭盒测量氡浓度时湿度对灵敏度的影响.所用的活性炭盒为圆柱形,每一个盒内装80 g活性炭.活性炭盒在相对湿度为68%、80%、88%和96%环境中(28℃)暴露48 h和72 h.在盒内被吸收的氡的量用氡子体214 Ph和214 Bi的γ射线计数确定.实验结果表明,在湿度相同情况下,计数随湿度的增高而降低,两个变量之间呈现负线性相关.在相对湿度68%到96%之间,湿度每增加1%,吸收氡的灵敏度减少约2.4%.在高湿度环境中,活性炭盒的暴露时间不宜超过3天.

  16. Multi-objective optimization for orientator of airdropping launch canister%某运发箱定向器多目标优化设计研究

    Institute of Scientific and Technical Information of China (English)

    胡建国; 仲健林; 马大为; 乐贵高; 周晓和; 杨风波

    2014-01-01

    为实现空投储运发射箱轻量化、保证空投安全性,对定向器进行多目标优化设计。在有限元参数化模型基础上,结合Isight优化平台建立运发箱定向器冲击动力学多目标优化框架;以复合材料定向器铺层厚度、角度为输入变量,定向器质量、最大位移、药柱最大应力为输出变量,采用最优拉丁超立方设计和径向基神经网络方法建立数学近似模型;采用非劣排序遗传算法及模糊集合理论,得到多目标优化模型的Pareto解集及选优方案,并与原方案进行了对比。结果表明:优化方案实现了储运发射箱的轻量化设计,并提高了空投安全性。%To realize the lightweight and guarantee the safety for airdropping launch canister,multi-objective optimization for the orientator was proposed. Firstly,based on the finite element parameterized model,the impact dynamics multi-objective optimization framework was built with Isight platform. Secondly,with layer thickness,layer angle as the input variables and the mass,maximum displacement of orientator,maximum stress of grain as the output variables,the mathematical approximate model was proposed with optimal latin hypercube design and radial basis function neural network method. Finally,on the basis of non-dominated sorting genetic algorithm and fuzzy set theory,the Pareto solutions and the optimum choice were obtained,and compared with the old scheme. The results show that the optimum choice can guarantee the mass of orientator and increase the safety of airdropping.

  17. A natural analogue for copper waste canisters: The copper-uranium mineralised concretions in the Permian mudrocks of south Devon, United Kingdom

    Energy Technology Data Exchange (ETDEWEB)

    Milodowski, A.E.; Styles, M.T.; Hards, V.L. [Natural Environment Research Council (United Kingdom). British Geological Survey

    2000-08-01

    This report presents the results of a small-scale pilot study of the mineralogy and alteration characteristics of unusual sheet-like native copper occurring together with uraniferous and vanadiferous concretions in mudstones and siltstones of the Permian Littleham Mudstone Formation, at Littleham Cove, south Devon, England. The host mudstones and siltstones are smectitic and have been compacted through deep Mesozoic burial. The occurrence of native copper within these rocks represents a natural analogue for the long-term behaviour of copper canisters, sealed in a compacted clay (bentonite) backfill, that will be used for the deep geological disposal of high-level radioactive waste by the SKB. The study was undertaken by the British Geological Survey (BGS) on behalf of SKB between November 1999 and June 2000. The study was based primarily on archived reference material collected by the BGS during regional geological and mineralogical surveys of the area in the 1970's and 1980's. However, a brief visit was made to Littleham Cove in January 2000 to try to examine the native copper in situ and to collect additional material. Unfortunately, recent landslips and mudflows obscured much of the outcrop, and only one new sample of native copper could be collected. The native copper occurs as thin plates, up to 160 mm in diameter, which occur parallel to bedding in the Permian Littleham Mudstone Formation at Littleham Cove (near Budleigh Salterton) in south Devon. Each plate is made up of composite stacks of individual thin copper sheets each 1-2 mm thick. The copper is very pure (>99.4% Cu) but is accompanied by minor amounts of native silver (also pure - >99%) which occurs as small inclusions within the native copper. Detailed mineralogical and petrological studies of the native copper sheets, using optical petrography, backscattered scanning electron microscopy, X-ray diffraction analysis and electron probe microanalytical techniques, reveal a complex history of

  18. A natural analogue for copper waste canisters: The copper-uranium mineralised concretions in the Permian mudrocks of south Devon, United Kingdom

    Energy Technology Data Exchange (ETDEWEB)

    Milodowski, A.E.; Styles, M.T.; Hards, V.L. [Natural Environment Research Council (United Kingdom). British Geological Survey

    2000-08-01

    This report presents the results of a small-scale pilot study of the mineralogy and alteration characteristics of unusual sheet-like native copper occurring together with uraniferous and vanadiferous concretions in mudstones and siltstones of the Permian Littleham Mudstone Formation, at Littleham Cove, south Devon, England. The host mudstones and siltstones are smectitic and have been compacted through deep Mesozoic burial. The occurrence of native copper within these rocks represents a natural analogue for the long-term behaviour of copper canisters, sealed in a compacted clay (bentonite) backfill, that will be used for the deep geological disposal of high-level radioactive waste by the SKB. The study was undertaken by the British Geological Survey (BGS) on behalf of SKB between November 1999 and June 2000. The study was based primarily on archived reference material collected by the BGS during regional geological and mineralogical surveys of the area in the 1970's and 1980's. However, a brief visit was made to Littleham Cove in January 2000 to try to examine the native copper in situ and to collect additional material. Unfortunately, recent landslips and mudflows obscured much of the outcrop, and only one new sample of native copper could be collected. The native copper occurs as thin plates, up to 160 mm in diameter, which occur parallel to bedding in the Permian Littleham Mudstone Formation at Littleham Cove (near Budleigh Salterton) in south Devon. Each plate is made up of composite stacks of individual thin copper sheets each 1-2 mm thick. The copper is very pure (>99.4% Cu) but is accompanied by minor amounts of native silver (also pure - >99%) which occurs as small inclusions within the native copper. Detailed mineralogical and petrological studies of the native copper sheets, using optical petrography, backscattered scanning electron microscopy, X-ray diffraction analysis and electron probe microanalytical techniques, reveal a complex history of

  19. 基于有限元动力学的复合材料发射筒多目标优化设计%Multi-objective Optimization of Composites Launch Canister Based on Finite Element Dynamics

    Institute of Scientific and Technical Information of China (English)

    朱孙科; 马大为; 罗天洪; 李士军

    2012-01-01

    To solve the problem of the contact impact between missile and launch canister, we use the implicit scheme in the finite element's (FE) statics and the explicit scheme in FE dynamics to simulate a missile's charging and launching respectively. We use the composites launch canister's ply thickness and its lamination angle as the optimal design variables and define the launch canister's maximum displacement during missile launching and the launch canister's mass as optimization objectives. Taking into account the FE numerical calculation results, we use the Python language to develop the programs and the non-dominant sequencing genetic algorithm II (NSGA-II) which is based on the Pareto strategy to establish the multi-objective optimization model. With the optimization model, we obtain the Pareto front surface curve and the optimal solution. We use the specific stiffness structural ef- ficiency to' compare the performance of the launch canister before and after the optimization ; the comparison results show that optimal design of the launch canister is effective.%针对弹筒接触碰撞问题,分别采用有限元静力学隐式格式和动力学显式格式,模拟了导弹装填及发射过程。并以复合材料发射筒的复合铺层厚度和铺层角度为优化设计变量,定义发射筒在导弹发射过程中筒口产生的最大位移和发射筒质量为优化目标,结合有限元数值计算结果,采用Python语言进行编程,运用基于Pareto策略的改进的非支配排序遗传算法(NSGA-II),建立了多目标优化模型。通过优化求解,获得了Pareto前沿面曲线和最优解,运用比刚度结构效能对比分析了优化前后的发射筒性能,表明对发射筒的优化设计是有效的。

  20. 新型同心筒自力发射热环境优化设计%Optimization design for thermal environment of a new roadbed concentric canister launcher

    Institute of Scientific and Technical Information of China (English)

    杨风波; 马大为; 任杰; 乐贵高; 蔡德咏

    2015-01-01

    针对新型路基同心筒自力发射热环境评估与优化设计问题,依托弹性变形和域动分层结合的动网格技术,求解了二维轴对称N-S方程,分析了“中段导流”同心筒动态热环境特性,确定了热环境评价指标;通过建立以优化拉丁超立方实验设计和四阶响应面为理论基础的近似数学模型,解决了CFD自动建模困难、直接寻优计算量大的难点;利用多岛遗传和梯度优化算法搭建组合优化策略平台,克服了流场在不同热结构条件下的强非线性问题,并构建了支持近似数学模型的热环境优化构架。对比数值结果表明,倒吸进入内筒的低温气体有力改善了同心筒热环境;建立的近似数学模型精度较高,满足工程需求;优化后,热环境特性发生良性变化,导弹总体热环境得到显著改善。%Based on dynamic mesh technology with spring based smoothing method,and laying based zone moving method, the axisymmetric N-S equations were solved numerically, the dynamic thermal environment characteristics were obtained to deal with the thermal environment evaluating and optimization design problem of the new “middle diversion” concentric canister launcher (CCL),and evaluation index of thermal environment was also determined.The approximate mathematic model was established by op-timal latin hypercube design and fourth-order response surface method to solve the automatic modeling problem of CFD and compen-sate for the shortcoming of large amount of calculation for direct optimization. A combinatorial optimization strategy platform based on Multi-Island Genetic Algorithm and Sequential Quadratic Programming was established to overcome the problem of strong nonlin-ear characteristic of the flow field parameters under different thermal structure conditions,and the optimization design system of ther-mal environment for missile which supports approximate mathematic model was also built

  1. L/E coupling numerical simulation of pressure field near launch canister outlet for underwater vehicle vertical launch%水下航行体垂直发射筒口压力场 L/E耦合数值模拟

    Institute of Scientific and Technical Information of China (English)

    张晓乐; 卢丙举; 胡仁海; 杨兴林

    2016-01-01

    The fluid pressure field near canister outlet for underwater-launched vehicle vertical launch was simulated, using a 3D symmetric model based on the coupling of Lagrange structure mesh and Euler fluid mesh. The water horizontal relative motion and the process of vehicle motion in the launch canister were considered in the model. The characteristics of bubble pulsation were achieved through the simulation. The shape of the two primary pressure waves are approximately identical between simulation results and test results. It shows that, simulation model which consider the lateral flow can be more accurate than the model without lateral flow. The research method and its conclusions are good kind of reference to analysis of pressure field near launch canister.%采用 Lagrange 结构网格和 Euler 流场网格(L/E)耦合的数值仿真方法,对潜射航行器出筒后筒口压力场进行三维数值仿真分析。计算模型考虑海水的横向来流和航行体出筒过程对筒口气泡脉动的影响。仿真计算获得了筒口压力场气泡脉动主要特征,试验与计算的2个主要压力波峰曲线形状基本一致。通过对比计算表明,考虑横向流作用可以有效减少筒口压力场计算偏差。本文的计算方法及结论可为发射筒口压力场分析提供有益指导。

  2. Sealed Planetary Return Canister (SPRC) Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Sample return missions have primary importance in future planetary missions. A basic requirement is that samples be returned in pristine, uncontaminated condition,...

  3. Multi-canister overpack design report

    Energy Technology Data Exchange (ETDEWEB)

    SMITH, K.E.

    1999-05-12

    Revision 2 incorporates changes to reflect a 150 psig pressure rating for the mechanically closed MCO and 450 psig pressure rating with the cover cap welded in place, per the MCO Performance Specification, HNF-S-0426, Rev. 5 .

  4. 42 CFR 84.1154 - Canister and cartridge requirements.

    Science.gov (United States)

    2010-10-01

    ..., and Mist; Pesticide; Paint Spray; Powered Air-Purifying High Efficiency Respirators and Combination... are used in parallel, their resistance to airflow shall be essentially equal. (b) The color...

  5. Phase 2 fire hazard analysis for the canister storage building

    Energy Technology Data Exchange (ETDEWEB)

    Sadanaga, C.T., Westinghouse Hanford

    1996-07-01

    The fire hazard analysis assesses the risk from fire in a facility to ascertain whether the fire protection policies are met. This document provides a preliminary FHA for the CSB facility. Open items have been noted in the document. A final FHA will be required at the completion of definitive design, prior to operation of the facility.

  6. Estimates of particulate mass in multi-canister overpacks

    Energy Technology Data Exchange (ETDEWEB)

    SLOUGHTER, J.P.

    1999-02-25

    High, best estimate, and low values are developed for particulate inventories within MCO baskets that have been loaded with freshly cleaned fuel assemblies and scrap. These per-basket estimates are then applied to all anticipated MCO payload configurations to identify which configurations are bounding for each type of particulate. Finally the resulting bounding and nominal values for residual particulates are combined with corresponding values [from other documents] for particulate that may be generated by corrosion of exposed uranium after the fuel has been cleaned. The resulting rounded nominal estimate for a typical MCO after 40 years of storage is 8 kg. The estimate for a bounding total particulate case MCO is that it may contain up to 64 kg of particulate after 40 years of storage.

  7. Estimates of Particulate Mass in Multi Canister Overpacks (MCO)

    Energy Technology Data Exchange (ETDEWEB)

    SLOUGHTER, J.P.

    2000-02-16

    High, best estimate, and low values are developed for particulate inventories within MCO baskets that have been loaded with freshly cleaned fuel assemblies and scrap. These per-basket estimates are then applied to all anticipated MCO payload configurations to identify which configurations are bounding for each type of particulate. Finally the resulting bounding and nominal values for residual particulates are combined with corresponding values [from other documents] for particulates that may be generated by corrosion of exposed uranium after the fuel has been cleaned. The resulting rounded nominal estimate for a typical MCO after 40 years of storage is 8 kg. The estimate for a bounding total particulate case MCO is that it may contain up to 64 kg of particulate after 40 years of storage.

  8. Modellization of Metal Hydride Canister for Hydrogen Storage

    Directory of Open Access Journals (Sweden)

    Rocio Maceiras

    2015-06-01

    Full Text Available Hydrogen shows very interesting features for its use on-board applications as fuel cell vehicles. This paper presents the modelling of a tank with a metal hydride alloy for on-board applications, which provides good performance under ambient conditions. The metal hydride contained in the tank is Ti0.98Zr0.02V0.43Fe0.09Cr0.05Mn1.5. A two-dimensional model has been performed for the refuelling process (absorption and the discharge process (desorption. For that, individual models of mass balance, energy balance, reaction kinetics and behaviour of hydrogen gas has been modelled. The model has been developed under Matlab / Simulink© environment. Finally, individual models have been integrated into a global model, and simulated under ambient conditions. With the aim to analyse the temperature influence on the state of charge and filling and emptying time, other simulations were performed at different temperatures. The obtained results allow to conclude that this alloy offers a good behaviour with the discharge process under normal ambient conditions. Keywords: Hydrogen storage; metal hydrides; fuel cell; simulation; board applications

  9. Clay Generic Disposal System Model - Sensitivity Analysis for 32 PWR Assembly Canisters (+2 associated model files).

    Energy Technology Data Exchange (ETDEWEB)

    Morris, Edgar [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-10-01

    The Used Fuel Disposition Campaign (UFDC), as part of the DOE Office of Nuclear Energy’s (DOE-NE) Fuel Cycle Technology program (FCT) is investigating the disposal of high level radioactive waste (HLW) and spent nuclear fuela (SNF) in a variety of geologic media. The feasibility of disposing SNF and HLW in clay media has been investigated and has been shown to be promising [Ref. 1]. In addition the disposal of these wastes in clay media is being investigated in Belgium, France, and Switzerland. Thus, Argillaceous media is one of the environments being considered by UFDC. As identified by researchers at Sandia National Laboratory, potentially suitable formations that may exist in the U.S. include mudstone, clay, shale, and argillite formations [Ref. 1]. These formations encompass a broad range of material properties. In this report, reference to clay media is intended to cover the full range of material properties. This report presents the status of the development of a simulation model for evaluating the performance of generic clay media. The clay Generic Disposal System Model (GDSM) repository performance simulation tool has been developed with the flexibility to evaluate not only different properties, but different waste streams/forms and different repository designs and engineered barrier configurations/ materials that could be used to dispose of these wastes.

  10. Impact of Aluminum on Anticipated Corrosion in a Flooded SNF Multi Canister Overpack (MCO)

    Energy Technology Data Exchange (ETDEWEB)

    DUNCAN, D.R.

    1999-07-06

    Corrosion reactions in a flooded MCO are examined to determine the impact of aluminum corrosion products (from aluminum basket grids and spacers) on bound water estimates and subsequent fuel/environment reactions during storage. The mass and impact of corrosion products were determined to be insignificant, validating the choice of aluminum as an MCO component and confirming expectations that no changes to the Technical Databook or particulate mass or water content are necessary.

  11. The Use of One-Sample Prediction Intervals for Estimating CO2 Scrubber Canister Durations

    Science.gov (United States)

    2012-10-01

    pp. 299–301. 5. R. L. Scheaffer and J. T. McClave, Probability and Statistics, 3rd ed. (Boston, MA: PWS-Kent, 1990), pp. 291–292. 6. G. G...1990), p. 82. 2. Ibid., p. 206. 14 B-1 Appendix B: Textbook Examples 1. Scheaffer and McClave (1990), p. 292. Ten independent observations...next observation will lie between 16.076 and 16.124. 2. Scheaffer and McClave (1990), p. 293. A particular subcompact automobile has been tested

  12. Canister storage building compliance assessment DOE Order 6430.1A, General Design Criteria

    Energy Technology Data Exchange (ETDEWEB)

    BLACK, D.M.

    1999-08-12

    This document presents the Project's position on compliance with DOE Order 6430.1A ''General Design Criteria.'' No non-compliances are shown. The compliance statements have been reviewed and approved by DOE. Open items are scheduled to be closed prior to project completion.

  13. 40 CFR 86.153-98 - Vehicle and canister preconditioning; refueling test.

    Science.gov (United States)

    2010-07-01

    ... part). Fifteen seconds after the engine starts, place the transmission in gear. Twenty seconds after... controlled to 50±25 grains of water vapor per pound of dry air) maintained at a nominal flow rate of 0.8 cfm... vapor per pound of dry air) maintained at a nominal flow rate of 0.8 cfm. In this case, the...

  14. Canister storage building compliance assessment SNF project NRC equivalency criteria - HNF-SD-SNF-DB-003

    Energy Technology Data Exchange (ETDEWEB)

    BLACK, D.M.

    1999-08-11

    This document presents the Project's position on compliance with the SNF Project NRC Equivalency Criteria--HNF-SD-SNF-DE-003, Spent Nuclear Fuel Project Path Forward Additional NRC Requirements. No non-compliances are shown The compliance statements have been reviewed and approved by DOE. Open items are scheduled to be closed prior to project completion.

  15. Stress corrosion cracking in canistered waste package containers: Welds and base metals

    Energy Technology Data Exchange (ETDEWEB)

    Huang, J.S.

    1998-03-01

    The current design of waste package containers include outer barrier using corrosion allowable material (CAM) such as A516 carbon steel and inner barrier of corrosion resistant material (CRM) such as alloy 625 and C22. There is concern whether stress corrosion cracking would occur at welds or base metals. The current memo documents the results of our analysis on this topic.

  16. A REVIEW OF THE US EPA'S SINGLE BREATH CANISTER (SBC) METHOD FOR EXHALED VOLATILE ORGANIC BIOMARKERS

    Science.gov (United States)

    Exhaled alveolar breath can provide a great deal of information about an individual?s health and previous exposure to potentially harmful xenobiotic materials. Because breath can be obtained noninvasively and its constituents directly reflect concentrations in the blood, its us...

  17. Developing a structural health monitoring system for nuclear dry cask storage canister

    Science.gov (United States)

    Sun, Xiaoyi; Lin, Bin; Bao, Jingjing; Giurgiutiu, Victor; Knight, Travis; Lam, Poh-Sang; Yu, Lingyu

    2015-03-01

    Interim storage of spent nuclear fuel from reactor sites has gained additional importance and urgency for resolving waste-management-related technical issues. In total, there are over 1482 dry cask storage system (DCSS) in use at US plants, storing 57,807 fuel assemblies. Nondestructive material condition monitoring is in urgent need and must be integrated into the fuel cycle to quantify the "state of health", and more importantly, to guarantee the safe operation of radioactive waste storage systems (RWSS) during their extended usage period. A state-of-the-art nuclear structural health monitoring (N-SHM) system based on in-situ sensing technologies that monitor material degradation and aging for nuclear spent fuel DCSS and similar structures is being developed. The N-SHM technology uses permanently installed low-profile piezoelectric wafer sensors to perform long-term health monitoring by strategically using a combined impedance (EMIS), acoustic emission (AE), and guided ultrasonic wave (GUW) approach, called "multimode sensing", which is conducted by the same network of installed sensors activated in a variety of ways. The system will detect AE events resulting from crack (case for study in this project) and evaluate the damage evolution; when significant AE is detected, the sensor network will switch to the GUW mode to perform damage localization, and quantification as well as probe "hot spots" that are prone to damage for material degradation evaluation using EMIS approach. The N-SHM is expected to eventually provide a systematic methodology for assessing and monitoring nuclear waste storage systems without incurring human radiation exposure.

  18. Status report. Characterization of Weld Residual Stresses on a Full-Diameter SNF Interim Storage Canister Mockup.

    Energy Technology Data Exchange (ETDEWEB)

    Enos, David [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Bryan, Charles R. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

    2015-08-01

    This report documents the mockup specifications and manufacturing processes; the initial cutting of the mockup into three cylindrical pieces for testing and the measured strain changes that occurred during the cutting process; and the planned weld residual stress characterization activities and the status of those activities.

  19. Inspection of copper canisters for spent nuclear fuel by means of ultrasound. Ultrasonic imaging of EB weld, theory of harmonic imaging of welds, NDE of cast iron

    Energy Technology Data Exchange (ETDEWEB)

    Stepinski, T.; Lingvall, F.; Ping Wu [Uppsala Univ. (Sweden). Dept. of Materials Science

    2001-07-01

    The objective of task presented in the first chapter, ultrasonic imaging of EB weld is to investigate imaging methods capable of improving ultrasonic imaging of defects in EB-welds. Algorithms based on ideas from ultrasonic tomography were examined as the first step. After a concise review of literature in the field of tomography the attention is focused on synthetic focusing and particularly on using linear phased array systems for imaging. Synthetic focusing is a technique where the focusing is performed by software after gathering the ultrasonic data. General principles of synthetic aperture focusing technique (SAFT) - a synthetic focusing technique especially suitable for linear ultrasonic arrays are presented. Problems related to the application of SAFT to ultrasonic transducers with large apertures are identified and the solution is proposed. It appears that when the probe becomes larger (i.e., cannot be regarded as a point source) the ultrasonic pulses that it generates will be smeared by its spatial impulse response (SIR). This impairs the spatial resolution achieved for the finite aperture probes comparing to the point source. Thus, a proper application of synthetic focusing requires taking into account the spatially varying probe's SIR. The SIR has to be calculated (measured) in the interesting points of space and than deconvoluted. A technique for deconvoluting the SIR based on Wiener filter is proposed and illustrated by experimental results. Some preliminary results from immersion testing of copper blocks using the ALLIN system in our lab facility are presented. Nonlinear propagation of plane waves in fluids based on the Burgers equation is investigated in the second chapter. The presented method is basically adopted from the existing literature although some modification has been made to adapt to our situation. The solution has been re-derived and two alternative forms feasible for computer calculation are given and some numerical results are presented. The calculated results show how the harmonics evolve as the plane wave propagates. It should be noted that the work presented here is at its preliminary stage, the goal of the present and future work is to build a simulating tool for material harmonic imaging technology. The theory of phase conjugation is presented and different methods of wave phase conjugation (WPC) are reviewed and characterized in the third chapter. The ability of WPC to self-adaptive focus ultrasonic waves in inhomogeneous media makes it interesting in the application to the inspection of as EB welds. The WPC can be performed either in time or frequency domain. Time domain method, known as time reversal mirrors is reviewed in some detail with focus on its applications to NDT. Frequency domain techniques use nonlinear piezoelectric or magnetic materials. The choice of magneto-acoustic phase conjugation, performed in nonlinear magnetic ceramics as a candidate for the feasibility demonstration is motivated. Details of the preliminary experiment with high frequency NDE application (10 MHz) are presented. NDE methods suitable for the characterization of cast iron are reviewed in the fourth chapter. Two groups of methods that could be used in an industrial environment, those based on ultrasound and on eddy current measurement are presented in some detail. The review is focused on sensing the interaction of elastic waves with the microstructure of cast iron. It is explained how three different features of ultrasound, the sound velocity, the attenuation and the backscattering, can be used for the characterization.

  20. Waste disposal package

    Science.gov (United States)

    Smith, M.J.

    1985-06-19

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

  1. Corrosion resistant storage container for radioactive material

    Science.gov (United States)

    Schweitzer, Donald G.; Davis, Mary S.

    1990-01-01

    A corrosion resistant long-term storage container for isolating radioactive waste material in a repository. The container is formed of a plurality of sealed corrosion resistant canisters of different relative sizes, with the smaller canisters housed within the larger canisters, and with spacer means disposed between judxtaposed pairs of canisters to maintain a predetermined spacing between each of the canisters. The combination of the plural surfaces of the canisters and the associated spacer means is effective to make the container capable of resisting corrosion, and thereby of preventing waste material from leaking from the innermost canister into the ambient atmosphere.

  2. Determination of VOCs in In-door Smoking Air by GC/MS with Canister Sampling%SUMMA罐采样-GC/MS法测定吸烟室内空气中挥发性有机物

    Institute of Scientific and Technical Information of China (English)

    杨丽莉; 王美飞; 胡恩宇

    2011-01-01

    采用空气预浓缩与气相色谱/质谱联用技术对空气中59种痕量挥发性有机化合物进行定性与定量分析,应用研究的技术对吸烟室烟草空气中的挥发性有机物成分定性解析,对59种常见挥发性有机污染物定量检测.室内环境烟草空气中检出多种挥发性有机污染物,主要有烯烃、烷烃、苯系物等有害成分,不仅对被动吸烟人群造成危害,同时也影响大气环境质量.%A determination method of 59 volatile organic compounds ( VOCs) in ambient air by air pre-con-centration and gas chromatography-mass spectrometry has been studied. VOCs in air of smoking room was qualitatively analyzed and 59 VOCs was quantitatively detected. Some VOCs in the air were hazardous pollutants such as olefins, alkanes, and aromatic hydrocarbons. These compounds not only harmed to passive smoking people but also affected the atmospheric environmental quality.

  3. 同心筒发射装置导弹燃气流热效应数值模拟%Numerical Simulation of Thermal Effect of Missile Combustion-gas Flow for Concentric Canister Launcher

    Institute of Scientific and Technical Information of China (English)

    蔺翠郎; 毕世华

    2008-01-01

    针对某型舰载导弹同心筒发射装置,采用动态网格技术,建立了导弹发射过程中燃气流与发射筒之间的流固耦合换热模型.模型中考虑了导弹在发射筒内运动引起的燃气流场边界的变化.通过对燃气流非稳态传热的数值模拟,得到了发射筒内燃气流及筒壁的温度分布规律,为发射筒的热强度设计提供依据.

  4. 微重力下高温固液相变蓄热容器内空穴分布%INITIAL STUDY OF VOID FORMATION OF HIGH TEMPERATURE SOLID-LIQUID PHASE CHANGE THERMAL ENERGY STORAGE CANISTER IN MICROGRAVITY

    Institute of Scientific and Technical Information of China (English)

    邢玉明; 崔海亭; 袁修干; 粟卫芳

    2003-01-01

    高温固液相变蓄热容器是空间太阳能动力装置吸热-储热器的关键部件.作为相变材料(PCM)的氟盐在凝固时体积收缩很大,从而在PCM容器内形成空穴.空穴的存在增大了传热热阻,还可能使PCM容器产生"热斑"和"热松脱"现象.该文建立了微重力下基于焓法形式的二维数学模型和一个改进的空穴模型,提出了计算相变过程中空穴体积变化及空穴调整的算法.预测了PCM容器在一个轨道周期内的空穴分布.计算结果有助于解决PCM容器的"热斑"和"热松脱"问题.

  5. Thermal Effects of Structural Parameters of PCM Canister on Heat Receiver under a Certain Working Condition%PCM容器的结构参数对蓄热单元热性能的影响

    Institute of Scientific and Technical Information of China (English)

    应歌; 杜朝辉; 王平阳; 钱中

    2006-01-01

    适合空间设备的主要电力来源就是太阳能热动力发电系统.为了深入考察蓄热容器(PCM容器)的结构参数对空间太阳能热动力发电系统的关键部件之一,即吸热蓄热器热性能的影响,建立了PCM容器的二维热分析模型,并在两种工作参数条件下对不同径向高度的PCM容器进行了数值计算,结果表明PCM容器的外径对吸热蓄热器热性能具有重要影响.研究结果可为提高PCM容器的功率质量比提供参考依据.

  6. Alternative technical summary report for immobilized disposition in deep boreholes: Immobilized disposal of plutonium in coated ceramic pellets in grout without canisters, Version 4.0. Fissile materials disposition program

    Energy Technology Data Exchange (ETDEWEB)

    Wijesinghe, A.M.

    1996-08-23

    This paper summarizes and compares the immobilized and direct borehole disposition alternatives previously presented in the alternative technical summary. The important design concepts, facility features and operational procedures are first briefly described. This is followed by a discussion of the issues that affect the evaluation of each alternative against the programmatic assessment criteria that have been established for selecting the preferred alternatives for plutonium disposition.

  7. Fissile Material Disposition Program: Deep Borehole Disposal Facility PEIS data input report for direct disposal. Direct disposal of plutonium metal/plutonium dioxide in compound metal canisters. Version 3.0

    Energy Technology Data Exchange (ETDEWEB)

    Wijesinghe, A.M.; Shaffer, R.J.

    1996-01-15

    The US Department of Energy (DOE) is examining options for disposing of excess weapons-usable nuclear materials [principally plutonium (Pu) and highly enriched uranium (HEU)] in a form or condition that is substantially and inherently more difficult to recover and reuse in weapons production. This report is the data input report for the Programmatic Environmental Impact Statement (PEIS). The PEIS examines the environmental, safety, and health impacts of implementing each disposition alternative on land use, facility operations, and site infrastructure; air quality and noise; water, geology, and soils; biotic, cultural, and paleontological resources; socioeconomics; human health; normal operations and facility accidents; waste management; and transportation. This data report is prepared to assist in estimating the environmental effects associated with the construction and operation of a Deep Borehole Disposal Facility, an alternative currently included in the PEIS. The facility projects under consideration are, not site specific. This report therefore concentrates on environmental, safety, and health impacts at a generic site appropriate for siting a Deep Borehole Disposal Facility.

  8. Fissile Material Disposition Program: Deep borehole disposal Facility PEIS date input report for immobilized disposal. Immobilized disposal of plutonium in coated ceramic pellets in grout with canisters. Version 3.0

    Energy Technology Data Exchange (ETDEWEB)

    Wijesinghe, A.M.; Shaffer, R.J.

    1996-01-15

    Following President Clinton`s Non-Proliferation Initiative, launched in September, 1993, an Interagency Working Group (IWG) was established to conduct a comprehensive review of the options for the disposition of weapons-usable fissile materials from nuclear weapons dismantlement activities in the United States and the former Soviet Union. The IWG review process will consider technical, nonproliferation, environmental budgetary, and economic considerations in the disposal of plutonium. The IWG is co-chaired by the White House Office of Science and Technology Policy and the National Security Council. The Department of Energy (DOE) is directly responsible for the management, storage, and disposition of all weapons-usable fissile material. The Department of Energy has been directed to prepare a comprehensive review of long-term options for Surplus Fissile Material (SFM) disposition, taking into account technical, nonproliferation, environmental, budgetary, and economic considerations.

  9. Development of the Virtual Instrument for the Active Charcoal Canister Used in Vehicle with Petrol Engine%汽油车用活性炭罐性能测试系统的虚拟仪器

    Institute of Scientific and Technical Information of China (English)

    邴冰; 刘长良; 许善珍; 耿松亮

    2007-01-01

    为了更好地执行国家标准HBC 32-2004《环境保护产品认定技术要求汽油车燃油蒸发污染物控制系统》,利用虚拟仪器开发平台LabVIEW7.1研制一套汽油车用活性炭罐性能测试系统,实现了系统的自动测控.

  10. Evaluation of design and operation basis of the smear test station

    Energy Technology Data Exchange (ETDEWEB)

    Hutsell, D.J.

    2000-02-29

    The purpose of the WTC-STS is to provide final verification that the external canister surface is free of transferable contamination before transporting the canister to the Glass Waste Storage Building for onsite storage.

  11. Storage, transportation and disposal system for used nuclear fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Scaglione, John M.; Wagner, John C.

    2017-01-10

    An integrated storage, transportation and disposal system for used fuel assemblies is provided. The system includes a plurality of sealed canisters and a cask sized to receive the sealed canisters in side by side relationship. The plurality of sealed canisters include an internal basket structure to receive a plurality of used fuel assemblies. The internal basket structure includes a plurality of radiation-absorbing panels and a plurality of hemispherical ribs generally perpendicular to the canister sidewall. The sealed canisters are received within the cask for storage and transportation and are removed from the cask for disposal at a designated repository. The system of the present invention allows the handling of sealed canisters separately or collectively, while allowing storage and transportation of high burnup fuel and damaged fuel to the designated repository.

  12. Assaying Benefits of Poly[styrene-4-(trimethylammonium)methyl Triiodide] in Respiratory Protection Devices

    Science.gov (United States)

    2009-12-01

    Bronze—a clip-on device containing a PSTI-coated nonwoven medium that attached to the front face of a commercial off-the-shelf (COTS) canister...Silver—a fabricated plastic canister that placed a PSTI-coated nonwoven layer in front of the components of the COTS canister) • Gold— a redesigned...humidity. Accommodation of the mouse model was a necessary aspect of the design and construction processes. This system will be used in the

  13. Visual examinations of K west fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Pitner, A.L., Fluor Daniel Hanford

    1997-02-03

    Over 250 fuel assemblies stored in sealed canisters in the K West Basin were extracted and visually examined for damage. Substantial damage was expected based on high cesium levels previously measured in water samples taken from these canisters. About 11% of the inner elements and 45% of the outer elements were found to be failed in these examinations. Canisters that had cesium levels of I curie or more generally had multiple instances of major fuel damage.

  14. In-Place Filter Tester Instrument for Nuclear Material Containers.

    Science.gov (United States)

    Brown, Austin D; Moore, Murray E; Runnels, Joel T; Reeves, Kirk

    2016-05-01

    A portable instrument was developed to determine filter clogging and container leakage of in-place nuclear material storage canisters. This paper describes the development of an in-place filter tester for determining the "as found" condition of unopened canisters. The U.S. Department of Energy uses several thousand canisters for nuclear material storage, and air filters in the canister lids allow gases to escape while maintaining an equilibrated pressure without release of radioactive contamination. Diagnosing the filter condition and canister integrity is important for ensuring worker and public safety. Customized canister interfaces were developed for suction clamping (during tests) to two of the canister types in use at Los Alamos National Laboratory. Experimental leakage scenarios included: O-rings fouled with dust, cracked O-rings, and loose canister lids. The prototype tester has a measurement range for air leakage rates from 8.2 × 10 mL s up to 3.0 × 10 mL s. This is sufficient to measure a leak rate of 3.4 × 10 mL s, which is the Los Alamos helium leak criterion for post-drop tested canisters. The In-Place-Filter-Tester cannot measure to the lower value of the helium leak criterion for pre-drop tested canisters (1.0 × 10 mL s). However, helium leak testing requires canister disassembly, while the new in-place filter tester is able to assess the assembled condition of as-found and in-situ canisters.

  15. 327 SNF fuel return to K-Basin quality process plan

    Energy Technology Data Exchange (ETDEWEB)

    Ham, J.E.

    1998-09-22

    The B and W Hanford Company`s (BWHC) 327 Facility, in the 300 Area of the Hanford Site, contains Spent Nuclear Fuel (SNF) single fuel element canisters (SFEC) and fuel remnant canisters (FRC) which are to be returned to K-Basin. Seven shipments of up to six fuel canisters will be loaded into the CNS 1-13G Cask and transported to 105-KE.

  16. Flow Visualization of Forced and Natural Convection in Internal Cavities

    Energy Technology Data Exchange (ETDEWEB)

    John Crepeau; Hugh M. Mcllroy,Jr.; Donald M. McEligot; Keith G. Condie; Glenn McCreery; Randy Clarsean; Robert S. Brodkey; Yann G. Guezennec

    2002-01-31

    The report descries innovative flow visualization techniques, fluid mechanics measurements and computational models of flows in a spent nuclear fuel canister. The flow visualization methods used a fluid that reacted with a metal plate to show how a local reaction affects the surrounding flow. A matched index of refraction facility was used to take mean flow and turbulence measurements within a generic spent nuclear fuel canister. Computational models were also made of the flow in the canister. It was determined that the flow field in the canister was very complex, and modifications may need to be made to ensure that the spent fuel elements are completely passivated.

  17. Visual examinations of K east fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Pitner, A.L., Fluor Daniel Hanford

    1997-02-03

    Selected fuel elements stored in both ``good fuel`` and ``bad fuel`` canisters in K East Basin were extracted and visually examined full length for damage. Lower end damage in the ``bad fuel`` canisters was found to be more severe than expected based on top end appearances. Lower end damage for the ``good fuel`` canisters, however, was less than expected based on top end observations. Since about half of the fuel in K East Basin is contained in ``good fuel`` canisters based on top end assessments, the fraction of fuel projected to be intact with respect to IPS processing considerations remains at 50% based on these examination results.

  18. Basalt near-surface test facility test plans

    Energy Technology Data Exchange (ETDEWEB)

    Krug, A.D.

    1979-06-01

    The NSTF is under construction at Gable Mountain for in-situ testing, which will be conducted in two phases: Phase I, using electric heaters to simulate nuclear waste canisters in order to study the thermomechanical response of basalt; and Phase II, using spent fuel canisters. The tests planned for Phases I and II are described. (DLC)

  19. Observations during the first K West fuel shipping campaign

    Energy Technology Data Exchange (ETDEWEB)

    Makenas, B.J.

    1995-11-01

    Three fuel elements were shipped to the 300 Area hotcells during the first characterization shipping campaign from K West Basin. This document summarizes observations made during this campaign including the gas, liquid, and sludge content of the observed canisters. Included in an appendix is a detailed evaluation of fuel element condition for each canister opened

  20. System Specification for Immobilized High-Level Waste Interim Storage

    Energy Technology Data Exchange (ETDEWEB)

    CALMUS, R.B.

    2000-12-27

    This specification establishes the system-level functional, performance, design, interface, and test requirements for Phase 1 of the IHLW Interim Storage System, located at the Hanford Site in Washington State. The IHLW canisters will be produced at the Hanford Site by a Selected DOE contractor. Subsequent to storage the canisters will be shipped to a federal geologic repository.

  1. Assessment of alternative disposal concepts

    Energy Technology Data Exchange (ETDEWEB)

    Autio, J.; Saanio, T.; Tolppanen, P. [Saanio and Riekkola Consulting Engineers, Helsinki (Finland); Raiko, H.; Vieno, T. [VTT Energy, Espoo (Finland); Salo, J.P. [Posiva Oy, Helsinki (Finland)

    1996-12-01

    Four alternative repository designs for the disposal of spent nuclear in the Finnish crystalline bedrock were assessed in the study. The alternatives were: (1) the basic KBS-3 design in which copper canisters are emplaced in vertical deposition holes bored in the floors of horizontal tunnels, (2) the KBS-3-2C design with two canisters in a deposition hole, (3) Short Horizontal Holes (SHH) in the side walls of the tunnels, and (4) the Medium Long Holes (MLH) concept in which approximately 25 canisters are emplaced in a horizontal deposition hole about 200 metres in length bored between central and side tunnels. In all the alternatives considered, the thickness of the layer of compacted bentonite between copper canister and bedrock is 35 cm. Two different copper canister designs were also assessed. Technical feasibility and flexibility, post-closure safety and repository cost were assessed for each of the alternative canister and repository designs. On the basis of this assessment it is recommended that further development and studies should focus on the vacuum- or inert gas-filled cast insert type copper canister and the basic KBS-3 type repository design with a single canister in a vertical deposition hole. The KBS-3 design is robust and flexible and provides excellent post-closure safety. The transfer, emplacement and sealing operations are technically uncomplicated. The alternative options assessed do not offer any significant benefits in safety or cost over the basic design, but they are technically more complex and also in some respects more vulnerable to malfunction during the emplacement of canisters and buffer, as well as common mode failures. (60 refs.).

  2. Effects of brine migration on waste storage systems. Final report. [Thermomechanical effects

    Energy Technology Data Exchange (ETDEWEB)

    Gaffney, E.S.; Nickell, R.E.

    1979-05-15

    Processes which can lead to mobilization of brine adjacent to spent fuel or nuclear waste canisters and some of the thermomechanical consequences have been investigated. Velocities as high as 4 x 10/sup -7/ m s/sup -1/ (13 m y/sup -1/) are calculated at the salt/canister boundary. As much as 40 liters of pure NaCl brine could accumulate around each canister during a 10-year storage period. Accumulations of bittern brines would probably be less, in the range of 2 to 5 liters. With 0.5% water, NaCl brine accumulation over a 10-year storage cycle around a spent fuel canister producing 0.6 kW of heat is expected to be less than 1 liter for centimeter-size inclusions and less than 0.5 liter for millimeter-size inclusions. For bittern brines, about 25 years would be required to accumulate 0.4 liter. The most serious mechanical consequence of brine migration would be the increased mobility of the waste canister due to pressure solution. In pressure solution enhanced deformation, the existence of a thin film of fluid either between grains or between media (such as between a canister and the salt) provides a pathway by which the salt can be redistributed leading to a marked increase in strain rates in wet rock relative to dry rock. In salt, intergranular water will probably form discontinuous layers rather than films so that they would dominate pressure solution. A mathematical model of pressure solution indicates that pressure solution will not lead to appreciable canister motions except possibly in fine grained rocks (less than 10/sup -4/ m). In fine grained salts, details of the contact surface between the canister and the salt bed may lead to large pressure solution motions. A numerical model indicates that heat transfer in the brine layer surrounding a spent fuel canister is not conduction dominated but has a significant convective component.

  3. Hydride heat pump with heat regenerator

    Science.gov (United States)

    Jones, Jack A. (Inventor)

    1991-01-01

    A regenerative hydride heat pump process and system is provided which can regenerate a high percentage of the sensible heat of the system. A series of at least four canisters containing a lower temperature performing hydride and a series of at least four canisters containing a higher temperature performing hydride is provided. Each canister contains a heat conductive passageway through which a heat transfer fluid is circulated so that sensible heat is regenerated. The process and system are useful for air conditioning rooms, providing room heat in the winter or for hot water heating throughout the year, and, in general, for pumping heat from a lower temperature to a higher temperature.

  4. Novel Long-Term CO2 Removal System Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Current Technology for CO2 removal from enclosed air of spacecraft utilizes LiOH canisters for CO2 absorption. This absorption is irreversible so longer flights...

  5. CO2 Removal from Mars EMU Project

    Data.gov (United States)

    National Aeronautics and Space Administration — CO2 control for during ExtraVehicular Activity (EVA) on mars is challenging. Lithium hydroxide (LiOH) canisters have impractical logistics penalties, and regenerable...

  6. 75 FR 12315 - Pacific Gas and Electric Company; Diablo Canyon Independent Spent Fuel Storage Installation...

    Science.gov (United States)

    2010-03-15

    ... MPC-32 canisters and, to allow linear interpolation for some enrichments consistent with the Holtec... conditions in the annular gap between the MPC and the transfer cask depending on which drying process is...

  7. Numerical analysis of thermal process in the near field around vertical disposal of high-level radioactive waste

    Institute of Scientific and Technical Information of China (English)

    H.G. Zhao; H. Shao; H. Kunz; J. Wang; R. Su; Y.M. Liu

    2014-01-01

    For deep geological disposal of high-level radioactive waste (HLW) in granite, the temperature on the HLW canisters is commonly designed to be lower than 100◦C. This criterion dictates the dimension of the repository. Based on the concept of HLW disposal in vertical boreholes, thermal process in the near field (host rock and buffer) surrounding HLW canisters has been simulated by using different methods. The results are drawn as follows:(a) the initial heat power of HLW canisters is the most important and sensitive parameter for evolution of temperature field;(b) the thermal properties and variations of the host rock, the engineered buffer, and possible gaps between canister and buffer and host rock are the additional key factors governing the heat transformation;(c) the gaps width and the filling by water or air determine the temperature offsets between them.

  8. Conceptual design criteria for facilities for geologic disposal of radioactive wastes in salt formations

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    The facility design requirements and criteria discussed are: general codes, standards, specifications, and regulations; site criteria; land improvements criteria, low-level waste facility criteria; canistered waste facility criteria; support facilities criteria; and utilities and services criteria. (LK)

  9. Study of a Fractured Full Scale Inactive R7T7 Type Glass

    Institute of Scientific and Technical Information of China (English)

    GIN; Stephane

    2008-01-01

    <正>The nuclear glass block is fractured after it is poured into the canister. Aqueous alteration of glass involves essentially surface mechanisms; hence it is great importance to determine the surface area of the

  10. 42 CFR 84.1131 - Respirators; required components.

    Science.gov (United States)

    2010-10-01

    ... SAFETY AND HEALTH RESEARCH AND RELATED ACTIVITIES APPROVAL OF RESPIRATORY PROTECTIVE DEVICES Dust, Fume... noseclip, hood, or helmet; (2) Filter unit, canister with filter, or cartridge with filter; (3) Harness;...

  11. Numerical analysis of thermal process in the near field around vertical disposal of high-level radioactive waste

    Directory of Open Access Journals (Sweden)

    H.G. Zhao

    2014-02-01

    Full Text Available For deep geological disposal of high-level radioactive waste (HLW in granite, the temperature on the HLW canisters is commonly designed to be lower than 100 °C. This criterion dictates the dimension of the repository. Based on the concept of HLW disposal in vertical boreholes, thermal process in the near field (host rock and buffer surrounding HLW canisters has been simulated by using different methods. The results are drawn as follows: (a the initial heat power of HLW canisters is the most important and sensitive parameter for evolution of temperature field; (b the thermal properties and variations of the host rock, the engineered buffer, and possible gaps between canister and buffer and host rock are the additional key factors governing the heat transformation; (c the gaps width and the filling by water or air determine the temperature offsets between them.

  12. Environmental permits and approvals plan for high-level waste interim storage, Project W-464

    Energy Technology Data Exchange (ETDEWEB)

    Deffenbaugh, M.L.

    1998-05-28

    This report discusses the Permitting Plan regarding NEPA, SEPA, RCRA, and other regulatory standards and alternatives, for planning the environmental permitting of the Canister Storage Building, Project W-464.

  13. Reactive Rendezvous and Docking Sequencer Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Mars Sample Return poses some of the most challenging operational activities of any NASA deep space mission. Rendezvous of a vehicle with a sample canister in order...

  14. The Subject Headings of the Morris Swett Library, USAFAS. Revised.

    Science.gov (United States)

    1980-05-15

    Rifle. Ammunition, Beacon. See Iluminants - Beacon shells. AMMUNITION, BIOLOGICAL. AMMUNITION, BLANK. x Hlank armunition. AMMUNITION, CANISTER. (UL 400...illumination x Star shells - Beacon shells x Ammunition, Beacon - Candles - Flares - Flares, Aeroplanes x Aeroplanes - Flares - Iluminating shells (UL 440.1

  15. 50 CFR 32.28 - Florida.

    Science.gov (United States)

    2010-10-01

    ... report card and place it in an entrance fee canister each day prior to exiting the refuge. 12. All youth... after the Florida State Central Management Zone General Gun (antlered deer and wild hog) season...

  16. Research program to study the gamma radiation effects in Spanish bentonites; Programa de investigacion para estudiar los efectos de la radiacion gamma en bentonitas calcicas espanolas

    Energy Technology Data Exchange (ETDEWEB)

    Dies, J.; Tarrasa, F. [Universidad Politecnica de Catalunya (Spain); Cuevas de las, C.; Miralles, L.; Pueyo, J. J. [Universidad de Barcelona (Spain)

    2000-07-01

    The engineering barrier of a radioactive waste underground disposal facility, placed in a granitic host rock, will consist of a backfill of compacted bentonite blocks. At first, this material will be subjected to a gamma radiation field, from the waste canister, and heat from the spent fuel inside the canister. Moreover, any groundwater that reaches the repository will saturate the bentonite. For these reasons the performance of the engineered barrier must be carefully assessed in laboratory experiments. (Author)

  17. Dry spent fuel storage with the MACSTOR system

    Energy Technology Data Exchange (ETDEWEB)

    Pare, F. [Atomic Energy of Canada Ltd., Montreal, PQ (Canada). CANDU Operations

    1996-10-01

    Atomic Energy of Canada Limited (AECL), and Transnuclear Inc. (TNI) began in 1989 the development of the concrete spent fuel storage system, called MACSTOR (Modular Air-Cooled Canister STORage) for use with LWR spent fuel assemblies. It is a hybrid system which combines the operational economies of metal cask technology with the capital economies of concrete technology. The MACSTOR Module is a monolithic, shielded concrete vault structure that can accommodate up to 20 spent fuel canisters. Each canister typically holds up to 21 PWR or 44 BWR spent fuel assemblies with a nominal fuel burn up rate of 40,000 MWD/MTU and a 7 year minimum cooling period. The structure is passively cooled by natural convection through an array of inlet and outlet gratings and galleries serving a central plenum where the (vertically) stored canisters are located. The canisters are continuously monitored by means of a pressure monitoring system developed by TNI. Thus, the utility can be assured of both positive cooling of the fuel and verification of the integrity of the fuel confinement boundary. The structure is seismically designed and is capable of withstanding site design basis accident events. The MACSTOR system includes the storage module(s), an overhead gantry system for cask handling, a transfer cask for moving fuel from wet to dry storage and a cask transporter. The canister and transfer cask designs are based on Transnuclear transport cask designs and proven hot cell transfer cask technology, adapted to requirements for on-site spent fuel storage. The MACSTOR system can economically address a wide range of storage capacity requirements. The modular concept allows for flexibility in determining each module`s capacity. Starting with 8 canisters, the capacity can be increased by increments of 4 up to 20 canisters. The MACSTOR system is also flexible in accommodating the various spent fuel types from such reactors as VVER-440, VVER-1000 and RBMK 1500. (J.P.N.)

  18. MACSTOR{trademark}: Dry spent fuel storage for the nuclear power industry

    Energy Technology Data Exchange (ETDEWEB)

    Pare, F.E.; Pattantyus, P. [AECL Candu, Montreal, Quebec (Canada); Hanson, A.S. [Transnuclear, Inc., Hawthorne, NY (United States)

    1993-12-31

    Safe storage of spent fuel has long been an area of critical concern for the nuclear power industry. As fuel pools fill up and re-racking possibilities become exhausted, power plant operators will find that they must ship spent fuel assemblies off-site or develop new on-site storage options. Many utility companies are turning to dry storage for their spent fuel assemblies. The MACSTOR (Modular Air-cooled Canister STORage) concept was developed with this in mind. Derived from AECL`s successful vertical loading, concrete silo program for storing CANDU nuclear spent fuel, MACSTOR was developed for light water reactor spent fuel and was subjected to full scale thermal testing. The MACSTOR Module is a monolithic, shielded concrete vault structure than can accommodate up to 24 spent fuel canisters. Each canister holds 12 PWR or 32 PWR previously cooled spent fuel assemblies with burn-up rates as high as 45,000 MWD/MTU. The structure is passively cooled by natural convection through an array of inlet and outlet gratings and galleries serving a central plenum where the (vertically) stored canisters are located. The canisters are continuously monitored by means of a pressure monitoring system developed by TNI. The MACSTOR system includes the storage module(s), an overhead gantry system for cask handling, a transfer cask for moving fuel from wet to dry storage and a cask transporter. The canister and transfer cask designs are based on Transnuclear transport cask designs and proven hot cell transfer cask technology, adapted to requirements for on-site spent fuel storage. This Modular Air Cooled System has a number of inherent advantages: efficient use of construction materials and site space; cooling is virtually impossible to impede; has the ability to monitor fuel confinement boundary integrity during storage; the fuel canisters may be used for both storage and transport and canisters utilize a flanged, ASME-III closure system that allows for easy inspection.

  19. General Purpose Satellites: a concept for affordable low earth orbit vehicles

    OpenAIRE

    Boyd, Austin W.; Fuhs, Allen E.

    1997-01-01

    A general purpose satellite has been designed which will be launched from the Space Shuttle using a NASA Get-Away-Special (GAS) canister. The design is based upon the use of a new extended GAS canister and a low profile launch mechanism. The satellite is cylindrical. measuring 19 inches in diameter and 35 inches long. The maximum vehicle weight is 250 pounds, of which 50 pounds is dedicated to user payloads. The remaining 200 pounds encompasses the satellite structure and support ...

  20. Treatment and final disposal of nuclear waste. Programme for encapsulation, deep geological disposal, and research, development and demonstration

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    Programs for RD and D concerning disposal of radioactive waste are presented. Main topics include: Design, testing and manufacture of canisters for the spent fuels; Design of equipment for deposition of waste canisters; Material and process for backfilling rock caverns; Evaluation of accuracy and validation of methods for safety analyses; Development of methods for defining scenarios for the safety analyses. 471 refs, 67 figs, 21 tabs.

  1. Treatment and disposal of radioactive wastes from nuclear power plants. Program for encapsulation, deep geologic deposition and research, development and demonstration; Kaernkraftavfallets behandling och slutfoervaring. Program foer inkapsling, geologisk djupfoervaring samt forskning, utveckling och demonstration

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    Programs for RD and D concerning disposal of radioactive waste are presented. Main topics include: Design, testing and manufacture of canisters for the spent fuels; Design of equipment for deposition of waste canisters; Material and process for backfilling rock caverns; Evaluation of accuracy and validation of methods for safety analyses; Development of methods for defining scenarios for the safety analyses. 471 refs, 67 figs, 21 tabs.

  2. Final report spent nuclear fuel retrieval system primary cleaning development testing

    Energy Technology Data Exchange (ETDEWEB)

    Ketner, G.L.; Meeuwsen, P.V.

    1997-09-01

    Developmental testing of the primary cleaning station for spent nuclear fuel (SNF) and canisters is reported. A primary clean machine will be used to remove the gross sludge from canisters and fuel while maintaining water quality in the downstream process area. To facilitate SNF separation from canisters and minimize the impact to water quality, all canisters will be subjected to mechanical agitation and flushing with the Primary Clean Station. The Primary Clean Station consists of an outer containment box with an internally mounted, perforated wash basket. A single canister containing up to 14 fuel assemblies will be loaded into the wash basket, the confinement box lid closed, and the wash basket rotated for a fixed cycle time. During this cycle, basin water will be flushed through the wash basket and containment box to remove and entrain the sludge and carry it out of the box. Primary cleaning tests were performed to provide information concerning the removal of sludge from the fuel assemblies while in the basin canisters. The testing was also used to determine if additional fuel cleaning is required outside of the fuel canisters. Hydraulic performance and water demand requirements of the cleaning station were also evaluated. Thirty tests are reported in this document. Tests demonstrated that sludge can be dislodged and suspended sufficiently to remove it from the canister. Examination of fuel elements after cleaning suggested that more than 95% of the exposed fuel surfaces were cleaned so that no visual evidence of remained. As a result of testing, recommendations are made for the cleaning cycle. 3 refs., 16 figs., 4 tabs.

  3. Results for the Aboveground Configuration of the Boiling Water Reactor Dry Cask Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-09-30

    The thermal performance of commercial nuclear spent fuel dry storage casks is evaluated through detailed numerical analysis. These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and also by increasing the internal convection through greater canister helium pressure. These same canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and belowground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of aboveground and belowground canistered dry cask systems. The purpose of the current investigation was to produce data sets that can be used to test the validity of the assumptions associated with the calculations used to determine steady-state cladding temperatures in modern dry casks that utilize elevated helium pressure in the sealed canister in an aboveground configuration.

  4. Analysis of SKB MiniCan - Experiment 3

    Energy Technology Data Exchange (ETDEWEB)

    Smart, Nick; Rance, Andy; Reddy, Bharti; Fennell, Paul [AMEC (UK); Winsley, Robert [NDA (Country unknown/Code not available)

    2012-11-15

    To ensure the safe encapsulation of spent nuclear fuel rods for geological disposal, SKB of Sweden is planning to use a system that consists of an outer copper canister and a cast iron insert (the KBS-3 concept). In 2007 Serco1 completed the set up of five model canister experiments at SKB's Aespoe laboratory and monitoring has continued since. The original aim of the model canister experiments was to examine how corrosion of the cast iron insert inside a copper canister would evolve with time, if water ingress through a small defect in the copper canister were to occur. Serco arranged manufacture and installation of five miniature copper canisters containing cast iron inserts, with 1 mm defects deliberately machined into the copper shell. The experiments use five small-scale model canisters (300 mm long x 150 mm diameter) that simulate the main features of the SKB canister design (hence the project name, 'MiniCan'). The main aim of the work is to examine how corrosion of the cast iron insert will evolve if a leak is present in the outer copper canister The experiments also included electrochemical equipment to monitor the corrosion behaviour of the model canisters in situ. In 2011 one of the experiments, Experiment 3, was removed f analysis. This report presents details of the procedures that were applied and the findings that were obtained from the analysis that was carried out on Experiment 3. To minimise exposure to air and to keep the contents of the experiment wet until the analysis was carried out, Experiment 3 was extracted from its borehole in August 2011 directly into a transfer tank that was filled with deaerated groundwater and placed in a purpose-built, water-filled and deoxygenated transfer flask. The transfer flask was then transported to the UK for dismantling and examination in a purpose-built anoxic glovebox that contained the appropriate lifting and cutting equipment for handling and sectioning the copper canister and the cast iron

  5. Radiological consequences of accidents during disposal of spent nuclear fuel in a deep borehole

    Energy Technology Data Exchange (ETDEWEB)

    Grundfelt, Bertil [Kemakta Konsult AB, Stockholm (Sweden)

    2013-07-15

    In this report, an analysis of the radiological consequences of potential accidents during disposal of spent nuclear fuel in deep boreholes is presented. The results presented should be seen as coarse estimates of possible radiological consequences of a canister being stuck in a borehole during disposal rather than being the results of a full safety analysis. In the concept for deep borehole disposal of spent nuclear fuel developed by Sandia National Laboratories, the fuel is assumed to be encapsulated in mild steel canisters and stacked between 3 and 5 km depth in boreholes that are cased with perforated mild steel casing tubes. The canisters are joined together by couplings to form strings of 40 canisters and lowered into the borehole. When a canister string has been emplaced in the borehole, a bridge plug is installed above the string and a 10 metres long concrete plug is cast on top of the bridge plug creating a floor for the disposal of the next sting. In total 10 canister strings, in all 400 canisters, are assumed to be disposed of at between 3 and 5 kilometres depth in one borehole. An analysis of potential accidents during the disposal operations shows that the potentially worst accident would be that a canister string is stuck above the disposal zone of a borehole and cannot be retrieved. In such a case, the borehole may have to be sealed in the best possible way and abandoned. The consequences of this could be that one or more leaking canisters are stuck in a borehole section with mobile groundwater. In the case of a leaking canister being stuck in a borehole section with mobile groundwater, the potential radiological consequences are likely to be dominated by the release of the so-called Instant Release Fraction (IRF) of the radionuclide inventory, i.e. the fraction of the radionuclides that as a consequence of the in-core conditions are present in the annulus between the fuel pellets and the cladding or on the grain boundaries of the UO{sub 2} matrix

  6. SR 97. Alternative models project. Stochastic continuum modelling of Aberg

    Energy Technology Data Exchange (ETDEWEB)

    Widen, H. [Kemakta AB, Stockholm (Sweden); Walker, D. [INTERA KB/DE and S (Sweden)

    1999-08-01

    As part of studies into the siting of a deep repository for nuclear waste, Swedish Nuclear Fuel and Waste Management Company (SKB) has commissioned the Alternative Models Project (AMP). The AMP is a comparison of three alternative modelling approaches to bedrock performance assessment for a single hypothetical repository, arbitrarily named Aberg. The Aberg repository will adopt input parameters from the Aespoe Hard Rock Laboratory in southern Sweden. The models are restricted to an explicit domain, boundary conditions and canister location to facilitate the comparison. The boundary conditions are based on the regional groundwater model provided in digital format. This study is the application of HYDRASTAR, a stochastic continuum groundwater flow and transport-modelling program. The study uses 34 realisations of 945 canister locations in the hypothetical repository to evaluate the uncertainty of the advective travel time, canister flux (Darcy velocity at a canister) and F-ratio. Several comparisons of variability are constructed between individual canister locations and individual realisations. For the ensemble of all realisations with all canister locations, the study found a median travel time of 27 years, a median canister flux of 7.1 x 10{sup -4} m/yr and a median F-ratio of 3.3 x 10{sup 5} yr/m. The overall pattern of regional flow is preserved in the site-scale model, as is reflected in flow paths and exit locations. The site-scale model slightly over-predicts the boundary fluxes from the single realisation of the regional model. The explicitly prescribed domain was seen to be slightly restrictive, with 6% of the stream tubes failing to exit the upper surface of the model. Sensitivity analysis and calibration are suggested as possible extensions of the modelling study.

  7. SR 97 - Alternative models project. Discrete fracture network modelling for performance assessment of Aberg

    Energy Technology Data Exchange (ETDEWEB)

    Dershowitz, B.; Eiben, T. [Golder Associates Inc., Seattle (United States); Follin, S.; Andersson, Johan [Golder Grundteknik KB, Stockholm (Sweden)

    1999-08-01

    As part of studies into the siting of a deep repository for nuclear waste, Swedish Nuclear Fuel and Waste Management Company (SKB) has commissioned the Alternative Models Project (AMP). The AMP is a comparison of three alternative modeling approaches for geosphere performance assessment for a single hypothetical site. The hypothetical site, arbitrarily named Aberg is based on parameters from the Aespoe Hard Rock Laboratory in southern Sweden. The Aberg model domain, boundary conditions and canister locations are defined as a common reference case to facilitate comparisons between approaches. This report presents the results of a discrete fracture pathways analysis of the Aberg site, within the context of the SR 97 performance assessment exercise. The Aberg discrete fracture network (DFN) site model is based on consensus Aberg parameters related to the Aespoe HRL site. Discrete fracture pathways are identified from canister locations in a prototype repository design to the surface of the island or to the sea bottom. The discrete fracture pathways analysis presented in this report is used to provide the following parameters for SKB's performance assessment transport codes FARF31 and COMP23: * F-factor: Flow wetted surface normalized with regards to flow rate (yields an appreciation of the contact area available for diffusion and sorption processes) [TL{sup -1}]. * Travel Time: Advective transport time from a canister location to the environmental discharge [T]. * Canister Flux: Darcy flux (flow rate per unit area) past a representative canister location [LT{sup -1}]. In addition to the above, the discrete fracture pathways analysis in this report also provides information about: additional pathway parameters such as pathway length, pathway width, transport aperture, reactive surface area and transmissivity, percentage of canister locations with pathways to the surface discharge, spatial pattern of pathways and pathway discharges, visualization of pathways, and

  8. Test Plan for the Boiling Water Reactor Dry Cask Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-11-01

    The thermal performance of commercial nuclear spent fuel dry storage casks are evaluated through detailed numerical analysis . These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. These same vertical, canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and below-ground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of above-ground and below-ground canistered dry cask systems. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern vertical, canistered dry cask systems. The BWR cask simulator (BCS) has been designed in detail for both the above-ground and below-ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 deg C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the

  9. Technical note: Can the sulfur hexafluoride tracer gas technique be used to accurately measure enteric methane production from ruminally cannulated cattle?

    Science.gov (United States)

    Beauchemin, K A; Coates, T; Farr, B; McGinn, S M

    2012-08-01

    An experiment was conducted to determine whether using ruminally cannulated cattle affects the estimate of enteric methane (CH(4)) emissions when using the sulfur hexafluoride (SF(6)) tracer technique with samples taken from a head canister. Eleven beef cattle were surgically fitted with several types of ruminal cannula (2C, 3C, 3C+washer, 9C; Bar Diamond, Parma, ID). The 2C and 3C models (outer and inner flanges with opposite curvature) had medium to high leakage, whereas the 9C models (outer and inner flanges with the same curvature) provided minimum to moderate leakage of gas. A total of 48 cow-day measurements were conducted. For each animal, a permeation tube containing sulfur hexafluoride (SF(6)) was placed in the rumen, and a sample of air from around the nose and mouth was drawn through tubing into an evacuated canister (head canister). A second sample of air was collected from outside the rumen near the cannula into another canister (cannula canister). Background concentrations were also monitored. The methane (CH(4)) emission was estimated from the daily CH(4) and SF(6) concentrations in the head canister (uncorrected). The permeation SF(6) release rate was then partitioned based on the proportion of the SF(6) concentration measured in the head vs. the cannula canister. The CH(4) emissions at each site were calculated using the two release rates and the two CH(4):SF(6) concentration ratios. The head and cannula emissions were summed to obtain the total emission (corrected). The difference (corrected - uncorrected) in CH4 emission was attributed to the differences in CH(4):SF(6) ratio at the 2 exit locations. The proportions of CH(4) and SF(6) recovered at the head were greater (P 0.05; 2C, 6% and 4%; 3C, 17% and 15%; 3C+washer, 19% and 14%). Uncorrected CH(4) emissions were ± 10% of corrected emissions for 53% of the cow-day measurements. Only when more than 80% of the SF(6) escaped via the rumen did the difference between the uncorrected and corrected

  10. Preliminary conceptual designs for advanced packages for the geologic disposal of spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Westerman, R.E.

    1979-04-01

    The present study assumes that the spent fuel will be disposed of in mined repositories in continental geologic formations, and that the post-emplacement control of the radioactive species will be accomplished independently by both the natural barrier, i.e., the geosphere, and the engineered barrier system, i.e., the package components consisting of the stabilizer, the canister, and the overpack; and the barrier components external to the package consisting of the hole sleeve and the backfill medium. The present document provides an overview of the nature of the spent fuel waste; the general approach to waste containment, using the defense-in-depth philosophy; material options, both metallic and nonmetallic, for the components of the engineered barrier system; a set of strawman criteria to guide the development of package/engineered barrier systems; and four preliminary concepts representing differing approaches to the solution of the containment problem. These concepts use: a corrosion-resistant meta canister in a special backfill (2 barriers); a mild steel canister in a corrosion-resistant metallic or nonmetallic hole sleeve, surrounded by a special backfill (2 barriers); a corrosion-resistant canister and a corrosion-resistant overpack (or hole sleeve) in a special backfill (3 barriers); and a mild steel canister in a massive corrosion-resistant bore sleeve surrounded by a polymer layer and a special backfill (3 barriers). The lack of definitive performance requirements makes it impossible to evaluate these concepts on a functional basis at the present time.

  11. Design of spent-fuel concrete pit dry storage and handling system

    Energy Technology Data Exchange (ETDEWEB)

    Tamaki, H.; Natsume, T.; Maruoka, K.; Yokoyama, T. [Mitsubishi Heavy Industries Ltd., Yokohama (Japan)

    1998-07-01

    An advanced dry storage system design with highly improved storage efficiency of spent nuclear fuel has been developed. The new concept 'Concrete Pit Dry Storage System' realizes a safe and economical solution to an increasing demand of storing spent fuel assemblies (SFAs) generated from commercial nuclear power reactors. The system is basically composed of a large mass concrete module which has densely arranged pit boreholes, sealed canisters containing spent fuel assemblies and a canister handling system. The system is characterized by the following advantages compared with the existing concrete module type storage systems: higher storage efficiency can be achieved by the storage module filled with concrete which also gives a high shielding performance; simple handling technology is used for transfer and installation of the canisters at the storage facility as well as the transport cask of the canisters, surface contamination of the canister is prevented; lower radiation around the storage area is provided to reduce radiation exposure during handling and storage; high structural integrity of the facility is maintained by the concrete module with a simple construction ; the ventilation gallery introducing cooling air air to the bit borehole has an enough draft height to improve cooling performance of the system; a result of the design concept, the storage system can store higher burn-up SFAs with a short cooling period. (authors)

  12. Efficacy of backfilling and other engineered barriers in a radioactive waste repository in salt

    Energy Technology Data Exchange (ETDEWEB)

    Claiborne, H.C.

    1982-09-01

    In the United States, investigation of potential host geologic formations was expanded in 1975 to include hard rocks. Potential groundwater intrusion is leading to very conservative and expensive waste package designs. Recent studies have concluded that incentives for engineered barriers and 1000-year canisters probably do not exist for reasonable breach scenarios. The assumption that multibarriers will significantly increase the safety margin is also questioned. Use of a bentonite backfill for surrounding a canister of exotic materials was developed in Sweden and is being considered in the US. The expectation that bentonite will remain essentially unchanged for hundreds of years for US repository designs may be unrealistic. In addition, thick bentonite backfills will increase the canister surface temperature and add much more water around the canister. The use of desiccant materials, such as CaO or MgO, for backfilling seems to be a better method of protecting the canister. An argument can also be made for not using backfill material in salt repositories since the 30-cm-thick space will provide for hole closure for many years and will promote heat transfer via natural convection. It is concluded that expensive safety systems are being considered for repository designs that do not necessarily increase the safety margin. It is recommended that the safety systems for waste repositories in different geologic media be addressed individually and that cost-benefit analyses be performed.

  13. Retrievable storage concept designs. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Nickell, R.E.

    1979-03-01

    Three tasks related to the reference design of retrievable storage canisters for radioactive waste have been completed. The three tasks consist of the reference design itself, the definition of failure modes most appropriate for structural integrity determinations for the reference canister, and the development of a failure methodology for the structural integrity of the containers. The reference design is a sealed storage canister concept based upon the waste isolation pilot plant (WIPP) design, with slight modifications. The modifications consist of an alternate lifting yoke arrangement for the top head and a revised bottom head design for absorption of impact energy. Welded closures provide the seal at each end. Overpacking is considered as a possibility, but is not included in the preliminary reference design. The four failure modes that are deemed the most appropriate for the design of the reference canister are: (i) a loss of functional capability; (ii) ductile rupture of the canister; (iii) buckling of the structural members; and (iv) stress corrosion cracking. Failure scenarios are provided for each of the relevant failure modes. In addition, a failure methodology based upon the distribution of demand and the distribution of capacity for the structural members, with respect to each failure mode, is proffered.

  14. Draft Geologic Disposal Requirements Basis for STAD Specification

    Energy Technology Data Exchange (ETDEWEB)

    Ilgen, Anastasia G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hardin, Ernest [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-03-25

    This document provides the basis for requirements in the current version of Performance Specification for Standardized Transportation, Aging, and Disposal Canister Systems, (FCRD-NFST-2014-0000579) that are driven by storage and geologic disposal considerations. Performance requirements for the Standardized Transportation, Aging, and Disposal (STAD) canister are given in Section 3.1 of that report. Here, the requirements are reviewed and the rationale for each provided. Note that, while FCRD-NFST-2014-0000579 provides performance specifications for other components of the STAD storage system (e.g. storage overpack, transfer and transportation casks, and others), these have no impact on the canister performance during disposal, and are not discussed here.

  15. Thermal and mechanical analysis for the detailed model using submodel

    Energy Technology Data Exchange (ETDEWEB)

    Kuh, Jung Eui; Kang, Chul Hyung; Park, Jeong Hwa [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-11-01

    A very big model is required for the TM analysis for HLRW repository, and also very small size of mesh is needed to simulate precisely main parts of analysis, e.g., canister, buffer, etc. However, it is practically impossible due to high memory size and computing time. In this report, a submodel concept in ABAQUS is used to handle this difficulty. A submodel concept is the part interested only is performed detailed modelling and this result is used as a boundary condition of full scale model. To follow this kind of computation procedure temperature distribution in buffer and canister could be computed precisely. This approach can be applied to TM analysis of buffer and canister, or a finite size of repository. 12 refs., 28 figs., 9 tabs. (Author)

  16. CASTOR GSF packaging design criteria

    Energy Technology Data Exchange (ETDEWEB)

    Burnside, M.E.

    1996-08-06

    Encapsulated vitrified materials (Isotopic Heat Sources) are currently stored in the Pacific Northwest National Laboratories (PNNL) 324 Building located in the 300 Area. As part of the 324 Building transition program, the vitrified material, encapsulated in stainless steel canisters, must be removed. These canisters were originally intended to be used by the German government, but are no longer desired. As part of an agreement with the German government, the Germans are providing the U.S. Department of Energy (DOE) with six (6) CASTOR GSF and four (4) GNS-12 casks.The canisters will be transported onsite in CASTOR GSF and GNS-12 casks for interim storage until final disposition of the material is determined.

  17. Radiation Heat Transfer Modeling Improved for Phase-Change, Thermal Energy Storage Systems

    Science.gov (United States)

    Kerslake, Thomas W.; Jacqmin, David A.

    1998-01-01

    Spacecraft solar dynamic power systems typically use high-temperature phase-change materials to efficiently store thermal energy for heat engine operation in orbital eclipse periods. Lithium fluoride salts are particularly well suited for this application because of their high heat of fusion, long-term stability, and appropriate melting point. Considerable attention has been focused on the development of thermal energy storage (TES) canisters that employ either pure lithium fluoride (LiF), with a melting point of 1121 K, or eutectic composition lithium-fluoride/calcium-difluoride (LiF-20CaF2), with a 1040 K melting point, as the phase-change material. Primary goals of TES canister development include maximizing the phase-change material melt fraction, minimizing the canister mass per unit of energy storage, and maximizing the phase-change material thermal charge/discharge rates within the limits posed by the container structure.

  18. Numerical simulation of temperature field in deep penetration laser welding of 5A06 aluminum cylinder

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    Deep penetration laser welding temperature field of 5A06 aluminum alloy canister structure was simulated using the surface-body combination heat source model by ANSYS, which was made up of Gauss surface heat source model and Gauss revolved body heat source model. Convection, radiation and conduction were all considered during the simulation process. The thermal cycle curves of the points both on the shell outer surface and in the seam thickness direction were calculated. Simulated results agreed well with the experiment results. It concluded that the surface-body combination heat source model was fit for the temperature field simulation of deep penetration laser welding of the aluminum alloy canister structure. This method was proved to be an efficient way to predict the shape and dimension of welded joint for deep penetration laser welding of the aluminum alloy canister structure.

  19. TMI Fuel Characteristics for Disposal Criticality Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Larry L. Taylor

    2003-09-01

    This report documents the reported contents of the Three Mile Island Unit 2 (TMI-2) canisters. proposed packaging, and degradation scenarios expected in the repository. Most fuels within the U.S. Department of Energy spent nuclear fuel inventory deal with highly enriched uranium, that in most cases require some form of neutronic poisoning inside the fuel canister. The TMI-2 fuel represents a departure from these fuel forms due to its lower enrichment (2.96% max.) values and the disrupted nature of the fuel itself. Criticality analysis of these fuel canisters has been performed over the years to reflect conditions expected during transit from the reactor to the Idaho National Engineering and Environmental Laboratory, water pool storage,1 and transport/dry-pack storage at Idaho Nuclear Technology and Engineering Center.2,3 None of these prior analyses reflect the potential disposal conditions for this fuel inside a postclosure repository.

  20. Temperature field due to time-dependent heat sources in a large rectangular grid - Derivation of analytical solution

    Energy Technology Data Exchange (ETDEWEB)

    Claesson, J.; Probert, T. [Lund Univ. (Sweden). Dept. of Building Physics and Mathematical Physics

    1996-01-01

    The temperature field in rock due to a large rectangular grid of heat releasing canisters containing nuclear waste is studied. The solution is by superposition divided into different parts. There is a global temperature field due to the large rectangular canister area, while a local field accounts for the remaining heat source problem. The global field is reduced to a single integral. The local field is also solved analytically using solutions for a finite line heat source and for an infinite grid of point sources. The local solution is reduced to three parts, each of which depends on two spatial coordinates only. The temperatures at the envelope of a canister are given by a single thermal resistance, which is given by an explicit formula. The results are illustrated by a few numerical examples dealing with the KBS-3 concept for storage of nuclear waste. 8 refs.

  1. Human factors analysis and design methods for nuclear waste retrieval systems: Human factors design methodology and integration plan

    Science.gov (United States)

    Casey, S. M.

    1980-06-01

    The nuclear waste retrieval system intended to be used for the removal of storage canisters (each canister containing a spent fuel rod assembly) located in an underground salt bed depository is discussed. The implementation of human factors engineering principles during the design and construction of the retrieval system facilities and equipment is reported. The methodology is structured around a basic system development effort involving preliminary development, equipment development, personnel subsystem development, and operational test and evaluation. Examples of application of the techniques in the analysis of human tasks, and equipment required in the removal of spent fuel canisters is provided. The framework for integrating human engineering with the rest of the system development effort is documented.

  2. Acoustic emission and ultrasonic monitoring results from deposition hole DA3545G01 in the Prototype Repository between April 2009 and September 2009

    Energy Technology Data Exchange (ETDEWEB)

    Haycox, Jon; Pettitt, Will [ASC, Applied Seismology Consultants, Shrewsbury, Shropshire (United Kingdom)

    2009-12-15

    This report describes results from acoustic emission (AE) and ultrasonic monitoring around a canister deposition hole (DA3545G01) in the Prototype Repository Experiment at SKB's Hard Rock Laboratory (HRL), Sweden. The monitoring aims to examine changes in the rock mass caused by an experimental repository environment, in particular due to thermal stresses induced from canister heating and changes in pore pressures induced from tunnel sealing. Monitoring of this volume has previously been performed during excavation (Pettitt et al. 1999), and during stages of canister heating and tunnel pressurisation (Haycox and Pettitt 2005a, b, 2006a, b, Zolezzi et al. 2007, 2008, Duckworth et al. 2008, 2009, Haycox and Duckworth 2009). Further information on the previous monitoring undertaken can be found in Appendix 1. This report covers the period between 1st April 2009 and 30th September 2009 and is the ninth 6-monthly processing and interpretation of the results from the experiment.

  3. Acoustic emission and ultrasonic monitoring results from deposition hole DA3545G01 in the Prototype Repository between October 2008 and March 2009

    Energy Technology Data Exchange (ETDEWEB)

    Haycox, Jon; Duckworth, Damion [ASC, Applied Seismology Consultants, Shrewsbury, Shropshire (United Kingdom)

    2009-06-15

    This report describes results from acoustic emission (AE) and ultrasonic monitoring around a canister deposition hole (DA3545G01) in the Prototype Repository Experiment at SKB's Hard Rock Laboratory (HRL), Sweden. The monitoring aims to examine changes in the rock mass caused by an experimental repository environment, in particular due to thermal stresses induced from canister heating and changes in pore pressures induced from tunnel sealing. Monitoring of this volume has previously been performed during excavation (Pettitt et al. 1999), and during stages of canister heating and tunnel pressurisation (Haycox and Pettitt 2005a, b, 2006a, b, Zolezzi et al. 2007, 2008, Duckworth et al. 2008, 2009). Further information on the previous monitoring undertaken can be found in Appendix 1. This report covers the period between 1st October 2008 and 31st March 2009 and is the eighth 6-monthly processing and interpretation of the results from the experiment.

  4. Aespoe Hard Rock Laboratory. Prototype Repository. Acoustic emission and ultrasonic monitoring results from deposition hole DA3545G01 in the Prototype Repository between April 2008 and September 2008

    Energy Technology Data Exchange (ETDEWEB)

    Duckworth, D.; Haycox, J.; Pettitt, W.S. (Applied Seismology Consultants, Shrewsbury (United Kingdom))

    2009-03-15

    This report describes results from acoustic emission (AE) and ultrasonic monitoring around a canister deposition hole (DA3545G01) in the Prototype Repository Experiment at SKB's Hard Rock Laboratory (HRL), Sweden. The monitoring aims to examine changes in the rock mass caused by an experimental repository environment, in particular due to thermal stresses induced from canister heating and pore pressures induced from tunnel sealing. Monitoring of this volume has previously been performed during excavation [Pettitt et al., 1999], and during stages of canister heating and tunnel pressurisation [Haycox et al., 2005a and 2005b; Haycox et al., 2006a and 2006b; Zolezzi et al., 2007 and Duckworth et al., 2008]. Further information on this monitoring can be found in Appendix I. This report covers the period between 1st April 2008 and 30th September 2008 and is the seventh instalment of the 6-monthly processing and interpretation of the results from the experiment.

  5. Influence of void ratio on phase change of thermal energy storage for heat pipe receiver

    Directory of Open Access Journals (Sweden)

    Xiaohong Gui

    2015-01-01

    Full Text Available In this paper, influence of void ratio on phase change of thermal storage unit for heat pipe receiver under microgravity is numerically simulated. Accordingly, mathematical model is set up. A solidification-melting model upon the enthalpy-porosity method is specially provided to deal with phase changes. The liquid fraction distribution of thermal storage unit of heat pipe receiver is shown. The fluctuation of melting ratio in PCM canister is indicated. Numerical results are compared with experimental ones in Japan. The results show that void cavity prevents the process of phase change greatly. PCM melts slowly during sunlight periods and freezes slowly during eclipse periods as void ratio increases. The utility ratio of PCM during both sunlight periods and eclipse periods decreases obviously with the improvement of void ratio. The thermal resistance of void cavity is much higher than that of PCM canister wall. Void cavity prevents the heat transfer between PCM zone and canister wall.

  6. Spent nuclear fuel project product specification

    Energy Technology Data Exchange (ETDEWEB)

    Pajunen, A.L.

    1998-01-30

    Product specifications are limits and controls established for each significant parameter that potentially affects safety and/or quality of the Spent Nuclear Fuel (SNF) packaged for transport to dry storage. The product specifications in this document cover the spent fuel packaged in MultiCanister Overpacks (MCOs) to be transported throughout the SNF Project. The SNF includes N Reactor fuel and single-pass reactor fuel. The FRS removes the SNF from the storage canisters, cleans it, and places it into baskets. The MCO loading system places the baskets into MCO/Cask assembly packages. These packages are then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the MCO cask packages are transferred to the Canister Storage Building (CSB), where the MCOs are removed from the casks, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The key criteria necessary to achieve these goals are documented in this specification.

  7. Influence of phosphorus on the tensile stress strain curves in copper

    Science.gov (United States)

    Sandström, Rolf

    2016-03-01

    Copper canisters are planned to be used for final disposal of spent nuclear fuel in Sweden. The canisters will be exposed to slow plastic straining over extensive periods of time. To be able to predict the mechanical properties a range of basic models have previously been developed for copper with and without phosphorus (Cu-OFP, Cu-OF). Already with the small amount of phosphorus added in the canisters (60 wt. ppm) dramatic improvements in the measured creep strength and the creep ductility are found. The basic models are further developed in the present paper. The influence of phosphorus on slow strain rate tests is analysed. It is shown that the main effect of phosphorus is that it prevents brittle rupture, which is modelled by taking creep cavitation into account.

  8. A new approach to helicopter rotor blade research instrumentation

    Science.gov (United States)

    Knight, V. H., Jr.

    1978-01-01

    A rotor-blade-mounted telemetry instrumentation system developed and used in flight tests by the NASA/Langley Research Center is described. The system uses high-speed digital techniques to acquire research data from miniature pressure transducers on advanced rotor airfoils which are flight tested using an AH-1G helicopter. The system employs microelectronic PCM multiplexer-digitizer stations located remotely on the blade and in a hub-mounted metal canister. The electronics contained in the canister digitizes up to 16 sensors, formats this data with serial PCM data from the remote stations, and transmits the data from the canister which is above the plane of the rotor. Data is transmitted over an RF link to the ground for real-time monitoring and to the helicopter fuselage for tape recording.

  9. Acoustic emission and ultrasonic monitoring results from deposition hole DA3545G01 in the Prototype Repository between April 2010 and September 2010

    Energy Technology Data Exchange (ETDEWEB)

    Haycox, Jon (ASC Applied Seismology Consultants (United Kingdom))

    2011-05-15

    This report describes results from acoustic emission (AE) and ultrasonic monitoring around a canister deposition hole (DA3545G01) in the Prototype Repository Experiment at SKB's Hard Rock Laboratory (HRL), Sweden. The monitoring aims to examine changes in the rock mass caused by an experimental repository environment, in particular due to thermal stresses induced from canister heating and changes in pore pressures induced from tunnel sealing. Monitoring of this volume has previously been performed during excavation (Pettitt et al. 1999), and during stages of canister heating and tunnel pressurisation (Haycox and Pettitt 2005a, b, 2006a, b, Zolezzi et al. 2007, 2008, Duckworth et al. 2008, 2009, Haycox et al. 2009a, b, 2010). Appendix I contains further information about previous monitoring periods. This report covers the period between 1st April 2010 and 30th September 2010 and is the eleventh 6-monthly processing and interpretation of the results from the experiment

  10. Acoustic emission and ultrasonic monitoring results from deposition hole DA3545G01 in the Prototype Repository between October 2009 and March 2010

    Energy Technology Data Exchange (ETDEWEB)

    Haycox, Jon; Andrews, Jennifer [ASC, Applied Seismology Consultants, Shrewsbury, Shropshire (United Kingdom)

    2010-09-15

    This report describes results from acoustic emission (AE) and ultrasonic monitoring around a canister deposition hole (DA3545G01) in the Prototype Repository Experiment at SKB's Hard Rock Laboratory (HRL), Sweden. The monitoring aims to examine changes in the rock mass caused by an experimental repository environment, in particular due to thermal stresses induced from canister heating and changes in pore pressures induced from tunnel sealing. Monitoring of this volume has previously been performed during excavation (Pettitt et al. 1999), and during stages of canister heating and tunnel pressurisation (Haycox and Pettitt 2005a, b, 2006a, b, 2009, Zolezzi et al. 2007, 2008, Duckworth et al. 2008, 2009, Haycox and Duckworth 2009). Further information on the previous monitoring periods can be found in Appendix 1. This report covers the period between 1st October 2009 and 31st March 2010 and is the tenth 6-monthly processing and interpretation of the results from the experiment.

  11. Effect Of Up-Scaling On The Study Of The Steel/Bentonite Interface In A Deep Geological Repository

    Energy Technology Data Exchange (ETDEWEB)

    Torres Alvarez, Elena; Turrero, Maria Jesus; Martin, Pedro Luis; Escribano, Alicia [CIEMAT, Avda. Complutense 22, 28040, Madrid (Spain)

    2008-07-01

    Deep geological disposal is the most accepted management option for High Level Nuclear Wastes. The multi-barrier system for the isolation of high-level radioactive waste includes the concept of the spent fuel encapsulated in canisters of carbon steel. Corrosion phenomena affect the integrity of the canister and can modify the chemical environment either at the interface or in the bentonite pore water. The experimental studies conducted by CIEMAT are focused on the iron canister corrosion products interaction with the bentonite system and are based on a series of short term and medium term experiments conceived at different scales, from conventional laboratory experiments and experiments in cylindrical cells, to those specifically designed 3D mock up experiments, the so called 'GAME (Geochemical Mock up experiments) scale'. The results obtained from the up-scaling could be a useful tool to understand the key processes at the steel/bentonite interface and the later modelling work. (authors)

  12. Long-term non-isothermal reactive transport model of compacted bentonite, concrete and corrosion products in a HLW repository in clay

    Science.gov (United States)

    Mon, Alba; Samper, Javier; Montenegro, Luis; Naves, Acacia; Fernández, Jesús

    2017-02-01

    Radioactive waste disposal in deep geological repositories envisages engineered barriers such as carbon-steel canisters, compacted bentonite and concrete liners. The stability and performance of the bentonite barrier could be affected by the corrosion products at the canister-bentonite interface and the hyper-alkaline conditions caused by the degradation of concrete at the bentonite-concrete interface. Additionally, the host clay formation could also be affected by the hyper-alkaline plume at the concrete-clay interface. Here we present a non-isothermal multicomponent reactive transport model of the long-term (1 Ma) interactions of the compacted bentonite with the corrosion products of a carbon-steel canister and the concrete liner of the engineered barrier of a high-level radioactive waste repository in clay. Model results show that magnetite is the main corrosion product. Its precipitation reduces significantly the porosity of the bentonite near the canister. The degradation of the concrete liner leads to the precipitation of secondary minerals and the reduction of the porosity of the bentonite and the clay formation at their interfaces with the concrete liner. The reduction of the porosity becomes especially relevant at t = 104 years. The zones affected by pore clogging at the canister-bentonite and concrete-clay interfaces at 1 Ma are approximately equal to 1 and 3.3 cm thick, respectively. The hyper-alkaline front (pH > 8.5) spreads 2.5 cm into the clay formation after 1 Ma. Our simulation results share the key features of the models reported by others for engineered barrier systems at similar chemical conditions, including: 1) Pore clogging at the canister-bentonite and concrete-clay interfaces; 2) Narrow alteration zones; and 3) Limited smectite dissolution after 1 Ma.

  13. Long-term non-isothermal reactive transport model of compacted bentonite, concrete and corrosion products in a HLW repository in clay.

    Science.gov (United States)

    Mon, Alba; Samper, Javier; Montenegro, Luis; Naves, Acacia; Fernández, Jesús

    2017-02-01

    Radioactive waste disposal in deep geological repositories envisages engineered barriers such as carbon-steel canisters, compacted bentonite and concrete liners. The stability and performance of the bentonite barrier could be affected by the corrosion products at the canister-bentonite interface and the hyper-alkaline conditions caused by the degradation of concrete at the bentonite-concrete interface. Additionally, the host clay formation could also be affected by the hyper-alkaline plume at the concrete-clay interface. Here we present a non-isothermal multicomponent reactive transport model of the long-term (1Ma) interactions of the compacted bentonite with the corrosion products of a carbon-steel canister and the concrete liner of the engineered barrier of a high-level radioactive waste repository in clay. Model results show that magnetite is the main corrosion product. Its precipitation reduces significantly the porosity of the bentonite near the canister. The degradation of the concrete liner leads to the precipitation of secondary minerals and the reduction of the porosity of the bentonite and the clay formation at their interfaces with the concrete liner. The reduction of the porosity becomes especially relevant at t=10(4)years. The zones affected by pore clogging at the canister-bentonite and concrete-clay interfaces at 1Ma are approximately equal to 1 and 3.3cm thick, respectively. The hyper-alkaline front (pH>8.5) spreads 2.5cm into the clay formation after 1Ma. Our simulation results share the key features of the models reported by others for engineered barrier systems at similar chemical conditions, including: 1) Pore clogging at the canister-bentonite and concrete-clay interfaces; 2) Narrow alteration zones; and 3) Limited smectite dissolution after 1Ma.

  14. Copper corrosion under expected conditions in a deep geologic repository

    Energy Technology Data Exchange (ETDEWEB)

    King, F. [Integrity Corrosion Consulting Ltd, Calgary, Alberta (Canada); Ahonen, L. [Geological Survey of Finland, Espoo (Finland); Taxen, C. [Swedish Corrosion Inst., Stockholm (Sweden); Vuorinen, U. [VTT Chemical Technology, Espoo (Finland); Werme, L. [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden)

    2001-12-01

    Copper has been the corrosion barrier of choice for the canister in the Swedish and Finnish, nuclear waste disposal programmes for over 20 years. During that time many studies have been carried out on the corrosion behaviour of copper under conditions likely to exist in an underground nuclear disposal repository located in he Fenno-Scandian bedrock. This review is a summary of what has been learnt about the long- term behaviour of the corrosion barrier during this period and what the implications of this knowledge are for the predicted service life of the canisters. The review is based on the existing knowledge from various nuclear waste management programs around the world and from the open literature.Various areas are considered: the expected evolution of the geochemical conditions in the groundwater and of the repository environment, the thermodynamics of copper corrosion, corrosion before and during saturation of the compacted bentonite buffer by groundwater, general and localized corrosion following saturation of the compacted bentonite buffer, stress corrosion cracking, radiation effects, the implications of corrosion on the service life of the canister, and areas for further study. Much has been learnt about the long-term corrosion behaviour of copper canisters over the past 20 years. The majority of the information reviewed here is drawn from the Swedish/Finnish and Canadian programmes. Despite differences in scientific approach, and canister and repository design, the results of these two programmes both suggest that copper provides an excellent corrosion barrier in an underground repository. The conclusion drawn from this review is that the original prediction made in 1978 of canister lifetimes exceeding 100,000 years remains valid.

  15. Copper corrosion under expected conditions in a deep geologic repository

    Energy Technology Data Exchange (ETDEWEB)

    King, F.; Ahonen, L.; Taxen, C.; Vuorinen, U.; Werme, L

    2002-01-01

    Copper has been the corrosion barrier of choice for the canister in the Swedish and Finnish, nuclear waste disposal programmes for over 20 years. During that time many studies have been carried out on the corrosion behaviour of copper under conditions likely to exist in an underground nuclear disposal repository located in the Fenno-Scandian bedrock. This review is a summary of what has been learnt about the long-term behaviour of the corrosion barrier during this period and what the implications of this knowledge are for the predicted service life of the canisters. The review is based on the existing knowledge from various nuclear waste management programs around the world and from the open literature. Various areas are considered: the expected evolution of the geochemical conditions in the groundwater and of the repository environment, the thermodynamics of copper corrosion, corrosion before and during saturation of the compacted bentonite buffer by groundwater, general and localized corrosion following saturation of the compacted bentonite buffer, stress corrosion cracking, radiation effects, the implications of corrosion on the service life of the canister, and areas for further study. Much has been learnt about the long-term corrosion behaviour of copper canisters over the past 20 years. The majority of the information reviewed here is drawn from the Swedish/Finnish and Canadian programmes. Despite differences in scientific approach, and canister and repository design, the results of these two programmes both suggest that copper provides an excellent corrosion barrier in an underground repository. The conclusion drawn from this review is that the original prediction made in 1978 of canister lifetimes exceeding 100,000 years remains valid. (orig.)

  16. Packaging design criteria for the MCO cask

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, W.S.

    1997-01-30

    Approximately 2,100 metric tons of unprocessed, irradiated nuclear fuel elements are presently stored in the K Basins. To permit cleanup of the K Basins and fuel conditioning, the fuel will be transported from the K Basins to a Canister Storage Building in the 200 East Area. The purpose of this packaging design criteria is to provide criteria for the design, fabrication, and use of a packaging system to transport the large quantities of irradiated nuclear fuel elements positioned within Multiple Canister Overpacks.

  17. 结构参数对吸热器热性能的影响%EFFECT OF STRUCTURAL PARAMETER ON HEAT TRANSFER IN HEAT RECEIVER

    Institute of Scientific and Technical Information of China (English)

    崔海亭; 袁修干; 邢玉明

    2004-01-01

    The heat receiver is a key part of Solar Dynamic Power System. The structural parameter affects the heat transfer of the heat receiver. The enthalpy method was used to calculate the three - dimensional phase change process of the heat exchange tube. The maximum canister temperature, average canister temperature , the receiver gas exit temperature and liquid PCM fraction in many kinds of length diameter ratio and aperture were calculated. The results showed that the output temperature of the gas from the working fluid tubes could meet the expected demand during the sunlight and eclipse period. The maximum temperature of the PCM container was within the safe range.

  18. Survey of creep properties of copper intended for nuclear waste disposal

    Energy Technology Data Exchange (ETDEWEB)

    Andersson-Oestling, Henrik C.M. (Swerea KIMAB AB, Stockholm (Sweden)); Sandstroem, Rolf (Materials Science and Engineering, School of Industrial Engineering and Management, Royal Inst. of Technology (KTH), Stockholm (Sweden))

    2009-12-15

    Creep in copper for application in canisters for nuclear waste disposal is surveyed. The importance of phosphorus doping to obtain adequate properties is demonstrated experimentally as well as explained theoretically. Creep tests results for electron beam and friction stir welds are compared. The latter type of welds has properties that are close to those of parent metal. The relation between slow strain rate tensile and creep is described. Fundamental constitutive equations are presented that are suitable for finite element modelling. These equations are used to simulate creep deformation in canisters

  19. Aespoe Hard Rock Laboratory. Prototype Repository. Acoustic emission and ultrasonic monitoring results from deposition hole DA3545G01 in the Prototype Repository between October 2007 and March 2008

    Energy Technology Data Exchange (ETDEWEB)

    Duckworth, D.; Haycox, J.; Pettitt, W.S. (Applied Seismology Consultants, Shrewsbury (United Kingdom))

    2008-12-15

    This report describes results from acoustic emission (AE) and ultrasonic monitoring around a canister deposition hole (DA3545G01) in the Prototype Repository Experiment at SKB's Hard Rock Laboratory (HRL), Sweden. The experiment has been designed to simulate a disposal tunnel in a real deep repository environment for storage of high-level radioactive waste. The test consists of a 90 m long, 5 m diameter subhorizontal tunnel excavated in dioritic granite. The monitoring aims to examine changes in the rock mass caused by an experimental repository environment, in particular due to thermal stresses induced from canister heating and pore pressures induced from tunnel sealing.

  20. Cold Vacuum Drying (CVD) Facility Technical Safety Requirements

    Energy Technology Data Exchange (ETDEWEB)

    KRAHN, D.E.

    2000-08-08

    The Technical Safety Requirements (TSRs) for the Cold Vacuum Drying Facility define acceptable conditions, safe boundaries, bases thereof, and management or administrative controls required to ensure safe operation during receipt of multi-canister overpacks (MCOs) containing spent nuclear fuel. removal of free water from the MCOs using the cold vacuum drying process, and inerting and testing of the MCOs before transport to the Canister Storage Building. Controls required for public safety, significant defense in depth, significant worker safety, and for maintaining radiological and toxicological consequences below risk evaluation guidelines are included.

  1. Dry Transfer Systems for Used Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Brett W. Carlsen; Michaele BradyRaap

    2012-05-01

    The potential need for a dry transfer system (DTS) to enable retrieval of used nuclear fuel (UNF) for inspection or repackaging will increase as the duration and quantity of fuel in dry storage increases. This report explores the uses for a DTS, identifies associated general functional requirements, and reviews existing and proposed systems that currently perform dry fuel transfers. The focus of this paper is on the need for a DTS to enable transfer of bare fuel assemblies. Dry transfer systems for UNF canisters are currently available and in use for transferring loaded canisters between the drying station and storage and transportation casks.

  2. Apollo Telescope Mount Spar Assembly

    Science.gov (United States)

    1969-01-01

    The Apollo Telescope Mount (ATM), designed and developed by the Marshall Space Flight Center, served as the primary scientific instrument unit aboard the Skylab. The ATM contained eight complex astronomical instruments designed to observe the Sun over a wide spectrum from visible light to x-rays. This image shows the ATM spar assembly. All solar telescopes, the fine Sun sensors, and some auxiliary systems are mounted on the spar, a cruciform lightweight perforated metal mounting panel that divides the 10-foot long canister lengthwise into four equal compartments. The spar assembly was nested inside a cylindrical canister that fit into the rack, a complex frame, and was protected by the solar shield.

  3. Mineral formation on metallic copper in a `future repository site environment`

    Energy Technology Data Exchange (ETDEWEB)

    Amcoff, Oe.; Holenyi, K.

    1996-04-01

    Since reducing conditions are expected much effort has been concentrated on Cu-sulfides and CuFe-sulfides. However, oxidizing conditions are also discussed. A list of copper minerals are included. It is concluded that mineral formation and mineral transitions on the copper canister surface will be governed by kinetics and metastabilities rather than by stability relations. The sulfides formed are less likely to form a passivating layer, and the rate of sulfide growth will probably be governed by the rate of transport of reacting species to the canister surface. A series of tests are recommended, in an environment resembling the initial repository site conditions. 82 refs, 8 figs.

  4. Analysis of Dust Samples Collected from an Unused Spent Nuclear Fuel Interim Storage Container at Hope Creek, Delaware.

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Enos, David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-03-01

    In July, 2014, the Electric Power Research Institute and industry partners sampled dust on the surface of an unused canister that had been stored in an overpack at the Hope Creek Nuclear Generating Station for approximately one year. The foreign material exclusion (FME) cover that had been on the top of the canister during storage, and a second recently - removed FME cover, were also sampled. This report summarizes the results of analyses of dust samples collected from the unused Hope Creek canister and the FME covers. Both wet and dry samples of the dust/salts were collected, using SaltSmart(TM) sensors and Scotch - Brite(TM) abrasive pads, respectively. The SaltSmart(TM) samples were leached and the leachate analyzed chemically to determine the composition and surface load per unit area of soluble salts present on the canister surface. The dry pad samples were analyzed by X-ray fluorescence and by scanning electron microscopy to determine dust texture and mineralogy; and by leaching and chemical analysis to deter mine soluble salt compositions. The analyses showed that the dominant particles on the canister surface were stainless steel particles, generated during manufacturing of the canister. Sparse environmentally - derived silicates and aluminosilicates were also present. Salt phases were sparse, and consisted of mostly of sulfates with rare nitrates and chlorides. On the FME covers, the dusts were mostly silicates/aluminosilicates; the soluble salts were consistent with those on the canister surface, and were dominantly sulfates. It should be noted that the FME covers were w ashed by rain prior to sampling, which had an unknown effect of the measured salt loads and compositions. Sulfate salts dominated the assemblages on the canister and FME surfaces, and in cluded Ca - SO4 , but also Na - SO4 , K - SO4 , and Na - Al - SO4 . It is likely that these salts were formed by particle - gas conversion reactions, either

  5. Thermal modeling of a vertical dry storage cask for used nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Li, Jie, E-mail: jieli@anl.gov; Liu, Yung Y., E-mail: yyliu@anl.gov

    2016-05-15

    Graphical abstract: - Highlights: • Thermal performance of a 3-D vertical dry cask under various conditions has been numerically studied by using ANSYS/FLUENT code. • The simulation was validated by comparing the results against data obtained from the temperature measurements of a commercial cask. • The results indicated that the basket with higher thermal conductivity dissipates decay heat out of the cask more efficiently than that with a lower thermal conductivity (aluminum composite vs. stainless steel). A heavier cooling gas is also helpful to enhance heat transfer via enhanced natural convection (N{sub 2} vs. He). • Coolant release from the fuel canister results in temperature change of the canister external surfaces. The simulation shows that such a change is large enough and detectable, which can provide a mechanism for leak detection by continuously monitoring this temperature change at the top center of the canister surface. • Partial blockage of the cask air inlets affects the temperature profiles marginally for both the fuel canister and those components inside. In contrast, fully blocked air inlets will lead to remarkable increases of the component temperatures. - Abstract: Thermal modeling of temperature profiles of dry casks has been identified as a high-priority item in a U.S. Department of Energy gap analysis. In this work, a three-dimensional model of a vertical dry cask has been constructed for computer simulation by using the ANSYS/FLUENT code. The vertical storage cask contains a welded canister for 32 Pressurized Water Reactor (PWR) used-fuel assemblies with a total decay heat load of 34 kW. To simplify thermal calculations, an effective thermal conductivity model for a 17 × 17 PWR used (or spent)-fuel assembly was developed and used in the simulation of thermal performance. The effects of canister fill gas (helium or nitrogen), internal pressure (1–6 atm), and basket material (stainless steel or aluminum alloy) were studied to

  6. Estimation of rock movements due to future earthquakes at four candidate sites for a spent fuel repository in Finland

    Energy Technology Data Exchange (ETDEWEB)

    Pointe, P. La [Golder Associates Inc. (United States); Hermanson, J. [Golder Associates AB (Sweden)

    2002-02-01

    Numerical simulations of the displacements of fractures intersecting canisters due to future seismicity have been carried out for the Olkiluoto, Kivetty, Hastholmen and Romuvaara sites. The numerical simulation process uses stochastic models of repository-scale natural fracturing, as well as the magnitude and frequency of future earthquakes within a 100 km radius of the repositories. Future seismicity is based on the extrapolation of the current earthquake catalog for each region. Earthquakes with magnitudes greater than ML = 5.5 (up to 7 - 8) are relegated to the post glacial period at the end of the 100 000 year simulation interval. There are uncertainties related to these assumptions, as the magnitude-frequency distribution of seismicity is likely different in the post-glacial period. Future earthquakes are located on lineament segments of sufficient size for the specified magnitude that have been mapped within the 100 km circular region. The orientation of the lineament and the sense of slip on the fault are selected based upon current seismicity, and depend upon whether the future earthquake is post-glacial or not. The repository-scale fracture model is constructed from the analysis of borehole and outcrop fracture data, as well as lineament data at scales ranging from 1:20 000 to 1:1 000 000. Scaling analysis indicates that both lineament and outcrop scale data can be combined to construct the repository scale fracture model. The results of the fracture analyses are numerically implemented as Discrete Fracture Network (DFN) models. One hundred stochastic realizations of both the repository-scale DFN model and future seismicity were combined with canister layouts for each site to provide the input data for earthquake simulations using a three-dimensional linearly elastic fracture mechanics code. The amount of induced slip on fractures intersecting canisters was tabulated for all of the stochastic realizations for each of the four sites. Statistics on canister

  7. 46 CFR 160.031-4 - Equipment for shoulder gun type line-throwing appliance.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 6 2010-10-01 2010-10-01 false Equipment for shoulder gun type line-throwing appliance... Appliance, Shoulder Gun Type (and Equipment) § 160.031-4 Equipment for shoulder gun type line-throwing... the gun. The line canister shall be secured by clamps or brackets below the barrel of the gun. (c)...

  8. Preliminary analysis of species partitioning in the DWPF melter. Sludge batch 7A

    Energy Technology Data Exchange (ETDEWEB)

    Choi, A. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Smith III, F. G. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McCabe, D. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-01-01

    The work described in this report is preliminary in nature since its goal was to demonstrate the feasibility of estimating the off-gas carryover from the Defense Waste Processing Facility (DWPF) melter based on a simple mass balance using measured feed and glass pour stream (PS) compositions and time-averaged melter operating data over the duration of one canister-filling cycle.

  9. Proceedings of the fifth annual NEA-seabed working group meeting

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, D. R. [ed.

    1980-09-01

    European Communities, the Federal Republic of Germany, France, Japan, Netherlands, Switzerland, the United Kingdom, and the United States on national policies and positions on seabed disposal are summarized. Task group reports on systems analysis, site assessment, canisters, waste forms, sediment and rocks, physical oceanography, and biology are presented. (DMC)

  10. View of the firing of a single engine OMS

    Science.gov (United States)

    1985-01-01

    View of a single engine orbital maneuvering system (OMS) firing on the Discovery. The payload bay is open and the protective canisters for the AUSSAT communications satellite (open) and the ASC-1 are visible. A cloudy Earth's horizon can be seen above the orbiter.

  11. 10 CFR 75.4 - Definitions.

    Science.gov (United States)

    2010-01-01

    ... graphite, irradiated fuel casks and canisters, reactor control rods, criticality safe tanks and vessels...; reactors; critical facilities; reprocessing of nuclear fuel; and processing of intermediate or high-level... nuclear material, such as an independent spent fuel storage installation (ISFSI) or a...

  12. 46 CFR 151.50-32 - Ammonia, anhydrous.

    Science.gov (United States)

    2010-10-01

    ... copper bearing alloys shall not be used as materials of construction for tanks, pipelines, valves... have close at hand at all times a canister mask approved for ammonia or each person shall carry on his... provide respiratory protection for emergency escape from a contaminated area resulting from cargo...

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

    Energy Technology Data Exchange (ETDEWEB)

    Ladkany, S.G.

    1995-11-15

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

  14. Monitoring instrumentation spent fuel management program. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1979-01-01

    Preliminary monitoring system methodologies are identified as an input to the risk assessment of spent fuel management. Conceptual approaches to instrumentation for surveillance of canister position and orientation, vault deformation, spent fuel dissolution, temperature, and health physics conditions are presented. In future studies, the resolution, reliability, and uncertainty associated with these monitoring system methodologies will be evaluated.

  15. Final Report - Spent Nuclear Fuel Retrieval System Manipulator System Cold Validation Testing

    Energy Technology Data Exchange (ETDEWEB)

    D.R. Jackson; G.R. Kiebel

    1999-08-24

    Manipulator system cold validation testing (CVT) was performed in support of the Fuel Retrieval System (FRS) Sub-Project, a subtask of the Spent Nuclear Fuel Project at the Hanford Site in Richland, Washington. The FRS will be used to retrieve and repackage K-Basin Spent Nuclear Fuel (SNF) currently stored in old K-Plant storage basins. The FRS is required to retrieve full fuel canisters from the basin; clean the fuel elements inside the canister to remove excessive uranium corrosion products (or sludge); remove the contents from the canisters; and sort the resulting debris, scrap, and fuel for repackaging. The fuel elements and scrap will be collected in fuel storage and scrap baskets in preparation for loading into a multi canister overpack (MCO), while the debris is loaded into a debris bin and disposed of as solid waste. The FRS is composed of three major subsystems. The Manipulator Subsystem provides remote handling of fuel, scrap, and debris; the In-Pool Equipment subsystem performs cleaning of fuel and provides a work surface for handling materials; and the Remote Viewing Subsystem provides for remote viewing of the work area by operators. There are two complete and identical FRS systems, one to be installed in the K-West basin and one to be installed in the K-East basin. Another partial system will be installed in a cold test facility to provide for operator training.

  16. Comparative Evaluation of Inoculation of Urine Samples with the Copan WASP® and BD Kiestra™ InoqulA™ Instruments

    DEFF Research Database (Denmark)

    Iversen, Jesper; Stendal, Gitta; Gerdes, Cecilie Marie;

    2016-01-01

    This study evaluated quantitative results as well as the quality of the inoculation patterns on urine specimens produced by two automated instruments, the Copan WASP® and the BD InoqulA™. 526 urine samples submitted in 10 ml canisters containing boric acid were processed within 30 minutes...

  17. Technical assessment of processes to enable recycling of low-level contaminated metal waste

    Energy Technology Data Exchange (ETDEWEB)

    Reimann, G.A.

    1991-10-01

    Accumulations of metal waste exhibiting low levels of radioactivity (LLCMW) have become a national burden, both financially and environmentally. Much of this metal could be considered as a resource. The Department of Energy was assigned the task of inventorying and classifying LLCMW, identifying potential applications, and applying and/or developing the technology necessary to enable recycling. One application for recycled LLCMW is high-quality canisters for permanent repository storage of high-level waste (HLW). As many as 80,000 canisters will be needed by 2035. Much of the technology needed to decontaminate LLCMW has already been developed, but no integrated process has been described, even on a pilot scale, for recycling LLCMW into HLW canisters. This report reviews practices for removal of radionuclides and for producing low carbon stainless steel. Contaminants that readily form oxides may be reduced to below de minimis levels and combined with a slag. Most of the radioactivity remaining in the ingot is concentrated in the inclusions. Radionuclides that chemically resemble the elements that comprise stainless steel can not be removed effectively. Slag compositions, current melting practices, and canister fabrication techniques were reviewed.

  18. A simple method of monitoring carbon dioxide output in anaesthetized patients.

    Science.gov (United States)

    Christensen, K N

    1977-01-01

    The mean CO2 output during anaesthesia in paralyzed patients can be monitored by continuous capnographic analysis of the total exhaled gases, the latter being mechanically integrated by pumice canisters. The gas is evacuated from the Hafnia A circuit via an ejector flowmeter. The results are not influenced by the flow rates employed.

  19. Basalt Waste Isolation Project. Quarterly report, July 1, 1979-September 30, 1979

    Energy Technology Data Exchange (ETDEWEB)

    Deju, R.A.

    1979-10-01

    Progress in various areas of the Basalt Waste Isolation Project during the last quarter is reported. Systems integration, licensing, geologic activities, hydrology, borehole studies, geophysical logging, engineered barriers, test facilities, testing of canisters, and selection process for architect-engineer services for repository conceptual design are discussed. (DC)

  20. Sharing the Dragon’s Teeth: Terrorist Groups and the Exchange of New Technologies

    Science.gov (United States)

    2007-01-01

    film canisters. As one Bangkok-based journalist put it, “This is sophisticated stuff and certainly not something that you would typically associate...the joint U.S.-Filipino Balikitan Exercises (first initiated in 2002 and then reenacted in 2004) and are currently directed at flushing out residual

  1. STS 133 Return Samples: Air Quality Aboard Shuttle (STS-133) and International Space Station (ULFS)

    Science.gov (United States)

    James, John T.

    2011-01-01

    The toxicological assessments of 2 canisters (mini-GSC or GSCs) from the Shuttle are reported. Analytical methods have not changed from earlier reports. The percent recoveries of the 3 surrogates (C-13-acetone, fluorobenzene, and chlorobenzene) from the 2 Shuttle GSCs averaged 86, 100, and 87, respectively. Based on the end-of-mission sample, the Shuttle atmosphere was acceptable for human respiration.

  2. Low activity aluminum blanket

    Energy Technology Data Exchange (ETDEWEB)

    Benenati, R.; Tichler, P.; Powell, J.R.

    1976-03-01

    The basic design of the breeding blanket consists of cylindrical aluminium canisters filled with a ceramic bed of moderating, shielding, and breeding materials all suitably cooled. A technical analysis of the blanket for an EPR design is given. Activation studies are presented. The effect of pulsed magnetic fields on module structure is investigated. (MOW)

  3. Spontaneous emergence, imitation and spread of alternative foraging techniques among groups of vervet monkeys.

    Directory of Open Access Journals (Sweden)

    Erica van de Waal

    Full Text Available Animal social learning has become a subject of broad interest, but demonstrations of bodily imitation in animals remain rare. Based on Voelkl and Huber's study of imitation by marmosets, we tested four groups of semi-captive vervet monkeys presented with food in modified film canisters ("aethipops'. One individual was trained to take the tops off canisters in each group and demonstrated five openings to them. In three groups these models used their mouth to remove the lid, but in one of the groups the model also spontaneously pulled ropes on a canister to open it. In the last group the model preferred to remove the lid with her hands. Following these spontaneous differentiations of foraging techniques in the models, we observed the techniques used by the other group members to open the canisters. We found that mouth opening was the most common technique overall, but the rope and hands methods were used significantly more in groups they were demonstrated in than in groups where they were not. Our results show bodily matching that is conventionally described as imitation. We discuss the relevance of these findings to discoveries about mirror neurons, and implications of the identity of the model for social transmission.

  4. Removing Solids From Supercritical Water

    Science.gov (United States)

    Hong, Glenn T.

    1992-01-01

    Apparatus removes precipitated inorganic salts and other solids in water-recycling process. Designed for use with oxidation in supercritical water which treats wastes and yields nearly pure water. Heating coils and insulation around vessel keep it hot. Locking bracket seals vessel but allows it to be easily opened for replacement of filled canisters.

  5. Safety and Suitability for Service Assessment Testing for Surface and Underwater Launched Munitions

    Science.gov (United States)

    2014-12-05

    data, is required accounting for different vehicles, stowage configurations (e.g., in racks , launch tubes, canisters, or rails on a turret) and launch...PACKAGING. The munition test configuration should be tailored to the appropriate shipping, handling, storage, and operational deployment ( stowage ...item configuration examples. For many munitions, the shipping and storage container serves as the stowage and launch tube, hereafter designated as

  6. A rotor-mounted digital instrumentation system for helicopter blade flight research measurements

    Science.gov (United States)

    Knight, V. H., Jr.; Haywood, W. S., Jr.; Williams, M. L.

    1978-01-01

    A rotor mounted flight instrumentation system developed for helicopter rotor blade research is described. The system utilizes high speed digital techniques to acquire research data from miniature pressure transducers on advanced rotor airfoils which are flight tested on an AH-1G helicopter. The system employs microelectronic pulse code modulation (PCM) multiplexer digitizer stations located remotely on the blade and in a hub mounted metal canister. As many as 25 sensors can be remotely digitized by a 2.5 mm thick electronics package mounted on the blade near the tip to reduce blade wiring. The electronics contained in the canister digitizes up to 16 sensors, formats these data with serial PCM data from the remote stations, and transmits the data from the canister which is above the plane of the rotor. Data are transmitted over an RF link to the ground for real time monitoring and to the helicopter fuselage for tape recording. The complete system is powered by batteries located in the canister and requires no slip rings on the rotor shaft.

  7. Conceptual design report for immobilized high-level waste interim storage facility (Phase 1)

    Energy Technology Data Exchange (ETDEWEB)

    Burgard, K.C.

    1998-04-09

    The Hanford Site Canister Storage Building (CSB Bldg. 212H) will be utilized to interim store Phase 1 HLW products. Project W-464, Immobilized High-Level Waste Interim Storage, will procure an onsite transportation system and retrofit the CSB to accommodate the Phase 1 HLW products. The Conceptual Design Report establishes the Project W-464 technical and cost basis.

  8. Conceptual design report for immobilized high-level waste interim storage facility (Phase 1)

    Energy Technology Data Exchange (ETDEWEB)

    Burgard, K.C.

    1998-06-02

    The Hanford Site Canister Storage Building (CSB Bldg. 212H) will be utilized to interim store Phase 1 HLW products. Project W-464, Immobilized High-Level Waste Interim Storage, will procure an onsite transportation system and retrofit the CSB to accommodate the Phase 1 HLW products. The Conceptual Design Report establishes the Project W-464 technical and cost basis.

  9. Disposability Assessment: Aluminum-Based Spent Nuclear Fuel Forms

    Energy Technology Data Exchange (ETDEWEB)

    Vinson, D.W.

    1998-11-06

    This report provides a technical assessment of the Melt-Dilute and Direct Al-SNF forms in disposable canisters with respect to meeting the requirements for disposal in the Mined Geologic Disposal System (MGDS) and for interim dry storage in the Treatment and Storage Facility (TSF) at SRS.

  10. Interim report on safety assessment of spent fuel disposal TILA-96

    Energy Technology Data Exchange (ETDEWEB)

    Vieno, T.; Nordman, H. [VTT Energy, Espoo (Finland)

    1996-12-01

    The TILA-96 study, a continuation and update of the TVO-92 safety analysis for Finnish radioactive waste disposal, confirms that the planned system for spent fuel disposal fulfills the proposed safety criteria. Provided that no major disruptive event hits the repository, initially intact copper canisters preserve their integrity for millions of years and no significant amount of radioactive substances will ever escape from the repository. Impacts of potential canister failures have been analysed employing conservative assumptions, models and data. In the case of single canister failures, the results show that the margin to the proposed regulatory criteria is more than three orders of magnitude in the dose rate and more than four orders of magnitude in the release rates into the biosphere. Even in the extreme cases, where all 1500 canisters are assumed to be initially defective or to disappear simultaneously at 10 000 years in the worst possible location in the repository, all the proposed safety criteria would be passed. When realistic modelling and data are used in the consequence analyses, the results show negligible releases and doses. (refs.).

  11. 77 FR 38789 - Notice of Availability of Draft Waste Incidental to Reprocessing Evaluation for the Concentrator...

    Science.gov (United States)

    2012-06-29

    ... vitrifying waste from reprocessing of spent nuclear fuel and certain treatment material at the West Valley... canisters where the mixture hardened into a solid glass waste form. DOE operated the vitrification system... of Chapter IV of DOE Manual 435.1-1, provided the waste will be incorporated in a solid physical...

  12. The transportation of fine arts materials aboard the space shuttle Columbia. GAS payload No. 481: Vertical horizons

    Science.gov (United States)

    Kurtz, Ellery; Wishnow, Howard

    1988-01-01

    The Vertical Horizons experiment represents an initial investigation into the transportation of fine arts materials aboard a space shuttle. Within the confines of a GAS canister, artist quality fine arts materials were packaged and exposed to the rigors of space flight in an attempt to identify adverse effects.

  13. An improved, automated whole air sampler and gas chromatography mass spectrometry analysis system for volatile organic compounds in the atmosphere

    Science.gov (United States)

    Lerner, Brian M.; Gilman, Jessica B.; Aikin, Kenneth C.; Atlas, Elliot L.; Goldan, Paul D.; Graus, Martin; Hendershot, Roger; Isaacman-VanWertz, Gabriel A.; Koss, Abigail; Kuster, William C.; Lueb, Richard A.; McLaughlin, Richard J.; Peischl, Jeff; Sueper, Donna; Ryerson, Thomas B.; Tokarek, Travis W.; Warneke, Carsten; Yuan, Bin; de Gouw, Joost A.

    2017-01-01

    Volatile organic compounds were quantified during two aircraft-based field campaigns using highly automated, whole air samplers with expedited post-flight analysis via a new custom-built, field-deployable gas chromatography-mass spectrometry instrument. During flight, air samples were pressurized with a stainless steel bellows compressor into electropolished stainless steel canisters. The air samples were analyzed using a novel gas chromatograph system designed specifically for field use which eliminates the need for liquid nitrogen. Instead, a Stirling cooler is used for cryogenic sample pre-concentration at temperatures as low as -165 °C. The analysis system was fully automated on a 20 min cycle to allow for unattended processing of an entire flight of 72 sample canisters within 30 h, thereby reducing typical sample residence times in the canisters to less than 3 days. The new analytical system is capable of quantifying a wide suite of C2 to C10 organic compounds at part-per-trillion sensitivity. This paper describes the sampling and analysis systems, along with the data analysis procedures which include a new peak-fitting software package for rapid chromatographic data reduction. Instrument sensitivities, uncertainties and system artifacts are presented for 35 trace gas species in canister samples. Comparisons of reported mixing ratios from each field campaign with measurements from other instruments are also presented.

  14. An update of the state-of-the-art report on the corrosion of copper under expected conditions in a deep geologic repository

    Energy Technology Data Exchange (ETDEWEB)

    King, F. [Integrity Corrosion Consulting Limited (Canada); Lilja, C. [Svensk Kaernbraenslehantering AB, Stockholm (Sweden); Pedersen, K. [Microbial Analytics Sweden AB, Molnlycke (Sweden); Pitkaenen, P.; Vaehaenen, M.

    2012-07-15

    Copper has been the corrosion barrier of choice for the canister in the Swedish and Finnish, nuclear waste disposal programmes for over 30 years. During that time many studies have been carried out on the corrosion behaviour of copper under conditions likely to exist in an underground nuclear waste repository located in the Fenno-Scandian bedrock. This review is a summary of what has been learnt about the long-term behaviour of the corrosion barrier during this period and what the implications of this knowledge are for the predicted service life of the canisters. The review is based on the existing knowledge from various nuclear waste management programs around the world and from the open literature. Various areas are considered: the expected evolution of the geochemical and microbiological conditions in the groundwater and of the repository environment, the thermodynamics of copper corrosion, corrosion during the operational phase and in the bentonite prior to saturation of the buffer by groundwater, general and localised corrosion following saturation of the compacted bentonite buffer, stress corrosion cracking, radiation effects, the implications of corrosion on the service life of the canister, and areas for further study. This report is an updated version of that originally published in 2001/2002. The original material has been supplemented by information from studies carried out over the last decade. The conclusion drawn from this review is that the original prediction made in 1978 of canister lifetimes exceeding 100,000 years remains valid. (orig.)

  15. Mobile System for Precise Aero Delivery with Global Reach Network Capability

    Science.gov (United States)

    2009-08-30

    autonomous powered paraglider (LEAPP) developed under contract with DARPA (Fig.9b). a) b) Fig. 9. Onyx ML with mock sensor payload release...canister (a) and powered paraglider LEAPP (b). Compared to the Snowflake ADS / Arcturus UAS this system has: - Moth mode control for a parafoil

  16. Field trials with plant products to protect stored cowpea against insect damage

    NARCIS (Netherlands)

    Boeke, S.J.; Kossou, D.K.; Huis, van A.; Loon, van J.J.A.; Dicke, M.

    2004-01-01

    Plant products were evaluated under field conditions for their efficacy as insecticides against the cowpea beetle, Callosobruchus maculatus, on stored cowpea. Seeds, mixed with finely ground clay and three volatile oils were stored in air-tight jerry-cans and canisters. Pods were treated with leaf p

  17. Spent nuclear fuel project product specification

    Energy Technology Data Exchange (ETDEWEB)

    PAJUNEN, A.L.

    1999-02-25

    This document establishes the limits and controls for the significant parameters that could potentially affect the safety and/or quality of the Spent Nuclear Fuel (SNF) packaged for processing, transport, and storage. The product specifications in this document cover the SNF packaged in Multi-Canister Overpacks to be transported throughout the SNF Project.

  18. Spent Nuclear Fuel (SNF) Project Product Specification

    Energy Technology Data Exchange (ETDEWEB)

    PAJUNEN, A.L.

    2000-01-20

    This document establishes the limits and controls for the significant parameters that could potentially affect the safety and/or quality of the Spent Nuclear Fuel (SNF) packaged for processing, transport, and storage. The product specifications in this document cover the SNF packaged in Multi-Canister Overpacks to be transported throughout the SNF Project.

  19. Geochemical assessment of nuclear waste isolation. Report of activities during fiscal year 1982

    Energy Technology Data Exchange (ETDEWEB)

    1983-07-01

    The status of the following investigations is reported: canister/overpack-backfill chemical interactions and mechanisms; backfill and near-field host rock chemical interactions mechanisms; far-field host rock geochemical interactions; verification and improvement of predictive algorithms for radionuclide migration; and geologic systems as analogues for long-term radioactive waste isolation.

  20. 200 Area Interim Storage Area Technical Safety Requirements

    Energy Technology Data Exchange (ETDEWEB)

    CARRELL, R.D.

    2000-03-15

    The 200 Area Interim Storage Area Technical Safety Requirements define administrative controls and design features required to ensure safe operation during receipt and storage of canisters containing spent nuclear fuel. This document is based on the 200 Area Interim Storage Area, Annex D, Final Safety Analysis Report which contains information specific to the 200 Area Interim Storage Area.

  1. U.S. Army CERDEC Field Evaluation and Testing of Soldier and Man-Portable Fuel Cell Power Sources. CERDEC C2D Army Power Division, Power Sources Branch

    Science.gov (United States)

    2009-11-19

    60W SOFC Developed with CERDEC and DARPA Rated 60W continuous (100 W Peak) Solid Oxide Fuel Cell ( SOFC ) Fuel: Commercial Propane Canisters Dimensions...Technology Current Efforts Protonex P-125a SOFC Developed with CERDEC & ARO Rated 100W continuous Solid Oxide Fuel Cell ( SOFC ) Fuel: 100% Pure Propane

  2. Microbial analyses of groundwater and surfaces during the retrieval of experiment 3, A04, in MINICAN

    Energy Technology Data Exchange (ETDEWEB)

    Hallbeck, Lotta; Edlund, Johanna; Eriksson, Lena [Microbial Analytics Sweden AB, Moelnlycke (Sweden)

    2011-12-15

    The MINICAN project is located at the depth of 450 m in the Aespoe Hard Rock Laboratory (HRL) research tunnel. The aim of the project was to study corrosion of the cast iron inserts if a hole is introduced in the outer copper-canister. The experimental part of MINICAN started in 2007 and consists of five different experiment canisters (Table 1.1), denoted experiment A02-A06. Four of the MINICAN test copper canisters are surrounded by bentonite in a support steel cage, of which the bentonite in experiment A05 is fully compacted according to the KBS-3 approach (dry density 1,600 kg m{sup -3}) and experiments A02-A04 are compacted with bentonite to a lower density than will be used (dry density 1,300 kg m{sup -3}). Experiment A06 has no bentonite. In all the MINICAN copper canisters, holes with a diameter of 1 mm have been drilled to allow Aspo groundwater to come in contact with the interior cast iron inserts. This is done to mimic real accidental leakage during the KBS-3 type of long-time spent nuclear fuel storage. The project has been described in 1068871- Project Plan MINICAN, in AP TD F77.3-05-001, AP TD F77.3.08-44 and in AP TD F77.3.

  3. Technical support for GEIS: radioactive waste isolation in geologic formations. Volume 22. Nuclear considerations for repository design

    Energy Technology Data Exchange (ETDEWEB)

    1978-04-01

    This volume, Y/OWI/TM-36/22, ''Nuclear Considerations for Repository Design,'' is one of a 23-volume series, ''Technical Support for GEIS: Radioactive Waste Isolation in Geologic Formations,'' Y/OWI/TM-36, which supplements the ''Contribution to Draft Generic Environmental Impact Statement on Commercial Waste Management: Radioactive Waste Isolation in Geologic Formations,'' Y/OWI/TM-44. The series provides a more complete technical basis for the preconceptual designs, resource requirements, and environmental source terms associated with isolating commercial LWR wastes in underground repositories in salt, granite, shale and basalt. Wastes are considered from three fuel cycles: uranium and plutonium recycling, no recycling of spent fuel and uranium-only recycling. Included in this volume are baseline design considerations such as characteristics of canisters, drums, casks, overpacks, and shipping containers; maximum allowable and actual decay-heat levels; and canister radiation levels. Other topics include safeguard and protection considerations; occupational radiation exposure including ALARA programs; shielding of canisters, transporters and forklift trucks; monitoring considerations; mine water treatment; canister integrity; and criticality calculations.

  4. High-level waste vitrification by spray calcination/in-can melting

    Energy Technology Data Exchange (ETDEWEB)

    Larson, D.E.; Bonner, W.F. (comp.)

    1976-11-01

    Federal regulations require that high-level liquid waste (HLLW) be converted to a solid for custody in a Federal repository. The Spray Solidification/In-Can Melting process has been developed and is being demonstrated for commercial application. The bases used are similar to those of the NFS plant and to anticipated regulations for waste canister receipt at a Federal repository. The reference NFS flowsheet combines plant HA Column Wastes, Low-Level Wastes, and various HLLW process recycle streams to produce a borosilicate glass. After the canister is filled and sealed, the lid weld will be inspected and decontaminated. Equipment and instrumentation for feed supply to the calciner, calcination, melting, welding, weld inspection, canister decontamination, and in-cell canister storage are being designed and demonstrated. Preliminary facility layouts, equipment design data, and instrumentation needs are provided for major process equipment systems. Additional demonstration work is being performed to verify and complete the plant scale equipment design, including full-scale nonradioactive equipment testing, nonradioactive facility mockup for equipment remote operation and maintenance demonstration, and pilot plant production of waste glass from commercial fuel HLLW. The technology for spray calcination and in-can melting is ready for commercial application. Required additional work is described. A preliminary evaluation is made of materials that may be released from the process from normal and abnormal operations in the facility. 34 figures, 20 tables. (DLC)

  5. Evaluation of closed incision management with negative pressure wound therapy (CIM): hematoma/seroma and involvement of the lymphatic system.

    Science.gov (United States)

    Kilpadi, Deepak V; Cunningham, Mark R

    2011-01-01

    The objective of this porcine study was to evaluate the effect of closed incision management with negative pressure wound therapy (CIM) on hematoma/seroma formation, fluid removal into the CIM canister, and involvement of the lymphatic system. In each swine (n = 8), two sets of ventral contralateral subcutaneous dead spaces with overlying sutured incisions were created. Stable isotope-labeled nanospheres were introduced into each subcutaneous dead space. Each contralateral incision was assigned to CIM (continuous -125 mmHg negative pressure) and control (semipermeable film dressing), respectively. Following 4 days of therapy, hematoma/seroma was weighed, total fluid volume in canisters was measured, five pre-identified lymph nodes were harvested, and five key organs were biopsied. There was 25 ± 8 g (standard error [SE]) (63%) less hematoma/seroma in CIM sites compared to control sites (p = 0.002), without any fluid collection in the CIM canister. In lymph nodes, there were ∼60 μg (∼50%) more 30- and 50-nm nanospheres from CIM sites than from control sites (p = 0.04 and 0.05, respectively). There was significantly greater nanosphere incidence from CIM sites than from control sites in lungs, liver, and spleen (p CIM significantly decreased hematoma/seroma levels without fluid collection in the canister, which may be explained by increased lymph clearance.

  6. Modeling transient heat transfer in nuclear waste repositories.

    Science.gov (United States)

    Yang, Shaw-Yang; Yeh, Hund-Der

    2009-09-30

    The heat of high-level nuclear waste may be generated and released from a canister at final disposal sites. The waste heat may affect the engineering properties of waste canisters, buffers, and backfill material in the emplacement tunnel and the host rock. This study addresses the problem of the heat generated from the waste canister and analyzes the heat distribution between the buffer and the host rock, which is considered as a radial two-layer heat flux problem. A conceptual model is first constructed for the heat conduction in a nuclear waste repository and then mathematical equations are formulated for modeling heat flow distribution at repository sites. The Laplace transforms are employed to develop a solution for the temperature distributions in the buffer and the host rock in the Laplace domain, which is numerically inverted to the time-domain solution using the modified Crump method. The transient temperature distributions for both the single- and multi-borehole cases are simulated in the hypothetical geological repositories of nuclear waste. The results show that the temperature distributions in the thermal field are significantly affected by the decay heat of the waste canister, the thermal properties of the buffer and the host rock, the disposal spacing, and the thickness of the host rock at a nuclear waste repository.

  7. Science, Society, and America's Nuclear Waste: The Waste Management System, Unit 4. Teacher Guide. Second Edition.

    Science.gov (United States)

    Department of Energy, Washington, DC. Office of Civilian Radioactive Waste Management, Washington, DC.

    This guide is Unit 4 of the four-part series, Science, Society, and America's Nuclear Waste, produced by the U.S. Department of Energy's Office Civilian Radioactive Waste Management. The goal of this unit is to explain how transportation, a geologic repository, and the multi-purpose canister will work together to provide short-term and long-term…

  8. Copper Corrosion in Nuclear Waste Disposal: A Swedish Case Study on Stakeholder Insight

    Science.gov (United States)

    Andersson, Kjell

    2013-01-01

    The article describes the founding principles, work program, and accomplishments of a Reference Group with both expert and layperson stakeholders for the corrosion of copper canisters in a proposed deep repository in Sweden for spent nuclear fuel. The article sets the Reference Group as a participatory effort within a broader context of…

  9. Natural analogues for expansion due to the anaerobic corrosion of ferrous materials

    Energy Technology Data Exchange (ETDEWEB)

    Smart, N.R.; Adams, R. [Serco Assurance, Culham Science Centre (United Kingdom)

    2006-10-15

    In Sweden, spent nuclear fuel will be encapsulated in sealed cylindrical canisters, consisting of a cast iron insert and a copper outer container. The canisters will be placed in a deep geologic repository and surrounded by bentonite. If a breach of the outer copper container were to occur the cast iron insert would undergo anaerobic corrosion, forming a magnetite film whose volume would be greater than that of the base metal. In principle there is a possibility that accumulation of iron corrosion product could cause expansion of the copper canister. Anaerobic corrosion rates are very slow, so in the work described in this report reference was made to analogous materials that had been corroding for long periods in natural anoxic aqueous environments. The report considers the types of naturally occurring environments that may give rise to anoxic environments similar to deep geological groundwater and where ferrous materials may be found. Literature information regarding the corrosion of iron archaeological artefacts is summarised and a number of specific archaeological artefacts containing iron and copper corroding in constrained geometries in anoxic natural waters are discussed in detail. No evidence was obtained from natural analogues which would suggest that severe damage is likely to occur to the SKB waste canister design as a result of expansive corrosion of cast iron under repository conditions.

  10. A Structural Analytic Evaluation of a Connote Pad In a Spent Fuel Dry Storage Cask

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong Hak; Seo, Ki Seog; Lee, Ju Chan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Yeon Do; Cho, Chun Hyung; Lee, Dae Ki [Nuclear Environment Technology Institute, Daejeon (Korea, Republic of)

    2006-06-15

    A spent fuel storage cask is required to prove the safety of a canister under a hypothetical accidental drop condition. A hypothetical accidental drop condition means that a canister is assumed to be a lee drop on to a pad of the storage cask during loading it into a storage cask. A pad of the storage cask absorbs shock to maintain the structural integrities of a canister under a hypothetical accidental drop condition. In this paper a finite element analysis for various pad structures was carried out to improve the structural integrity of a canister under a hypothetical accidental drop condition. A pad of a storage cask was designed a steel structure with concrete. The 1/4 height of a pad was modified with a structure composed of a steel and a polyurethane foam as a impact limiter. The effect of a shape of a steel structure was studied. The effects of the thickness of a steel structure and the density of a polyurethane foam was also studied.

  11. 77 FR 64765 - Airworthiness Directives; Airbus Airplanes

    Science.gov (United States)

    2012-10-23

    ..., Operational Check of Centre Tank Fuel Pump GFI, of the Airbus A318/A319/A320 Aircraft Maintenance Manual. Note...: We propose to adopt a new airworthiness directive (AD) for all Airbus Model A318, A319, and A320... on A320 family aeroplanes, for which the canisters become uncovered during normal operation,...

  12. 40 CFR 86.1824-08 - Durability demonstration procedures for evaporative emissions.

    Science.gov (United States)

    2010-07-01

    ... accumulation must be conducted using the SRC or any road cycle approved under the provisions of § 86.1823(e)(1... causes of evaporative emission deterioration: (1) Cycling of canister loading due to diurnal and... guidance see 40 CFR 86.1824-01(d). (i) If EPA determines based on IUVP data or other information that...

  13. 40 CFR 86.1825-08 - Durability demonstration procedures for refueling emissions.

    Science.gov (United States)

    2010-07-01

    ... must be conducted using the SRC or a road cycle approved under the provisions of § 86.1823(e)(1). (2...: (1) Cycling of canister loading due to diurnal and refueling events; (2) Use of various commercially... paragraph. (h) Emission component durability. . For guidance see 40 CFR 86.1845-01 (e). (i) If...

  14. Comparison of volatiles and mosquito capture efficacy for three carbohydrate sources in a yeast-fermentation CO2 generator

    Science.gov (United States)

    Mosquito surveillance in remote areas with limited access to canisters of CO2 or dry ice will benefit from an effective alternative CO2 source. In this study, we document the differences in mosquito and non-mosquito capture rates from CO2 baited (dry ice or yeast fermentation of carbohydrates) CDC t...

  15. Comparison of carbohydrate sources in yeast-fermentation CO2 generators for mosquito surveillance

    Science.gov (United States)

    Mosquito surveillance in remote areas with limited access to canisters of CO2 or dry ice will benefit from an effective alternative CO2 source, such as the natural production of CO2 from yeast fermentation of several carbohydrate sources. In this study, we document the differences in mosquito and n...

  16. STS 119 Return Samples: Assessment of Air Quality aboard the Shuttle (STS-119) and International Space Station (15A)

    Science.gov (United States)

    James, John T.

    2009-01-01

    The toxicological assessments of 2 grab sample canisters (GSCs) from the Shuttle are reported. Analytical methods have not changed from earlier reports. The recoveries of the 3 surrogates (C-13-acetone, fluorobenzene, and chlorobenzene) from the 2 GSCs averaged 106, 106, and 101 %,respectively. Based on the end-of-mission sample, the Shuttle atmosphere was acceptable for human respiration.

  17. STS 127 Return Samples: Assessment of Air Quality aboard the Shuttle (STS-127) and International Space Station (2J/A)

    Science.gov (United States)

    James, John T.

    2010-01-01

    The toxicological assessments of 2 grab sample canisters (GSCs) from the Shuttle are reported. The toxicological assessment of 9 GSCs and 6 pairs of formaldehyde badges from the ISS is also reported. Other than a problem with traces of acrolein in the samples, the air quality was acceptable for respiration.

  18. STS 120 Return Samples: Assessment of Air Quality Aboard the Shuttle (STS-120) and International Space Station (10A)

    Science.gov (United States)

    James, John T.

    2008-01-01

    The toxicological assessments of 2 grab sample canisters (GSCs) from the Shuttle are reported. Formaldehyde badges were not used. Analytical methods have not changed from earlier reports. The recoveries of the 3 surrogates (C-13-acetone, fluorobenzene, and chlorobenzene) from the 2 GSCs averaged 111, 82, and 78%, respectively. The Shuttle atmosphere was acceptable for human respiration.

  19. RD and D-Programme 2001. Programme for research, development and demonstration of methods for the management and disposal of nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-09-01

    An overall goal for SKB is to start the initial operation of a deep repository for spent fuel in 2015. This presumes that site investigations have been commenced at the beginning of 2002 and that the different phases have been executed without major changes. The encapsulation plant should be ready to start roughly one year before the deep repository is finished. Spent fuel is the waste that is to be isolated in the deep repository. Various processes will with time alter the conditions in the fuel and in the voids of the canister. Many of these process only occur if the isolation of the canister is breached and water enters the canister. Radiolysis of water is an example of such a process, which can in turn influence the chemical conditions in the canister. Water in the canister can also cause corrosion of the fuel's cladding tubes. If water comes into contact with the fuel it can lead to dissolution of radionuclides. Dissolved radionuclides can diffuse in the water and thereby escape from a damaged canister. Fuel dissolution is a priority area in RDandD Programme 2001. Large resources are being devoted to studies of copper corrosion and stress corrosion cracking in the copper canister. SKB will also investigate the long-term safety of a canister type with a slightly thinner shell but a heavier-duty insert. The buffer of bentonite clay is supposed to protect the canister mechanically against minor rock movements. It is also supposed to retard solute transport. The initial evolution of the buffer is studied in the Aespoe HRL and by means of models. The long-term evolution of the backfill is controlled by largely the same processes as in the buffer. The backfill is more sensitive to saline water than the more compacted buffer. Several processes in the geosphere are important for the safety assessment, such as groundwater flow, earthquakes, microbial processes and matrix diffusion. The models for groundwater flow will be further refined in order to handle the

  20. SLUDGE TREATMENT PROJECT KOP CONCEPTUAL DESIGN CONTROL DECISION REPORT

    Energy Technology Data Exchange (ETDEWEB)

    CARRO CA

    2010-03-09

    This control decision addresses the Knock-Out Pot (KOP) Disposition KOP Processing System (KPS) conceptual design. The KPS functions to (1) retrieve KOP material from canisters, (2) remove particles less than 600 {micro}m in size and low density materials from the KOP material, (3) load the KOP material into Multi-Canister Overpack (MCO) baskets, and (4) stage the MCO baskets for subsequent loading into MCOs. Hazard and accident analyses of the KPS conceptual design have been performed to incorporate safety into the design process. The hazard analysis is documented in PRC-STP-00098, Knock-Out Pot Disposition Project Conceptual Design Hazard Analysis. The accident analysis is documented in PRC-STP-CN-N-00167, Knock-Out Pot Disposition Sub-Project Canister Over Lift Accident Analysis. Based on the results of these analyses, and analyses performed in support of MCO transportation and MCO processing and storage activities at the Cold Vacuum Drying Facility (CVDF) and Canister Storage Building (CSB), control decision meetings were held to determine the controls required to protect onsite and offsite receptors and facility workers. At the conceptual design stage, these controls are primarily defined by their safety functions. Safety significant structures, systems, and components (SSCs) that could provide the identified safety functions have been selected for the conceptual design. It is anticipated that some safety SSCs identified herein will be reclassified based on hazard and accident analyses performed in support of preliminary and detailed design.

  1. An update of the state-of-the-art report on the corrosion of copper under expected conditions in a deep geologic repository

    Energy Technology Data Exchange (ETDEWEB)

    King, Fraser (Integrity Corrosion Consulting Limited (Canada)); Lilja, Christina (Svensk Kaernbraenslehantering AB (Sweden)); Pedersen, Karsten (Microbial Analytics Sweden AB (Sweden)); Pitkaenen, Petteri; Vaehaenen, Marjut (Posiva Oy (Finland))

    2010-12-15

    Copper has been the corrosion barrier of choice for the canister in the Swedish and Finnish, nuclear waste disposal programmes for over 30 years. During that time many studies have been carried out on the corrosion behaviour of copper under conditions likely to exist in an underground nuclear waste repository located in the Fenno-Scandian bedrock. This review is a summary of what has been learnt about the long-term behaviour of the corrosion barrier during this period and what the implications of this knowledge are for the predicted service life of the canisters. The review is based on the existing knowledge from various nuclear waste management programs around the world and from the open literature. Various areas are considered: the expected evolution of the geochemical and microbiological conditions in the groundwater and of the repository environment, the thermodynamics of copper corrosion, corrosion during the operational phase and in the bentonite prior to saturation of the buffer by groundwater, general and localised corrosion following saturation of the compacted bentonite buffer, stress corrosion cracking, radiation effects, the implications of corrosion on the service life of the canister, and areas for further study. This report is an updated version of that originally published in 2001/2002. The original material has been supplemented by information from studies carried out over the last decade. The conclusion drawn from this review is that the original prediction made in 1978 of canister lifetimes exceeding 100,000 years remains valid

  2. Full perimeter intersection criteria. Definitions and implementations in SR-Site

    Energy Technology Data Exchange (ETDEWEB)

    Munier, Raymond

    2010-03-15

    Large fractures intersecting canisters have potential to reactivate due to nearby earthquakes and thereby jeopardising the canister/buffer integrity. The use of full perimeter intersection (FPI) as a proxy for fracture size has been explored within SKB since late 2004. A methodology to use FPI as a deposition hole rejection criterion was first reported in and the concept has successively matured ever since. As a response to feedback obtained from various instances of our organisation and external reviewers, additional analyses and benchmarks were reported in 2007. An analytical solution to the canister/fracture intersection probability was introduced, which enabled us to benchmark various aspects of the FPI simulations. The methodology was first applied within the framework of SR-CAN to compute the number of potentially critical canister positions and soon after, based on preliminary DFN models, as one of many prerequisites for repository design. The methodology and simulation logic has evolved substantially ever since it was originally reported in 2006 and the present report is intended to entirely replace previous reports on this subject, to thereby provide the interested reader with an description of the modelling procedure, prerequisites and limitations. As a consequence thereof, major portions of previous reports are repeated herein, though we occasionally refer to these reports for comparative purposes. Furthermore, as the final versions of the site descriptive models have been reported we find it convenient to, within this report, also apply the methodology using the most actual site specific fracture data

  3. Staging and storage facility feasibility study. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Swenson, C.E. [Westinghouse Hanford Co., Richland, WA (United States)

    1995-02-01

    This study was performed to investigate the feasibility of adapting the design of the HWVP Canister Storage Building (CSB) to meet the needs of the WHC Spent Nuclear Fuel Project for Staging and Storage Facility (SSF), and to develop Rough Order of Magnitude (ROM) cost and schedule estimates.

  4. Project JADE. Comparison of repository systems. Executive summary of results

    Energy Technology Data Exchange (ETDEWEB)

    Sandstedt, H. [Scandiaconsult Sverige AB, Stockholm (Sweden); Pers, K.; Birgersson, Lars [Kemakta Konsult AB, Stockholm (Sweden); Ageskog, L. [SWECO VBB VIAK AB, Stockholm (Sweden); Munier, R. [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden)

    2001-12-01

    KBS-3 has since 1984 been the reference method for disposal of spent fuel in Sweden. Several other methods like WP-Cave, Very Deep Holes and Very Long Holes have been evaluated and compared with KBS-3. Though the methods have been judged to have a high safety potential, KBS-3 has been shown to provide advantages in the combined judgement of 'long-term performance and safety', 'technology' and 'costs'. In the present study, different variants of the KBS-3 method have been analysed and compared with the reference concept KBS-3 V (V for vertical). The variants are: KBS-3 H (H for horizontal) and MLH (medium long holes) - with canisters in a horizontal position, single or in a row respectively. The comparison has been carried out separately for the interim items 'technology', 'long-term performance and safety' and 'costs' respectively. The outcome in each of these comparisons have finally been combined in a ranking. This ranking placed KBS-3 V in the top followed by MLH and KBS-3 H. Vertical deposition of a single canister in one deposition hole, KBS-3 V, is robust as gravity is used for lowering the canister and the bentonite into the deposition hole and since each canister has its own barrier in the near field, which reduces the risk for interference between canisters. The drawback for MLH is the uncertainty about the emplacement technique as well as the impact of weak rock and water leakage into a long deposition hole for several canisters. The advantage is that a smaller volume of rock has to be excavated. This is positive regarding the long-term performance and safety, environmental impact and costs. KBS-3 H does not have the same positive potential. The conclusion of the JADE study is that KBS-3 V should remain as reference concept, and that MLH should be studied further with the aim of clarifying the technical feasibility of emplacement and the means of handling water inflow. It is recommended that KBS

  5. FUEL HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    Energy Technology Data Exchange (ETDEWEB)

    C.E. Sanders

    2005-06-30

    The purpose of this design calculation is to perform a criticality evaluation of the Fuel Handling Facility (FHF) and the operations and processes performed therein. The current intent of the FHF is to receive transportation casks whose contents will be unloaded and transferred to waste packages (WP) or MGR Specific Casks (MSC) in the fuel transfer bays. Further, the WPs will also be prepared in the FHF for transfer to the sub-surface facility (for disposal). The MSCs will be transferred to the Aging Facility for storage. The criticality evaluation of the FHF features the following: (I) Consider the types of waste to be received in the FHF as specified below: (1) Uncanistered commercial spent nuclear fuel (CSNF); (2) Canistered CSNF (with the exception of horizontal dual-purpose canister (DPC) and/or multi-purpose canisters (MPCs)); (3) Navy canistered SNF (long and short); (4) Department of Energy (DOE) canistered high-level waste (HLW); and (5) DOE canistered SNF (with the exception of MCOs). (II) Evaluate the criticality analyses previously performed for the existing Nuclear Regulatory Commission (NRC)-certified transportation casks (under 10 CFR 71) to be received in the FHF to ensure that these analyses address all FHF conditions including normal operations, and Category 1 and 2 event sequences. (III) Evaluate FHF criticality conditions resulting from various Category 1 and 2 event sequences. Note that there are currently no Category 1 and 2 event sequences identified for FHF. Consequently, potential hazards from a criticality point of view will be considered as identified in the ''Internal Hazards Analysis for License Application'' document (BSC 2004c, Section 6.6.4). (IV) Assess effects of potential moderator intrusion into the fuel transfer bay for defense in depth. The SNF/HLW waste transfer activity (i.e., assembly and canister transfer) that is being carried out in the FHF has been classified as safety category in the &apos

  6. Groundwater chemistry of a nuclear waste reposoitory in granite bedrock

    Energy Technology Data Exchange (ETDEWEB)

    Rydberg, J.

    1981-09-01

    This report concerns the prediction of the maximum dissolution rate for nuclear waste stored in the ground. That information is essential in judging the safety of a nuclear waste repository. With a limited groundwater flow, the maximum dissolution rate coincides with the maximum solubility. After considering the formation and composition of deep granite bedrock groundwater, the report discusses the maximum solubility in such groundwater of canister materials, matrix materials and waste elements. The parameters considered are pH, Eh and complex formation. The use of potential-pH (Pourbaix) diagrams is stressed; several appendixes are included to help in analyzing such diagrams. It is repeatedly found that desirable basic information on solution chemistry is lacking, and an international cooperative research effort is recommended. The report particularly stresses the lack of reliable data about complex formation and hydrolysis of the actinides. The Swedish Nuclear Fuel Safety (KBS) study has been used as a reference model. Notwithstanding the lack of reliable chemical data, particularly for the actinides and some fission products, a number of essential conclusions can be drawn about the waste handling model chosen by KBS. (1) Copper seems to be highly resistant to groundwater corrosion. (2) Lead and titanium are also resistant to groundwater, but inferior to copper. (3) Iron is not a suitable canister material. (4) Alumina (Al/sub 2/O/sub 3/) is not a suitable canister material if groundwater pH goes up to or above 10. Alumina is superior to copper at pH < 9, if there is a risk of the groundwater becoming oxidizing. (5) The addition of vivianite (ferrous phosphate) to the clay backfill around the waste canisters improves the corrosion resistance of the metal canisters, and reduces the solubility of many important waste elements. This report does not treat the migration of dissolved species through the rock.

  7. Characterization and dosimetry of a practical X-ray alternative to self-shielded gamma irradiators

    Science.gov (United States)

    Mehta, Kishor; Parker, Andrew

    2011-01-01

    The Insect Pest Control Laboratory of the Joint FAO/IAEA Division of Nuclear Techniques in Food and Agriculture recently purchased an X-ray irradiator as part of their programme to develop the sterile insect technique (SIT). It is a self-contained type with a maximum X-ray beam energy of 150 keV using a newly developed 4 π X-ray tube to provide a very uniform dose to the product. This paper describes the results of our characterization study, which includes determination of dose rate in the centre of a canister as well as establishing absorbed dose distribution in the canister. The irradiation geometry consists of five canisters rotating around an X-ray tube—the volume of each canister being 3.5 l. The dose rate at the maximum allowed power of the tube (about 6.75 kW) in the centre of a canister filled with insects (or a simulated product) is about 14 Gy min -1. The dose uniformity ratio is about 1.3. The dose rate was measured using a Farmer type 0.18-cm 3 ionization chamber calibrated at the relevant low photon energies. Routine absorbed dose measurement and absorbed dose mapping can be performed using a Gafchromic® film dosimetry system. The radiation response of Gafchromic film is almost independent of X-ray energy in the range 100-150 keV, but is very sensitive to the surrounding material with which it is in immediate contact. It is important, therefore, to ensure that all absorbed dose measurements are performed under identical conditions to those used for the calibration of the dosimetry system. Our study indicates that this X-ray irradiator provides a practical alternative to self-shielded gamma irradiators for SIT programmes. Food and Agriculture Organization/International Atomic Energy Agency.

  8. International program to study subseabed disposal of high-level radioactive wastes

    Energy Technology Data Exchange (ETDEWEB)

    Carlin, E.M.; Hinga, K.R.; Knauss, J.A.

    1984-01-01

    This report provides an overview of the international program to study seabed disposal of nuclear wastes. Its purpose is to inform legislators, other policy makers, and the general public as to the history of the program, technological requirements necessary for feasibility assessment, legal questions involved, international coordination of research, national policies, and research and development activities. Each of these major aspects of the program is presented in a separate section. The objective of seabed burial, similar to its continental counterparts, is to contain and to isolate the wastes. The subseabed option should not be confuesed with past practices of ocean dumping which have introduced wastes into ocean waters. Seabed disposal refers to the emplacement of solidified high-level radioactive waste (with or without reprocessing) in certain geologically stable sediments of the deep ocean floor. Specially designed surface ships would transport waste canisters from a port facility to the disposal site. Canisters would be buried from a few tens to a few hundreds of meters below the surface of ocean bottom sediments, and hence would not be in contact with the overlying ocean water. The concept is a multi-barrier approach for disposal. Barriers, including waste form, canister, ad deep ocean sediments, will separate wastes from the ocean environment. High-level wastes (HLW) would be stabilized by conversion into a leach-resistant solid form such as glass. This solid would be placed inside a metallic canister or other type of package which represents a second barrier. The deep ocean sediments, a third barrier, are discussed in the Feasibility Assessment section. The waste form and canister would provide a barrier for several hundred years, and the sediments would be relied upon as a barrier for thousands of years. 62 references, 3 figures, 2 tables.

  9. SPE5 Sub-Scale Test Series Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Vandersall, Kevin S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Reeves, Robert V. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); DeHaven, Martin R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Strickland, Shawn L. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-01-14

    A series of 2 SPE5 sub-scale tests were performed to experimentally confirm that a booster system designed and evaluated in prior tests would properly initiate the PBXN-110 case charge fill. To conduct the experiments, a canister was designed to contain the nominally 50 mm diameter booster tube with an outer fill of approximately 150 mm diameter by 150 mm in length. The canisters were filled with PBXN-110 at NAWS-China Lake and shipped back to LLNL for testing in the High Explosives Applications Facility (HEAF). Piezoelectric crystal pins were placed on the outside of the booster tube before filling, and a series of piezoelectric crystal pins along with Photonic Doppler Velocimetry (PDV) probes were placed on the outer surface of the canister to measure the relative timing and magnitude of the detonation. The 2 piezoelectric crystal pins integral to the booster design were also utilized along with a series of either piezoelectric crystal pins or piezoelectric polymer pads on the top of the canister or outside case that utilized direct contact, gaps, or different thicknesses of RTV cushions to obtain time of arrival data to evaluate the response in preparation for the large-scale SPE5 test. To further quantify the margin of the booster operation, the 1st test (SPE5SS1) was functioned with both detonators and the 2nd test (SPE5SS2) was functioned with only 1 detonator. A full detonation of the material was observed in both experiments as observed by the pin timing and PDV signals. The piezoelectric pads were found to provide a greater measured signal magnitude during the testing with an RTV layer present, and the improved response is due to the larger measurement surface area of the pad. This report will detail the experiment design, canister assembly for filling, final assembly, experiment firing, presentation of the diagnostic results, and a discussion of the results.

  10. Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Evolution report

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Paul; Johnson, Lawrence; Snellman, Margit; Pastina, Barbara; Gribi, Peter

    2008-01-15

    The KBS-3 method, based on multiple barriers, is the proposed spent fuel disposal method both in Sweden and Finland. KBS-3H and KBS-3V are the two design alternatives of the KBS-3 method. Posiva and SKB have conducted a joint research, demonstration and development (RDandD) programme in 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to the reference alternative KBS-3V. The overall objectives of the present phase covering the period 2004-2007, have been to demonstrate that the horizontal deposition alternative is technically feasible and to demonstrate that it fulfils the same long-term safety requirements as KBS-3V. The safety studies conducted as part of this programme include a safety assessment of a preliminary design of a KBS-3H repository for spent nuclear fuel located about 400 m underground at the Olkiluoto site, which is the proposed site for a spent fuel repository in Finland. In the KBS-3H design alternative, each canister, with a surrounding layer of bentonite clay, is pre-packaged in a perforated steel cylinder prior to emplacement in the deposition drift; the entire assembly is called the supercontainer. Several supercontainers are positioned along parallel, 100-300 m long deposition drifts, which are sealed following waste emplacement using drift end plugs. Bentonite distance blocks separate the supercontainers, one from another, along the drift. Steel compartment plugs can be used to seal off drift sections with higher inflow, thus isolating the different compartments within the drift. The present report describes the repository evolution in successive time frames, including key uncertainties. The description of evolution starts with the initial conditions at the time of emplacement of the first canisters. The repository evolves through an early, transient phase to a state where evolution is far slower. Particular attention is given to describing the transient phase, since this is where most of the

  11. DESIGN ANALYSIS FOR THE DEFENSE HIGH-LEVEL WASTE DISPOSAL CONTAINER

    Energy Technology Data Exchange (ETDEWEB)

    G. Radulesscu; J.S. Tang

    2000-06-07

    The purpose of ''Design Analysis for the Defense High-Level Waste Disposal Container'' analysis is to technically define the defense high-level waste (DHLW) disposal container/waste package using the Waste Package Department's (WPD) design methods, as documented in ''Waste Package Design Methodology Report'' (CRWMS M&O [Civilian Radioactive Waste Management System Management and Operating Contractor] 2000a). The DHLW disposal container is intended for disposal of commercial high-level waste (HLW) and DHLW (including immobilized plutonium waste forms), placed within disposable canisters. The U.S. Department of Energy (DOE)-managed spent nuclear fuel (SNF) in disposable canisters may also be placed in a DHLW disposal container along with HLW forms. The objective of this analysis is to demonstrate that the DHLW disposal container/waste package satisfies the project requirements, as embodied in Defense High Level Waste Disposal Container System Description Document (SDD) (CRWMS M&O 1999a), and additional criteria, as identified in Waste Package Design Sensitivity Report (CRWMS M&Q 2000b, Table 4). The analysis briefly describes the analytical methods appropriate for the design of the DHLW disposal contained waste package, and summarizes the results of the calculations that illustrate the analytical methods. However, the analysis is limited to the calculations selected for the DHLW disposal container in support of the Site Recommendation (SR) (CRWMS M&O 2000b, Section 7). The scope of this analysis is restricted to the design of the codisposal waste package of the Savannah River Site (SRS) DHLW glass canisters and the Training, Research, Isotopes General Atomics (TRIGA) SNF loaded in a short 18-in.-outer diameter (OD) DOE standardized SNF canister. This waste package is representative of the waste packages that consist of the DHLW disposal container, the DHLW/HLW glass canisters, and the DOE-managed SNF in disposable

  12. High-level waste processing at the Savannah River Site: An update

    Energy Technology Data Exchange (ETDEWEB)

    Marra, J.E.; Bennett, W.M.; Elder, H.H.; Lee, E.D.; Marra, S.L.; Rutland, P.L.

    1997-09-01

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) in Aiken, SC mg began immobilizing high-level radioactive waste in borosilicate glass in 1996. Currently, the radioactive glass is being produced as a ``sludge-only`` composition by combining washed high-level waste sludge with glass frit. The glass is poured in stainless steel canisters which will eventually be disposed of in a permanent, geological repository. To date, DWPF has produced about 100 canisters of vitrified waste. Future processing operations will, be based on a ``coupled`` feed of washed high-level waste sludge, precipitated cesium, and glass frit. This paper provides an update of the processing activities completed to date, operational/flowsheet problems encountered, and programs underway to increase production rates.

  13. Epidemic septic arthritis caused by Serratia marcescens and associated with a benzalkonium chloride antiseptic.

    Science.gov (United States)

    Nakashima, A K; McCarthy, M A; Martone, W J; Anderson, R L

    1987-01-01

    During a 6-week period, 10 patients were admitted to a hospital for treatment of knee or shoulder joint infections due to Serratia species. Isolates from eight patients were identified as Serratia marcescens with identical biochemical characteristics and antibiotic susceptibility patterns. Before the onset of infections, all patients had been treated by two orthopedic surgeons who shared an office. Studies revealed that infections were associated with previous joint injections (P = 4.44 X 10(-5] of methylprednisolone and lidocaine. Environmental cultures revealed that a canister of cotton balls soaked in aqueous benzalkonium chloride and two multiple-dose vials of methylprednisolone previously used by office personnel were contaminated with the epidemic strain of S. marcescens. The canister may have served as a potential reservoir for contamination of sterile solutions and equipment used for joint injections, of skin at the injection site, and of hands of personnel. No further cases occurred after the use of aqueous benzalkonium chloride was discontinued. PMID:3298308

  14. Design package test weights for fuel retrieval system (OCRWM)

    Energy Technology Data Exchange (ETDEWEB)

    TEDESCHI, D.J.

    1999-10-26

    This is a design package that documents the development of test weights used in the Spent Nuclear Fuels subproject Fuel Retrieval System. The K Basins Spent Nuclear Fuel (SNF) project consists of the safe retrieval, preparation, and repackaging of the spent fuel stored at the K East (KE) and K West (KW) Basins for interim safe storage in the Canister Storage Building (CSB). Multi-Canister Overpack (MCO) scrap baskets and fuel baskets will be loaded and weighed under water. The equipment used to weigh the loaded fuel baskets requires daily calibration checks, using test weights traceable to National Institute of Standards Testing (NIST) standards. The test weights have been designated as OCRWM related in accordance with HNF-SD-SNF-RF'T-007 (McCormack).

  15. Waste disposal options report. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    Russell, N.E.; McDonald, T.G.; Banaee, J.; Barnes, C.M.; Fish, L.W.; Losinski, S.J.; Peterson, H.K.; Sterbentz, J.W.; Wenzel, D.R.

    1998-02-01

    Volume 2 contains the following topical sections: estimates of feed and waste volumes, compositions, and properties; evaluation of radionuclide inventory for Zr calcine; evaluation of radionuclide inventory for Al calcine; determination of k{sub eff} for high level waste canisters in various configurations; review of ceramic silicone foam for radioactive waste disposal; epoxides for low-level radioactive waste disposal; evaluation of several neutralization cases in processing calcine and sodium-bearing waste; background information for EFEs, dose rates, watts/canister, and PE-curies; waste disposal options assumptions; update of radiation field definition and thermal generation rates for calcine process packages of various geometries-HKP-26-97; and standard criteria of candidate repositories and environmental regulations for the treatment and disposal of ICPP radioactive mixed wastes.

  16. Modelling the field behaviour of a granular expansive barrier

    Science.gov (United States)

    Alonso, Eduardo; Hoffmann, Christian

    The large scale “Engineered Barrier” (EB) experiment, performed at the Mont Terri Underground Laboratory is described. A coupled hydromechanical model is then used to simulate the test performance. Constitutive parameters for the bentonite granular backfill are based on experimental work described in a companion paper. An elastoplastic model describes the granular fill, while the host rock is simulated by a damage model. Predictions of EDZ development around the tunnel are compared with some indirect measurements. Calculated evolutions of relative humidity and stresses within the buffer are compared with sensor records. Good agreement was found for the bentonite blocks supporting the canister. The granular expansive fill exhibit a more irregular behavior. Calculated displacements of the canister agree in absolute terms with actual measurements.

  17. Swelling of the buffer of KBS-3V deposition hole

    Energy Technology Data Exchange (ETDEWEB)

    Lempinen, A. [Marintel Ky, Turku (Finland)

    2006-12-15

    At the time of the installation of spent nuclear fuel canister in the KBS-3V deposition hole, empty space is left around bentonite buffer for technical reasons. The gap between the buffer and the canister is about 10 mm, and the gap between the buffer and the rock is 30 to 35 mm. In this study, the swelling of the buffer to fill the gaps was simulated, when the gaps are initially filled with water and no external water is available. The model used here is a thermodynamical model for swelling clay, with parameters determined for bentonite. The simulations presented here were performed with Freefem++ software, which is a finite element application for partial differential equations. These equations come from the material model. The simulation results show that the swelling fills the outer gaps in few years, but no significant swelling pressure is generated. For swelling pressure, external water supply is required. (orig.)

  18. Thermal evaluation facility for LMFBR spent fuel transport

    Energy Technology Data Exchange (ETDEWEB)

    Wesley, D.A.

    1980-04-01

    A full-scale mock-up of a 217 pin breeder reactor fuel assembly in a cylindrical pipe was initially designed and constructed by Oak Ridge National Laboratory (ORNL). It was transferred to Sandia where it was extensively redesigned and modified. The 217 pin hexagonal core assembly was installed in a smaller diameter stainless steel pipe which more closely represents the diameter of a shipping canister or shipping cask basket wall. Two-hundred four of the tubes are electrically heated over an active length of 4-feet and the remaining thirteen are instrumented with multiple junction thermocouples which can be traversed axially. Thermocouples and heat-flux gauges are located on the hex core and canister perimeters at several axial locations.

  19. Stochastic image reconstruction for a dual-particle imaging system

    Energy Technology Data Exchange (ETDEWEB)

    Hamel, M.C., E-mail: mchamel@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 2355 Bonisteel Blvd, Ann Arbor, MI 48109 (United States); Polack, J.K., E-mail: kpolack@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 2355 Bonisteel Blvd, Ann Arbor, MI 48109 (United States); Poitrasson-Rivière, A., E-mail: alexispr@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 2355 Bonisteel Blvd, Ann Arbor, MI 48109 (United States); Flaska, M., E-mail: mflaska@psu.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 2355 Bonisteel Blvd, Ann Arbor, MI 48109 (United States); Department of Mechanical and Nuclear Engineering, Pennsylvania State University, 137 Reber Building, University Park, PA 16802 (United States); Clarke, S.D., E-mail: clarkesd@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 2355 Bonisteel Blvd, Ann Arbor, MI 48109 (United States); Pozzi, S.A., E-mail: pozzisa@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 2355 Bonisteel Blvd, Ann Arbor, MI 48109 (United States); Tomanin, A., E-mail: alice.tomanin@jrc.ec.europa.eu [European Commission, Joint Research Centre, Institute for Transuranium Elements, 21027 Ispra, VA (Italy); Lainsa-Italia S.R.L., via E. Fermi 2749, 21027 Ispra, VA (Italy); Peerani, P., E-mail: paolo.peerani@jrc.ec.europa.eu [European Commission, Joint Research Centre, Institute for Transuranium Elements, 21027 Ispra, VA (Italy)

    2016-02-21

    Stochastic image reconstruction has been applied to a dual-particle imaging system being designed for nuclear safeguards applications. The dual-particle imager (DPI) is a combined Compton-scatter and neutron-scatter camera capable of producing separate neutron and photon images. The stochastic origin ensembles (SOE) method was investigated as an imaging method for the DPI because only a minimal estimation of system response is required to produce images with quality that is comparable to common maximum-likelihood methods. This work contains neutron and photon SOE image reconstructions for a {sup 252}Cf point source, two mixed-oxide (MOX) fuel canisters representing point sources, and the MOX fuel canisters representing a distributed source. Simulation of the DPI using MCNPX-PoliMi is validated by comparison of simulated and measured results. Because image quality is dependent on the number of counts and iterations used, the relationship between these quantities is investigated.

  20. Development of solid amine CO2 control systems for extended duration missions

    Science.gov (United States)

    Dresser, K. J.; Cusick, R. J.

    1984-01-01

    This paper briefly discusses the development history of solid amine CO2 control systems, describes two distinct CO2 control system concepts, and presents the performance characteristics for both system concepts. The first concept (developed under NASA Contract NAS9-13624) incorporates a solid amine canister, an automatic microprocessor controller, and an accumulator to collect CO2 and to provide regulated CO2 delivery to an oxygen recovery system. This system is currently operating in the Crew Systems Division's Advanced Life Support Development Laboratory (ALSDL). The second system concept (being developed under NASA Contract NAS9-16978) employs multiple solid amine canisters, an advanced automatic controller and system status display, the ability to regulate CO2 delivery for oxygen recovery, and energy saving features that allow system operation at lower power levels than the first concept.

  1. Analysis of Corrosion Residues Collected from the Aluminum Basket Rails of the High-Burnup Demonstration Cask.

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-03-01

    On September, 2015, an inspection was performed on the TN-32B cask that will be used for the high-burnup demonstration project. During the survey, wooden cribbing that had been placed within the cask eleven years earlier to prevent shifting of the basket during transport was removed, revealing two areas of residue on the aluminum basket rails, where they had contacted the cribbing. The residue appeared to be a corrosion product, and concerns were raised that similar attack could exist at more difficult-to-inspect locations in the canister. Accordingly, when the canister was reopened, samples of the residue were collected for analysis. This report presents the results of that assessment, which determined that the corrosion was due to the presence of the cribbing. The corrosion was associated with fungal material, and fungal activity likely contributed to an aggressive chemical environment. Once the cask has been cleaned, there will be no risk of further corrosion.

  2. Skylab Apollo Telescope Mount Spar and Sun End

    Science.gov (United States)

    1971-01-01

    The Apollo Telescope Mount (ATM) was designed and developed by the Marshall Space Flight Center and served as the primary scientific instrument unit aboard Skylab (1973-1979). The ATM contained eight complex astronomical instruments designed to observe the Sun over a wide spectrum from visible light to x-rays. This image depicts the sun end and spar of the ATM flight unit showing individual telescopes. All solar telescopes, the fine Sun sensors, and some auxiliary systems are mounted on the spar, a cruciform lightweight perforated metal mounting panel that divides the canister lengthwise into four equal compartments. The spar assembly was nested inside a cylindrical canister that fit into a complex frame named the rack, and was protected by the solar shield.

  3. KSC-05PD-0724

    Science.gov (United States)

    2005-01-01

    KENNEDY SPACE CENTER, FLA. In the Space Station Processing Facility, technician Fred Parisi is attaching the CELA (Cargo Element Lifting Assembly) down rods to the External Stowage Platform -2 (ESP-2) in preparation for lifting the ESP-2 from its transportation container and installing it into the Payload Transportation Canister. At right, above, is Scott Kisner, task leader. The ESP-2 will travel to Launch Pad 39B with its fellow payload elements the Lightweight MPESS Carrier (Multi-Purpose Experiment Support Structure) and Multi-Purpose Logistics Module Raffaello in the canister. Once at the pad, the three payloads will be transferred to the payload bay of Discovery for flight. The ESP-2 is carrying replacement parts to the International Space Station. The platform will be deployed and attached to the Stations airlock and used as a permanent spare parts facility. STS-114 is targeted for launch during a window that extends from May 15 through June 3.

  4. Safety case for the disposal of spent nuclear fuel at Olkiluoto. Assessment of radionuclide release scenarios for the repository system 2012

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-12-15

    Assessment of Radionuclide Release Scenarios sits within Posiva Oy's Safety Case 'TURVA-2012' report portfolio and has the objective of presenting an assessment of the repository system scenarios leading to radionuclide releases that have been identified in Formulation of Radionuclide Release Scenarios. A base scenario, variant scenarios and disturbance scenarios are considered. For each scenario, a range of calculation cases, also identified in Formulation of Radionuclide Release Scenarios, has been analysed, complemented by Monte Carlo simulations, a probabilistic sensitivity analysis and other supporting calculations. The calculation cases and analyses take into account major uncertainties in the initial state of the barriers and possible paths for the evolution of the repository system identified in Performance Assessment. Quality control and assurance measures have been adopted to ensure transparency and traceability of the calculations performed and hence to promote confidence in the analysis of the calculation cases. The calculation cases each consider a single, failed canister, where three possible modes of failure are addressed: (1) the presence of an initial penetrating defect in the copper overpack of the canister, (2) corrosion of the copper overpack, which occurs most rapidly in scenarios in which buffer density is reduced, e.g. by erosion, (3) shear movement on a fracture intersecting a deposition hole. The likelihood and consequences of multiple canister failure occurring during the assessment time frame are also considered. In particular, the analyses consider: The likelihood and consequences of there being multiple canisters with initial penetrating defects; The consequences if canister failure due to corrosion following buffer erosion were to occur; and The low annual probability of there being an earthquake large enough to give rise to canister failure due to rock shear movements and the potential consequences of such an earthquake

  5. Spent fuel handling and packaging program. Quarterly report, April-June 1980

    Energy Technology Data Exchange (ETDEWEB)

    Durrill, D C

    1980-07-01

    This document is a report of activities performed by Westinghouse Advanced Energy Systems Division-Nevada Operations at the E-MAD Facility, Area 25, Nevada Test Site, in meeting subtask objectives during the third quarter of FY 1980. Activities during this period included completion of encapsulation and preparation for shipment of 11 spent fuel assemblies to be tested at the Climax test site by Lawrence Livermore Laboratories; calorimetry of two fuel assemblies; repeat of three 1 kW Fuel Temperature Test runs; acquisition of gas samples from fueled canisters; removal of ten R-MAD shielding windows; and assembly and checkout of the canister cutter, which was received from AESD-Large.

  6. Trade study for the disposition of cesium and strontium capsules

    Energy Technology Data Exchange (ETDEWEB)

    Claghorn, R.D.

    1996-03-01

    This trade study analyzes alternatives for the eventual disposal of cesium and strontium capsules currently stored at the Waste Encapsulation and Storage Facility as by-product. However, for purposes of this study, it is assumed that at some time in the future, the capsules will be declared high-level waste and therefore will require disposal at an offsite geologic repository. The study considered numerous alternatives and selected three for detailed analysis: (1) overpack and storage at high-level waste canister storage building, (2) overpack at the high-level waste vitrification facility followed by storage at a high-level waste canister storage building, and (3) blend capsule contents with other high-level waste feed streams and vitrify at the high-level waste vitrification facility.

  7. Thermal-Hydrology Simulations of Disposal of High-Level Radioactive Waste in a Single Deep Borehole

    Energy Technology Data Exchange (ETDEWEB)

    Hadgu, Teklu [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Stein, Emily [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hardin, Ernest [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Freeze, Geoffrey A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hammond, Glenn Edward [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-11-01

    Simulations of thermal-hydrology were carried out for the emplacement of spent nuclear fuel canisters and cesium and strontium capsules using the PFLOTRAN simulator. For the cesium and strontium capsules the analysis looked at disposal options such as different disposal configurations and surface aging of waste to reduce thermal effects. The simulations studied temperature and fluid flux in the vicinity of the borehole. Simulation results include temperature and vertical flux profiles around the borehole at selected depths. Of particular importance are peak temperature increases, and fluxes at the top of the disposal zone. Simulations of cesium and strontium capsule disposal predict that surface aging and/or emplacement of the waste at the top of the disposal zone reduces thermal effects and vertical fluid fluxes. Smaller waste canisters emplaced over a longer disposal zone create the smallest thermal effect and vertical fluid fluxes no matter the age of the waste or depth of emplacement.

  8. Prototypical Rod Construction Demonstration Project. Phase 3, Final report: Volume 1, Cold checkout test report, Book 3

    Energy Technology Data Exchange (ETDEWEB)

    1993-05-01

    The objective of Phase 3 of the Prototypical Rod consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod consolidation System as described in the NUS Phase 2 Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase 3 effort the system was tested on a component, subsystem, and system level. This volume 1, discusses the PRCDP Phase 3 Test Program that was conducted by the HALLIBURTON NUS Environmental Corporation under contract AC07-86ID12651 with the United States Department of Energy. This document, Volume 1, Book 3 discusses the following topics: Downender Test Results and Analysis Report; NFBC Canister Upender Test Results and Analysis Report; Fuel Assembly Handling Fixture Test Results and Analysis Report; and Fuel Canister Upender Test Results and Analysis Report.

  9. Statement of work for the immobilized high-level waste transportation system, Project W-464

    Energy Technology Data Exchange (ETDEWEB)

    Mouette, P.

    1998-06-24

    The objective of this Statement of Work (SOW) is to present the scope, the deliverables, the organization, the technical and schedule expectations for the development of a Package Design Criteria (PDC), cost and schedule estimate for the acquisition of a transportation system for the Immobilized High-Level Waste (IHLW). This transportation system which includes the truck, the trailer, and a shielded cask will be used for on-site transportation of the IHLW canisters from the private vendor vitrification facility to the Hanford Site interim storage facility, i.e., vaults 2 and 3 of the Canister Storage Building (CSB). This Statement of Work asks Waste Management Federal Services, Inc., Northwest Operations, to provide Project W-464 with a Design Criteria Document, plus a life-cycle schedule and cost estimate for the acquisition of a transportation system (shielded cask, truck, trailer) for IHLW on-site transportation.

  10. Radon adsorbed in activated charcoal—a simple and safe radiation source for teaching practical radioactivity in schools and colleges

    Science.gov (United States)

    Al-Azmi, Darwish; Mustapha, Amidu O.; Karunakara, N.

    2012-07-01

    Simple procedures for teaching practical radioactivity are presented in a way that attracts students' attention and does not make them apprehensive about their safety. The radiation source is derived from the natural environment. It is based on the radioactivity of radon, a ubiquitous inert gas, and the adsorptive property of activated charcoal. Radon gas from ambient air in the laboratory was adsorbed into about 70 g of activated charcoal inside metallic canisters. Gamma radiation was subsequently emitted from the canisters, following the radioactive decay of radon and its progenies. The intensities of the emitted gamma-rays were measured at suitable intervals using a NaI gamma-ray detector. The counts obtained were analysed and used to demonstrate the radioactive decay law and determine the half-life of radon. In addition to learning the basic properties of radioactivity the students also get practical experience about the existence of natural sources of radiation in the environment.

  11. Modelling the thermo-mechanical volume change behaviour of compacted expansive clays

    CERN Document Server

    Tang, Anh-Minh; 10.1680/geot.2009.59.3.185

    2009-01-01

    Compacted expansive clays are often considered as a possible buffer material in high-level deep radioactive waste disposals. After the installation of waste canisters, the engineered clay barriers are subjected to thermo-hydro-mechanical actions in the form of water infiltration from the geological barrier, heat dissipation from the radioactive waste canisters, and stresses generated by clay swelling under almost confined conditions. The aim of the present work is to develop a constitutive model that is able to describe the behaviour of compacted expansive clays under these coupled thermo-hydro-mechanical actions. The proposed model is based on two existing models: one for the hydro-mechanical behaviour of compacted expansive clays and another for the thermo-mechanical behaviour of saturated clays. The elaborated model has been validated using the thermo-hydro-mechanical test results on the compacted MX80 bentonite. Comparison between the model prediction and the experimental data show that this model is able...

  12. UoGAS - A Get Away Special Satellite with Orbit-Raising Capability

    OpenAIRE

    Lorenz, Ralph

    1988-01-01

    The low cost of satellite deployment from Shuttle GAS canister makes it an attractive launch option. However, the low deployment altitude severely constrains lifetime so the UoGAS (University of Surrey Get Away Special) spacecraft will incorporate a propulsion system. Lifetime extension methods are considered and a start-of-mission orbit-raising manoeuvre is selected. An orbit dynamics simulation method (taking into account the atmospheric drag) is discussed and results presented. Mission pro...

  13. Spent nuclear fuel retrieval system fuel handling development testing. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Jackson, D.R.; Meeuwsen, P.V.

    1997-09-01

    Fuel handling development testing was performed in support of the Fuel Retrieval System (FRS) Sub-Project, a subtask of the Spent Nuclear Fuel Project at the Hanford Site in Richland, Washington. The FRS will be used to retrieve and repackage K-Basin Spent Nuclear Fuel (SNF) currently stored in old K-Plant storage basins. The FRS is required to retrieve full fuel canisters from the basin, clean the fuel elements inside the canister to remove excessive uranium corrosion products (or sludge), remove the contents from the canisters and sort the resulting debris, scrap, and fuel for repackaging. The fuel elements and scrap will be collected in fuel storage and scrap baskets in preparation for loading into a multi canister overpack (MCO), while the debris is loaded into a debris bin and disposed of as solid waste. This report describes fuel handling development testing performed from May 1, 1997 through the end of August 1997. Testing during this period was mainly focused on performance of a Schilling Robotic Systems` Conan manipulator used to simulate a custom designed version, labeled Konan, being fabricated for K-Basin deployment. In addition to the manipulator, the camera viewing system, process table layout, and fuel handling processes were evaluated. The Conan test manipulator was installed and fully functional for testing in early 1997. Formal testing began May 1. The purposes of fuel handling development testing were to provide proof of concept and criteria, optimize equipment layout, initialize the process definition, and identify special needs/tools and required design changes to support development of the performance specification. The test program was set up to accomplish these objectives through cold (non-radiological) development testing using simulated and prototype equipment.

  14. Performance of a new carbon dioxide absorbent, Yabashi lime® as compared to conventional carbon dioxide absorbent during sevoflurane anesthesia in dogs.

    Science.gov (United States)

    Kondoh, Kei; Atiba, Ayman; Nagase, Kiyoshi; Ogawa, Shizuko; Miwa, Takashi; Katsumata, Teruya; Ueno, Hiroshi; Uzuka, Yuji

    2015-08-01

    In the present study, we compare a new carbon dioxide (CO2) absorbent, Yabashi lime(®) with a conventional CO2 absorbent, Sodasorb(®) as a control CO2 absorbent for Compound A (CA) and Carbon monoxide (CO) productions. Four dogs were anesthetized with sevoflurane. Each dog was anesthetized with four preparations, Yabashi lime(®) with high or low-flow rate of oxygen and control CO2 absorbent with high or low-flow rate. CA and CO concentrations in the anesthetic circuit, canister temperature and carbooxyhemoglobin (COHb) concentration in the blood were measured. Yabashi lime(®) did not produce CA. Control CO2 absorbent generated CA, and its concentration was significantly higher in low-flow rate than a high-flow rate. CO was generated only in low-flow rate groups, but there was no significance between Yabashi lime(®) groups and control CO2 absorbent groups. However, the CO concentration in the circuit could not be detected (≤5ppm), and no change was found in COHb level. Canister temperature was significantly higher in low-flow rate groups than high-flow rate groups. Furthermore, in low-flow rate groups, the lower layer of canister temperature in control CO2 absorbent group was significantly higher than Yabashi lime(®) group. CA and CO productions are thought to be related to the composition of CO2 absorbent, flow rate and canister temperature. Though CO concentration is equal, it might be safer to use Yabashi lime(®) with sevoflurane anesthesia in dogs than conventional CO2 absorbent at the point of CA production.

  15. Radon-Instrumentation; Radon-Instrumentacion

    Energy Technology Data Exchange (ETDEWEB)

    Moreno y Moreno, A. [Departamento de Apoyo en Ciencias Aplicadas, Benemerita Universidad Autonoma de Puebla, 4 Sur 104, Centro Historico 72000 Puebla (Mexico)

    2003-07-01

    The presentation of the active and passive methods for radon, their identification and measure, instrumentation and characteristics are the objectives of this work. Active detectors: Active Alpha Cam Continuous Air Monitor, Model 758 of Victoreen, Model CMR-510 Continuous Radon Monitor of the Signature Femto-Tech. Passive detectors: SSNTD track detectors in solids Measurement Using Charcoal Canisters, disk of activated coal deposited in a metallic box Electrets Methodology. (Author)

  16. Sample Return Missions Where Contamination Issues are Critical: Genesis Mission Approach

    Science.gov (United States)

    Allton, Judith H.; Stansbery E. K.

    2011-01-01

    The Genesis Mission, sought the challenging analytical goals of accurately and precisely measuring the elemental and isotopic composition of the Sun to levels useful for planetary science, requiring sensitivities of ppm to ppt in the outer 100 nm of collector materials. Analytical capabilities were further challenged when the hard landing in 2004 broke open the canister containing the super-clean collectors. Genesis illustrates that returned samples allow flexibility and creativity to recover from setbacks.

  17. Spent nuclear fuel project design basis capacity study

    Energy Technology Data Exchange (ETDEWEB)

    Cleveland, K.J.

    1998-07-22

    A parametric study of the Spent Nuclear Fuel Project system capacity is presented. The study was completed using a commercially available software package to develop a summary level model of the major project systems. A base case, reflecting the Fiscal Year 1998 process configuration, is evaluated. Parametric evaluations are also considered, investigating the impact of higher fuel retrieval system productivity and reduced shift operations at the canister storage building on total project duration.

  18. A Revised Health Risk Assessment for the Use of Hexachloroethane Smoke on an Army Training Area.

    Science.gov (United States)

    1987-09-01

    gases listed in Table 2. Metals and metalloids quantified from actual HC smoke canisters are listed in Table 3. The basic chemical reaction of the HC mix...Conference Government Industrial Hygienists). F. McDonald and R. Porton, Toxicity of Zinc Chloride Smoke and Treatment With BAL, Report No. 2703 (September...tetrachloride, and arsenic in their water quality criteria documents. When it had been found that HC smoke contains material potentially hazardous upon

  19. Synfuels from fusion: using the tandem mirror reactor and a thermochemical cycle to produce hydrogen

    Energy Technology Data Exchange (ETDEWEB)

    Werner, R.W. (ed.)

    1982-11-01

    This study is concerned with the following area: (1) the tandem mirror reactor and its physics; (2) energy balance; (3) the lithium oxide canister blanket system; (4) high-temperature blanket; (5) energy transport system-reactor to process; (6) thermochemical hydrogen processes; (7) interfacing the GA cycle; (8) matching power and temperature demands; (9) preliminary cost estimates; (10) synfuels beyond hydrogen; and (11) thermodynamics of the H/sub 2/SO/sub 4/-H/sub 2/O system. (MOW)

  20. Pondering imponderables: occultism in the mirror of late classical physics

    NARCIS (Netherlands)

    Asprem, E.

    2011-01-01

    On a souvent caractérisé le fait d’ établir une ligne de partage entre la physique « classique » et la physique «moderne » comme une manière de rompre avec une conception mécaniste, matérialiste et réductionniste du monde, pour se tourner vers une autre, ouverte et fondamentalement incertaine. Selon

  1. Securing Nuclear and Radiological Material in the Homeland

    Science.gov (United States)

    2007-03-01

    measures at the facility, a ballistic vest, night vision goggles, a gas mask, handcuffs, and a pepper spray canister. Further investigation revealed...25 C. D. Ferguson , T. Kazi and J. Perera, Commercial Radioactive Sources: Surveying the Security Risks (Monterey Institute of...http://www.ewg.org/reports/nuclearwaste/faq/faq_terroristthreat_more.php (accessed 10 February 2006). Ferguson , C. D., T. Kazi, and J. Perera

  2. Collection of Human Wastes on Long Missions

    Science.gov (United States)

    Jennings, D. C.; Lewis, T. A.; Brose, H. F.

    1986-01-01

    Report evaluates and compares three alternative approaches to hygienic containment of human wastes. Three practical means of waste collection: filter-bag collection with compaction by fan suction, canister collection with compaction by force applied to compaction cups or disks, and sleeve collection with compaction by rollers and winding on reel. Potentially useful in airplanes, buses, boats, trains, and campers and temporary toilets for construction sites and outdoor gatherings.

  3. Evaluation of Coupled Thermo-Hydro-Mechanical Phenomena in the Near Field for Geological Diaposal of High-Level Radioactive waste

    OpenAIRE

    2000-01-01

    Geological disposal of high-level radioactive waste (HLW) in Japan is based on a multibarrier system composed of engineered and natural barriers. The engineered barriers are composed of vitrified waste confined within a canister, overpack and buffer material. Highly compacted bentonite clay is considered one of the most promising candidate buffer material mainly because of its low hydraunc conductivity and high adsorption capacity of radionuclides. In a repository for HLW, complex thermal, hy...

  4. SNF fuel retrieval sub project safety analysis document

    Energy Technology Data Exchange (ETDEWEB)

    BERGMANN, D.W.

    1999-02-24

    This safety analysis is for the SNF Fuel Retrieval (FRS) Sub Project. The FRS equipment will be added to K West and K East Basins to facilitate retrieval, cleaning and repackaging the spent nuclear fuel into Multi-Canister Overpack baskets. The document includes a hazard evaluation, identifies bounding accidents, documents analyses of the accidents and establishes safety class or safety significant equipment to mitigate accidents as needed.

  5. Collateral Damage to Satellites from an EMP Attack

    Science.gov (United States)

    2010-08-01

    tests were subsequently conducted on the ground and the failures were traced to a problem with certain npn transistors enclosed in nitrogen canisters...one command decoder • Intermittent recovery made via corrective procedures - power adjustments to affected transistors - continuous commanding...modified commands • 21 Feb 63 - complete failure of command system - end of mission • Lab tests confirm ionization damage to critical transistors

  6. Measurements of volatile organic compounds in rural area of Yangtze River Delta region: Measurement comparison and source characterization

    Science.gov (United States)

    Kudo, S.; Saito, S.; Tanimoto, H.; Inomata, S.; Kanaya, Y.; Yamaji, K.; Xiaole, P.; Wang, Z.

    2012-12-01

    Concentrations of non-methane volatile organic compounds (NMVOCs) in ambient air were measured by three different methods in a city of Rudong in May and June 2010. Intercomparison of VOCs measurements was made among in-situ measurements and canister analyses with a gas chromatography/flame ionization detection/mass spectrometry (GC/FID/MS) and proton transfer reaction-mass spectrometry (PTR-MS). For 18 VOCs measured by GC/FID/MS, canister analyses and in-situ measurements were in reasonably good agreement (R2 > 0.90). However, alkenes and aromatics in canister samples were found to be lower than in-situ measurements likely due to adsorption of low volatile compounds onto the wall surface inside canisters. For comparison of GC/MS with PTR-MS, the correlations for isoprene, benzene, C8 aromatics, and C9 aromatics were highly significant (R2 ≥ 0.93) with each other. However, there were quantitative differences between GC/MS and PTR-MS. For example, isoprene measured by PTR-MS indicates existence of interferences from C5 alcohols, C5 aldehydes, and furan. During the latter part of the field campaign, elevated concentrations of VOCs and CO were observed when intensive burning of crop residues took place near the sampling site. The concentrations of ethane, propane, ethane, isoprene, acetone, acetaldehyde, and aromatics varied in the range between 0 and 30 ppbv. The observed VOCs concentrations are compared to model results by a regional chemistry-transport model for Asia. The modeled concentrations underestimated the observed concentrations by a factor of 10 for NMHCs, 100 for aromatics, 10 for oxygenated VOCs, implying that current emissions inventories miss a number of sources for these VOCs.

  7. STS 131 Return Samples: Assessment of Air Quality Aboard the Shuttle (STS-131) and International Space Station (19A)

    Science.gov (United States)

    James, John T.

    2010-01-01

    The toxicological assessments of 1 grab sample canister (GSC) from the Shuttle are reported in Table 1. Analytical methods have not changed from earlier reports. The recoveries of the 3 surrogates (C-13-acetone, fluorobenzene, and chlorobenzene) from the Shuttle GSC were 100%, 93%, and 101%, respectively. Based on the historical experience using end-of-mission samples, the Shuttle atmosphere was acceptable for human respiration.

  8. STS 130 Return Samples: Assessment of Air Quality Aboard the Shuttle (STS-130) and International Space Station (20A)

    Science.gov (United States)

    James, John T.

    2010-01-01

    The toxicological assessments of 3 grab sample canisters (GSCs) from the Shuttle are reported in Table 1. Analytical methods have not changed from earlier reports. The recoveries of the 3 surrogates ( 13C-acetone, fluorobenzene, and chlorobenzene) from the 3 Shuttle GSCs averaged 96, 90, and 85 %, respectively. Based on the end-of-mission sample, the Shuttle atmosphere was acceptable for human respiration.

  9. Advanced Computer Simulations of Military Incinerators

    Science.gov (United States)

    2004-12-01

    models contain 3D furnace and canister geometries and all of the relevant physics and chemistry. The destruction of chemical agent is predicted using...computational chemistry methods, chemical kinetics have been developed that describe the incineration of organo -phosphorus nerve agent (GB, VX) and...States. The chemical warfare agents (CWA) consist of mustard gas and other blister agents as well as organo -phosphorus nerve agents. Incineration was

  10. The Art of War in Transition?

    Science.gov (United States)

    1992-05-12

    losses, The Holy Roman Empire disintegrated and that of Charlemagne restored, all as a result of seventy days’ campaigning and eight hours’ fighting. 3...They fired three types of ammunition: caseshot or canister, shell, or grapeshot. The Emperor im- proved on the artillery system fielded under Jean...consisted of three sections: 1. A personal staff, "the Raison," performed political as well as military functions for the Emperor . 2. The General Staff of

  11. Combating WMD Journal. Issue 2

    Science.gov (United States)

    2008-03-01

    and saws to enter a facility; the use of the (PINS) to con- duct non intrusive chemical ordnance assessment to characterize unex- ploded chemical...Nanotechnology has been brought to bear on IPE filter design. Especially in face-mask canisters, both vibration and adsorption of water- vapour ...compounds are utilised to improve activation (ASC-TEDA charcoal) to render it more effective against high vapour pressure com- pounds such as the blood

  12. Criticality in a high level waste repository. A review of some important factors and an assessment of the lessons that can be learned from the Oklo reactors

    Energy Technology Data Exchange (ETDEWEB)

    Oversby, V.M. [VMO Konsult, Stockholm (Sweden)

    1996-06-01

    The conditions and scenarios that might allow sufficient {sup 239}Pu and/or {sup 235}U to accumulate together with enough water to allow for moderation of neutron energies and thereby achieving a state where neutron-induced fission reactions could be sustained at a rate significantly above the natural rate of spontaneous fission is discussed. The uranium deposit in Oklo, Gabon, which was the site of naturally-occurring neutron-induced fission reactions approximately 2000 My ago is described. The chemistry, mineralogy, and conditions of the nuclear reactor operations are reviewed. Results of modelling the conditions for criticality at Oklo are used to estimate the amounts of spent fuel uranium that must be assembled in a favorable geometry in order to produce a similar reactive situation in a geologic repository. The amounts of uranium that must be transported and redeposited to reach a critical configuration are extremely large in relation to those that could be transported under any reasonably achievable conditions. In addition, transport and redeposition scenarios often require opposite chemical characteristics. It is concluded that the likelihood of achieving a critical condition due to accumulation of a critical mass of uranium outside the canisters after disposal is negligible. Criticality inside the canister is rendered impossible by the use of low-solubility materials inside the canisters that fill space and prevent the entry of enough water to allow moderation of neutron energies. Criticality due to plutonium outside the canister can be ruled out because it requires a series of processes, each of which has a vanishingly small probability. 25 refs, 9 tabs, 8 figs.

  13. MCO loading and cask loadout technical manual

    Energy Technology Data Exchange (ETDEWEB)

    PRAGA, A.N.

    1998-10-01

    A compilation of the technical basis for loading a multi-canister overpack (MCO) with spent nuclear fuel and then placing the MCO into a cask for shipment to the Cold Vacuum Drying Facility. The technical basis includes a description of the process, process technology that forms the basis for loading alternatives, process control considerations, safety considerations, equipment description, and a brief facility structure description.

  14. Freefem++ in THM analyses of KBS-3 deposition hole

    Energy Technology Data Exchange (ETDEWEB)

    Lempinen, A. [Marintel Ky, Turku (Finland)

    2006-12-15

    The applicability of Freefem++ as a software for thermo-hydro-mechanical analysis of KBS-3V deposition hole was evaluated. Freefem++ is software for multiphysical simulations with finite element method. A set of previously performed analyses were successfully repeated with Freefem++. The only significant problem was to impose unique values for variables at the canister surface. This problem can be circumvented with an iterative method, and it can possibly be solved later, since Freefem++ is opensource software. (orig.)

  15. Macstor system for spent fuel storage

    Energy Technology Data Exchange (ETDEWEB)

    Pattantyus, P. (Atomic Energy of Canada Ltd., Montreal, PQ (Canada). Power Projects)

    1993-01-01

    In 1989, Transnuclear Inc. and AECL jointly developed the conceptual design for the Modular Aircooled Canister Storage System (Macstor) for LWR fuel. The development effort has proceeded to the completion of successful full-scale thermal testing. In 1990, AECL adapted the Macstor System approach for use with Candu fuel. The adapted design, called Canstor, has also successfully completed full-scale thermal testing, and the final system design has been completed. (author) 1 fig.

  16. Swimming Performance of Bighead Carp and Silver Carp: Methodology, Metrics, and Management Applications

    Science.gov (United States)

    2012-08-01

    the Laurentian Great Lakes? A bioenergetic modeling exercise. Freshwater Biology 55: 2138-2152. Cornish, M., J. Crossland, L. Nelson, J. Pothoff...tubs, half- to three-quarters full, circulated with a Little Giant Water Wizard Model 5 MSP submersible water pump, filtered with a 200-L canister...tions. Endurance data are analyzed using regression models that describe linear or curvilinear relationships between water velocity (independent or

  17. A method to feed individual bees (Hymenoptera: Apiformes) known amounts of pesticides

    OpenAIRE

    Ladurner, Edith; Bosch, Jordi; Maini, Stefano; Kemp, William

    2003-01-01

    International audience; We devised a simple method ("flower") to feed bees individually, and compared it with two other methods commonly used ("film canister" and "glass vial"). We tested the three methods on two solitary species, Osmia lignaria and Megachile rotundata, and one social species, Apis mellifera, under four different light regimes (natural, artificial, plant growth and darkness). The flower method was the most effective for all three bee species: 90-95% of the bees fed under natu...

  18. An Annotated Bibliography of MANPRINT-Related Assessments and Evaluations Conducted by the U.S. Army, 2nd Edition: 1953 to 2009. Volume 2 - MANPRINT Assessment and Evaluations

    Science.gov (United States)

    2010-02-01

    Vehicle- Wheeled- MRAP- MPCV Buffalo Hatch MPCV Buffalo Hatch Buffalo Mine Protection Clearance Vehicle (MPCV) Roof Hatch 2007 Vehicle- Wheeled...hollow steel shell containing a canister and drogue parachute along with a primary expelling charge. A candle assembly, main parachute, delay and...PA160 Shipping and Storage Container, this one-man portable, cylindrical , 35 pound munition is designed to be set in either the Manual Mode or the

  19. A plunger lift and monitoring system for gas wells based on deployment-retrievement integration

    Directory of Open Access Journals (Sweden)

    Zheng Tong

    2015-11-01

    Full Text Available As a necessary step, removing liquid in the wellbore plays an important role during the production of gas wells. Plunger lift is a widely-used intermittent deliquification process for gas wells. However, the manual control way and wire logging are still utilized as a downhole monitoring way for plunger lift, which is not efficient in terms of interrupting the production. This paper presents an improved solution that logging instruments canister are deployed and retrieved by means of a new assembly. With the reciprocating plunger, logging instruments canister can be carried and deployed to the bottom of a gas well to carry out logging and sampling tasks on the production demand of a field. After the deployment and logging tasks are performed, logging instruments canister is carried back to the surface by the plunger and then data is transferred to the wellhead device near field wireless communication technology. This newly developed plunger lift system comprises plunger body, deployment sub-assembly, retrieve sub-assembly and logging instruments canister. The surface device comprises RF antenna, reader and writer. Based upon the method of deployment-retrieve integration, the new deliquification process is introduced and on-line monitoring of production dynamics can be performed including P/T measurement, downhole fluid sampling, pressure build-up, etc. without interrupting production. The general solution and engineering design parameters have been confirmed by research teams, while system prototype manufacture and workbench tests are being performed. The cost-effective way combining deliquification with dynamic monitoring is developed and contributes to increasing production and the stable productivity of gas wells. It is very significant for low-pressure and low-production gas fields to achieve automation production and management.

  20. Spent nuclear fuel discharges from U.S. reactors 1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-02-01

    Spent Nuclear Fuel Discharges from US Reactors 1994 provides current statistical data on fuel assemblies irradiated at commercial nuclear reactors operating in the US. This year`s report provides data on the current inventories and storage capacities at these reactors. Detailed statistics on the data are presented in four chapters that highlight 1994 spent fuel discharges, storage capacities and inventories, canister and nonfuel component data, and assembly characteristics. Five appendices, a glossary, and bibliography are also included. 10 figs., 34 tabs.

  1. Development of the near field geochemistry model; Desarrollo de un modelo geoquimico de campo proximo

    Energy Technology Data Exchange (ETDEWEB)

    Arcos, D.; Bruno, J.; Duro, L.; Grive, M.

    2000-07-01

    This report discusses in a quantitative manner the evolution of the near field geochemistry as a result of the interactions between two different introducing granitic groundwaters and the FEBEX bentonite as a buffer material. The two granitic groundwaters considered are: SR-5 water, sampled in a borehole at 500 m depth in Mina Ratones, and a mean composition of different granitic groundwaters from the iberian Massif. The steel canister has also been introduced by considering the iron corrosion in anoxic conditions. (Author)

  2. Seeking the Light: Gravity Without the Influence of Gravity

    Science.gov (United States)

    Sack, Fred; Kern, Volker; Reed, Dave; Etheridge, Guy (Technical Monitor)

    2002-01-01

    All living things sense gravity like humans might sense light or sound. The Biological Research In Canisters (BRIC-14) experiment, explores how moss cells sense and respond to gravity and light. This experiment studies how gravity influences the internal structure of moss cells and seeks to understand the influences of the spaceflight environment on cell growth. This knowledge will help researchers understand the role of gravity in the evolution of cells and life on earth.

  3. EMPLACEMENT DRIFT SHIELDING CALCULATION

    Energy Technology Data Exchange (ETDEWEB)

    A. Nielsen

    1999-10-13

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

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

    Energy Technology Data Exchange (ETDEWEB)

    S. Mastilovic

    1999-07-28

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

  5. Long-term geochemical evolution of the near field repository: Insights from reactive transport modelling and experimental evidences

    Science.gov (United States)

    Arcos, David; Grandia, Fidel; Domènech, Cristina; Fernández, Ana M.; Villar, María V.; Muurinen, Arto; Carlsson, Torbjörn; Sellin, Patrik; Hernán, Pedro

    2008-12-01

    The KBS-3 underground nuclear waste repository concept designed by the Swedish Nuclear Fuel and Waste Management Co. (SKB) includes a bentonite buffer barrier surrounding the copper canisters and the iron insert where spent nuclear fuel will be placed. Bentonite is also part of the backfill material used to seal the access and deposition tunnels of the repository. The bentonite barrier has three main safety functions: to ensure the physical stability of the canister, to retard the intrusion of groundwater to the canisters, and in case of canister failure, to retard the migration of radionuclides to the geosphere. Laboratory experiments (dissolution of calcite and dolomite, when present. The equilibrium of these minerals is deeply influenced by gypsum dissolution and cation exchange reactions in the smectite interlayer. If carbonate minerals are initially absent in bentonite, pH is then controlled by surface acidity reactions in the hydroxyl groups at the edge sites of the clay fraction, although its buffering capacity is not as strong as the equilibrium with carbonate minerals. The redox capacity of the bentonite pore water system is mainly controlled by Fe(II)-bearing minerals (pyrite and siderite). Changes in the groundwater composition lead to variations in the cation exchange occupancy, and dissolution-precipitation of carbonate minerals and gypsum. The most significant changes in the evolution of the system are predicted when ice-melting water, which is highly diluted and alkaline, enters into the system. In this case, the dissolution of carbonate minerals is enhanced, increasing pH in the bentonite pore water. Moreover, a rapid change in the population of exchange sites in the smectite is expected due to the replacement of Na for Ca.

  6. Foreign materials in a deep repository for spent nuclear fuels; Fraemmande material i ett djupfoervar foer anvaent kaernbraensle

    Energy Technology Data Exchange (ETDEWEB)

    Jones, C.; Christiansson, Aa.; Wiborgh, M. [Kemakta Konsult AB, Stockholm (Sweden)

    1999-12-01

    The effects of foreign substances introduced into a spent-fuel repository are reviewed. Possible impacts on processes and barrier-functions are examined, and the following areas are identified: Corrosion of the spent-fuel canister through the presence of sulfur and substances that favor microbial growth; impacts on the bentonite properties through the presence of cations as calcium, potassium and iron; radionuclide transport through the presence of complex-formers and surface-active substances.

  7. Expanded High-Level Waste Glass Property Data Development: Phase I

    Energy Technology Data Exchange (ETDEWEB)

    Schweiger, Michael J.; Riley, Brian J.; Crum, Jarrod V.; Hrma, Pavel R.; Rodriguez, Carmen P.; Arrigoni, Benjamin M.; Lang, Jesse B.; Kim, Dong-Sang; Vienna, John D.; Raszewski, F. C.; Peeler, David K.; Edwards, Tommy B.; Best, D. R.; Reamer, Irene A.; Riley, W. T.; Simmons, P. T.; Workman, R. J.

    2011-01-21

    Two separate test matrices were developed as part if the EM-21 Glass Matrix Crucible Testing. The first matrix, developed using a single component-at-a-time design method and covering glasses of interest primarily to Hanford, is addressed in this data package. This data package includes methods and results from glass fabrication, chemical analysis of glass compositions, viscosity, electrical conductivity, liquidus temperature, canister centerline cooling, product consistency testing, and the toxicity characteristic leach procedure.

  8. Savannah River Technology Center monthly report: August 1994

    Energy Technology Data Exchange (ETDEWEB)

    1994-08-01

    Short summaries are given for 45 projects concerned with tritium, separations, environmental, and general topics. Included in the general topics are the following: Burst test qualification analysis of Defense Waste Processing Facility canister-plug weld; Design and development of sampling plans for non-radioactive hazardous waste; Thermal fluids laboratory melter feed test; FRR spent fuel dry storage development; SRTC buildings fire hazards analysis; and SRTC plutonium vulnerability study.

  9. Reduction of 5in./54 Gun Blast Overpressure by Means of an Aqueous Foam- Filled Muzzle Device

    Science.gov (United States)

    1981-08-01

    W. Shea S. PERFORMING ORGANIZATION NAME AND ADDRESS 10. PROGRAM ELEMENT. PROJECT, TASK Naval Surface Weapons Center AREA & WORK UNIT NUMUERS Code N43...more than 15 dB was attained. Containment of the foam in a muzzle-mounted canister was investigated. This report addresses related areas of gun...the internal pressure distribucion could be established and in- formation could be provided about the strength requirement of the device. Pres- sure

  10. Job Language Performance Requirements for MOS 19D.

    Science.gov (United States)

    1982-10-01

    pliable chest position ..ildren procedure circulation pulse clear ratio clothing resistance compressed respiration continue rotate correct signal counts...card d: rce inspect insulation jacket Q. issued lens opening dirty perform M1 canister coupling eit’r...o r remove N12 anti glare eyeiens eir. O...standard reading woods super user vehicle 4 76 MOS: 19D TASK: 071-326-0513 TASK TITLE: SELECT TEMPORARY BA7TLEFYFLD POSIIIONS GENERAL BASIC TECHNICAL

  11. The BREMSAT Phase-Modulated Communications Link

    OpenAIRE

    Lefevre, Don

    1990-01-01

    BREMSAT is a small scientific satellite being constructed by the west German Zentrum fur angewandte Raumfahrttechnologie und Mikrogravitation (ZARM), at Bremen University. BREMSAT's payload consists of five scientific experiment packages. The satellite is scheduled for Get-Away-Special (GAS) Canister launch during the German D-2 shuttle mission in March 1992. Cynetics Corporation is constructing the TT&C link for the satellite, using its standard CMX9600 modem. This modem is the result of Cyn...

  12. Assessing tungsten transport in the vadose zone: from dissolution studies to soil columns.

    Science.gov (United States)

    Tuna, Gulsah Sen; Braida, Washington; Ogundipe, Adebayo; Strickland, David

    2012-03-01

    This study investigates the dissolution, sorption, leachability, and plant uptake of tungsten and alloying metals from canister round munitions in the presence of model, well characterized soils. The source of tungsten was canister round munitions, composed mainly of tungsten (95%) with iron and nickel making up the remaining fraction. Three soils were chosen for the lysimeter studies while four model soils were selected for the adsorption studies. Lysimeter soils were representatives of the typical range of soils across the continental USA; muck-peat, clay-loamy and sandy-quartzose soil. Adsorption equilibrium data on the four model soils were modeled with Langmuir and linear isotherms and the model parameters were obtained. The adsorption affinity of soils for tungsten follows the order: Pahokee peat>kaolinite>montmorillonite>illite. A canister round munition dissolution study was also performed. After 24 d, the measured dissolved concentrations were: 61.97, 3.56, 15.83 mg L(-1) for tungsten, iron and nickel, respectively. Lysimeter transport studies show muck peat and sandy quartzose soils having higher tungsten concentration, up to 150 mg kg(-1) in the upper layers of the lysimeters and a sharp decline with depth suggesting strong retardation processes along the soil profile. The concentrations of tungsten, iron and nickel in soil lysimeter effluents were very low in terms of posing any environmental concern; although no regulatory limits have been established for tungsten in natural waters. The substantial uptake of tungsten and nickel by ryegrass after 120 d of exposure to soils containing canister round munition suggests the possibility of tungsten and nickel entering the food chain.

  13. LFCM vitrification technology. Quarterly progress report, April-June 1985

    Energy Technology Data Exchange (ETDEWEB)

    Burkholder, H.C.; Jarrett, J.H.; Minor, J.E. (comps.)

    1986-01-01

    This report is compiled by the Nuclear Waste Treatment Program and the Hanford Waste Vitrification Program at Pacific Northwest Laboratory to document progress on liquid-fed ceramic melter (LFCM) vitrification technology. Progress in the following technical subject areas during the third quarter of FY 1985 is discussed: pretreatment systems, melting process chemistry and glass development, feed preparation and transfer systems, melter systems, canister filling and handling systems, off-gas systems, process/product modeling and control, and supporting studies.

  14. Source Physics Experiment: Research in Support of Verification and Nonproliferation

    Science.gov (United States)

    2011-09-01

    ample information to study dry and water-saturated fractures, local lithology and topography on the radiated seismic wavefield. Spallation on...topography on the radiated seismic wavefield. Spallation on SPE1 is predicted well by an empirical scaling relationship developed for nuclear explosions... drilled to a depth of 190 feet and stemmed up to place a canister of 100 kg of HE at 180 feet. Six monitoring boreholes were drilled at different azimuths

  15. Near-field chemistry of the spent nuclear fuel repository; Kemialliset vuorovaikutukset kaeytetyn ydinpolttoaineen loppusijoitustilan laehialueella

    Energy Technology Data Exchange (ETDEWEB)

    Kumpulainen, H.; Lehikoinen, J.; Muurinen, A.; Ollila, K. [VTT Chemical Technology, Espoo (Finland). Industrial Physics

    1998-07-01

    Factors affecting near-field chemistry of the spent nuclear fuel repository as well as the involved mutual interactions are described on the basis of literature. The most important processes in the near-field (spent-fuel, canister and bentonite) are presented. The related examples on near-field chemistry models shed light on the extensive problematics of near-field chemistry. (authors) 80 refs.

  16. Cold vacuum drying proof of performance (first article testing) test results

    Energy Technology Data Exchange (ETDEWEB)

    MCCRACKEN, K.J.

    1999-06-23

    This report presents and details the test results of the first of a kind process referred to as Cold Vacuum Drying (CVD). The test results are compiled from several months of testing of the first process equipment skid and ancillary components to de-water and dry Multi-Canister Overpacks (MCO) filled with Spent Nuclear Fuel (SNF). The tests results provide design verifications, equipment validations, model validation data, and establish process parameters.

  17. An analysis of plutonium immobilization versus the "spent fuel" standard

    Energy Technology Data Exchange (ETDEWEB)

    Gray, W L; McKibben, J M

    1998-06-16

    Safe Pu management is an important and urgent task with profound environmental, national, and international security implications. Presidential Policy Directive 13 and analyses by scientific, technical, and international policy organizations brought about a focused effort within the Department of Energy (DOE) to identify and implement long-term disposition paths for surplus Pu. The principal goal is to render surplus Pu as inaccessible and unattractive for reuse in nuclear weapons as Pu in spent reactor fuel. In the Programmatic Environmental Impact Statement and Record of Decision for the Storage and Disposition of Weapons- Usable Fissile Materials (1997), DOE announced pursuit of two disposition technologies: (1) irradiation of Pu as MOX fuel in existing reactors and (2) immobilization of Pu into solid forms containing fission products as a radiation barrier. DOE chose an immobilization approach that includes "use of the can-in-canister option.. . for a portion of the surplus, non-pit Pu material." In the can-in-canister approach, cans of glass or ceramic forms containing Pu are encapsulated within canisters of HLW glass. In support of the selection process, a technical evaluation of retrievability and recoverability of Pu from glass and ceramic forms by a host nation and by rogue nations or subnational groups was completed. The evaluation involved determining processes and flowsheets for Pu recovery, comparing these processes against criteria and metrics established by the Fissile Materials Disposition Program and then comparing the recovery processes against each other and against SNF processes.

  18. Neutronic performance of a benchmark 1-MW LPSS

    Energy Technology Data Exchange (ETDEWEB)

    Russell, G.J.; Pitcher, E.J.; Ferguson, P.D. [Los Alamos National Laboratory, NM (United States)

    1995-12-31

    We used split-target/flux-trap-moderator geometry in our 1-MW LPSS computational benchmark performance calculations because the simulation models were readily available. Also, this target/moderator arrangement is a proven LANSCE design and a good neutronic performer. The model has four moderator viewed surfaces, each with a 13x13 cm field-of-view. For our scoping neutronic-performance calculations, we attempted to get as much engineering realism into the target-system mockup as possible. In our present model, we account for target/reflector dilution by cooling; the D{sub 2}O coolant fractions are adequate for 1 MW of 800-MeV protons (1.25 mA). We have incorporated a proton beam entry window and target canisters into the model, as well as (partial) moderator and vacuum canisters. The model does not account for target and moderator cooling lines and baffles, entire moderator canisters, and structural material in the reflector.

  19. Respect distances. Rationale and means of computation

    Energy Technology Data Exchange (ETDEWEB)

    Munier, Raymond [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden); Hoekmark, Harald [Clay Technology, Lund (Sweden)

    2004-12-01

    Canisters with spent nuclear fuel can obviously not be located within deformation zones as this might jeopardise their long term mechanical stability and thereby constitute a potential hazard to the biosphere. Less apparent, but equally important, is the fact that earthquakes trigger reactivation, slip, of structures some distance from their hypocentres due to, among many other factors, stress redistribution. Fault slip across a deposition hole might damage the isolation capacity of the canister and thereby jeopardise the overall integrity of the barrier system. Therefore, the following question might be posed: What is the distance from a deformation zone beyond which a canister can be safely emplaced? This respect distance cannot be readily computed because, unknown future events aside, there are some complicated aspects that need to be addressed e.g. degree of conservatism, scale, our ability to model ice sheets and earthquakes, etc. In this report we discuss various aspects of the assignment of respect distances, propose a methodology for its assignment and apply the methodology to the Forsmark Site as a worked example. Our main concern, in the context discussed in this report, is the post glacial faults anticipated to occur after the next glaciations. To properly address conservativeness, analysis of risk, and its implementation in safety analysis, we provide an extensive compilation of our current knowledge on post glacial faults as an appendix.

  20. Aespoe Hard Rock Laboratory. Sensor Data Report No 23

    Energy Technology Data Exchange (ETDEWEB)

    Goudarzi, Reza; Johannesson, Lars-Erik (Clay Technology AB (Sweden))

    2010-11-15

    The Prototype Repository Test consists of two sections. The installation of the first Section of Prototype Repository was made during summer and autumn 2001 and Section 2 was installed in spring and summer 2003. This report presents data from measurements in the Prototype Repository during the period 20010917-20100601. The report is organized so that the actual measured results are shown in Appendix 1-10, where Appendix 8 deals with measurements of canister displacements (by AITEMIN), Appendix 9 deals with geo-electric measurements in the backfill (by GRS), Appendix 10 deals with stress and strain measurement in the rock (by AaF) and Appendix 11 deals with measurement of water pressure in the rock (by VBB/VIAK). The main report and Appendix 1-7 deal with the rest of the measurements. Section 1. The following measurements are made in the bentonite in each of the two instrumented deposition holes in Section 1 (1 and 3): Temperature is measured in 32 points, total pressure in 27 points, pore water pressure in 14 points and relative humidity in 37 points. Temperature is also measured by all relative humidity gauges. Every measuring point is related to a local coordinate system in the deposition hole. The following measurements are made in the backfill in Section 1. Temperature is measured in 20 points, total pressure in 18 points, pore water pressure in 23 points and relative humidity in 45 points. Temperature is also measured by all relative humidity gauges. Furthermore, water content is measured by an electric chain in one section. Every measuring point is related to a local coordinate system in the tunnel. The following measurements are made on the surface of the canisters in Section 1: Temperature is measured every meter along two fiber optic cables. Furthermore, displacements of the canister in hole 3 are measured with 6 gauges. The following measurements are made in the rock in Section 1: Temperature is measured in 37 points in boreholes in the floor. Water

  1. Determination of radon concentration in soil air and in the radioactive spring's bathroom air by passive method

    Energy Technology Data Exchange (ETDEWEB)

    Horiuchi, Kimiko [Otsuma Women' s Univ., School of Social Information Studies, Tama, Tokyo (Japan); Ishii, Tadashi [Yamanashi Medical Univ., Radioisotope Lab., Tamaho, Yamanashi (Japan)

    2002-06-01

    There are many kind of passive methods such as open vial method, {alpha}-track method and active carbon method (PICO-RAD detector) or so. The open vial method can determine radon in soil air, which dissolves easily into toluene and afterwards can be measured by the integral counting with a liquid scintillation counter. The {alpha}-track method can record {alpha} ray tracks of radon on thin cellulose nitrate film which can be rendered visible for counting after NaOH solution treatment. The PICO-RAD detectors are passive devices requiring no power. They are integrating detectors used to determine the average radon concentration in air where they are placed. The detectors consist of a plastic liquid scintillation vial which consists a porous canister held securely near the top of the vial. The porous canister contains a bed of a controlled weight of charcoal covered by a layer of desiccant. The securely capped canister has and indefinite shelf life. The {alpha} radioactivity of radon gas adsorbed in fine active charcoal exposed in the air is measured with a liquid scintillation counter. Using of those simple detectors, we have measured distribution of radon concentration in soil air and in the radioactive spring's bathroom air. (author)

  2. Continuing the Validation of CCIM Processability for Glass Ceramic HLLW Forms: Plan for Test AFY14CCIM-GC1

    Energy Technology Data Exchange (ETDEWEB)

    Vince Maio

    2014-04-01

    This test plan covers test AFY14CCIM-GC1which is the first of two scheduled FY-2014 test runs involving glass ceramic waste forms in the Idaho National Laboratory’s Cold Crucible Induction Melter Pilot Plant. The test plan is based on the successes and challenges of previous tests performed in FY-2012 and FY-2013. The purpose of this test is to continue to collect data for validating the glass ceramic High Level Liquid Waste form processability advantages using Cold Crucible Induction Melter technology. The major objective of AFYCCIM-GC1 is to complete additional proposed crucible pouring and post tapping controlled cooling experiments not completed during previous tests due to crucible drain failure. This is necessary to qualify that no heat treatments in standard waste disposal canisters are necessary for the operational scale production of glass ceramic waste forms. Other objectives include the production and post-test analysis of surrogate waste forms made from separate pours into the same graphite mold canister, testing the robustness of an upgraded crucible bottom drain and drain heater assembly, testing the effectiveness of inductive melt initiation using a resistive starter ring with a square wave configuration, and observing the tapped molten flow behavior in pans with areas identical to standard High Level Waste disposal canisters. Testing conditions, the surrogate waste composition, key testing steps, testing parameters, and sampling and analysis requirements are defined.

  3. Progress on Fuel Receiving and Storage Decontamination Work at the West Valley Demonstration Project

    Energy Technology Data Exchange (ETDEWEB)

    Jablonski, J. F.; Al-Daouk, A. M.; Moore, H. R.

    2003-02-25

    The West Valley Demonstration Project (WVDP) removed the last of its spent nuclear fuel assemblies from an on-site storage pool last year and is now decontaminating its Fuel Receiving and Storage (FRS) Facility. The decontamination project will reduce the long-lived curie inventory, associated radiological hazards, and the operational costs associated with the maintenance of this facility. Workers at the WVDP conducted the first phase of the FRS decontamination project in late 2001 by removing 149 canisters that previously contained spent fuel assemblies from the pool. Removal of the canisters from the pool paved the way for nuclear divers to begin removing canister storage racks and other miscellaneous material from the FRS pool in February 2002. This was only the third time in the history of the WVDP that nuclear divers were used to perform underwater work. After decontaminating the pool, it will be drained slowly until all of the water is removed. The water will be processed through an ion exchanger to remove radioactive contaminants as it is being drained, and a fixative will be applied to the walls above the water surface to secure residual contamination.

  4. Characteristics of borosilicate waste glass form for high-level radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Soo; Chun, Kwan Sik; Choi, Jong Won; Kang, Chul Hyung

    2001-03-01

    Basic data, required for the design and the performance assessment of a repository of HLW, suchas the chemical composition and the characteristics of the borosilicate waste glass have been identified according to the burn-ups of spent PWR fuels. The diemnsion of waste canister is 430mm in diameter and 1135mm in length, and the canister should hold less than 2kwatts of heat from their decay of radionuclides contained in the HLW. Based on the reprocessing of 5 years-cooled spent fuel, one canister could hold about 11.5wt.% and 10.8wt.% of oxidized HLW corresponding to their burn-ups of 45,000MWD/MTU and 55,000MWD/MTU, respectively. These waste forms have been recommanded as the reference waste forms of HLW. The characteristics of these wastes as a function of decay time been evaluated. However, after a specific waste form and a specific site for the disposal would be selected, the characteristics of the waste should be reevaluated under the consideration of solidification period, loaded waste, storage condition and duration, site circumstances for the repository system and its performance assessment.

  5. Spanish participation in the Haw Project: Laboratory investigations on Gamma irradiation effects in rock salt

    Energy Technology Data Exchange (ETDEWEB)

    Cuevas, C. de las; Miralles, L.; Teixidor, P.; Garcia Veigas, J.; Dies, X.; Ortega, X.; Pueyo, J.J.

    1993-12-31

    In order to prove the safe disposal of high-level radioactive waste (HAW) in salt rock, a five years test disposal of thirty highly radioactive radiation sources is planned in the Asse salt mine, in the Federal Republic of Germany. The thirty radiation sources consist of steel canisters containing the vitrified radionuclides Caesium 137 and Strontium 90 in quantities sufficient to cover the bandwidth of heat generation and gamma radiation of real HAW. The radiation sources will be emplaced in six boreholes located in two galleries at the 800 m level. Two electrical heater tests were already started in November 1988 and are continuosly surveyed in respect of the rock mass. Also the handling system necessary for the emplacement of the radioactive canisters was developed and succesfully tested. A laboratory investigation programme on radiation effects in salt is being performed in advance to the radioactive canister emplacement. This programme includes the investigation of thermally and radiolytically induced water and gas release from the rock salt and the radiolytical decomposition of salt minerals. Part of this programme has been carried out since 1988 at the University of Barcelona, basically what refers to colloidal sodium determinations by light absorption measurements and microstructural studies on irradiated salt samples. For gamma dose and dose rate measurements in the test field, measuring systems consisting of ionisation chambers as well as solid state dosemeters were developed and tested. Thermomechanical computer code validation is performed by calculational predictions and parallel investigation of the stress and displacement fields in the underground test field.

  6. Natural radiation in Tenerife (Canary Islands)

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez-Aldecoa, J.C.; Robayna, B.; Allende, A.; Hernandez-Armas, J. (La Laguna Univ., Tenerife (Spain). Dept. de Medicina Fisica y Farmacologia); Poffijn, A. (Ghent Rijksuniversiteit (Belgium). Lab. voor Kernfysica)

    1992-01-01

    Preliminary results of outdoor gamma radioactivity measurements, specific activities of radionuclides in the soil and indoor radon concentrations in Tenerife (Canary Islands) are presented here. The results were obtained using GM MC-71 detectors, HPIC RSS-112, intrinsic Ge detector, charcoal canister and etched track. Outdoor gamma radioactivity levels were determined in each of 103 sites into which the island was divided up. A soil sample was taken at each site to determine concentrations of [sup 226R]a, [sup 232]Th and [sup 40]K. The charcoal canisters were placed for 3 days in dwellings located in the most densely populated areas of the island. In 13 dwellings etched track detectors were also placed for 3 months. The mean gamma radioactivity level was 67 nGy.h[sup -1]. Specific activities of [sup 226]Ra, [sup 232]Th and [sup 40]K were 44 Bq.kg[sup -1], 54 Bq.kg[sup -1] and 665 Bq.kg[sup -1], respectively. The median values for radon concentrations were 58 Bq.m[sup -3] (etched track) and 37 Bq.m[sup -3] (charcoal canisters). A thorough survey of environmental radioactivity in the Canary Islands archipelago is to be undertaken. (author).

  7. Effects of silica redistribution on performance of high-level nuclear waste repositories in saturated geologic formations

    Energy Technology Data Exchange (ETDEWEB)

    Verma, A.; Pruess, K.

    1985-11-01

    Evaluation of the thermohydrological conditions near high-level waste packages is needed for the design of the waste canister and for overall repository design and performance assessment. Most available studies in this area have assumed that the hydrologic properties of the host rock do not change in response to the thermal, mechanical or chemical effects caused by waste emplacement. However, the ramifications of this simplifying assumption have not been substantiated. We have studied dissolution and precipitation of silica in thermally driven flow systems, including changes in formation porosity and permeability. Using numerical simulation, we compare predictions of thermohydrological conditions with and without inclusion of silica redistribution effects. Two cases were studied, namely, a canister-scale problem, a repository-wide thermal convection problem, and different pore models were employed for the permeable medium (fractures with uniform or non-uniform cross sections). We find that silica redistribution generally has insignificant effects on host rock and canister temperatures, pore pressures, or flow velocites.

  8. Mineral formation on metallic copper in a `Future repository site environment`: Textural considerations based on natural analogs

    Energy Technology Data Exchange (ETDEWEB)

    Amcoff, Oe. [Uppsala Univ. (Sweden). Inst. of Earth Sciences

    1998-01-01

    Copper mineral formation in the Swedish `repository site environment` is discussed. Special attention is given to ore mineral textures (=the spatial relation among minerals), with examples given from nature. It is concluded: By analogy with observations from natural occurrences, an initial coating of Cu-oxide on the canister surface (because of entrapped air during construction) will probably not hinder a later sulphidation process. Early formation of Cu-sulphides on the canister surface may be accompanied by formation of CuFe-sulphides. The latter phase(s) may form through replacement of the Cu-sulphides or, alternatively, by means of reaction between dissolved copper and fine-grained iron sulphide (pyrite) in the surrounding bentonite. Should for some reason the bentonite barrier fail and the conditions become strongly oxidizing, we can expect crustifications and rhythmic growths of Cu(II)-phases, like malachite (Cu{sub 2}(OH){sub 2}CO{sub 3}). A presence of Fe{sup 2} in the clay minerals making up the bentonite might prove to have an adverse effect on the canister stability, since, in this case, the bentonite might be expected to act as a sink for dissolved copper. The mode of mineral growth along the copper - bentonite interface remains an open question.

  9. Retrieval options study

    Energy Technology Data Exchange (ETDEWEB)

    1980-03-01

    This Retrieval Options Study is part of the systems analysis activities of the Office of Nuclear Waste Isolation to develop the scientific and technological bases for radioactive waste repositories in various geologic media. The study considers two waste forms, high level waste and spent fuel, and defines various classes of waste retrieval and recovery. A methodology and data base are developed which allow the relative evaluation of retrieval and recovery costs and the following technical criteria: safety; technical feasibility; ease of retrieval; probable intact retrieval time; safeguards; monitoring; criticality; and licensability. A total of 505 repository options are defined and the cost and technical criteria evaluated utilizing a combination of facts and engineering judgments. The repositories evaluated are selected combinations of the following parameters: Geologic Media (salt, granite, basalt, shale); Retrieval Time after Emplacement (5 and 25 years); Emplacement Design (nominal hole, large hole, carbon steel canister, corrosion resistant canister, backfill in hole, nominal sleeves, thick wall sleeves); Emplacement Configuration (single vertical, multiple vertical, single horizontal, multiple horizontal, vaults; Thermal Considerations; (normal design, reduced density, once-through ventilation, recirculated ventilation); Room Backfill; (none, run-of-mine, early, 5 year delay, 25 year delay, decommissioned); and Rate of Retrieval; (same as emplacement, variably slower depending on repository/canister condition).

  10. Architecture Study for a Fuel Depot Supplied from Lunar Assets

    Science.gov (United States)

    Perrin, Thomas M.; Casler, James G.

    2016-01-01

    This architecture study sought to determine the optimum architecture for a fuel depot supplied from lunar assets. Four factors - the location of propellant processing (on the Moon or on the depot), the depot location (on the Moon, L1, GEO, or LEO), the propellant transfer location (L1, GEO, or LEO), and the propellant transfer method (bulk fuel or canister exchange) were combined to identify 18 candidate architectures. Two design reference missions (DRMs) - a commercial satellite servicing mission and a Government cargo mission to Mars - created demand for propellants, while a propellant delivery DRM examined supply issues. The study concluded Earth-Moon L1 is the best location for an orbiting depot. For all architectures, propellant boiloff was less than anticipated, and was far overshadowed by delta-v requirements and resulting fuel consumption. Bulk transfer is the most flexible for both the supplier and customer. However, since canister exchange bypasses the transfer of bulk cryogens and necessary chilldown losses, canister exchange shows promise and merits further investigation. Overall, this work indicates propellant consumption and loss is an essential factor in assessing fuel depot architectures.

  11. Thermal performance sensitivity studies in support of material modeling for extended storage of used nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Cuta, Judith M.; Suffield, Sarah R.; Fort, James A.; Adkins, Harold E.

    2013-08-15

    The work reported here is an investigation of the sensitivity of component temperatures of a storage system, including fuel cladding temperatures, in response to age-related changes that could degrade the design-basis thermal behavior of the system. Three specific areas of interest were identified for this study. • degradation of the canister backfill gas from pure helium to a mixture of air and helium, resulting from postulated leakage due to stress corrosion cracking (SCC) of canister welds • changes in surface emissivity of system components, resulting from corrosion or other aging mechanisms, which could cause potentially significant changes in temperatures and temperature distributions, due to the effect on thermal radiation exchange between components • changes in fuel and basket temperatures due to changes in fuel assembly position within the basket cells in the canister The purpose of these sensitivity studies is to provide a realistic example of how changes in the physical properties or configuration of the storage system components can affect temperatures and temperature distributions. The magnitudes of these sensitivities can provide guidance for identifying appropriate modeling assumptions for thermal evaluations extending long term storage out beyond 50, 100, 200, and 300 years.

  12. Spray Calciner/In-Can Melter high-level waste solidification technical manual

    Energy Technology Data Exchange (ETDEWEB)

    Larson, D.E. (ed.)

    1980-09-01

    This technical manual summarizes process and equipment technology developed at Pacific Northwest Laboratory over the last 20 years for vitrification of high-level liquid waste by the Spray Calciner/In-Can Melter process. Pacific Northwest Laboratory experience includes process development and demonstration in laboratory-, pilot-, and full-scale equipment using nonradioactive synthetic wastes. Also, laboratory- and pilot-scale process demonstrations have been conducted using actual high-level radioactive wastes. In the course of process development, more than 26 tonnes of borosilicate glass have been produced in 75 canisters. Four of these canisters contained radioactive waste glass. The associated process and glass chemistry is discussed. Technology areas described include calciner feed treatment and techniques, calcination, vitrification, off-gas treatment, glass containment (the canister), and waste glass chemistry. Areas of optimization and site-specific development that would be needed to adapt this base technology for specific plant application are indicated. A conceptual Spray Calciner/In-Can Melter system design and analyses are provided in the manual to assist prospective users in evaluating the process for plant application, to provide equipment design information, and to supply information for safety analyses and environmental reports. The base (generic) technology for the Spray Calciner/In-Can Melter process has been developed to a point at which it is ready for plant application.

  13. Analysis of dust samples collected from spent nuclear fuel interim storage containers at Hope Creek, Delaware, and Diablo Canyon, California.

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R.; Enos, David George

    2014-07-01

    Potentially corrosive environments may form on the surface of spent nuclear fuel dry storage canisters by deliquescence of deposited dusts. To assess this, samples of dust were collected from in-service dry storage canisters at two near-marine sites, the Hope Creek and Diablo Canyon storage installations, and have been characterized with respect to mineralogy, chemistry, and texture. At both sites, terrestrially-derived silicate minerals, including quartz, feldspars, micas, and clays, comprise the largest fraction of the dust. Also significant at both sites were particles of iron and iron-chromium metal and oxides generated by the manufacturing process. Soluble salt phases were minor component of the Hope Creek dusts, and were compositionally similar to inland salt aerosols, rich in calcium, sulfate, and nitrate. At Diablo Canyon, however, sea-salt aerosols, occurring as aggregates of NaCl and Mg-sulfate, were a major component of the dust samples. The seasalt aerosols commonly occurred as hollow spheres, which may have formed by evaporation of suspended aerosol seawater droplets, possibly while rising through the heated annulus between the canister and the overpack. The differences in salt composition and abundance for the two sites are attributed to differences in proximity to the open ocean and wave action. The Diablo Canyon facility is on the shores of the Pacific Ocean, while the Hope Creek facility is on the shores of the Delaware River, several miles from the open ocean.

  14. Survival of bacteria in nuclear waste buffer materials. The influence of nutrients, temperature and water activity

    Energy Technology Data Exchange (ETDEWEB)

    Pedersen, K.; Motamedi, M. [Goeteborg Univ. (Sweden). Dept. of General and Marine Microbiology; Karnland, O. [Clay Technology AB, Lund (Sweden)

    1995-12-01

    The concept of deep geological disposal of spent fuel is common to many national nuclear waste programs. Long-lived radioactive waste will be encapsulated in canisters made of corrosion resistant materials e.g. copper and buried several hundred meters below ground in a geological formation. Different types of compacted bentonite clay, or mixtures with sand, will be placed as a buffer around the waste canisters. A major concern for the performance of the canisters is that sulphate-reducing bacteria (SRB) may be present in the clay and induce corrosion by production of hydrogen sulphide. This report presents data on viable counts of SRB in the bedrock of Aespoe hard rock laboratory. A theoretical background on the concept water activity is given, together with basic information about SRB. Some results on microbial populations from a full scale buffer test in Canada is presented. These results suggested water activity to be a strong limiting factor for survival of bacteria in compacted bentonite. As a consequence, experiments were set up to investigate the effect from water activity on survival of SRB in bentonite. Here we show that survival of SRB in bentonite depends on the availability of water and that compacting a high quality bentonite to a density of 2.0 g/cm{sup 3}, corresponding to a water activity (a{sub w}) of 0.96, prevented SRB from surviving in the clay. 24 refs.

  15. Kinetics and mechanisms of reactions between H2O2 and copper and copper oxides.

    Science.gov (United States)

    Björkbacka, Åsa; Yang, Miao; Gasparrini, Claudia; Leygraf, Christofer; Jonsson, Mats

    2015-09-28

    One of the main challenges for the nuclear power industry today is the disposal of spent nuclear fuel. One of the most developed methods for its long term storage is the Swedish KBS-3 concept where the spent fuel is sealed inside copper canisters and placed 500 meters down in the bedrock. Gamma radiation will penetrate the canisters and be absorbed by groundwater thereby creating oxidative radiolysis products such as hydrogen peroxide (H2O2) and hydroxyl radicals (HO˙). Both H2O2 and HO˙ are able to initiate corrosion of the copper canisters. In this work the kinetics and mechanism of reactions between the stable radiolysis product, H2O2, and copper and copper oxides were studied. Also the dissolution of copper into solution after reaction with H2O2 was monitored by ICP-OES. The experiments show that both H2O2 and HO˙ are present in the systems with copper and copper oxides. Nevertheless, these species do not appear to influence the dissolution of copper to the same extent as observed in recent studies in irradiated systems. This strongly suggests that aqueous radiolysis can only account for a very minor part of the observed radiation induced corrosion of copper.

  16. Analysis of dust samples collected from spent nuclear fuel interim storage containers at Hope Creek, Delaware, and Diablo Canyon, California

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Enos, David George [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-07-01

    Potentially corrosive environments may form on the surface of spent nuclear fuel dry storage canisters by deliquescence of deposited dusts. To assess this, samples of dust were collected from in-service dry storage canisters at two near-marine sites, the Hope Creek and Diablo Canyon storage installations, and have been characterized with respect to mineralogy, chemistry, and texture. At both sites, terrestrially-derived silicate minerals, including quartz, feldspars, micas, and clays, comprise the largest fraction of the dust. Also significant at both sites were particles of iron and iron-chromium metal and oxides generated by the manufacturing process. Soluble salt phases were minor component of the Hope Creek dusts, and were compositionally similar to inland salt aerosols, rich in calcium, sulfate, and nitrate. At Diablo Canyon, however, sea-salt aerosols, occurring as aggregates of NaCl and Mg-sulfate, were a major component of the dust samples. The seasalt aerosols commonly occurred as hollow spheres, which may have formed by evaporation of suspended aerosol seawater droplets, possibly while rising through the heated annulus between the canister and the overpack. The differences in salt composition and abundance for the two sites are attributed to differences in proximity to the open ocean and wave action. The Diablo Canyon facility is on the shores of the Pacific Ocean, while the Hope Creek facility is on the shores of the Delaware River, several miles from the open ocean.

  17. Critical review of the literature on the corrosion of copper by water

    Energy Technology Data Exchange (ETDEWEB)

    King, Fraser (Integrity Corrosion Consulting Limited (Canada))

    2010-12-15

    The conventional belief that copper is thermodynamically stable in oxygen-free water has been questioned by a research group from the Royal Inst. of Technology, Stockholm lead by Dr. Gunnar Hultquist. A critical review of the literature both in support of the proposed mechanism and that which argues against it has been conducted. The critical review has been supported by supplementary analyses, with particular focus on the scientific validity of the reported observations and their significance for the corrosion of a copper canister. It is found that: - the scientific evidence in support of the suggestion that water oxidises copper is not conclusive and there are many aspects which are unclear and contradictory, - despite a number of attempts, no other researchers have been able to reproduce the observations of Hultquist and co-workers, - even if correct, the mechanism is not important for copper canisters in a repository, both because of differences in the environmental conditions and because, even if corrosion did occur by this mechanism, it would quickly stop, and - there is no adverse impact on the lifetime of copper canisters due to this proposed, but unproven, mechanism

  18. A Superfluid Pulse Tube Refrigerator Without Moving Parts for Sub-Kelvin Cooling

    Science.gov (United States)

    Miller, Franklin K.

    2012-01-01

    A report describes a pulse tube refrigerator that uses a mixture of He-3 and superfluid He-4 to cool to temperatures below 300 mK, while rejecting heat at temperatures up to 1.7 K. The refrigerator is driven by a novel thermodynamically reversible pump that is capable of pumping the He-3 He-4 mixture without the need for moving parts. The refrigerator consists of a reversible thermal magnetic pump module, two warm heat exchangers, a recuperative heat exchanger, two cold heat exchangers, two pulse tubes, and an orifice. It is two superfluid pulse tubes that run 180 out of phase. All components of this machine except the reversible thermal pump have been demonstrated at least as proof-of-concept physical models in previous superfluid Stirling cycle machines. The pump consists of two canisters packed with pieces of gadolinium gallium garnet (GGG). The canisters are connected by a superleak (a porous piece of VYCOR glass). A superconducting magnetic coil surrounds each of the canisters.

  19. RD and D-Programme 2001. Programme for research, development and demonstration of methods for the management and disposal of nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-09-01

    An overall goal for SKB is to start the initial operation of a deep repository for spent fuel in 2015. This presumes that site investigations have been commenced at the beginning of 2002 and that the different phases have been executed without major changes. The encapsulation plant should be ready to start roughly one year before the deep repository is finished. Spent fuel is the waste that is to be isolated in the deep repository. Various processes will with time alter the conditions in the fuel and in the voids of the canister. Many of these process only occur if the isolation of the canister is breached and water enters the canister. Radiolysis of water is an example of such a process, which can in turn influence the chemical conditions in the canister. Water in the canister can also cause corrosion of the fuel's cladding tubes. If water comes into contact with the fuel it can lead to dissolution of radionuclides. Dissolved radionuclides can diffuse in the water and thereby escape from a damaged canister. Fuel dissolution is a priority area in RDandD Programme 2001. Large resources are being devoted to studies of copper corrosion and stress corrosion cracking in the copper canister. SKB will also investigate the long-term safety of a canister type with a slightly thinner shell but a heavier-duty insert. The buffer of bentonite clay is supposed to protect the canister mechanically against minor rock movements. It is also supposed to retard solute transport. The initial evolution of the buffer is studied in the Aespoe HRL and by means of models. The long-term evolution of the backfill is controlled by largely the same processes as in the buffer. The backfill is more sensitive to saline water than the more compacted buffer. Several processes in the geosphere are important for the safety assessment, such as groundwater flow, earthquakes, microbial processes and matrix diffusion. The models for groundwater flow will be further refined in order to handle the

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-08-01

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

  1. Project JADE. Description of the MLH-method; Projekt JADE. Beskrivning av MLH-metoden

    Energy Technology Data Exchange (ETDEWEB)

    Sandstedt, H.; Munier, R. [Scandiaconsult, Stockholm (Sweden); Wichmann, C. [Nitro Consult AB, Stockholm (Sweden); Isaksson, Therese [Royal Inst. of Tech., Stockholm (Sweden). Div. of Soil and Rock Mechanics

    2001-08-01

    This report constitutes a part of a series of reports within project JADE, comparison of deposition methods. A comparison of the deposition methods MLH (Medium Long Holes with approximately 25 copper canisters emplaced in a horizontal deposition hole about 200 metres in length bored between central and side tunnels) and KBS-3 (copper canisters are emplaced in vertical deposition holes bored in the floors of horizontal tunnels) has earlier been performed and KBS-3 was judged to be more advantageous than MLH. However, the prerequisites for the comparison have changed with time and an updated evaluation of MLH was therefore required. In this report, the current knowledge of MLH is summarized with focus on geological prerequisites, methods for boring long, horizontal deposition holes, reinforcement and sealing, deposition and cost. Comparisons with KBS-3 are performed sequentially. An MLH-repository is judged to be more sensitive to ingress of water to the deposition holes during the deposition process. This implies that a MLH repository based on today's knowledge is basically recommended for bedrock with fairly low water baring capacity. It has been demonstrated that MLH has considerable economic potential compared to KBS-3. However, the method is judged to be more technically immature than KBS-3. Particularly, methods and equipment for deposition of canisters need to be developed further. Methods and equipment for deposition can be developed, which fulfill the demands on function and safety, in the near future. MLH cannot therefore be rejected as deposition method.

  2. New X-ray testing methods of aerosol products for industrial radiography

    Science.gov (United States)

    Bozydar Knyziak, Adrian; Rzodkiewicz, Witold; Kaczorowska, Ewa; Derlacinski, Michal

    2017-02-01

    An amount of product in e.g. an aerosol canister is not difficult to estimate by weighing a filled can and subtracting the tare of packaging. In this way, we can obtain the net weight of the ingredients present in the can. Although, this does not indicate the volumetric content. Therefore, in the paper, the fundamental (the weight method and given by FEICA) and new methods (given by authors) related to the determination of the volumetric content of canister filled with aeorosol products are presented. The new methods are based on direct digital radiography (DR) using X-ray radiation. For the needs of new methods, the X-ray CCD-DR imaging system was built and developed in our Laboratory in Department of Radiation and Vibration at the Central Office of Measures. For comparison purposes, with regard to the volumetric content, a lot of metal cans of capacities 140, 185, 450, 700 ml were inspected. In future, computed tomography (CT) for industrial radiography in our laboratory will be used. Currently, an algorithm for CT is being tested. It will give us possibility for very precise measurements to determine volumetric content of examined canisters.

  3. Status update of the BWR cask simulator

    Energy Technology Data Exchange (ETDEWEB)

    Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Durbin, Samuel G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-09-01

    The performance of commercial nuclear spent fuel dry storage casks are typically evaluated through detailed numerical analysis of the system's thermal performance. These modeling efforts are performed by the vendor to demonstrate the performance and regulatory compliance and are independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Numerous studies have been previously conducted. Recent advances in dry storage cask designs have moved the storage location from above ground to below ground and significantly increased the maximum thermal load allowed in a cask in part by increasing the canister helium pressure. Previous cask performance validation testing did not capture these parameters. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern dry casks. These modern cask designs utilize elevated helium pressure in the sealed canister or are intended for subsurface storage. The BWR cask simulator (BCS) has been designed in detail for both the above ground and below ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the canister. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. Various configurations of outer concentric ducting will be used to mimic conditions for above and below ground storage configurations

  4. Dynamic Acquisition and Retrieval Tool (DART) for Comet Sample Return : Session: 2.06.Robotic Mobility and Sample Acquisition Systems

    Science.gov (United States)

    Badescu, Mircea; Bonitz, Robert; Kulczycki, Erick; Aisen, Norman; Dandino, Charles M.; Cantrell, Brett S.; Gallagher, William; Shevin, Jesse; Ganino, Anthony; Haddad, Nicolas; Walkemeyer, Phillip; Backes, Paul; Shiraishi, Lori

    2013-01-01

    The 2011 Decadal Survey for planetary science released by the National Research Council of the National Academies identified Comet Surface Sample Return (CSSR) as one of five high priority potential New Frontiers-class missions in the next decade. The main objectives of the research described in this publication are: develop a concept for an end-to-end system for collecting and storing a comet sample to be returned to Earth; design, fabricate and test a prototype Dynamic Acquisition and Retrieval Tool (DART) capable of collecting 500 cc sample in a canister and eject the canister with a predetermined speed; identify a set of simulants with physical properties at room temperature that suitably match the physical properties of the comet surface as it would be sampled. We propose the use of a dart that would be launched from the spacecraft to impact and penetrate the comet surface. After collecting the sample, the sample canister would be ejected at a speed greater than the comet's escape velocity and captured by the spacecraft, packaged into a return capsule and returned to Earth. The dart would be composed of an inner tube or sample canister, an outer tube, a decelerator, a means of capturing and retaining the sample, and a mechanism to eject the canister with the sample for later rendezvous with the spacecraft. One of the significant unknowns is the physical properties of the comet surface. Based on new findings from the recent Deep Impact comet encounter mission, we have limited our search of solutions for sampling materials to materials with 10 to 100 kPa shear strength in loose or consolidated form. As the possible range of values for the comet surface temperature is also significantly different than room temperature and testing at conditions other than the room temperature can become resource intensive, we sought sample simulants with physical properties at room temperature similar to the expected physical properties of the comet surface material. The chosen

  5. Toward Improvements in Inter-laboratory Calibration of Argon Isotope Measurements

    Science.gov (United States)

    Hemming, S. R.; Deino, A. L.; Heizler, M. T.; Hodges, K. V.; McIntosh, W. C.; Renne, P. R.; Swisher, C. C., III; Turrin, B. D.; Van Soest, M. C.

    2015-12-01

    It is important to continue to develop strategies to improve our ability to compare results between laboratories chronometers. The U-Pb community has significantly reduced inter-laboratory biases with the application of a community tracer solution and the distribution of synthetic zircon solutions. Inevitably sample selection and processing and even biases in interpretations will still lead to some disagreements in the assignment of ages. Accordingly natural samples that are shared will be important for achievement of the highest levels of agreement. Analogous improvements in quality and inter-laboratory agreement of analytical aspects of Ar-Ar can be achieved through development of synthetic age standards in gas canisters with multiple pipettes to deliver various controlled amounts of argon to the mass spectrometer. A preliminary proof-of concept comes from the inter-laboratory calibration experiment for the 40Ar/39Ar community. This portable Argon Pipette Intercalibration System (APIS) consists of three 2.7 L canisters each equipped with three pipettes of 0.1, 0.2 and 0.4 cc volumes. The currently traveling APIS has the three canisters filled with air and 40Ar*/39Ar of 1.73 and canister 2 has a 40Ar*/39Ar of 40.98 (~ Alder Creek and Fish Canyon in the same irradiation). With these pipettes it is possible to combine them to provide 0.1, 0.2, 0.3 (0.1+0.2), 0.4, 0.5 (0.1+0.4), 0.6 (0.2+0.4), and 0.7 (0.1+0.2+0.4) cc. The configuration allows a simple test for inter-laboratory biases and for volume/pressure dependent mass fractionation on the measured ratios for a gas with a single argon isotope composition. Although not yet tested, it is also possible to mix gas from any one of the three canisters in proportions of these increments, allowing even more tightly controlled calibration of measurements. We suggest that ultimately each EARTHTIME lab should be equipped with such a system permanently, with a community plan for a traveling system to periodically repeat the

  6. HLW Disposal System Development

    Energy Technology Data Exchange (ETDEWEB)

    Choi, J. W.; Choi, H. J.; Lee, J. Y. (and others)

    2007-06-15

    A KRS is suggested through design requirement analysis of the buffer and the canister which are the constituent of disposal system engineered barrier and HLW management plans are proposed. In the aspect of radionuclide retention capacity, the thickness of the buffer is determined 0.5m, the shape to be disc and ring and the dry density to be 1.6 g/cm{sup 3}. The maximum temperature of the buffer is below 100 .deg. which meets the design requirement. And bentonite blocks with 5 wt% of graphite showed more than 1.0 W/mK of thermal conductivity without the addition of sand. The result of the thermal analysis for proposed double-layered buffer shows that decrease of 7 .deg. C in maximum temperature of the buffer. For the disposal canister, the copper for the outer shell material and cast iron for the inner structure material is recommended considering the results analyzed in terms of performance of the canisters and manufacturability and the geochemical properties of deep groundwater sampled from the research area with granite, salt water intrusion, and the heavy weight of the canister. The results of safety analysis for the canister shows that the criticality for the normal case including uncertainty is the value of 0.816 which meets subcritical condition. Considering nation's 'Basic Plan for Electric Power Demand and Supply' and based on the scenario of disposing CANDU spent fuels in the first phase, the disposal system that the repository will be excavated in eight phases with the construction of the Underground Research Laboratory (URL) beginning in 2020 and commissioning in 2040 until the closure of the repository is proposed. Since there is close correlation between domestic HLW management plans and front-end/back-end fuel cycle plans causing such a great sensitivity of international environment factor, items related to assuring the non-proliferation and observing the international standard are showed to be the influential factor and acceptability

  7. Safety assessment of spent fuel disposal in Haestholmen, Kivetty, Olkiluoto and Romuvaara - TILA-99

    Energy Technology Data Exchange (ETDEWEB)

    Vieno, T.; Nordman, H. [VTT Energy (Finland)

    1999-03-01

    The spent fuel from the Finnish nuclear power plants is planned to be disposed of in copper-iron canisters emplaced in a KBS-3 type repository constructed at a depth of about 500 metres at one of the four candidate sites investigated. The disposal concept aims at long-term isolation of the spent fuel assemblies from the biosphere and even from the geosphere. The evaluation of the normal evolution of the disposal system accords with the conclusions of the previous Finnish, Swedish and Canadian safety assessments of similar disposal concepts. Subject to the influence of the expected, normal evolution of the repository, initially intact copper-iron canisters will most likely preserve their integrity for more than one million years at any of the candidate sites. Consequently, the best-estimate assessment is that there never will be any significant releases of radionuclides from the repository into the geosphere. Consequences of potential canister failures have been evaluated using conservative assumptions, models and data. The results show that at any of the sites a large number of canisters could be assumed to be initially defective or to `disappear` simultaneously after some time without that the proposed constraints for release rates into the biosphere or dose rates were exceeded. In most cases this conclusion is valid for all canisters failing simultaneously, even if rather pessimistic flow and transport data is used. In the sensitivity and `what if` analyses where very high flow rates of saline groundwater are assumed, highest release and dose rates are caused by weakly-sorbing cations Sr-90 and Ra-226. The most important differences between the sites are related to the coastal location and brackish/saline groundwater of Haestholmen and Olkiluoto, and on the other hand to the inland location and fresh groundwater of Kivetty and Romuvaara. Because of the ongoing postglacial land uplift at the coast of the Baltic Sea, Olkiluoto and Haestholmen, too, may become

  8. Evaluation of Codisposal Viability for Melt and Dilute DOE-Owned Fuel

    Energy Technology Data Exchange (ETDEWEB)

    H. Radulescu

    2001-07-31

    component criticality analyses. One or more addenda (validation reports) to the topical report will be required to establish the critical limit for DOE SNF form once sufficient critical benchmarks are identified and verified. The waste package design for MD ingots holds one 18-in.-outer diameter DOE standardized SNF canister containing the MD ingots, and five defense high-level radioactive waste (DHLW) glass canisters. The 5-DHLW/DOE SNF-short waste package is based on the Site Recommendation design of waste packages (Appendix A). The waste package design consists of two concentric cylindrical shells in which the waste forms will be placed. The outer shell is made of a corrosion resistant nickel-based alloy (Alloy 22). The inner shell is composed of stainless steel 316 NG (nuclear grade). The waste package design incorporates three lids at the one end of the waste package (one for the inner shell and two for the outer shell) and two lids at the other end of the waste package (one for each shell). The DOE SNF canister containing three to six MD ingots is placed in a carbon steel support tube that becomes the center of the waste package. The DOE SNF canister is surrounded by five 3-meter-long Savannah River Site DHLW glass canisters. The five DHLW glass canisters are evenly spaced around the DOE SNF canister. This report presents the results of analyzing the 5-DHLWDOE SNF-short waste package against various design criteria. Section 2.2 provides the criteria, and Section 2.3 provides the key assumptions for the various analyses.

  9. BN-350 unattended safeguards system current status and initial fuel movement data

    Energy Technology Data Exchange (ETDEWEB)

    Williams, Richard Brady [Los Alamos National Laboratory; Browne, Michael C [Los Alamos National Laboratory; Parker, Robert F [Los Alamos National Laboratory; Ingegneri, Maurizio [IAEA

    2009-01-01

    The Unattended and Remote Monitoring (UNARM) system at the BN-350 fast breeder reactor facility in Aktau, Kazakhstan continues to provide safeguards monitoring data as the spent fuel disposition project transitions from wet fuel storage to dry storage casks. Qualitative data from the initial cask loading procedures has been released by the International Atomic Energy Agency (IAEA) and is presented here for the first time. The BN-350 fast breeder reactor in Aktau, Kazakhstan, operated as a plutonium-producing facility from 1973 W1til 1999. Kazakhstan signed the Nonproliferation Treaty (NPT) in February 1994, and shortly afterwards the IAEA began safeguarding the reactor facility and its nuclear material. Slnce the cessation of reactor operations ten years ago, the chief proliferation concern has been the spent fuel assemblies stored in the pond on-site. By 2002, all fuel assemblies in wet storage had been repackaged into proliferation-resistant canisters. From the beginning, the IAEA's safeguards campaign at the BN-350 included a constant unattended sensor presence in the form of UNARM which monitors nuclear material activities at the facility in the absence of inspector presence. The UNARM equipment at the BN-350 was designed to be modular and extensible, allowing the system to adapt as the safeguards requirements change. This has been particularly important at the BN-350 due to the prolonged wet storage phase of the project. The primary function of the BN-350 UNARM system is to provide the IAEA with an independent, radiation-centric Containment and Surveillance (C&S) layer in addition to the standard seals and video systems. The UNARM system has provided continuous Continuity of Knowledge (COK) data for the BN-350's nuclear material storage areas in order to ensure the validity of the attended measurements during the lifetime of the project. The first of these attended measurements was characterization of the spent fuel assemblies. This characterization

  10. Prototype repository - Microbes in the retrieved outer section

    Energy Technology Data Exchange (ETDEWEB)

    Arlinger, Johanna; Bengtsson, Andreas; Edlund, Johanna; Eriksson, Lena; Johansson, Jessica; Lydmark, Sara; Rabe, Lisa; Pedersen, Karsten [Microbial Analytics Sweden, Moelnlycke (Sweden)

    2013-10-15

    The Prototype repository is an international project to build and study a full-scale model of the planned Swedish final repository for spent nuclear fuel. The Prototype consists of two sections with four and two full-scale copper canisters, respectively. In 2011, the outer section with two canisters (nos. 5 and 6) was excavated. Groundwater surrounding the Prototype has been demonstrated to include microorganisms such as iron-reducing bacteria (IRB) and sulphate-reducing bacteria (SRB) with the ability to affect the repository through reduction of structural Fe(III) in the buffer or by the production of sulphide, respectively. During excavation, samples were taken for microbiological and molecular biological analysis from backfill, buffer, and canister surfaces and analysed with an emphasis on microbial presence and number. The underground environment is anaerobic, but the construction of a repository will raise the oxygen levels. Oxygen is not favourable for the longevity of the copper canister, but oxygen levels will decrease over time, partly due to microbial activity that consumes oxygen. Therefore, evaluating the presence and numbers of the heterotrophic aerobic bacteria that consume oxygen as well as monitoring the oxygen levels are important. The oxygen content of the bentonite itself is also a primary concern, and a method for measuring how the oxygen diffuses through the clay has long been needed. In the work reported here, we performed two pilot studies to address this need. One of these studies tested a method for differentiating between oxygen saturation in aerobic versus anaerobic bentonite; this method has potential for further development. The tunnel above the Prototype canisters was backfilled with a mixture of bentonite and crushed rock. Sixty-three randomly chosen samples from a cross-section through the backfill were analysed for culturable heterotrophic aerobic bacteria. All but one exhibited growth, with four samples exhibiting numbers over 106

  11. Deep repository for spent nuclear fuel. SR-97-Post-closure safety. Main Report. Volume I and II

    Energy Technology Data Exchange (ETDEWEB)

    Hedin, A. [ed.

    1999-11-01

    In preparation for coming site investigations for siting of a deep repository for spent nuclear fuel, the Swedish Government and nuclear regulatory authorities have requested an assessment of the repository's long-term safety. The purpose is to demonstrate whether the risk of harmful effects in individuals in the vicinity of the repository complies with the acceptance criterion formulated by the Swedish regulatory authorities, i.e. that the risk may not exceed 10{sup -6} per year. Geological data are taken from three sites in Sweden to shed light on different conditions in Swedish granitic bedrock. The future evolution of the repository system is analyzed in the form of five scenarios. The first is a base scenario where the repository is postulated to be built entirely according to specifications and where present-day conditions in the surroundings are postulated to persist. The four other scenarios show how the evolution of the repository differs from that in the base scenario if the repository contains a few initially defective canisters, in the event of climate change, earthquakes, and future inadvertent human intrusion. The time horizon for the analyses is at most one million years, in accordance with preliminary regulations. By means of model studies and calculations, the base scenario analyzes how the radioactivity of the fuel declines with time, the repository's thermal evolution as a result of the decay heat in the fuel, the hydraulic evolution in buffer and backfill when they become saturated with water, and the long-term groundwater flow in the geosphere on the three sites. The overall conclusion of the analyses in the base scenario is that the copper canisters isolating capacity is not threatened by either the mechanical or chemical stresses to which it is subjected. The safety margins are great even in a million-year perspective. The internal evolution in initially defective canisters and the possible resultant migration of radionuclides in

  12. Site-scale groundwater flow modelling of Ceberg

    Energy Technology Data Exchange (ETDEWEB)

    Walker, D. [Duke Engineering and Services (United States); Gylling, B. [Kemakta Konsult AB, Stockholm (Sweden)

    1999-06-01

    The Swedish Nuclear Fuel and Waste Management Company (SKB) SR 97 study is a comprehensive performance assessment illustrating the results for three hypothetical repositories in Sweden. In support of SR 97, this study examines the hydrogeologic modelling of the hypothetical site called Ceberg, which adopts input parameters from the SKB study site near Gideaa, in northern Sweden. This study uses a nested modelling approach, with a deterministic regional model providing boundary conditions to a site-scale stochastic continuum model. The model is run in Monte Carlo fashion to propagate the variability of the hydraulic conductivity to the advective travel paths from representative canister locations. A series of variant cases addresses uncertainties in the inference of parameters and the model of conductive fracturezones. The study uses HYDRASTAR, the SKB stochastic continuum (SC) groundwater modelling program, to compute the heads, Darcy velocities at each representative canister position, and the advective travel times and paths through the geosphere. The volumetric flow balance between the regional and site-scale models suggests that the nested modelling and associated upscaling of hydraulic conductivities preserve mass balance only in a general sense. In contrast, a comparison of the base and deterministic (Variant 4) cases indicates that the upscaling is self-consistent with respect to median travel time and median canister flux. These suggest that the upscaling of hydraulic conductivity is approximately self-consistent but the nested modelling could be improved. The Base Case yields the following results for a flow porosity of {epsilon}{sub f} 10{sup -4} and a flow-wetted surface area of a{sub r} = 0.1 m{sup 2}/(m{sup 3} rock): The median travel time is 1720 years. The median canister flux is 3.27x10{sup -5} m/year. The median F-ratio is 1.72x10{sup 6} years/m. The base case and the deterministic variant suggest that the variability of the travel times within

  13. Corrosion calculations report for the safety assessment SR-Site

    Energy Technology Data Exchange (ETDEWEB)

    2010-12-15

    This report is a compilation of the quantitative assessments of corrosion of the copper canisters in a KBS-3 repository. The calculations are part of the safety assessment SR-Site that is the long-term safety assessment to support the license application for building a final repository for spent nuclear fuel at Forsmark, Sweden. The safety assessment methodology gives the frame for the structured and documented approach to assess all conceivable corrosion processes. The quantitative assessments are done in different ways depending on the nature of the process and on the implications for the long-term safety. The starting point for the handling of the corrosion processes is the description of all known corrosion processes for copper with the current knowledge base and applied to the specific system and geology. Already at this stage some processes are excluded for further analysis, for example if the repository environment is not a sufficient prerequisite for the process to occur. The next step is to identify processes where the extent of corrosion could be bounded, e.g. by a mass balance approach. For processes where a mass balance is not limiting, the mass transport of corrodants (or corrosion products) is taken into account. A simple approach would be just to calculate the diffusive transport of corrodants through the bentonite, but generally the transport resistance for the interface between groundwater in a rock fracture intersecting the deposition hole and the bentonite buffer is more important. In SR-Site, the concept of equivalent flowrate, Q{sub eq}, is used. This assessment is done integrated with the evaluation of the geochemical and hydrogeological evolution of the repository. For most of the corrosion processes analysed, the corrosion depth is much smaller than the copper shell thickness, even for the assessment time of 106 years. Several processes give corrosion depths less than 100 mum, but no process give corrosion depths larger than a few

  14. Mechanical interaction buffer/backfill. Finite element calculations of the upward swelling of the buffer against both dry and saturated backfill

    Energy Technology Data Exchange (ETDEWEB)

    Boergesson, Lennart (Clay Technology AB, Lund (Sweden)); Hernelind, Jan (5T-Engineering AB, Vaesteraas (Sweden))

    2009-10-15

    The mechanical interaction between the buffer material in the deposition hole and the backfill material in the deposition tunnel is an important process in the safety assessment since the primary function of the backfill is to keep the buffer in place and not allow it to expand too much and thereby loose too much of its density and barrier properties. In order to study the upwards swelling of the buffer and the subsequent density reduction a number of finite element calculations have been performed. The calculations have been done with the FE-program Abaqus with 3D-models of a deposition hole and the deposition tunnel. In order to refine the modelling only the two extreme cases of completely un-wetted (dry) and completely water saturated (wet) backfill have been modelled. For the wet case the influence of different factors has been studied while only one calculation of the dry case has been done. The calculated upwards swelling of the buffer varied between 2 and 15 cm for the different wet cases while it was about 10 cm for the dry case. In the wet reference case the E-modulus of the block and pellets fillings was 50 MPa and 3.24 MPa respectively, the friction angle between the buffer and the rock and canister was 8.7 deg and there were no swelling pressure from the backfill. There is a strong influence of the friction angle on both the upwards swelling and the canister heave. The friction is important for preventing especially canister displacements. The unrealistic case of no friction yielded strong unacceptable influence on the buffer with an upwards swelling of 15 cm and a strong heave of 5 cm of the canister. The influence of the backfill stiffness is as expected strong. Both buffer swelling and canister heave are twice as large at the E-modulus E = 25 MPa than at the E-modulus E = 100 MPa. The influence of the stiffness of the pellets filling is not strong since there are no pellets on the floor in the model used. The influence of the swelling pressure of the

  15. Aespoe Hard Rock Laboratory. Annual Report 2008

    Energy Technology Data Exchange (ETDEWEB)

    2009-07-15

    Buffer Materials is to study clay materials that in laboratory tests have shown to be conceivable buffer materials. Three test parcels with different combinations of clay materials are installed in boreholes at Aespoe HRL; The Backfill and Plug Test is a test of the hydraulic and mechanical function of different backfill materials, emplacement methods and a full-scale plug; The aim of the Canister Retrieval Test was to demonstrate readiness for recovering emplaced canisters even after the time when the surrounding bentonite buffer is fully saturated; The Temperature Buffer Test aims at improving our current understanding of the thermo-hydromechanical behaviour of buffers with a temperature around and above 100 deg C during the water saturation transient. The main goal of the project Sealing of Tunnel at Great Depth is to confirm that silica sol is a useful grout at the water pressures prevailing at repository level. To achieve this, the Tass-tunnel has been constructed at the -450 m level at Aespoe HRL. The objective of the project In situ Corrosion Testing of Miniature Canisters is to obtain a better understanding of the corrosion processes inside a failed canister. In Aespoe HRL in situ experiments are performed with miniature copper canisters with cast iron inserts. The Task Force on Engineered Barrier Systems addresses, in the first phase, two tasks: (1) THM processes and (2) gas migration in buffer material. However, at the end of 2006 it was decided to start a parallel Task Force that deals with geochemical processes in engineered barriers. During 2008, two Task Force meetings have been held. In Benchmark 1 (laboratory tests) the modelling of THM processes and gas breakthrough is finalised. In Benchmark 2 (large scale field tests) the main work has been within modelling of the Canister retrieval test at Aespoe HRL and the finalising of the modelling of the URL tests. Laboratory experiments and results from the experiment Long term test of buffer materials have been

  16. A Review of Beneficial Effects of Reducing Environment at the Near-Field of KBS-3 Repository%综述KBS-3处置库近场还原性环境对处置安全的裨益(英文)

    Institute of Scientific and Technical Information of China (English)

    崔大庆

    2011-01-01

    核能为减小温室气体排放引起的气候变化的风险起到了重要的不可替代的作用,但是自福岛核电站事故以来公众对核能安全越来越关注。为了消除人们的忧虑,确保核能的持续发展,国际社会将加强旨在提高整个核燃料循环各个领域安全的研究。核废物,特别是高水平放射性废物的处置安全,同核电反应堆运行安全一样是核能安全的一个重要环节。乏燃料占世界高放废物中的大部分,其中所含放射性核素需要与世隔绝十万年才能衰变到无害水平。因为很难预知此时段环境的变化,评价乏燃料中放射性核素从深地层处置库通过地质圈和生物圈向人类环境的迁移将具有很大的不%The recent research activities,i.e.relevant publications and the author's experiments on chemical behaviors of spent nuclear fuel(SNF) and canister materials at near-field of KBS-3 deep geological repository were reviewed.The advantages of reductive substances at KBS-3 repository to the spent fuel disposal safety were discussed.Using data from literatures and experiments,the author demonstrated the blocking effect of hydrogen generated for iron canister corrosion on SNF dissolution,and discussed the reaction mechanism.It is also proved that the γ radiation expected at the early stage of disposal and micro mole level oxidative species in water solution can only slightly enhance the corrosion rate of copper canister to μm/y level,still 103 times slower than that at air saturated conditions.During a long period of time after copper canister leaks,under combined effects of iron canister material,hydrogen and fission product alloy particle catalysts,SNF dissolution can be depressed or blocked,and most radiotoxic multivalent radionuclides U,Np,Tc and Se released from SNF can be reduced and precipitated.This paper supplies scientific bases for the sitting of a SNF repository at a stable reducing area,and designing of canisters

  17. Multi-Pack Disposal Concepts for Spent Fuel (Revision 1)

    Energy Technology Data Exchange (ETDEWEB)

    Hardin, Ernest [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Matteo, Edward N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hadgu, Teklu [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-01-01

    At the initiation of the Used Fuel Disposition (UFD) R&D campaign, international geologic disposal programs and past work in the U.S. were surveyed to identify viable disposal concepts for crystalline, clay/shale, and salt host media. Concepts for disposal of commercial spent nuclear fuel (SNF) and high-level waste (HLW) from reprocessing are relatively advanced in countries such as Finland, France, and Sweden. The UFD work quickly showed that these international concepts are all “enclosed,” whereby waste packages are emplaced in direct or close contact with natural or engineered materials . Alternative “open” modes (emplacement tunnels are kept open after emplacement for extended ventilation) have been limited to the Yucca Mountain License Application Design. Thermal analysis showed that if “enclosed” concepts are constrained by peak package/buffer temperature, that waste package capacity is limited to 4 PWR assemblies (or 9 BWR) in all media except salt. This information motivated separate studies: 1) extend the peak temperature tolerance of backfill materials, which is ongoing; and 2) develop small canisters (up to 4-PWR size) that can be grouped in larger multi-pack units for convenience of storage, transportation, and possibly disposal (should the disposal concept permit larger packages). A recent result from the second line of investigation is the Task Order 18 report: Generic Design for Small Standardized Transportation, Aging and Disposal Canister Systems. This report identifies disposal concepts for the small canisters (4-PWR size) drawing heavily on previous work, and for the multi-pack (16-PWR or 36-BWR).

  18. Multi-pack Disposal Concepts for Spent Fuel (Rev. 0)

    Energy Technology Data Exchange (ETDEWEB)

    Hadgu, Teklu [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hardin, Ernest [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Matteo, Edward N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-12-01

    At the initiation of the Used Fuel Disposition (UFD) R&D campaign, international geologic disposal programs and past work in the U.S. were surveyed to identify viable disposal concepts for crystalline, clay/shale, and salt host media (Hardin et al., 2012). Concepts for disposal of commercial spent nuclear fuel (SNF) and high-level waste (HLW) from reprocessing are relatively advanced in countries such as Finland, France, and Sweden. The UFD work quickly showed that these international concepts are all “enclosed,” whereby waste packages are emplaced in direct or close contact with natural or engineered materials . Alternative “open” modes (emplacement tunnels are kept open after emplacement for extended ventilation) have been limited to the Yucca Mountain License Application Design (CRWMS M&O, 1999). Thermal analysis showed that, if “enclosed” concepts are constrained by peak package/buffer temperature, waste package capacity is limited to 4 PWR assemblies (or 9-BWR) in all media except salt. This information motivated separate studies: 1) extend the peak temperature tolerance of backfill materials, which is ongoing; and 2) develop small canisters (up to 4-PWR size) that can be grouped in larger multi-pack units for convenience of storage, transportation, and possibly disposal (should the disposal concept permit larger packages). A recent result from the second line of investigation is the Task Order 18 report: Generic Design for Small Standardized Transportation, Aging and Disposal Canister Systems (EnergySolution, 2015). This report identifies disposal concepts for the small canisters (4-PWR size) drawing heavily on previous work, and for the multi-pack (16-PWR or 36-BWR).

  19. Optimization of Deep Borehole Systems for HLW Disposal

    Energy Technology Data Exchange (ETDEWEB)

    Driscoll, Michael [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Baglietto, Emilio [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Buongiorno, Jacopo [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Lester, Richard [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Brady, Patrick [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Arnold, B. W. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-09-09

    This is the final report on a project to update and improve the conceptual design of deep boreholes for high level nuclear waste disposal. The effort was concentrated on application to intact US legacy LWR fuel assemblies, but conducted in a way in which straightforward extension to other waste forms, host rock types and countries was preserved. The reference fuel design version consists of a vertical borehole drilled into granitic bedrock, with the uppermost kilometer serving as a caprock zone containing a diverse and redundant series of plugs. There follows a one to two kilometer waste canister emplacement zone having a hole diameter of approximately 40-50 cm. Individual holes are spaced 200-300 m apart to form a repository field. The choice of verticality and the use of a graphite based mud as filler between the waste canisters and the borehole wall liner was strongly influenced by the expectation that retrievability would continue to be emphasized in US and worldwide repository regulatory criteria. An advanced version was scoped out using zinc alloy cast in place to fill void space inside a disposal canister and its encapsulated fuel assembly. This excludes water and greatly improves both crush resistance and thermal conductivity. However the simpler option of using a sand fill was found adequate and is recommended for near-term use. Thermal-hydraulic modeling of the low permeability and porosity host rock and its small (≤ 1%) saline water content showed that vertical convection induced by the waste’s decay heat should not transport nuclides from the emplacement zone up to the biosphere atop the caprock. First order economic analysis indicated that borehole repositories should be cost-competitive with shallower mined repositories. It is concluded that proceeding with plans to drill a demonstration borehole to confirm expectations, and to carry out priority experiments, such as retention and replenishment of in-hole water is in order.

  20. Nondestructive Examination Guidance for Dry Storage Casks

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, Ryan M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Suffield, Sarah R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hirt, Evelyn H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Suter, Jonathan D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lareau, John P. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Zhuge, Jing Wei [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Qiao, Hong [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Moran, Traci L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ramuhalli, Pradeep [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-09-30

    In this report, an assessment of NDE methods is performed for components of NUHOMS 80 and 102 dry storage system components in an effort to assist NRC staff with review of license renewal applications. The report considers concrete components associated with the horizontal storage modules (HSMs) as well as metal components in the HSMs. In addition, the report considers the dry shielded canister (DSC). Scope is limited to NDE methods that are considered most likely to be proposed by licensees. The document, ACI 349.3R, Evaluation of Existing Nuclear Safety-Related Concrete Structures, is used as the basis for the majority of the NDE methods summarized for inspecting HSM concrete components. Two other documents, ACI 228.2R, Nondestructive Test Methods for Evaluation of Concrete in Structures, and ORNL/TM-2007/191, Inspection of Nuclear Power Plant Structure-Overview of Methods and Related Application, supplement the list with additional technologies that are considered applicable. For the canister, the ASME B&PV Code is used as the basis for NDE methods considered, along with currently funded efforts through industry (Electric Power Research Institute [EPRI]) and the U.S. Department of Energy (DOE) to develop inspection technologies for canisters. The report provides a description of HSM and DSC components with a focus on those aspects of design considered relevant to inspection. This is followed by a brief description of other concrete structural components such as bridge decks, dams, and reactor containment structures in an effort to facilitate comparison between these structures and HSM concrete components and infer which NDE methods may work best for certain HSM concrete components based on experience with these other structures. Brief overviews of the NDE methods are provided with a focus on issues and influencing factors that may impact implementation or performance. An analysis is performed to determine which NDE methods are most applicable to specific

  1. Hydrogeological Characteristics of Fractured Rocks around the In-DEBS Test Borehole at the Underground Research Facility (KURT)

    Science.gov (United States)

    Ko, Nak-Youl; Kim, Geon Young; Kim, Kyung-Su

    2016-04-01

    In the concept of the deep geological disposal of radioactive wastes, canisters including high-level wastes are surrounded by engineered barrier, mainly composed of bentonite, and emplaced in disposal holes drilled in deep intact rocks. The heat from the high-level radioactive wastes and groundwater inflow can influence on the robustness of the canister and engineered barrier, and will be possible to fail the canister. Therefore, thermal-hydrological-mechanical (T-H-M) modeling for the condition of the disposal holes is necessary to secure the safety of the deep geological disposal. In order to understand the T-H-M coupling phenomena at the subsurface field condition, "In-DEBS (In-Situ Demonstration of Engineered Barrier System)" has been designed and implemented in the underground research facility, KURT (KAERI Underground Research Tunnel) in Korea. For selecting a suitable position of In-DEBS test and obtaining hydrological data to be used in T-H-M modeling as well as groundwater flow simulation around the test site, the fractured rock aquifer including the research modules of KURT was investigated through the in-situ tests at six boreholes. From the measured data and results of hydraulic tests, the range of hydraulic conductivity of each interval in the boreholes is about 10-7-10-8 m/s and that of influx is about 10-4-10-1 L/min for NX boreholes, which is expected to be equal to about 0.1-40 L/min for the In-DEBS test borehole (diameter of 860 mm). The test position was determined by the data and availability of some equipment for installing In-DEBS in the test borehole. The mapping for the wall of test borehole and the measurements of groundwater influx at the leaking locations was carried out. These hydrological data in the test site will be used as input of the T-H-M modeling for simulating In-DEBS test.

  2. Initiating the Validation of CCIM Processability for Multi-phase all Ceramic (SYNROC) HLW Form: Plan for Test BFY14CCIM-C

    Energy Technology Data Exchange (ETDEWEB)

    Maio, Vince [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-08-01

    This plan covers test BFY14CCIM-C which will be a first–of–its-kind demonstration for the complete non-radioactive surrogate production of multi-phase ceramic (SYNROC) High Level Waste Forms (HLW) using Cold Crucible Induction Melting (CCIM) Technology. The test will occur in the Idaho National Laboratory’s (INL) CCIM Pilot Plant and is tentatively scheduled for the week of September 15, 2014. The purpose of the test is to begin collecting qualitative data for validating the ceramic HLW form processability advantages using CCIM technology- as opposed to existing ceramic–lined Joule Heated Melters (JHM) currently producing BSG HLW forms. The major objectives of BFY14CCIM-C are to complete crystalline melt initiation with a new joule-heated resistive starter ring, sustain inductive melting at temperatures between 1600 to 1700°C for two different relatively high conductive materials representative of the SYNROC ceramic formation inclusive of a HLW surrogate, complete melter tapping and pouring of molten ceramic material in to a preheated 4 inch graphite canister and a similar canister at room temperature. Other goals include assessing the performance of a new crucible specially designed to accommodate the tapping and pouring of pure crystalline forms in contrast to less recalcitrant amorphous glass, assessing the overall operational effectiveness of melt initiation using a resistive starter ring with a dedicated power source, and observing the tapped molten flow and subsequent relatively quick crystallization behavior in pans with areas identical to standard HLW disposal canisters. Surrogate waste compositions with ceramic SYNROC forming additives and their measured properties for inductive melting, testing parameters, pre-test conditions and modifications, data collection requirements, and sampling/post-demonstration analysis requirements for the produced forms are provided and defined.

  3. Spent Fuel Test-Climax: technical measurements data management system description and data presentation

    Energy Technology Data Exchange (ETDEWEB)

    Carlson, R.C.

    1985-08-01

    The Spent Fuel Test-Climax (SFT-C) was located 420 m below surface in the Climax Stock granite on the Nevada Test Site. The test was conducted under the technical direction of the Lawrence Livermore National Laboratory (LLNL) as part of the Nevada Nuclear Waste Storage Investigations (NNWSI) for the US Department of Energy. Eleven canisters of spent nuclear reactor fuel were emplaced, along with six electrical simulators, in April-May 1980. The spent fuel canisters were retrieved and the electrical simulators de-energized in March-April 1983. During the test, just over 1000 MW-hr of thermal energy was deposited in the site, causing temperature changes 100{sup 0}C near the canisters, and about 5{sup 0} in the tunnels. More than 900 channels of geotechnical, seismological, and test status data were recorded on nearly continuous basis for about 3-1/2 years, ending in September 1983. Most geotechnical instrumentation was known to be temperature sensitive, and thus would require temperature compensation before interpretation. Accordingly, a 10-in. reel of digital tape was off-loaded and shipped to Livermore every 4 to 8 weeks, where the data were verified, organized into 45 one-million-word files, and temperature corrected. The purpose of this report is to document the receipt and processing of the data by LLNL Livermore personnel, present facts about the history of the instruments which may be important to the interpretation of the data, present the data themselves in graphical form for each instrument over its operating lifetime, document the forms and locations in which the data will be archived, and offer the data to the geotechnical community for future use in understanding and predicting the effects of the storage of heat-generating waste in hard rocks such as granite.

  4. Human factors analysis and design methods for nuclear waste retrieval systems. Human factors design methodology and integration plan

    Energy Technology Data Exchange (ETDEWEB)

    Casey, S.M.

    1980-06-01

    The purpose of this document is to provide an overview of the recommended activities and methods to be employed by a team of human factors engineers during the development of a nuclear waste retrieval system. This system, as it is presently conceptualized, is intended to be used for the removal of storage canisters (each canister containing a spent fuel rod assembly) located in an underground salt bed depository. This document, and the others in this series, have been developed for the purpose of implementing human factors engineering principles during the design and construction of the retrieval system facilities and equipment. The methodology presented has been structured around a basic systems development effort involving preliminary development, equipment development, personnel subsystem development, and operational test and evaluation. Within each of these phases, the recommended activities of the human engineering team have been stated, along with descriptions of the human factors engineering design techniques applicable to the specific design issues. Explicit examples of how the techniques might be used in the analysis of human tasks and equipment required in the removal of spent fuel canisters have been provided. Only those techniques having possible relevance to the design of the waste retrieval system have been reviewed. This document is intended to provide the framework for integrating human engineering with the rest of the system development effort. The activities and methodologies reviewed in this document have been discussed in the general order in which they will occur, although the time frame (the total duration of the development program in years and months) in which they should be performed has not been discussed.

  5. Macstor dry spent fuel storage system

    Energy Technology Data Exchange (ETDEWEB)

    Pare, F. E. [Atomic Energy of Canada Limited, Montreal (Canada)

    1996-04-15

    AECL, a Canadian Grown Corporation established since 1952, is unique among the world's nuclear organizations. It is both supplier of research reactors and heavy water moderated CANDU power reactors as well as operator of extensive nuclear research facilities. As part of its mandate, AECL has developed products and conceptual designs for the short, intermediate and long term storage and disposal of spent nuclear fuel. AECL has also assumed leadership in the area of dry storage of spent fuel. This Canadian Crown Corporation first started to look into dry storage for the management of its spent nuclear fuel in the early 1970's. After developing silo-like structures called concrete canisters for the storage of its research reactor enriched uranium fuel, AECL went on to perfect that technology for spent CANDU natural uranium fuel. In 1989 AECL teamed up with Trans nuclear, Inc.,(TN), a US based member of the international Trans nuclear Group, to extend its dry storage technology to LWR spent fuel. This association combines AECL's expertise and many years experience in the design of spent fuel storage facilities with TN's proven capabilities of processing, transportation, storage and handling of LWR spent fuel. From the early AECL-designed unventilated concrete canisters to the advanced MACSTOR concept - Modular Air-Cooled Canister Storage - now available also for LWR fuel - dry storage is proving to be safe, economical, practical and, most of all, well accepted by the general public. AECL's experience with different fuels and circumstances has been conclusive.

  6. Evaluation of an international, perpetual, and retrievable facility for storage of vitrified radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Libby, L.M.; Whipple, C.G.; Wurtele, M.G.

    1982-10-01

    It is technically feasible to site a retrievable but permanent surface storage facility for vitrified radioactive wastes in the northwestern Egyptian desert. Present-day commercial vitrification plants are in England and France and produce glass cylinders in the shape of an annulus, about 9 ft high, clad in a stainless steel can, containing about 25% of fission product and actinide oxides, weighing about 10 tonnes, having a volume of about 70 ft/sup 3/, releasing about 1.8 X 10/sup 5/ Btu heat/h. The high-level waste (HLW) glass cylinders, in lead shipping casks, are to be shipped to European ports by truck, sent to Mersa Matruh on the Egyptian coast, about ten at a time in small barges, then offloaded and sent by train a short distance inland to the site. The storage facility envisaged at the site is a concrete-walled round house with a radial crane, equipped with recanning facilities in case of breakage of stainless steel canisters, with a shop for repair of the train as needed, and with a turntable for the engine. Cooling is provided by natural air draft resulting from the canister surface temperature of about 100/sup 0/C. If needed, backup cooling is provided by equipment for forced-air drafts and by tanks of water. The canister arrangement is that produced by coaxial vertical stacking; horizontal coaxial arrangements are yet to be analyzed. The site chosen is exposed hard rock close to the Mediterranean in the northwest corner of the Egyptian desert. Groundwater is found at about 100 m. The rainfall is about 4 in./yr so that flash floods sometimes occur and surface drains are needed. Meteorology, seismicity, agriculture, and wildlife are all favorable factors, and plane studies show no thermal or radioactive threat.

  7. Visualization experiment to investigate capillary barrier performance in the context of a Yucca Mountain emplacement drift.

    Science.gov (United States)

    Tidwell, Vincent C; Glass, Robert J; Chocas, Connie; Barker, Glenn; Orear, Lee

    2003-01-01

    The use of capillary barriers as engineered backfill systems to divert water away from radioactive waste potentially stored in a Yucca Mountain emplacement drift is investigated. We designed and conducted a flow visualization experiment to investigate capillary barrier performance in this context. A two-dimensional, thin slab, test system replicated the physical emplacement drift to one-quarter scale (1.4-m diameter) and included the simulated drift wall, waste canister, pedestal, capillary barrier backfill, and host-rock fracture system. Water was supplied at the top of the simulated drift and allowed to discharge by way of wicks located along the left wall of the cell (simulated fractures) or by a gravity drain at the bottom of the right side (simulated impermeable rock with floor drain). Photographs captured the migration of water and a blue dye tracer within the system, analytical balances measured the mass balance of water, while tensiometers measured the capillary pressure at numerous locations. Of particular concern to this test was the drainage of the capillary barrier, which terminates against the drift wall. We found that while the simulated fractures (left side) and drain (right side) each influenced the performance of the capillary barrier at early time, they had little differential affect at later times. Also of concern was the small disparity in capillary properties between the fine and coarse layer (limited by the need of a fine-grained material that would not filter into the coarse layer under dry conditions). While the capillary barrier was able to divert the majority of flow toward the edges of the system and away from the simulated waste canister, the barrier did not preclude flow in the coarse layer, which was noted to be visually wet next to the waste canister on day 92 and was continuing to take on water at termination on day 112.

  8. Processes and parameters involved in modeling radionuclide transport from bedded salt repositories. Final report. Technical memorandum

    Energy Technology Data Exchange (ETDEWEB)

    Evenson, D.E.; Prickett, T.A.; Showalter, P.A.

    1979-07-01

    The parameters necessary to model radionuclide transport in salt beds are identified and described. A proposed plan for disposal of the radioactive wastes generated by nuclear power plants is to store waste canisters in repository sites contained in stable salt formations approximately 600 meters below the ground surface. Among the principal radioactive wastes contained in these canisters will be radioactive isotopes of neptunium, americium, uranium, and plutonium along with many highly radioactive fission products. A concern with this form of waste disposal is the possibility of ground-water flow occurring in the salt beds and endangering water supplies and the public health. Specifically, the research investigated the processes involved in the movement of radioactive wastes from the repository site by groundwater flow. Since the radioactive waste canisters also generate heat, temperature is an important factor. Among the processes affecting movement of radioactive wastes from a repository site in a salt bed are thermal conduction, groundwater movement, ion exchange, radioactive decay, dissolution and precipitation of salt, dispersion and diffusion, adsorption, and thermomigration. In addition, structural changes in the salt beds as a result of temperature changes are important. Based upon the half-lives of the radioactive wastes, he period of concern is on the order of a million years. As a result, major geologic phenomena that could affect both the salt bed and groundwater flow in the salt beds was considered. These phenomena include items such as volcanism, faulting, erosion, glaciation, and the impact of meteorites. CDM reviewed all of the critical processes involved in regional groundwater movement of radioactive wastes and identified and described the parameters that must be included to mathematically model their behavior. In addition, CDM briefly reviewed available echniques to measure these parameters.

  9. Reference design and operations for deep borehole disposal of high-level radioactive waste.

    Energy Technology Data Exchange (ETDEWEB)

    Herrick, Courtney Grant; Brady, Patrick Vane; Pye, Steven; Arnold, Bill Walter; Finger, John Travis; Bauer, Stephen J.

    2011-10-01

    A reference design and operational procedures for the disposal of high-level radioactive waste in deep boreholes have been developed and documented. The design and operations are feasible with currently available technology and meet existing safety and anticipated regulatory requirements. Objectives of the reference design include providing a baseline for more detailed technical analyses of system performance and serving as a basis for comparing design alternatives. Numerous factors suggest that deep borehole disposal of high-level radioactive waste is inherently safe. Several lines of evidence indicate that groundwater at depths of several kilometers in continental crystalline basement rocks has long residence times and low velocity. High salinity fluids have limited potential for vertical flow because of density stratification and prevent colloidal transport of radionuclides. Geochemically reducing conditions in the deep subsurface limit the solubility and enhance the retardation of key radionuclides. A non-technical advantage that the deep borehole concept may offer over a repository concept is that of facilitating incremental construction and loading at multiple perhaps regional locations. The disposal borehole would be drilled to a depth of 5,000 m using a telescoping design and would be logged and tested prior to waste emplacement. Waste canisters would be constructed of carbon steel, sealed by welds, and connected into canister strings with high-strength connections. Waste canister strings of about 200 m length would be emplaced in the lower 2,000 m of the fully cased borehole and be separated by bridge and cement plugs. Sealing of the upper part of the borehole would be done with a series of compacted bentonite seals, cement plugs, cement seals, cement plus crushed rock backfill, and bridge plugs. Elements of the reference design meet technical requirements defined in the study. Testing and operational safety assurance requirements are also defined. Overall

  10. High-Level Waste Systems Plan. Revision 7

    Energy Technology Data Exchange (ETDEWEB)

    Brooke, J.N.; Gregory, M.V.; Paul, P.; Taylor, G.; Wise, F.E.; Davis, N.R.; Wells, M.N.

    1996-10-01

    This revision of the High-Level Waste (HLW) System Plan aligns SRS HLW program planning with the DOE Savannah River (DOE-SR) Ten Year Plan (QC-96-0005, Draft 8/6), which was issued in July 1996. The objective of the Ten Year Plan is to complete cleanup at most nuclear sites within the next ten years. The two key principles of the Ten Year Plan are to accelerate the reduction of the most urgent risks to human health and the environment and to reduce mortgage costs. Accordingly, this System Plan describes the HLW program that will remove HLW from all 24 old-style tanks, and close 20 of those tanks, by 2006 with vitrification of all HLW by 2018. To achieve these goals, the DWPF canister production rate is projected to climb to 300 canisters per year starting in FY06, and remain at that rate through the end of the program in FY18, (Compare that to past System Plans, in which DWPF production peaked at 200 canisters per year, and the program did not complete until 2026.) An additional $247M (FY98 dollars) must be made available as requested over the ten year planning period, including a one-time $10M to enhance Late Wash attainment. If appropriate resources are made available, facility attainment issues are resolved and regulatory support is sufficient, then completion of the HLW program in 2018 would achieve a $3.3 billion cost savings to DOE, versus the cost of completing the program in 2026. Facility status information is current as of October 31, 1996.

  11. Mechanisms of Copper Corrosion in Aqueous Environments. A report from the Swedish National Council for Nuclear Waste's scientific workshop, on November 16, 2009

    Energy Technology Data Exchange (ETDEWEB)

    2010-07-01

    In 2010 the Swedish Nuclear Fuel and Waste Management Company, SKB, plans to submit its license application for the final repository of spent nuclear fuel. The proposed method is the so-called KBS-3 method and implies placing the spent nuclear fuel in copper canisters, surrounded by a buffer of bentonite clay, at 500 m depth in the bedrock. The site selected by SKB to host the repository is located in the municipality of Oesthammar on the Swedish east coast. The copper canister plays a key role in the design of the repository for spent nuclear fuel in Sweden. The long-term physical and chemical stability of copper in aqueous environments is fundamental for the safety evolution of the proposed disposal concept. However, the corrosion resistance of copper has been questioned by results obtained under anoxic conditions in aqueous solution. These observations caused some head-lines in the Swedish newspapers as well as public and political concerns. Consequently, the Swedish National Council for Nuclear Waste organized a scientific workshop on the issue 'Mechanisms of Copper Corrosion in Aqueous Environments'. The purpose of the workshop was to address the fundamental understanding of the corrosion characteristics of copper regarding oxygen-free environments, and to identify what additional information is needed to assess the validity of the proposed corrosion mechanism and its implication on the containment of spent nuclear fuel in a copper canister. This seminar report is based on the presentations and discussions at the workshop. It also includes written statements by the members of the expert panel

  12. Water/rock interactions and mass transport within a thermal gradient Application to the confinement of high level nuclear waste; Interactions solide/solution et transferts de matiere dans un gradient de temperature. Application au confinement des dechets nucleaires de haute-activite

    Energy Technology Data Exchange (ETDEWEB)

    Poinssot, Ch. [CEA Saclay, 91 - Gif-sur-Yvette (France). Dept. d`Entreposage et de Stockage des Dechets]|[Ecole Normale Superieure, 92 - Fontenay-aux-Roses (France). Laboratoire de Geologie

    1998-12-31

    The initial stage of a high level nuclear waste disposal will be characterised by a large heat release within the near-field environment of the canisters. This heat flux caused by radioactive decay will lead to an increase of temperature and a subsequent thermal gradient between the `hot` canisters and the `cold`geological medium. In addition, this thermal gradient will decrease with time due to the heat decay although it could last hundred years. What will be the consequences of such a thermal field varying both on space and time for the alteration of the different constituents of the near field environment. In particular, what could be the effects on the radionuclides migration in the accidental case of an early breach of a canister during the thermal stage? This study brings significant answers to these questions in the light of a performance assessment study. This work is supported by a triple methodological approach involving experimental studies, modelling calculations and a natural analogues study. This complete work demonstrates that a thermal gradient leads to a large re-distribution of elements within the system: some elements are incorporated in the solid phases of the hot end (Si, Zr, Ca) whereas some others are in those of the cold end (Fe, Al, Zn). The confrontation of the results of very simple experiments with the results of a model built on equilibrium thermodynamics allow us to evidence the probable mechanisms causing this mass transport: out-of-equilibrium thermodiffusion processes coupled to irreversible precipitation. Moreover, the effects of the variation of temperatures with time is studied by the way of a natural system which underwent a similar temperature evolution as a disposal and which was initially rich in uranium: the Jurassic Alpine bauxites. In addition, part of the initial bauxite escaped this temperature transformations due to their incorporation in outer thrusting nappes. They are used as a reference. (author)

  13. How Weight Affects the Perceived Spacing between the Thumb and Fingers during Grasping.

    Directory of Open Access Journals (Sweden)

    Annie A Butler

    Full Text Available We know much about mechanisms determining the perceived size and weight of lifted objects, but little about how these properties of size and weight affect the body representation (e.g. grasp aperture of the hand. Without vision, subjects (n = 16 estimated spacing between fingers and thumb (perceived grasp aperture while lifting canisters of the same width (6.6cm but varied weights (300, 600, 900, and 1200 g. Lifts were performed by movement of either the wrist, elbow or shoulder to examine whether lifting with different muscle groups affects the judgement of grasp aperture. Results for perceived grasp aperture were compared with changes in perceived weight of objects of different sizes (5.2, 6.6, and 10 cm but the same weight (600 g. When canisters of the same width but different weights were lifted, perceived grasp aperture decreased 4.8% [2.2 ‒ 7.4] (mean [95% CI]; P < 0.001 from the lightest to the heaviest canister, no matter how they were lifted. For objects of the same weight but different widths, perceived weight decreased 42.3% [38.2 ‒ 46.4] from narrowest to widest (P < 0.001, as expected from the size-weight illusion. Thus, despite a highly distorted perception of the weight of objects based on their size, we conclude that proprioceptive afferents maintain a reasonably stable perception of the aperture of the grasping hand over a wide range of object weights. Given the small magnitude of this 'weight-grasp aperture' illusion, we propose the brain has access to a relatively stable 'perceptual ruler' to aid the manipulation of different objects.

  14. Performance of a new carbon dioxide absorbent, Yabashi lime® as compared to conventional carbon dioxide absorbent during sevoflurane anesthesia in dogs

    OpenAIRE

    KONDOH, Kei; ATIBA, Ayman; NAGASE, Kiyoshi; Ogawa, Shizuko; Miwa, Takashi; KATSUMATA, Teruya; Ueno, Hiroshi; UZUKA, Yuji

    2015-01-01

    In the present study, we compare a new carbon dioxide (CO2) absorbent, Yabashi lime® with a conventional CO2 absorbent, Sodasorb® as a control CO2 absorbent for Compound A (CA) and Carbon monoxide (CO) productions. Four dogs were anesthetized with sevoflurane. Each dog was anesthetized with four preparations, Yabashi lime® with high or low-flow rate of oxygen and control CO2 absorbent with high or low-flow rate. CA and CO concentrations in the anesthetic circuit, canister temperature and carb...

  15. Higher Magnification Imaging of the Polished Aluminum Collector Returned from the Genesis Mission

    Science.gov (United States)

    Rodriquez, Melissa C.; Burkett, P. J.; Allton, J. H.

    2011-01-01

    The polished aluminum collector (previously referred to as the polished aluminum kidney) was intended for noble gas analysis for the Gene-sis mission. The aluminum collector, fabricated from alloy 6061T, was polished for flight with alumina, then diamond paste. Final cleaning was performed by soak-ing and rinsing with hexane, then isopropanol, and last-ly megasonically energized ultrapure water prior to installation. It was mounted inside the collector canister on the thermal shield at JSC in 2000. The polished aluminum collector was not surveyed microscopically prior to flight.

  16. Increased CPC batch size study for Tank 42 sludge in the Defense Waste Processing Facility

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, W.E.

    2000-01-06

    A series of experiments have been completed at TNX for the sludge-only REDOX adjusted flowsheet using Tank 42 sludge simulant in response to the Technical Task Request HLW/DWPT/TTR-980013 to increase CPC batch sizes. By increasing the initial SRAT batch size, a melter feed batch at greater waste solids concentration can be prepared and thus increase melter output per batch by about one canister. The increased throughput would allow DWPF to dispose of more waste in a given time period thus shortening the overall campaign.

  17. State of Washington Department of Health radioactive air emission notice of construction phase 1 for spent nuclear fuel project - hot conditioning system annex, project W-484

    Energy Technology Data Exchange (ETDEWEB)

    Turnbaugh, J.E.

    1996-08-15

    This notice of construction (NOC) provides information regarding the source and the estimated annual possession quantity resulting from the operation of the Hot Conditioning System Annex (HCSA). This information will be discussed again in the Phase II NOC, providing additional details on emissions generated by the operation of the HCSA. This Phase I NOC is defined as construct in the substructure, including but limited to, pouring the concrete for the floor; construction of the process pits and exterior walls; making necessary interface connections to the Canister Storage Building (CSB) ventilation and utility systems for personnel comfort; and extending the multi-canister over-pack (MCO) handling machine rails into the HCSA. A Phase II NOC will be submitted for approval prior to installation and is defined as the completion of the HCSA, which will consist of installation of Hot Conditioning System Equipment (HCSA), air emissions control equipment, and emission monitoring equipment. About 80 percent of the U.S. Department of Energy`s spent nuclear fuel (SNF) inventory is stored under water in the Hanford Site K Basins. Spent nuclear fuel in the K West Basin is contained in closed canisters, while the SNF in the K East Basin is contained in open canisters, which allow free release of corrosion products to the K East Basin water. Storage in the K Basins was originally intended to be on an as-needed basis to sustain operation of the N Reactor while the Plutonium-Uranium Extraction (PUREX) Plant was refurbished and restarted. The decision in December 1992 to deactivate the PUREX Plant left approximately 2,300 MT (2,530 tons) of N Reactor SNF in the K Basins with no means for near-term removal and processing. The HCSA will be constructed as an addition to the CSB and will contain the HCSA. The hot conditioning system (HCS) will remove chemically-bound water and will passivate the exposed uranium surfaces associated,with the SNF. The HCSA will house seven hot

  18. OMS engine firing

    Science.gov (United States)

    1983-01-01

    An Orbital maneuvering system (OMS) engine firing caused this bright glow at the aft end of the shuttle Challenger during STS-7. Also visible in the open payload bay are parts of the Shuttle pallet satellite (SPAS-01), the experiment package for NASA's Office of Space and Terrestrial Applications (OSTA-2), the protective cradles for the Indonesian Palapa-B and Telesat Canada Anik C2 satellites, some getaway special (GAS) canisters and the Canadian built remote manipulator system (RMS). The earth's horizon can be seen above the orbiter.

  19. Radioactive Air Emissions Notice of Construction (NOC) for the Solid Waste Treatment Facility (T Plant) Fuel Removal Project

    Energy Technology Data Exchange (ETDEWEB)

    JOHNSON, R.E.

    2000-11-16

    This NOC describes the activities to remove all spent nuclear fuel (SNF) assemblies from the spent fuel pool in the T Plant Complex 221-T canyon for interim storage in the Canister Storage Building (CSB). The unabated total effective dose equivalent (TEDE) estimated for the public hypothetical maximally exposed individual (MEI) is 5.7 E-6 millirem (mrem) per year for this fuel removal NOC. The abated TEDE conservatively is estimated to account for 2.9 E-9 mrem per year to the MEI.

  20. High strength alloys

    Energy Technology Data Exchange (ETDEWEB)

    Maziasz, Phillip James [Oak Ridge, TN; Shingledecker, John Paul [Knoxville, TN; Santella, Michael Leonard [Knoxville, TN; Schneibel, Joachim Hugo [Knoxville, TN; Sikka, Vinod Kumar [Oak Ridge, TN; Vinegar, Harold J [Bellaire, TX; John, Randy Carl [Houston, TX; Kim, Dong Sub [Sugar Land, TX

    2010-08-31

    High strength metal alloys are described herein. At least one composition of a metal alloy includes chromium, nickel, copper, manganese, silicon, niobium, tungsten and iron. System, methods, and heaters that include the high strength metal alloys are described herein. At least one heater system may include a canister at least partially made from material containing at least one of the metal alloys. At least one system for heating a subterranean formation may include a tubular that is at least partially made from a material containing at least one of the metal alloys.