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Sample records for canisters

  1. Mechanical integrity of canisters

    International Nuclear Information System (INIS)

    This document constitutes the final report from 'SKBs reference group for mechanical integrity of canisters for spent nuclear fuel'. A complete list of all reports initiated by the reference group can be found in the summary report in this document. The main task of the reference group has been to advice SKB regarding the choice (ranking of alternatives) of canister type for different types of storage. The choice should be based on requirements of impermeability for a given time period and identification of possible limiting mechanisms. The main conclusions from the work were: From mechanical point of view, low phosphorous oxygen free copper (Cu-OFP) is a preferred canisters material. It exhibits satisfactory ductility both during tensile and creep testing. The residual stresses in the canisters are of such a magnitude that the estimated time to creep rupture with the data obtained for the Cu-OFP material is essentially infinite. Based on the present knowledge of stress corrosion cracking of copper there appears to be a small risk for such to occur in the projected environment. This risk need some further study. Rock shear movements of the size of 10 cm should pose no direct threat to the integrity of the canisters. Considering mechanical integrity, the composite copper/steel canister is an advantageous alternative. The recommendations for further research included continued studies of the creep properties of copper and of stress corrosion cracking. However, the studies should focus more directly on the design and fabrication aspect of the canister

  2. The concrete canister program

    International Nuclear Information System (INIS)

    In the spring of 1974, WNRE began development and demonstration of a dry storage concept, called the concrete canister, as a possible alternative to storage of irradiated CANDU fuel in water pools. The canister is a thick-walled concrete monolith containing baskets of fuel in the dry state. The decay heat from the fuel is dissipated to the environment by natural heat transfer. Four canisters were designed and constructed. Two canisters containing electric heaters have been subjected to heat loads of 2.5 times the design, ramp heat-load cycling, and simulated weathering tests. The other two canisters were loaded with irradiated fuel, one containing fuel bundles of uniform decay heat and the other containing bundles of non-uniform decay heat in a non-symmetrical radial and axial array. The collected data were used to verify the analytical tools for prediction of effectiveness of heat transfer and radiation shielding and to verify the design of the basket and canisters. The demonstration canisters have shown that this concept is a viable alternative to water pools for the storage of irradiated CANDU fuel. (author)

  3. 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

  4. Improved Air-Treatment Canister

    Science.gov (United States)

    Boehm, A. M.

    1982-01-01

    Proposed air-treatment canister integrates a heater-in-tube water evaporator into canister header. Improved design prevents water from condensing and contaminating chemicals that regenerate the air. Heater is evenly spiraled about the inlet header on the canister. Evaporator is brazed to the header.

  5. 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

  6. Status report, canister fabrication

    International Nuclear Information System (INIS)

    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

  7. Moisture insensitive charcoal canisters

    International Nuclear Information System (INIS)

    Continuous monitoring of 222Rn concentrations in the air in houses is the most appropriate approach for the real-time measurements, but this requires complex and expensive instruments and is not practical for large studies. Activated carbon canisters have been used extensively for determining the average concentration over a period of a few days. The ''open face'' charcoal detectors have an integration time constant of about 14 h so that they are sensitive to short-term transient changes in the radon concentration. In addition, water uptake at high relative humidities reduces the radon uptake by the charcoal. The addition of a diffusion barrier and a nylon screen results in a charcoal detector with an integration half-time ranging from 20 to 60 h and a reduced uptake of water at high humidities. Silicone rubber sheeting is relatively permeable to radon and impermeable to water vapor. It was the purpose of this study to evaluate the effect of a silicone barrier on the charcoal canister radon collective device. 3 refs

  8. Canister compatibility with Carlsbad salt

    International Nuclear Information System (INIS)

    No significant reaction was found when candidate canister alloys were heated with salt from Carlsbad, New Mexico, for up to 5000 hours in sealed capsules and for up to 10,000 hours in unsealed capsules at temperatures (80 to 2250C) that bracket the maximum temperature calculated for reference Savannah River Plant (SRP) waste containers at 20-foot spacings in salt. Additional tests were made at 6000C in sealed capsules to characterize reactions that may occur between candidate canister alloys and any component of the salt that is released when decrepitation occurs. Under these extreme conditions there was no significant attack of Type 304L stainless steel. But, there was up to 20-mils attack of the low-carbon steel

  9. Canister Transfer Facility Criticality Calculations

    Energy Technology Data Exchange (ETDEWEB)

    J.E. Monroe-Rammsy

    2000-10-13

    The objective of this calculation is to evaluate the criticality risk in the surface facility for design basis events (DBE) involving Department of Energy (DOE) Spent Nuclear Fuel (SNF) standardized canisters (Civilian Radioactive Waste Management System [CRWMS] Management and Operating Contractor [M&O] 2000a). Since some of the canisters will be stored in the surface facility before they are loaded in the waste package (WP), this calculation supports the demonstration of concept viability related to the Surface Facility environment. The scope of this calculation is limited to the consideration of three DOE SNF fuels, specifically Enrico Fermi SNF, Training Research Isotope General Atomic (TRIGA) SNF, and Mixed Oxide (MOX) Fast Flux Test Facility (FFTF) SNF.

  10. Spent fuel canister docking station

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-01-15

    The working report for the spent fuel canister docking station presents a design for the operation and structure of the docking equipment located in the fuel handling cell for the spent fuel in the encapsulation plant. The report contains a description of the basic requirements for the docking station equipment and their implementation, the operation of the equipment, maintenance and a cost estimate. In the designing of the equipment all the problems related with the operation have been solved at the level of principle, nevertheless, detailed designing and the selection of final components have not yet been carried out. In case of defects and failures, solutions have been considered for postulated problems, and furthermore, the entire equipment was gone through by the means of systematic risk analysis (PFMEA). During the docking station designing we came across with needs to influence the structure of the actual disposal canister for spent nuclear fuel, too. Proposed changes for the structure of the steel lid fastening screw were included in the report. The report also contains a description of installation with the fuel handling cell structures. The purpose of the docking station for the fuel handling cell is to position and to seal the disposal canister for spent nuclear fuel into a penetration located on the cell floor and to provide suitable means for executing the loading of the disposal canister and the changing of atmosphere. The designed docking station consists of a docking ring, a covering hatch, a protective cone and an atmosphere-changing cap as well as the vacuum technology pertaining to the changing of atmosphere and the inert gas system. As far as the solutions are concerned, we have arrived at rather simple structures and most of the actuators of the system are situated outside of the actual fuel handling cell. When necessary, the equipment can also be used for the dismantling of a faulty disposal canister, cut from its upper end by machining. The

  11. Inspection of disposal canisters components

    International Nuclear Information System (INIS)

    This report presents the inspection techniques of disposal canister components. Manufacturing methods and a description of the defects related to different manufacturing methods are described briefly. The defect types form a basis for the design of non-destructive testing because the defect types, which occur in the inspected components, affect to choice of inspection methods. The canister components are to nodular cast iron insert, steel lid, lid screw, metal gasket, copper tube with integrated or separate bottom, and copper lid. The inspection of copper material is challenging due to the anisotropic properties of the material and local changes in the grain size of the copper material. The cast iron insert has some acoustical material property variation (attenuation, velocity changes, scattering properties), which make the ultrasonic inspection demanding from calibration point of view. Mainly three different methods are used for inspection. Ultrasonic testing technique is used for inspection of volume, eddy current technique, for copper components only, and visual testing technique are used for inspection of the surface and near surface area

  12. 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.

  13. CANISTER HANDLING FACILITY DESCRIPTION DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    J.F. Beesley

    2005-04-21

    The purpose of this facility description document (FDD) is to establish requirements and associated bases that drive the design of the Canister Handling Facility (CHF), which will allow the design effort to proceed to license application. This FDD will be revised at strategic points as the design matures. This FDD identifies the requirements and describes the facility design, as it currently exists, with emphasis on attributes of the design provided to meet the requirements. This FDD is an engineering tool for design control; accordingly, the primary audience and users are design engineers. This FDD is part of an iterative design process. It leads the design process with regard to the flowdown of upper tier requirements onto the facility. Knowledge of these requirements is essential in performing the design process. The FDD follows the design with regard to the description of the facility. The description provided in this FDD reflects the current results of the design process.

  14. Conceptual designs of radioactive canister transporters

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-02-01

    This report covers conceptual designs of transporters for the vertical, horizontal, and inclined installation of canisters containing spent-fuel elements, high-level waste, cladding waste, and intermediate-level waste (low-level waste is not discussed). Included in the discussion are cask concepts; transporter vehicle designs; concepts for mechanisms for handling and manipulating casks, canisters, and concrete plugs; transporter and repository operating cycles; shielding calculations; operator radiation dosages; radiation-resistant materials; and criteria for future design efforts.

  15. Am/Cm canister temperature evaluation in CIM5

    International Nuclear Information System (INIS)

    To facilitate the evaluation of alternate canister designs, 2 canisters were outfitted with thermocouples at elevations of 1/2, 3 1/2, and 6 1/2 inches from the canister bottom. The canisters were fabricated from two inch diameter schedule 10 and two inch diameter schedule 40 stainless steel pipe. Each canister was filled with approximately 2 kilograms of 49 wt percent lanthanide (Ln) loaded 25SrABS glass during 5 inch Cylindrical Induction Melter (CIM5) runs for TTR Tasks 3.03 and 4.03. Melter temperature, total mass of glass poured, and the glass pour rates were almost identical in both runs. The schedule 40 canister has a slightly smaller ID compared to the schedule 10 canister and therefore filled to a level of 9.5 inches compared to 8.0 inches for the schedule 40 canister. The schedule 40 canister had an empty mass of 1906 grams compared to 919 grams for the schedule 10 canister. The schedule 10 canister was found to have a higher maximum surface temperature by about 50--100 C (depending on height) during the glass pour compared to the schedule 40 canister. The additional thermal mass of the schedule 40 canister accounts for this difference. Once filled with glass, each of the canisters cooled at about the same rate, taking about an hour to cool below a maximum surface temperature of 200 C. No significant deformation of the either of the canisters was visually observed

  16. CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    International Nuclear Information System (INIS)

    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 the current design of the CHF and may not reflect the ongoing design evolution of the facility

  17. 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.

  18. 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.)

  19. Chemical compatibility of DWPF canistered waste forms

    International Nuclear Information System (INIS)

    The Waste Acceptance Preliminary Specifications (WAPS) require that the contents of the canistered waste form are compatible with one another and the stainless steel canister. The canistered waste form is a closed system comprised of a stainless steel vessel containing waste glass, air, and condensate. This system will experience a radiation field and an elevated temperature due to radionuclide decay. This report discusses possible chemical reactions, radiation interactions, and corrosive reactions within this system both under normal storage conditions and after exposure to temperatures up to the normal glass transition temperature, which for DWPF waste glass will be between 440 and 460 degrees C. Specific conclusions regarding reactions and corrosion are provided. This document is based on the assumption that the period of interim storage prior to packaging at the federal repository may be as long as 50 years

  20. Studies of waste-canister compatibility

    International Nuclear Information System (INIS)

    Compatibility studies were conducted between 7 waste forms and 15 potential canister structural materials. The waste forms were Al-Si and Pb-Sn matrix alloys, FUETAP, glass, Synroc D, and waste particles coated with carbon or carbon plus silicon carbide. The canister materials included carbon steel (bare and with chromium or nickel coatings), copper, Monel, Cu-35% Ni, titanium (grades 2 and 12), several Inconels, aluminum alloy 5052, and two stainless steels. Tests of either 6888 or 8821 h were conducted at 100 and 3000C, which bracket the low and high limits expected during storage. Glass and FUETAP evolved sulfur, which reacted preferentially with copper, nickel, and alloys of these metals. The Pb-Sn matrix alloy stuck to all samples and the carbon-coated particles to most samples at 3000C, but the extent of chemical reaction was not determined. Testing for 0.5 h at 8000C was included because it is representative of a transportation accident and is required of casks containing nuclear materials. During these tests (1) glass and FUETAP evolved sulfur, (2) FUETAP evolved large amounts of gas, (3) Synroc stuck to titanium alloys, (4) glass was molten, and (5) both matrix alloys were molten with considerable chemical interactions with many of the canister samples. If this test condition were imposed on waste canisters, it would be design limiting in many waste storage concepts

  1. 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.

  2. 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.

  3. 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

  4. CANISTER HANDLING FACILITY - VENTILATION CONFINEMENT ZONING ANALYSIS

    International Nuclear Information System (INIS)

    The purpose of this calculation is to calculate the necessary airflow distribution used to size the HVAC equipment for the Canister Handling Facility. These results will be compared to the Heating and Cooling Load Calculation in detailed design. The calculations contained in this document were developed by DandE/Mechanical HVAC and are intended solely for the use of the DandE/Mechanical HVAC department in its work regarding the HVAC system for the Canister Handling Facility. Yucca Mountain Project personnel from the DandE/Mechanical HVAC department should be consulted before use of the calculations for purposes other than those stated herein or used by individuals other than authorized personnel in DandE/Mechanical HVAC department

  5. CANISTER HANDLING FACILITY - VENTILATION AIR CALCULATION

    International Nuclear Information System (INIS)

    The purpose of this analysis is to establish the preliminary Ventilation Confinement Zone for the Canister Handling Facility (CHF). The results of this document will be used to determine the air quantities for each VCZ that will eventually be reflected in the development of the Ventilation Flow Diagrams. The analyses contained in this document are developed by D and E/Mechanical HVAC and are intended solely for the use of the D and E/Mechanical HVAC in its work regarding Confinement Zoning Analysis for the Canister Handling Facility. Yucca Mountain Project personnel from D and E/Mechanical HVAC should be consulted before use of the analyses for purposes other than those stated herein or used by individuals other than authorized personnel in D and E/Mechanical HVAC

  6. COMSOL MULTIPHYSICS MODEL FOR DWPF CANISTER FILLING

    Energy Technology Data Exchange (ETDEWEB)

    Kesterson, M.

    2011-03-31

    The purpose of this work was to develop a model that can be used to predict temperatures of the glass in the Defense Waste Processing Facility (DWPF) canisters during filling and cooldown. Past attempts to model these processes resulted in large (>200K) differences in predicted temperatures compared to experimentally measured temperatures. This work was therefore intended to also generate a model capable of reproducing the experimentally measured trends of the glass/canister temperature during filling and subsequent cooldown of DWPF canisters. To accomplish this, a simplified model was created using the finite element modeling software COMSOL Multiphysics which accepts user defined constants or expressions to describe material properties. The model results were compared to existing experimental data for validation. A COMSOL Multiphysics model was developed to predict temperatures of the glass within DWPF canisters during filling and cooldown. The model simulations and experimental data were in good agreement. The largest temperature deviations were {approx}40 C for the 87inch thermocouple location at 3000 minutes and during the initial cooldown at the 51 inch location occurring at approximately 600 minutes. Additionally, the model described in this report predicts the general trends in temperatures during filling and cooling observed experimentally. However, the model was developed using parameters designed to fit a single set of experimental data. Therefore, Q-loss is not currently a function of pour rate and pour temperature. Future work utilizing the existing model should include modifying the Q-loss term to be variable based on flow rate and pour temperature. Further enhancements could include eliminating the Q-loss term for a user defined convection where Navier-Stokes does not need to be solved in order to have convection heat transfer.

  7. Canister storage building hazard analysis report

    International Nuclear Information System (INIS)

    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

  8. 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.

  9. 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

  10. 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.

  11. 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.

  12. CANISTER HANDLING FACILITY WORKER DOSE ASSESSMENT

    International Nuclear Information System (INIS)

    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

  13. Canister storage building trade study. Final report

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-05-01

    This study was performed to evaluate the impact of several technical issues related to the usage of the Canister Storage Building (CSB) to safely stage and store N-Reactor spent fuel currently located at K-Basin 100KW and 100KE. Each technical issue formed the basis for an individual trade study used to develop the ROM cost and schedule estimates. The study used concept 2D from the Fluor prepared ``Staging and Storage Facility (SSF) Feasibility Report`` as the basis for development of the individual trade studies.

  14. Canister storage building trade study. Final report

    International Nuclear Information System (INIS)

    This study was performed to evaluate the impact of several technical issues related to the usage of the Canister Storage Building (CSB) to safely stage and store N-Reactor spent fuel currently located at K-Basin 100KW and 100KE. Each technical issue formed the basis for an individual trade study used to develop the ROM cost and schedule estimates. The study used concept 2D from the Fluor prepared ''Staging and Storage Facility (SSF) Feasibility Report'' as the basis for development of the individual trade studies

  15. 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

  16. Impact testing of simulated high-level waste glass canisters

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, M.E.; Alzheimer, J.M.; Slate, S.C.

    1985-01-01

    Three Savannah River Laboratory reference high-level waste canisters were subjected to impact tests at the Pacific Northwest Laboratory in Richland, Washington, in June 1983. The purpose of the test was to determine the integrity of the canister, nozzle, and final closure weld and to assess the effects of impacts on the glass. Two of the canisters were fabricated from 304L stainless steel and the third canister from titanium. The titanium canister was subjected to two drops. The first drop was vertical from 9.14 m onto an unyielding surface with the bottom corner of the canister receiving the impact. No failure occurred during this drop. The second drop was vertical from 9.14 m onto an unyielding surface with the corner of the fill nozzle receiving the impact. A large breach in the canister occurred in the region where the fill nozzle joins the dished head. The first stainless steel canister was dropped with the corner of the fill nozzle receiving the impact. The canister showed significant strain with no rupturing in the region where the fill nozzle joins the dished head. The second canister was dropped with the bottom corner receiving the impact and also, dropped horizontally onto an unyielding vertical solid steel cylinder in a puncture test. The bottom drop did not damage the weld and the puncture test did not rupture the canister body. The glass particles in the damaged zone of these canisters were sampled and analyzed for particle size. A comparison was made with control canister in which no impact had occurred. The particle size distribution for the control canisters and the zones of damaged glass were determined down to 1.5 ..mu..m. The quantity of glass fines, smaller than 10 ..mu..m, which must be determined for transportation safety studies, was found to be the largest in the bottom-damaged zone. The total amount of fines smaller than 10 ..mu..m after impact was less than 0.01 wt % of the total amount of glass in the canister.

  17. Design analysis report for the canister

    International Nuclear Information System (INIS)

    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

  18. 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.

  19. Design basis for the copper/steel canister

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-02-01

    The development of the copper/iron canister which has been proposed by SKB for the containment of high level nuclear waste has been studied from the point of view of choice of materials, manufacturing technology and quality assurance. This report describes the observations on progress which have been made between March 1995 and Feb 1996 and the result of further literature studies. A first trial canister has been produced using a fabricated steel liner and an extruded copper tubular, a second one using a fabricated tubular is at an advanced stage. A change from a fabricated steel inner canister to a proposed cast canister has been justified by a criticality argument but the technology for producing a cast canister is at present untried. The microstructure achieved in the extruded copper tubular for the first canister is unacceptable. Similar problems exist with plate used for the fabricated tubular, but some more favourable structures have been achieved already by this route. Seam welding of the first tubular failed through a suspected material problem. The second fabricated tubular welded without difficulty. Welding of lids and bottoms to the copper canister is problematical.There is as yet no satisfactory non destructive test procedures for the parent metal or the welds in the copper canister material, partly due to the coarse grain size which arise in the proposed material processed by the proposed routes. Further studies are also required on crevice corrosion, galvanic attack and stress corrosion cracking in the copper 50 ppm phosphorus alloy. 28 refs.

  20. 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.

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

    International Nuclear Information System (INIS)

    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

  2. Design basis for the copper/steel canister

    International Nuclear Information System (INIS)

    The development of the copper/iron canister which has been proposed by SKB for the containment of high level nuclear waste has been studied from the point of view of choice of materials, manufacturing technology and quality assurance. This report describes the observations on progress which have been made between March 1995 and Feb 1996 and the result of further literature studies. A first trial canister has been produced using a fabricated steel liner and an extruded copper tubular, a second one using a fabricated tubular is at an advanced stage. A change from a fabricated steel inner canister to a proposed cast canister has been justified by a criticality argument but the technology for producing a cast canister is at present untried. The microstructure achieved in the extruded copper tubular for the first canister is unacceptable. Similar problems exist with plate used for the fabricated tubular, but some more favourable structures have been achieved already by this route. Seam welding of the first tubular failed through a suspected material problem. The second fabricated tubular welded without difficulty. Welding of lids and bottoms to the copper canister is problematical.There is as yet no satisfactory non destructive test procedures for the parent metal or the welds in the copper canister material, partly due to the coarse grain size which arise in the proposed material processed by the proposed routes. Further studies are also required on crevice corrosion, galvanic attack and stress corrosion cracking in the copper 50 ppm phosphorus alloy. 28 refs

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

    International Nuclear Information System (INIS)

    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

  4. 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.

  5. Canister storage building design basis accident analysis documentation

    International Nuclear Information System (INIS)

    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

  6. Description of DWPF reference waste form and canister

    International Nuclear Information System (INIS)

    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 requireiajps. The rabarajca saspa bkri is bkrksilicapa class cojtaining 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% SiO2 and 15% B2O3. This composition results in a low average leachability in the waste form of approximately 5 x 10-9 g/cm2-day based on 137Cs over 365 days in 250C 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 104 rem/hour at 1 cm

  7. 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.

  8. 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

  9. 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

  10. Canister Storage Building (CSB) Hazard Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

    POWERS, T.B.

    2000-03-16

    This report describes the methodology used in conducting the Canister Storage Building (CSB) Hazard Analysis to support the final CSB Safety Analysis Report and documents the results. 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 process identified hazardous conditions and material-at-risk, determined causes for potential accidents, identified preventive and mitigative features, and qualitatively estimated the frequencies and consequences of specific occurrences. The hazard analysis was performed by a team of cognizant CSB operations and design personnel, safety analysts familiar with the CSB, and technical experts in specialty areas. The material included in this report documents the final state of a nearly two-year long process. Attachment A provides two lists of hazard analysis team members and describes the background and experience of each. The first list is a complete list of the hazard analysis team members that have been involved over the two-year long process. The second list is a subset of the first list and consists of those hazard analysis team members that reviewed and agreed to the final hazard analysis documentation. The material included in this report documents the final state of a nearly two-year long process involving formal facilitated group sessions and independent hazard and accident analysis work. The hazard analysis process led to the selection of candidate accidents for further quantitative analysis. New information relative to the hazards, discovered during the accident analysis, was incorporated into the hazard analysis data in order to compile a complete profile of facility hazards. Through this process, the results of the hazard and accident analyses led directly to the identification of safety structures, systems, and components, technical safety requirements, and other

  11. 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

  12. Reference commercial high-level waste glass and canister definition

    International Nuclear Information System (INIS)

    This report presents technical data and performance characteristics of a high-level waste glass and canister intended for use in the design of a complete waste encapsulation package suitable for disposal in a geologic repository. The borosilicate glass contained in the stainless steel canister represents the probable type of high-level waste product that will be produced in a commercial nuclear-fuel reprocessing plant. Development history is summarized for high-level liquid waste compositions, waste glass composition and characteristics, and canister design. The decay histories of the fission products and actinides (plus daughters) calculated by the ORIGEN-II code are presented

  13. 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.

  14. Retrievability of spent nuclear fuel canisters; Kaeytetyn ydinpolttoaineen loppusijoituskapseleiden palautettavuus

    Energy Technology Data Exchange (ETDEWEB)

    Saanio, T. [Saanio and Riekkola Oy, Helsinki (Finland); Raiko, H. [VTT Energy, Espoo (Finland)

    1999-03-01

    As a part of the designing process of the Finnish spent nuclear fuel repository, a preliminary study has been carried out to investigate how the canisters could technically be retrieved to the ground surface. Possibility of retrieving a canister has been investigated in different phases of the disposal project. Retrievability has not been a design goal for the spent fuel repository. However, design of the repository includes some features that may ease the retrieval of canisters in the future. Spent fuel elements are packaged in massive copper-iron canisters, which are mechanically strong and long-lived. The repository consists of excavated tunnels in hard rock which are supposed to be very long-lived making the removal of the tunnel backfilling technically possible also in the future. As long as the bentonite buffer has not been installed the canister can be returned to the ground surface using the same equipment as was used when the canister was brought down to the repository and lowered into the hole. In the encapsulation station the spent fuel elements can be packaged in the other canister or in the transport cask. After a deposition tunnel has been backfilled and closed, the retrieval consists of tearing down the concrete structure at the entry of the deposition tunnel, removal of the tunnel backfilling, removal of the bentonite from the disposal hole and lifting up of the canister. Various methods, e.g., flushing the bentonite with saline solutions, can be used to detach the canister from a hole with fully saturated bentonite. Recovery will be technically possible also after closing of the disposal facility. Backfilling of the shafts and tunnels will be removed and additional new structures and systems will have to be built in the repository. After that canisters can be transported to the ground surface as described above. In addition, handling of the canisters at the ground surface will require additional facilities. Canisters can be packaged in the

  15. Evaluation of remote smearing of DWPF canistered waste forms

    International Nuclear Information System (INIS)

    The Savannah River Site (SRS) is evaluating the variables of the remote smearing process for monitoring transferable contamination on the waste glass canisters at the Defense Waste Processing Facility (DWPF). Smearing for transferable contamination is typically done by hand, but in this case, due to the nature of the high level waste within the canisters, remote smearing is required. The effectiveness of the smear pad was determined under varying conditions (distance traveled, force applied, and canister surface), as well as the relative importance of these factors. It was concluded that the remote smear is more reliable than the hand smear

  16. 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

  17. Technical note 4. Corrosion of copper canister

    International Nuclear Information System (INIS)

    Objectives of the project: In this review assignment, SKB's treatment of copper corrosion processes or mechanisms in SR-Site shall be reviewed both for the anticipated oxic and anoxic repository environments. The reviewer(s) shall consider if corrosion and corrosion mechanisms of the copper canisters in different possible evolutionary repository environments have been properly described. The objectives of this initial review phase in the area of copper corrosion is to achieve a broad coverage of SR-Site and its supporting references and in particular identify the need for complementary information and clarifications to be delivered by SKB. Summary by the authors: It is expected that the inflow of ground water to the deposition holes and tunnels in the Forsmark repository will be very slow. Thus, it might take some few hundred years up to thousand years before the deposition holes are filled with ground water and it might take 6000 years or more before the bentonite buffer is fully water saturated and pressurized. The copper canisters will therefore meet to two completely different environments: 1. An initial period of several hundreds of years when copper is exposed to gaseous corrosion. 2. And then to aqueous corrosion. From a corrosion point of view the first 1000 years are the most critical for the copper canister since pure, or phosphorus alloyed copper, is not designed to cope with corrosion at elevated temperatures. The outer copper surface temperature is expected to reach 100 deg C within some decades after closure of the repository and then slowly cool down to around 50 deg C after 1000 years. The gaseous corrosion is treated in SKB's safety assessment as being only dependent on oxygen gas and thus easily estimated by an oxygen mass-balance calculation. This simple model has no scientific support since several corrosive trace gases, such as sulphurous and nitrous compounds, operates together with water molecules (moisture) and the corrosion product consists

  18. Thermal dimensioning of the deep repository. Influence of canister spacing, canister power, rock thermal properties and nearfield design on the maximum canister surface temperature

    Energy Technology Data Exchange (ETDEWEB)

    Hoekmark, Harald; Faelth, Billy [Clay Technology AB, Lund (Sweden)

    2003-12-01

    The report addresses the problem of the minimum spacing required between neighbouring canisters in the deep repository. That spacing is calculated for a number of assumptions regarding the conditions that govern the temperature in the nearfield and at the surfaces of the canisters. The spacing criterion is that the temperature at the canister surfaces must not exceed 100 deg C .The results are given in the form of nomographic charts, such that it is in principle possible to determine the spacing as soon as site data, i.e. the initial undisturbed rock temperature and the host rock heat transport properties, are available. Results of canister spacing calculations are given for the KBS-3V concept as well as for the KBS-3H concept. A combination of numerical and analytical methods is used for the KBS-3H calculations, while the KBS-3V calculations are purely analytical. Both methods are described in detail. Open gaps are assigned equivalent heat conductivities, calculated such that the conduction across the gaps will include also the heat transferred by radiation. The equivalent heat conductivities are based on the emissivities of the different gap surfaces. For the canister copper surface, the emissivity is determined by back-calculation of temperatures measured in the Prototype experiment at Aespoe HRL. The size of the different gaps and the emissivity values are of great importance for the results and will be investigated further in the future.

  19. Multi-Canister Overpack (MCO) Design Report

    International Nuclear Information System (INIS)

    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

  20. 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.

  1. 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.

  2. Canister design for deep borehole disposal of nuclear waste

    OpenAIRE

    Hoag, Christopher Ian.

    2006-01-01

    The objective of this thesis was to design a canister for the disposal of spent nuclear fuel and other high-level waste in deep borehole repositories using currently available and proven oil, gas, and geothermal drilling technology. The canister is suitable for disposal of various waste forms, such as fuel assemblies and vitrified waste. The design addresses real and perceived hazards of transporting and placing high-level waste, in the form of spent reactor fuel, into a deep igneous rock env...

  3. Spent nuclear fuel canister storage building conceptual design report

    International Nuclear Information System (INIS)

    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

  4. Drop tests of the Three Mile Island knockout canister

    International Nuclear Information System (INIS)

    A type of Three Mile Island Unit 2 (TMI-2) defueling canister, called a ''knockout'' canister, was subjected to a series of drop tests at the Oak Ridge National Laboratory's Drop Test Facility. These tests were designed to confirm the structural integrity of internal fixed neutron poisons in support of a request for NRC licensing of this type of canister for the shipment of TMI-2 reactor fuel debris to the Idaho National Engineering Laboratory (INEL) for the Core Examination R and D Program. Work conducted at the Oak Ridge National Laboratory included (1) precise physical measurements of the internal poison rod configuration before assembly, (2) canister assembly and welding, (3) nondestructive examination (an initial hydrostatic pressure test and an x-ray profile of the internals before and after each drop test), (4) addition of a simulated fuel load, (5) instrumentation of the canister for each drop test, (6) fabrication of a cask simulation vessel with a developed and tested foam impact limiter, (7) use of refrigeration facilities to cool the canister to well below freezing prior to three of the drops, (8) recording the drop test with still, high-speed, and normal-speed photography, (9) recording the accelerometer measurements during impact, (10) disassembly and post-test examination with precise physical measurements, and (11) preparation of the final report

  5. Physical properties of encapsulate spent fuel in canisters

    International Nuclear Information System (INIS)

    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)

  6. 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.

  7. Corrosion evaluation of fuel canister crusher rigging

    International Nuclear Information System (INIS)

    A fuel canister crusher with attached rigging is located in the 105 K-East Basin discharge chute. This equipment is slated to be moved as part of seismic mitigation to prevent a major basin leak through a construction joint located in the base of the chute. This corrosion analysis assessed the load-bearing ability of the rigging, which consists of shackles and thimble-spliced wire rope. The K-East Basin demineralized water results in corrosion rates of <2 mil/year (<0.05 mm/year) for carbon, low-alloy carbon, and stainless steels. The galvanized carbon steel shackles (with low-alloy steel anchor pins) have experienced negligible corrosion and are judged to be mechanically unaffected by their water exposure. The carbon steel wire rope and stainless steel thimbles have undergone minimal corrosion. Due to the small amount of corrosion products (as seen from video inspection), the absence of wire breakage, and a Factor of Safety calculation, it is judged that the wire rope and thimbles would withstand the proposed relocation activities

  8. 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.

  9. Canister filling materials -- Design requirements and evaluation of candidate materials

    International Nuclear Information System (INIS)

    SKB has been evaluating a copper/steel canister for use in the disposal of spent nuclear reactor fuel. Once the canister is breached by corrosion, it is possible that the void volume inside the canister might fill with water. Water inside the canister would moderate the energy of the neutrons emitted by spontaneous fission in the fuel. It the space in the canister between and around the fuel pins is occupied by canister filling materials, the potential for criticality is avoided. The authors have developed a set of design requirements for canister filling material for the case where it is to be used alone, with no credit for burnup of the fuel or other measures, such as the use of neutron absorbers. Requirements were divided into three classes: essential requirements, desirable features, and undesirable features. The essential requirements are that the material fill at least 60% of the original void space, that the solubility of the filling material be less than 100 mg/l in pure water or expected repository waters at 50 C, and that the material not compact under its own weight by more than 10%. In this paper they review the reasons for these requirements, the desirable and undesirable features, and evaluate 11 candidate materials with respect to the design requirements and features. The candidate materials are glass beads, lead shot, copper spheres, sand, olivine, hematite, magnetite, crushed rock, bentonite, other clays, and concrete. Emphasis is placed on the determination of whether further work is needed to eliminate uncertainties in the evaluation of the ability of a particular filling material to be successfully used under actual conditions, and on the ability to predict the long-term performance of the material under the repository conditions

  10. Vitrification of high level wastes: a review of the computer thermal analyses for storage canisters

    International Nuclear Information System (INIS)

    CANIST, a two-dimensional (r and THETA) computer program that solves the unsteady-state, heat conduction equation was used to model the thermal behavior of canisters filled with waste glass. CANIST has been found to be a valuable analytical tool for predicting the temperature profile of a waste storage canister as a function of several variables, including the diameter of the canister, the placement of internal fins, the heat generation rate of the waste glass, and the thermophysical properties of the canister and the waste glass. Thus, temperature dependent processes that may affect the integrity of the glass/canister unit, for example cracking, can be investigated using an analytical approach. In the present study, the canister temperature profiles predicted by CANIST were compared to canister temperatures measured during full-scale non-radioactive waste immobilization tests conducted at Pacific Northwest Laboratory. The agreement between experimental and predicted temperatures was good, particularly considering the fact that the thermophysical properties of the waste glass modeled have not yet been accurately determined. Examination of some glass-filled canisters has revealed cracking to have occurred in the glass. However, the comparison between measured and CANIST predicted temperatures suggests that cracking does not significantly influence the heat-transfer process. CANIST was also used to evaluate different ways of reducing the centerline temperature of a canister, and to predict the centerline temperature as a function of the heat generation rate of the waste glass and the type of interim storage, i.e., air or water

  11. 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

  12. 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.

  13. Safety Analysis Report for the PWR Spent Fuel Canister

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Heui Joo; Choi, Jong Won; Cho, Dong Keun; Chun, Kwan Sik; Lee, Jong Youl; Kim, Seong Ki; Kim, Seong Soo; Lee, Yang

    2005-11-15

    This report outlined the results of the safety assessment of the canisters for the PWR spent fuels which will be used in the KRS. All safety analyses including criticality and radiation shielding analyses, mechanical analyses, thermal analyses, and containment analyses were performed. The reference PWR spent fuels were in the 17x17 and determined to have 45,000 MWD/MTU burnup. The canister consists of copper outer shell and nodular cast iron inner structure with diameter of 102 cm and height of 483 cm. Criticality safety was checked for normal and abnormal conditions. It was assumed that the integrity of engineered barriers is preserved and saturated with water of 1.0g/cc for normal condition. For the abnormal condition container and bentonite was assumed to disappear, which allows the spent fuel to be surrounded by water with the most reactive condition. In radiation shielding analysis it was investigated that the absorbed dose at the surface of the canister met the safety limit. The structural analysis was conducted considering three load conditions, normal, extreme, and rock movement condition. Thermal analysis was carried out for the case that the canister with four PWR assemblies was deposited in the repository 500 meter below the surface with 40 m tunnel spacing and 6 m deposition hole spacing. The results of the safety assessment showed that the proposed KDC-1 canister met all the safety limits.

  14. Design, production and initial state of the canister

    International Nuclear Information System (INIS)

    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

  15. Decontamination of DWPF canisters by glass frit blasting

    International Nuclear Information System (INIS)

    High-level radioactive waste at the Savannah River Plant will be incorporated in borosilicate glass for permanent disposal. The waste glass will be encapsulated in a 304L stainless steel canister. During the filling operation the outside of the canister will become contaminated. This contamination must be reduced to an accepable level before the canister leaves the Defense Waste Processing Facility (DWPF). Tests with contaminated coupons have demonstrated that this decontamination can be accomplished by blasting the surface with glass frit. The contaminated glass frit byproduct of this operation is used as a feedstock for the waste glass process, so no secondary waste is created. Three blasting techniques, using glass frit as the blasting medium, were evaluated. Air-injected slurry blasting was the most promising and was chosen for further development. The optimum parametric values for this process were determined in tests using coupon weight loss as the output parameter. 1 reference, 13 figures, 3 tables

  16. 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.

  17. Chemical stability of copper-canisters in deep repository

    International Nuclear Information System (INIS)

    The spent fuel from Finnish nuclear reactors is planned to be encapsulated in thick-walled copper-iron canisters and placed deep into the bedrock. The copper wall of the canister provides a long-time shield against corrosion, preventing the high-level nuclear fuel from contact with ground water. In the report, stability of metallic copper and its possible corrosion reactions in the conditions of deep bedrock are evaluated by means of thermo-dynamic calculations. (90 refs., 28 figs., 11 tabs.)

  18. 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.

  19. 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

  20. 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.

  1. 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.

  2. Study on the methods for analysis of the chemical poison in canister by neutron activity

    International Nuclear Information System (INIS)

    The method that is used to analyse the poison gases in canister by neutron activity is proposed. Through theory analysis and experimental measurement, the feasibility for analysis of the poison gases in a canister by neutron activity has been demonstrated, and it is proved that the method itself do not result in radioactive problem to use again the canister. (authors)

  3. 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.

  4. Analysis of water from K west basin canisters (second campaign)

    Energy Technology Data Exchange (ETDEWEB)

    Trimble, D.J., Fluor Daniel Hanford

    1997-03-06

    Gas and liquid samples have been obtained 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. The data will provide 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 in 1995, and 50 canisters were sampled in a second campaign in 1996. The analysis results for the gas and liquid samples of the first campaign have been reported (Trimble 1995a; Trimble 1995b; Trimble 1996a; Trimble 1996b). An analysis of cesium-137 (137CS ) data from the second campaign samples was reported (Trimble and Welsh 1997), and the gas sample results are documented in Trimble 1997. This report documents the results of all analytes of liquid samples from the second campaign.

  5. High level waste canister emplacement and retrieval concepts study

    International Nuclear Information System (INIS)

    Several concepts are described for the interim (20 to 30 years) storage of canisters containing high level waste, cladding waste, and intermediate level-TRU wastes. It includes requirements, ground rules and assumptions for the entire storage pilot plant. Concepts are generally evaluated and the most promising are selected for additional work. Follow-on recommendations are made

  6. Corrosion resistance of a copper canister for spent nuclear fuel

    International Nuclear Information System (INIS)

    The report presents an evaluation of copper as canister material for spent nuclear fuel. The evaluation is made from the viewpoint of corrosion and applies to a concept of 1977. Supplementary corrosion studies have been performed. The report includes 9 appendices which deal with experimental data. (G.B.)

  7. 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.

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

    International Nuclear Information System (INIS)

    This publication consists of two parts: OCRWM (Office of Civilian Radioactive Waste Management) Bulletin; and Of Mountains ampersand 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

  9. Heat propagation from a radioactive waste repository. SKB 91 reference canister

    International Nuclear Information System (INIS)

    A study of heat propagation around a hypothetical radioactive waste repository is presented. The investigated flow domain was limited to a quarter of the flow domain around a single canister due to symmetry by vertical planes passing through the centre of the canister, half distance between the adjacent tunnels and the adjacent canisters. Strictly speaking, such an approach is applicable to a repository of infinite extent. However, from a practical point of view this assumption applies to all canisters but the ones close to the edge of the repository. The following different material regions were considered: (a.) Canister containing the spent fuel, (b.) Buffer (bentonite) around the canister, (c.) Backfilled (mixture of bentonite and sand) tunnels, and (d.) host Rock. The canister material was presented by a 'homogenized' medium obtained by weighted averaging of the main constituents of the canister, viz. spent fuel, copper and lead. A geothermal gradient of 13 degrees C/km was assumed. The initial heat effect per canister was 1066 W. The total vertical extent of the flow domain considered was about 1500 meters. The base case, with 6.2 m canister spacing and 30 m tunnel spacing, resulted in a maximum temperature at the canister/buffer interface of about 66 degrees C (corresponding to a temperature rise of about 54 degrees C), and about 50 degrees C (about 38 degrees C temperature rise) in the rock. (au)

  10. Pressure tests of two KBS-3 canister mock-ups

    International Nuclear Information System (INIS)

    The Swedish concept for geological disposal of spent nuclear fuel, the so-called KBS-3 concept, relies on a multibarrier system with the copper/cast iron canister as the first barrier. The canister is designed to retain its integrity for at least 100,000 years, which means that future glaciations need to be considered. A 3 km thick ice block together with hydrostatic pressure from groundwater and swelling of the buffer material would produce hydrostatic compressive stresses of maximum 44 MPa (440 bar). Although the canister is loaded globally in compression, tensile stresses develop at fuel channel surface with increasing load. Tensile tests of the insert material in the development phase of the KBS-3 canister indicated a large scatter and relatively low values of the inserts' ductility. An important issue was whether this could lead to mechanical failure of canisters at the 44 MPa iso-static load either by plastic collapse or fracture from the defects in the regions with tensile stresses. SKB therefore initiated a project together with the European commission's Joint Research Centre (JRC) Institute of Energy in Petten and a number of Swedish partners to evaluate the probability of mechanical failure during glaciation. Three inserts manufactured by different Swedish foundries and referred to as 1, 125 and 126 were used in the project. A large statistical test programme was developed to determine statistical distributions of various material parameters and defect distributions. These data were subsequently used in probabilistic analysis to determine the probability for local plastic collapse or fracture. The main conclusion was that the failure probability is extremely low at the design load (44 MPa) provided some basic geometrical requirements are fulfilled. In parallel to the statistical test programme and the associated analysis, the group decided also to perform two pressure tests of canister mock-ups to demonstrate the actual safety margins. The fractographic

  11. Drying tests conducted on Three Mile Island fuel canisters containing simulated debris

    Energy Technology Data Exchange (ETDEWEB)

    Palmer, A.J.

    1995-12-31

    Drying tests were conducted on TMI-2 fuel canisters filled with simulated core debris. During these tests, canisters were dried by heating externally by a heating blanket while simultaneously purging the canisters` interior with hot, dry nitrogen. Canister drying was found to be dominated by moisture retention properties of a concrete filler material (LICON) used for geometry control. This material extends the drying process 10 days or more beyond what would be required were it not there. The LICON resides in a nonpurgeable chamber separate from the core debris, and because of this configuration, dew point measurements on the exhaust stream do not provide a good indication of the dew point in the canisters. If the canisters are not dried, but rather just dewatered, 140-240 lb of water (not including the LICON water of hydration) will remain in each canister, approximately 50-110 lb of which is pore water in the LICON and the remainder unbound water.

  12. Desludging of N Reactor fuel canisters: Analysis, Test, and data requirements

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, A.B. Jr.

    1996-01-01

    The N Reactor fuel is currently stored in canisters in the K East (KE) and K West (KW) Basins. In KE, the canisters have open tops; in KW, the cans have sealed lids, but are vented to release gases. Corrosion products have formed on exposed uranium metal fuel, on carbon steel basin component surfaces, and on aluminum alloy canister surfaces. Much of the corrosion product is retained on the corroding surfaces; however, large inventories of particulates have been released. Some of the corrosion product particulates form sludge on the basin floors; some particulates are retained within the canisters. The floor sludge inventories are much greater in the KE Basin than in the KW Basin because KE Basin operated longer and its water chemistry was less controlled. Another important factor is the absence of lids on the KE canisters, allowing uranium corrosion products to escape and water-borne species, principally iron oxides, to settle in the canisters. The inventories of corrosion products, including those released as particulates inside the canisters, are only beginning to be characterized for the closed canisters in KW Basin. The dominant species in the KE floor sludge are oxides of aluminum, iron, and uranium. A large fraction of the aluminum and uranium floor sludge particulates may have been released during a major fuel segregation campaign in the 1980s, when fuel was emptied from 4990 canisters. Handling and jarring of the fuel and aluminum canisters seems likely to have released particulates from the heavily corroded surfaces. Four candidate methods are discussed for dealing with canister sludge emerged in the N Reactor fuel path forward: place fuel in multi-canister overpacks (MCOs) without desludging; drill holes in canisters and drain; drill holes in canisters and flush with water; and remove sludge and repackage the fuel.

  13. Spent nuclear fuel Canister Storage Building CDR Review Committee report

    International Nuclear Information System (INIS)

    The Canister Storage Building (CSB) is a subproject under the Spent Nuclear Fuels Major System Acquisition. This subproject is necessary to design and construct a facility capable of providing dry storage of repackaged spent fuels received from K Basins. The CSB project completed a Conceptual Design Report (CDR) implementing current project requirements. A Design Review Committee was established to review the CDR. This document is the final report summarizing that review

  14. Spent nuclear fuel Canister Storage Building CDR Review Committee report

    Energy Technology Data Exchange (ETDEWEB)

    Dana, W.P.

    1995-12-01

    The Canister Storage Building (CSB) is a subproject under the Spent Nuclear Fuels Major System Acquisition. This subproject is necessary to design and construct a facility capable of providing dry storage of repackaged spent fuels received from K Basins. The CSB project completed a Conceptual Design Report (CDR) implementing current project requirements. A Design Review Committee was established to review the CDR. This document is the final report summarizing that review

  15. Analysis of probability of defects in the disposal canisters

    International Nuclear Information System (INIS)

    This report presents a probability model for the reliability of the spent nuclear waste final disposal canister. Reliability means here that the welding of the canister lid has no critical defects from the long-term safety point of view. From the reliability point of view, both the reliability of the welding process (that no critical defects will be born) and the non-destructive testing (NDT) process (all critical defects will be detected) are equally important. In the probability model, critical defects in a weld were simplified into a few types. Also the possibility of human errors in the NDT process was taken into account in a simple manner. At this moment there is very little representative data to determine the reliability of welding and also the data on NDT is not well suited for the needs of this study. Therefore calculations presented here are based on expert judgements and on several assumptions that have not been verified yet. The Bayesian probability model shows the importance of the uncertainty in the estimation of the reliability parameters. The effect of uncertainty is that the probability distribution of the number of defective canisters becomes flat for larger numbers of canisters compared to the binomial probability distribution in case of known parameter values. In order to reduce the uncertainty, more information is needed from both the reliability of the welding and NDT processes. It would also be important to analyse the role of human factors in these processes since their role is not reflected in typical test data which is used to estimate 'normal process variation'.The reported model should be seen as a tool to quantify the roles of different methods and procedures in the weld inspection process. (orig.)

  16. BRIC-60: Biological Research in Canisters (BRIC)-60

    Science.gov (United States)

    Richards, Stephanie E. (Compiler); Levine, Howard G.; Romero, Vergel

    2016-01-01

    The Biological Research in Canisters (BRIC) is an anodized-aluminum cylinder used to provide passive stowage for investigations evaluating the effects of space flight on small organisms. Specimens flown in the BRIC 60 mm petri dish (BRIC-60) hardware include Lycoperscion esculentum (tomato), Arabidopsis thaliana (thale cress), Glycine max (soybean) seedlings, Physarum polycephalum (slime mold) cells, Pothetria dispar (gypsy moth) eggs and Ceratodon purpureus (moss).

  17. 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.

  18. 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.

  19. 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.

  20. Stress analysis of high-level waste canisters: methods, applications, and design data

    International Nuclear Information System (INIS)

    An overview of stress analysis methods, structural design procedures, and design data is presented for canisters used to package solidified wastes, particularly borosilicate glass. In addition, waste processing, canister materials, fabrication and inspection methods, and performance testing are summarized. Sources of stress in canisters are lifting and handling loads, internal pressure, high-temperature filling operations, transient heating and cooling, differential thermal expansions of canisters and glass, and impact loadings from low-probability accidents. Results of case studies that illustrate applicable methods of stress analyses are presented for these sources of stress. Existing sections of ASME Boiler and Pressure Vessel Code are applicable to canister fabrication, but the code does not cover many aspects of canister service loadings. Specialized criteria for minimum wall thicknesses to sustain filling stresses are proposed in this report. Results of a test program to measure the creep strength of candidate canister materials are described. Methods to predict residual stresses in the walls of waste canisters are described; predicted residual stress levels agree with measured stress levels. The consequences of these residual stresses are reviewed, and stress-corrosion cracking is identified as the mode of canister failure affected by residual stresses. Canister-closure design is covered in detail, particularly the welding and inspection of the final closure seal-weld. It is shown that the methods of fracture mechanics and fatigue-crack-growth analyses are valuable tools for evaluating the performance of closure welds in the presence of crack-like defects. Canister performance in process trials at PNL shows the ability of canisters to survive high temperatures and loadings during processing. Impact tests show that a suitably designed canister can sustain severe impacts without loss of intergrity

  1. Test manufacturing of copper canisters with cast inserts. Assessment report

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, C.G

    1998-08-01

    The current design of canisters for the deep repository for spent nuclear fuel consists of an outer corrosion-protective copper casing in the form of a tubular section with lid and bottom and an inner pressure-resistant insert. The insert is designed to be manufactured by casting and inside are channels in which the fuel assemblies are to be placed. Over the last years, a number of full-scale manufacturing tests of all canister components have been carried out. The purpose has been to determine and develop the best manufacturing technique and to establish long-term contacts with the best suppliers of material and technology. Part of the work has involved the developing and implementing of a quality assurance system in accordance with ISO 9001, covering the whole chain from suppliers of material up to and including the delivery of assembled canisters. This report consists of a description of the design of the canister together with current drawings and complementary technical specifications stipulating, among other things, requirements placed on different materials. The different manufacturing methods that have been used are also described and commented on in both text and illustrations. For the manufacturing of copper tubes, the roll-forming of rolled plate to tube halves and longitudinal welding is a method that has been tested on a relatively large number of tubes by now, and that probably can be developed into a functioning production method. However, the very promising outcome of performed tests on seamless tube manufacturing, has resulted in a change in direction in tube manufacturing, focusing on continued testing of extrusion as well as pierce and draw processing in the immediate future. In connection with ongoing operations, new manufacturing tests of tubes with less material thickness will be carried out. Test manufacturing of cast inserts has resulted in the choice of nodular iron as material in the continued work. This improvement in design has resulted

  2. Test manufacturing of copper canisters with cast inserts. Assessment report

    International Nuclear Information System (INIS)

    The current design of canisters for the deep repository for spent nuclear fuel consists of an outer corrosion-protective copper casing in the form of a tubular section with lid and bottom and an inner pressure-resistant insert. The insert is designed to be manufactured by casting and inside are channels in which the fuel assemblies are to be placed. Over the last years, a number of full-scale manufacturing tests of all canister components have been carried out. The purpose has been to determine and develop the best manufacturing technique and to establish long-term contacts with the best suppliers of material and technology. Part of the work has involved the developing and implementing of a quality assurance system in accordance with ISO 9001, covering the whole chain from suppliers of material up to and including the delivery of assembled canisters. This report consists of a description of the design of the canister together with current drawings and complementary technical specifications stipulating, among other things, requirements placed on different materials. The different manufacturing methods that have been used are also described and commented on in both text and illustrations. For the manufacturing of copper tubes, the roll-forming of rolled plate to tube halves and longitudinal welding is a method that has been tested on a relatively large number of tubes by now, and that probably can be developed into a functioning production method. However, the very promising outcome of performed tests on seamless tube manufacturing, has resulted in a change in direction in tube manufacturing, focusing on continued testing of extrusion as well as pierce and draw processing in the immediate future. In connection with ongoing operations, new manufacturing tests of tubes with less material thickness will be carried out. Test manufacturing of cast inserts has resulted in the choice of nodular iron as material in the continued work. This improvement in design has resulted

  3. 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.

  4. 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.

  5. 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.

  6. Plutonium Can-In-Canister-Design Basis Event Analysis

    International Nuclear Information System (INIS)

    The purpose of this document is to perform a preliminary design basis event (DBE) analysis of the immobilized plutonium (can-in-canister) waste form to be referred to in this analysis as high level waste/plutonium (HLW/Pu). The objective of the analysis is to determine any preclosure safety impacts of the waste form on the Monitored Geologic Repository (MGR). The scope of this analysis is to determine the offsite dose consequences and associated frequencies of selected DBEs for systems handling disposable canisters that bound all surface and subsurface off-normal events, and to compare these results against regulatory limits. The results of this work are preliminary and are intended to be used to establish a set of preliminary MGR and waste form requirements, to identify mitigation or prevention options that may be required to meet regulatory limits, and to provide input to the Site Recommendation (SR) report. This document is prepared in accordance with the associated development plan (Civilian Radioactive Waste Management System Management and Operating Contractor [CRWMS M and O] 1999e)

  7. 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.

  8. 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.

  9. 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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  12. SPENT NUCLEAR FUEL (SNF) PROJECT CANISTER STORAGE BUILDING (CSB) MULTI CANISTER OVERPACK (MCO) SAMPLING SYSTEM VALIDATION (OCRWM)

    Energy Technology Data Exchange (ETDEWEB)

    BLACK, D.M.; KLEM, M.J.

    2003-11-17

    Approximately 400 Multi-canister overpacks (MCO) containing spent nuclear fuel are to be interim stored at the Canister Storage Building (CSB). Several MCOs (monitored MCOs) are designated to be gas sampled periodically at the CSB sampling/weld station (Bader 2002a). The monitoring program includes pressure, temperature and gas composition measurements of monitored MCOs during their first two years of interim storage at the CSB. The MCO sample cart (CART-001) is used at the sampling/weld station to measure the monitored MCO gas temperature and pressure, obtain gas samples for laboratory analysis and refill the monitored MCO with high purity helium as needed. The sample cart and support equipment were functionally and operationally tested and validated before sampling of the first monitored MCO (H-036). This report documents the results of validation testing using training MCO (TR-003) at the CSB. Another report (Bader 2002b) documents the sample results from gas sampling of the first monitored MCO (H-036). Validation testing of the MCO gas sampling system showed the equipment and procedure as originally constituted will satisfactorily sample the first monitored MCO. Subsequent system and procedural improvements will provide increased flexibility and reliability for future MCO gas sampling. The physical operation of the sampling equipment during testing provided evidence that theoretical correlation factors for extrapolating MCO gas composition from sample results are unnecessarily conservative. Empirically derived correlation factors showed adequate conservatism and support use of the sample system for ongoing monitored MCO sampling.

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

    International Nuclear Information System (INIS)

    '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 bentonite is

  14. 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

  15. Reliability in sealing of canister for spent nuclear fuel

    International Nuclear Information System (INIS)

    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

  16. 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

  17. 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

  18. Analysis of sludge from Hanford K East Basin canisters

    International Nuclear Information System (INIS)

    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

  19. STS-100 MPLM Raffaello is moved to the payload canister

    Science.gov (United States)

    2001-01-01

    KENNEDY SPACE CENTER, Fla. - Workers inside the payload canister wait for the Multi-Purpose Logistics Module Raffaello to be lowered inside. It joins the Canadian robotic arm, SSRMS, already in place. Both elements are part of the payload on mission STS- 100 to the International Space Station. Raffaello carries six system racks and two storage racks for the U.S. Lab. The arm has seven motorized joints and is capable of handling large payloads and assisting with docking the Space Shuttle. The SSRMS is self- relocatable with a Latching End Effector so it can be attached to complementary ports spread throughout the Station'''s exterior surfaces. Launch of STS-100 is scheduled for April 19, 2001 at 2:41 p.m. EDT from Launch Pad 39A.

  20. Acoustic monitoring techniques for corrosion degradation in cemented waste canisters

    International Nuclear Information System (INIS)

    This report describes work carried out to investigate acoustic emission as a monitor of corrosion and degradation of wasteforms where the waste is potentially reactive metal. Electronic monitoring equipment has been designed, built and tested to allow long-term monitoring of a number of waste packages simultaneously. Acoustic monitoring experiments were made on a range of 1 litre cemented Magnox and aluminium samples cast into canisters comparing the acoustic events with hydrogen gas evolution rates and electrochemical corrosion rates. The attenuation of the acoustic signals by the cement grout under a range of conditions has been studied to determine the volume of wasteform that can be satisfactorily monitored by one transducer. The final phase of the programme monitored the acoustic events from full size (200 litre) cemented, inactive, simulated aluminium swarf wastepackages prepared at the AEA waste cementation plant at Winfrith. (Author)

  1. 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.

  2. Procedural development for nuclear waste canister impact testing

    International Nuclear Information System (INIS)

    Double containment requirements for transporting nuclear waste in glass form are costly and may not be necessary for some waste forms. To allow single containment, a procedure for examining particle size distribution and the amount of respirable particles generated under accident conditions was needed. A statistically designed experiment was conducted to examine the effects of glass temperature, fill rate and canister drop orientation upon the amount of sub-ten micron particles generated under simulated accident conditions. Measuring such small particles is somewhat inaccurate because of material loss in handling. By assuming a lognormal particle size distribution, the amount of sub-ten micron particles was estimated from the results for the larger measurable particles. Analyses revealed no temperature or fill rate effect but indicated that the amount of respirable particles is affected by drop orientation. This led to identification of a worst case drop orientation to be used in qualification testing. 4 refs., 2 figs

  3. 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.

  4. Choices of canisters and elements for the first fuel shipment from K West Basin

    Energy Technology Data Exchange (ETDEWEB)

    Makenas, B.J.

    1995-03-01

    Twenty-two canisters (10 prime and 12 backup candidates) in the K West Basin have been identified as containing fuel which, when examined, will satisfy the Data Quality Objectives for the first fuel shipment from this basin. These were chosen as meeting criteria such as containing relatively long fuel elements, locking bar integrity, and the availability of gas/liquid interface level measurements for associated canister gas traps. Two canisters were identified as having reported broken fuel on initial loading. Usage and interpretation of canister cesium concentration measurements have also been established and levels of maximum and minimum acceptable cesium concentration (from a data optimization point of view) for decapping have been determined although other operational cesium limits may also apply. Criteria for picking particular elements, once a canister is opened, are reviewed in this document. A pristine, a slightly damaged, and a badly damaged element are desired. The latter includes elements with end caps removed but does not include elements which have large amounts of swelling or split cladding that might interfere with handling tools. Finally, operational scenarios have been suggested to aid in the selections of canisters and elements in a way that utilizes anticipated canister gas sampling and leads to a correct and quick choice of elements which will supply the desired data.

  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.

    2000-06-07

    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.

  6. Plutonium Immobilization Project - Can-In-Canister Hardware Development/Selection

    International Nuclear Information System (INIS)

    This paper covers the design, development and testing of the magazines (cylinders containing cans of plutonium-ceramic pucks) and the rack that holds them in place inside the waste glass canister. Several magazine and rack concepts were evaluated to produce a design that gives the optimal balance between resistance to thermal degradation and facilitation of remote handling. This paper also reviews the effort to develop a jointed robotic arm that can remotely load seven magazines into defined locations inside a stationary canister working only through the 4 inch (102mm) diameter canister throat

  7. Preliminary design specification for Department of Energy standardized spent nuclear fuel canisters. Volume 2: Rationale document

    International Nuclear Information System (INIS)

    This document (Volume 2) is a companion document to a preliminary design specification for the design of canisters to be used during the handling, storage, transportation, and repository disposal of Department of Energy (DOE) spent nuclear fuel (SNF). This document contains no procurement information, such as the number of canisters to be fabricated, explicit timeframes for deliverables, etc. However, this rationale document does provide background information and design philosophy in order to help engineers better understand the established design criteria (contained in Volume 1 respectively) necessary to correctly design and fabricate these DOE SNF canisters

  8. 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.

  9. Demonstration of a Solution Film Leak Test Technique and Equipment for the S00645 Canister Closure

    International Nuclear Information System (INIS)

    The purpose of this effort was to demonstrate that the SFT technique, when adapted to a DWPF canister nozzle, is capable of detecting leaks not meeting the Waste Acceptance Product Specifications (WAPS) acceptance criterion

  10. Demonstration of a Solution Film Leak Test Technique and Equipment for the S00645 Canister Closure

    Energy Technology Data Exchange (ETDEWEB)

    Cannell, G.R.

    1999-10-07

    The purpose of this effort was to demonstrate that the SFT technique, when adapted to a DWPF canister nozzle, is capable of detecting leaks not meeting the Waste Acceptance Product Specifications (WAPS) acceptance criterion.

  11. Coupled transport/reaction modelling of copper canister corrosion aided by microbial processes

    International Nuclear Information System (INIS)

    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 sulphide 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 intruding groundwater, 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 is included in the model. The local depth of copper canister corrosion is calculated by the model. (orig.)

  12. Coupled transport/reaction modelling of copper canister corrosion aided by microbial processes

    Energy Technology Data Exchange (ETDEWEB)

    Liu Jinsong; Neretnieks, I. [Dept. of Chemical Engineering and Technology, Royal Inst. of Tech., Stockholm (Sweden)

    2004-07-01

    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 sulphide 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 intruding groundwater, 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 is included in the model. The local depth of copper canister corrosion is calculated by the model. (orig.)

  13. High-level waste canister storage final design, installation, and testing. Topical report

    Energy Technology Data Exchange (ETDEWEB)

    Connors, B.J.; Meigs, R.A.; Pezzimenti, D.M.; Vlad, P.M.

    1998-04-01

    This report is a description of the West Valley Demonstration Project`s radioactive waste storage facility, the Chemical Process Cell (CPC). This facility is currently being used to temporarily store vitrified waste in stainless steel canisters. These canisters are stacked two-high in a seismically designed rack system within the cell. Approximately 300 canisters will be produced during the Project`s vitrification campaign which began in June 1996. Following the completion of waste vitrification and solidification, these canisters will be transferred via rail or truck to a federal repository (when available) for permanent storage. All operations in the CPC are conducted remotely using various handling systems and equipment. Areas adjacent to or surrounding the cell provide capabilities for viewing, ventilation, and equipment/component access.

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

    International Nuclear Information System (INIS)

    The purpose of this report is to provide an evaluation of the Canister Storage Building (CSB) design criteria, the design's compliance with the applicable criteria, and the basis for authorization to proceed with construction of the CSB

  15. Canister design concepts for disposal of spent fuel and high level waste

    International Nuclear Information System (INIS)

    As part of its long-term plans for development of a repository for spent fuel (SF) and high level waste (HLW), Nagra is exploring various options for the selection of materials and design concepts for disposal canisters. The selection of suitable canister options is driven by a series of requirements, one of the most important of which is providing a minimum 1000 year lifetime without breach of containment. One candidate material is carbon steel, because of its relatively low corrosion rate under repository conditions and because of the advanced state of overall technical maturity related to construction and fabrication. Other materials and design options are being pursued in parallel studies. The objective of the present study was to develop conceptual designs for carbon steel SF and HLW canisters along with supporting justification. The design process and outcomes result in design concepts that deal with all key aspects of canister fabrication, welding and inspection, short-term performance (handling and emplacement) and long-term performance (corrosion and structural behaviour after disposal). A further objective of the study is to use the design process to identify the future work that is required to develop detailed designs. The development of canister designs began with the elaboration of a number of design requirements that are derived from the need to satisfy the long-term safety requirements and the operational safety requirements (robustness needed for safe handling during emplacement and potential retrieval). It has been assumed based on radiation shielding calculations that the radiation dose rate at the canister surfaces will be at a level that prohibits manual handling, and therefore a hot cell and remote handling will be needed for filling the canisters and for final welding operations. The most important canister requirements were structured hierarchically and set in the context of an overall design methodology. Conceptual designs for SF canisters

  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. Data quality objectives for gas and liquid samples from sealed K Basin canisters

    Energy Technology Data Exchange (ETDEWEB)

    Makenas, B.J.

    1995-03-01

    Data Quality Objectives (DQOS) for gas and liquid sampling from the sealed canisters in K West Basin have been developed and are presented in this document. The scope of this document is limited primarily to the initial sampling effort. This sampling campaign either supports the selection of canisters to provide fuel for hotcell examinations, supports the demonstration of sampling equipment capabilities or provides an initial assessment of gas/liquid chemistry for comparison to the results of fuel element hotcell examinations. No sampling of canisters has occurred since 1983. It is proposed here that samples of gas and water be analyzed for constituents such as cesium, fission gases, and hydrogen which are markers for corrosion of uranium in a water environment. These data will allow an assessment of the risks involved when particular canisters are opened to retrieve fuel. This sampling campaign will also ensure that canisters with some failed fuel elements are included in the population that is opened for retrieval of fuel for hotcell examinations. Additionally, valuable correlations between the macroscopic visible condition of fuel, hotcell examinations, and the gases generated in canisters will be possible. The analysis of other chemical species in the gas and liquid will allow assessments of the performance of the previously added corrosion inhibitor and possibly assessments of radiolysis. Sampling of canisters will be performed with equipment that opens the valves in the canister lid and draws a 15 ml sample of either gas or water. This work will most likely be performed in one of the pits-associated with the K West Basin.

  18. 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.

  19. Effects of annular air gaps surrounding an emplaced nuclear waste canister in deep geologic storage

    International Nuclear Information System (INIS)

    Annular air spaces surrounding an emplaced nuclear waste canister in deep geologic storage will have significant effects on the long-term performance of the waste form. Addressed specifically in this analysis is the influence of a gap on the thermal response of the waste package. Three dimensional numerical modeling predicts temperature effects for a series of parameter variations, including the influence of gap size, surface emissivities, initial thermal power generation of the canister, and the presence/absence of a sleeve. Particular emphasis is placed on determining the effects these variables have on the canister surface temperature. We have identified critical gap sizes at which the peak transient temperature occurs when gap widths are varied for a range of power levels. It is also shown that high emissivities for the heat exchanging surfaces are desirable, while that of the canister surface has the greatest influence. Gap effects are more pronounced, and therefore more effort should be devoted to optimal design, in situations where the absolute temperature of the near field medium is high. This occurs for higher power level emplacements and in geomedia with low thermal conductivities. Finally, loosely inserting a sleeve in the borehole effectively creates two gaps and drastically raises the canister peak temperature. It is possible to use these results in the design of an optimum package configuration which will maintain the canister at acceptable temperature levels. A discussion is provided which relates these findings to NRC regulatory considerations

  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. Corrosion of iron: A study for radioactive waste canisters

    Science.gov (United States)

    Lagha, S. Ben; Crusset, D.; Mabille, I.; Tran, M.; Bernard, M. C.; Sutter, E.

    2007-05-01

    The purpose of this study is to examine the risks of atmospheric corrosion of steel waste canisters following their deep geological disposal in the temperature range from 303 to 363 K. The work was performed using iron samples deposited as thin films on a quartz crystal microbalance (QCM) and disposed in a climatic chamber. The experiments showed that, in the temperature under study (298-363 K), the mass increase due to the formation of oxide/hydroxide rose sharply above 70% RH, as is commonly observed at room temperatures, indicating that the phenomenon remains electrochemical in nature. Ex situ Raman spectrometric analyses indicate the formation of magnetite, maghemite and oxyhydroxides species in the 298-363 K temperature range, and for oxygen contents above 1 vol.%, whereas only Fe3O4 and γ-Fe2O3 are detected at 363 K. In this work, the kinetics of the rust growth is discussed, on the bases of the rate of mass increase and of the composition of the rust, as a function of the climatic parameters and the oxygen content of the atmosphere.

  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. Nanomembrane Canister Architectures for the Visualization and Filtration of Oxyanion Toxins with One-Step Processing.

    Science.gov (United States)

    Aboelmagd, Ahmed; El-Safty, Sherif A; Shenashen, Mohamed A; Elshehy, Emad A; Khairy, Mohamed; Sakaic, Masaru; Yamaguchi, Hitoshi

    2015-11-01

    Nanomembrane canister-like architectures were fabricated by using hexagonal mesocylinder-shaped aluminosilica nanotubes (MNTs)-porous anodic alumina (PAA) hybrid nanochannels. The engineering pattern of the MNTs inside a 60 μm-long membrane channel enabled the creation of unique canister-like channel necks and cavities. The open-tubular canister architecture design provides controllable, reproducible, and one-step processing patterns of visual detection and rejection/permeation of oxyanion toxins such as selenite (SeO3(2-)) in aquatic environments (i.e., in ground and river water sources) in the Ibaraki Prefecture of Japan. The decoration of organic ligand moieties such as omega chrome black blue (OCG) into inorganic Al2O3@tubular SiO2/Al2O3 canister membrane channel cavities led to the fabrication of an optical nanomembrane sensor (ONS). The OCG ligand was not leached from the canister as observed in washing, sensing, and recovery assays of selenite anions in solution, which enabled its multiple reuse. The ONS makes a variety of alternate processing analyses of selective quantification, visual detection, rejection/permeation, and recovery of toxic selenite quick and simple without using complex instrumentation. Under optimal conditions, the ONS canister exhibited a high selectivity toward selenite anions relative to other ions and a low-level detection limit of 0.0093 μM. Real analytical data showed that approximately 96% of SeO3(2-) anions can be recovered from aquatic and wastewater samples. The ONS canister holds potential for field recovery applications of toxic selenite anions from water. PMID:26178184

  6. 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

  7. Manufacturing of the canister shells T54 and T55

    International Nuclear Information System (INIS)

    This report constitutes a summary of the manufacturing test of the disposal canister copper shells T54 and T55. The copper billets were manufactured at Luvata Pori Oy, Finland. The hot-forming and machining of the copper shells were made at Vallourec and Mannesmann Tubes, Reisholz mill, Germany. The shells were manufactured with the pierce and draw method. Both of the pipes were manufactured separately in two phases. The first phase consisted of following steps: preheating of the billet, upsetting, piercing and the first draw with mandrel through drawing ring. After cooling down the block is measured and machined in case of excessive eccentricity or surface defects. In the second phase the block is heated up again and expanded and drawn in 6 sequences. In this process the pipe inside dimension is expanded and the length is increased in each step. Before the last, the 6th step, the bottom of the pipe is deformed in a sequence of special processes. During the manufacture of the first pipe, T54, some difficulties were detected with the centralization of the billet before upsetting. For the second manufacture of the T55, an additional steering ring was made and the result was remarkably more coaxial. After the manufacture and non-destructive inspections the shells were cut in pieces and three parts of each shell were taken for destructive testing. The three inspected parts were the bottom plate, a ring from the middle of the cylinder and a ring from the top of the cylinder. The destructive testing was made by Luvata Pori Oy. In spite of some practical difficulties and accidents during the manufacturing process, the results of the examinations showed that both of the test produced copper shells fulfilled all the specified requirements as for soundness (integrity), mechanical properties, chemical composition, dimensions, hardness and grain size. (orig.)

  8. Recommendations for codes and standards to be used for design and fabrication of high level waste canister

    International Nuclear Information System (INIS)

    This study identifies codes, standards, and regulatory requirements for developing design criteria for high-level waste (HLW) canisters for commercial operation. It has been determined that the canister should be designed as a pressure vessel without provision for any overpressure protection type devices. It is recommended that the HLW canister be designed and fabricated to the requirements of the ASME Section III Code, Division 1 rules, for Code Class 3 components. Identification of other applicable industry and regulatory guides and standards are provided in this report. Requirements for the Design Specification are found in the ASME Section III Code. It is recommended that design verification be conducted principally with prototype testing which will encompass normal and accident service conditions during all phases of the canister life. Adequacy of existing quality assurance and licensing standards for the canister was investigated. One of the recommendations derived from this study is a requirement that the canister be N stamped. In addition, acceptance standards for the HLW waste should be established and the waste qualified to those standards before the canister is sealed. A preliminary investigation of use of an overpack for the canister has been made, and it is concluded that the use of an overpack, as an integral part of overall canister design, is undesirable, both from a design and economics standpoint. However, use of shipping cask liners and overpack type containers at the Federal repository may make the canister and HLW management safer and more cost effective. There are several possible concepts for canister closure design. These concepts can be adapted to the canister with or without an overpack. A remote seal weld closure is considered to be one of the most suitable closure methods; however, mechanical seals should also be investigated

  9. A review of the possible effects of hydrogen on lifetime of carbon steel nuclear waste canisters

    International Nuclear Information System (INIS)

    In Switzerland, the National Cooperative for the Disposal of Radioactive Waste (Nagra) is responsible for developing an effective method for the safe disposal of vitrified high level waste (HLW) and spent fuel. One of the options for disposal canisters is thick-walled carbon steel. The canisters, which would have a diameter of about 1 m and a length of about 3 m (HLW) or about 5 m (spent fuel), will be embedded in horizontal tunnels and surrounded with bentonite clay. The regulatory requirement for the minimum canister lifetime is 1000 years but demonstration of a minimum lifetime of 10,000 years would be desirable. The pore-water to which the canister will be exposed is of marine origin with about 0.1-0.3 M Cl-. Since hydrogen is generated during the corrosion process, it is necessary to assess the probability of hydrogen assisted cracking modes and to make recommendations to eliminate that probability. To that aim, key reports detailing projections for the local environment and associated corrosion rate of the waste canister have been evaluated with the focus on the implication for the absorbed hydrogen concentration in the steel. Simple calculations of hydrogen diffusion and accumulation in the inner compartment of the sealed canister indicate that a pressure equivalent to that for gas pockets external to the canister (envisaged to be about 10 MPa) may be attained in the proposed exposure time, an important consideration since it is not possible to modify the internal surface of the closure weld. Current ideas on mechanisms of hydrogen assisted cracking are assessed from which it is concluded that the mechanistic understanding and associated models of hydrogen assisted cracking are insufficient to provide a framework for quantitative prediction for this application. The emphasis then was to identify threshold conditions for cracking and to evaluate the likelihood that these may be exceeded over the lifetime of the containment. Based on an analysis of data in the

  10. Numerical Modelling of Mechanical Integrity of the Copper-Cast Iron Canister. A Literature Review

    International Nuclear Information System (INIS)

    This review article presents a summary of the research works on the numerical modelling of the mechanical integrity of the composite copper-cast iron canisters for the final disposal of Swedish nuclear wastes, conducted by SKB and SKI since 1992. The objective of the review is to evaluate the outstanding issues existing today about the basic design concepts and premises, fundamental issues on processes, properties and parameters considered for the functions and requirements of canisters under the conditions of a deep geological repository. The focus is placed on the adequacy of numerical modelling approaches adopted in regards to the overall mechanical integrity of the canisters, especially the initial state of canisters regarding defects and the consequences of their evolution under external and internal loading mechanisms adopted in the design premises. The emphasis is the stress-strain behaviour and failure/strength, with creep and plasticity involved. Corrosion, although one of the major concerns in the field of canister safety, was not included

  11. A preliminary assessment of gas migration from the copper/steel canister

    International Nuclear Information System (INIS)

    A preliminary assessment has been carried out of the consequences of hydrogen gas generation in the copper/steel canister, a new concept that is being considered by SKB, Sweden, for the encapsulation of spent fuel for geological disposal. The principal aims of the study were as follows: a. to determine the mechanisms by which gas generated by anaerobic corrosion will migrate from a canister; b. to identify the possible consequences of gas generation, for example overpressurization of the canisters and effects on water movement; c. to carry out studies to assess the magnitudes of the consequences of gas generation and the way in which they are influenced by the mechanisms and ease of gas migration; d. to determine the likely fate of the gas produced in the repository; for example whether the gas will eventually be dissolved in the groundwater as it moves away from the canister or whether it will collect as free gas in the tunnel or elsewhere; e. to identify the potential benefits of using computer modelling techniques for estimating hydrogen generation rates within disposal canisters during the post-emplacement period

  12. Effects of stabilizers on the heat transfer characteristics of a nuclear waste canister

    International Nuclear Information System (INIS)

    This report summarizes the feasibility and the effectiveness of using stabilizers (internal metal structural components) to augment the heat transfer characteristics of a nuclear waste canister. The problem was modeled as a transient two-dimensional heat transfer in two physical domains - the stabilizer and the wedge (a 30-degree-angle canister segment), which includes the heat-producing spent-fuel rods. This problem is solved by a simultaneous and interrelated numerical investigation of the two domains in cartesian and polar coordinate systems. The numerical investigations were performed for three cases. In the first case, conduction was assumed to be the dominant mechanism for heat transfer. The second case assumed that radiation was the dominant mechanism, and in the third case both radiation and conduction were considered as mechanisms of heat transfer. The results show that for typical conditions in a waste package design, the stabilizers are quite effective in reducing the overall temperature in a waste canister. Furthermore, the results show that increasing the stabilizer thickness over the thickness specified in the present design has a negligible effect on the temperature distribution in the canister. Finally, the presence of the stabilizers was found to shift the location of the peak temperature areas in the waste canister

  13. 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.

  14. Corrosion of the copper canister in the repository environment

    International Nuclear Information System (INIS)

    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 there is

  15. 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

  16. 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.

  17. Design, Manufacturing, and Performance estimation of a Disposal Canister for the Ceramic Waste from Pyroprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Min Soo; Choi, Heui Joo; Lee, Jong Youl; Choi, Jong Won [Korea Atomic Energy Institute, Daejeon (Korea, Republic of)

    2012-09-15

    A pyroprocess is currently being developed by KAERI to cope with a highly accumulated spent nuclear fuel in Korea. The pyroprocess produces a certain amount of high-level radioactive waste (HLW), which is solidified by a ceramic binder. The produced ceramic waste will be confined in a secure disposal canister and then placed in a deep geologic formation so as not to contaminate human environment. In this paper, the development of a disposal canister was overviewed by discussing mainly its design premises, constitution, manufacturing methods, corrosion resistance in a deep geologic environment, radiation shielding, and structural stability. The disposal canister should be safe from thermal, chemical, mechanical, and biological invasions for a very long time so as not to release any kind of radionuclides.

  18. Design, Manufacturing, and Performance estimation of a Disposal Canister for the Ceramic Waste from Pyroprocessing

    International Nuclear Information System (INIS)

    A pyroprocess is currently being developed by KAERI to cope with a highly accumulated spent nuclear fuel in Korea. The pyroprocess produces a certain amount of high-level radioactive waste (HLW), which is solidified by a ceramic binder. The produced ceramic waste will be confined in a secure disposal canister and then placed in a deep geologic formation so as not to contaminate human environment. In this paper, the development of a disposal canister was overviewed by discussing mainly its design premises, constitution, manufacturing methods, corrosion resistance in a deep geologic environment, radiation shielding, and structural stability. The disposal canister should be safe from thermal, chemical, mechanical, and biological invasions for a very long time so as not to release any kind of radionuclides.

  19. Enhanced Earthquake-Resistance on the High Level Radioactive Waste Canister

    International Nuclear Information System (INIS)

    In this paper, the earthquake-resistance type buffer was developed with the method protecting safely about the earthquake. The main parameter having an effect on the earthquake-resistant performance was analyzed and the earthquake-proof type buffer material was designed. The shear analysis model was developed and the performance of the earthquake-resistance buffer material was evaluated. The dynamic behavior of the radioactive waste disposal canister was analyzed in case the earthquake was generated. In the case, the disposal canister gets the serious damage. In this paper, the earthquake-resistance buffer material was developed in order to prevent this damage. By putting the buffer in which the density is small between the canister and buffer, the earthquake-resistant performance was improved about 80%

  20. STS-45 ATLAS-1 pallets and SSBUV canisters in OV-104's payload bay (PLB)

    Science.gov (United States)

    1992-01-01

    STS-45 payload bay (PLB) configuration onboard Atlantis, Orbiter Vehicle (OV) 104, includes the Shuttle Solar Backscatter Ultraviolet 4 (SSBUV-4) and Atmospheric Laboratory for Applications and Science 1 (ATLAS-1) instruments. The SSBUV get away special (GAS) canisters are mounted on a GAS adapter beam on the starboard PLB sill longeron. THE SSBUV support canister is in the foreground and the SSBUV instrument canister with motorized door assembly (MDA) is next to it. ATLAS-1 equipment includes the igloo (center - decorated with several insignias), the Space Experiments with Particle Accelerators (SEPAC) spheres, and additional instruments mounted on unpressurized spacelab pallets. In the background, are the orbital maneuvering system (OMS) pods and vertical tail highlighted against the cloud-covered surface of the Earth.

  1. Mechanical analysis of cylindrical part of canisters for spent nuclear fuel

    International Nuclear Information System (INIS)

    This report describes mechanical analyses of cylindrical part of the VVER 440-, BWR and EPR-type canisters for spent nuclear fuel. The task was first to evaluate the stresses at maximum design pressure and further by increasing pressure load to determine the limit collapse load and corresponding safety factor. Maximum design pressure 44 MPa is a sum of the hydrostatic pressure 30 MPa caused by 3 km ice layer, 7 MPa caused by ground water pressure at the deepest disposal depth of 700 m and 7 MPa from bentonite swelling pressure. The analysis presented in this report concern the middle area of the canisters, where the cast iron insert is considered to be more critical than in the ends of the canister. For the model a piece from the middle area of the canister was separated by two planes perpendicular to the axis of the canister. This piece was studied first by two-dimensional plane strain model, where the planes are constrained and no elongation of the canister takes place. In the second model one of the planes was constrained and the other plane was allowed to displace in axial direction, which remains as a plane during deformation and to which axial pressure force is directed. This analysis, which corresponds better the real condition in the canister, was performed as threedimensional. The analyses gave however practically equal results due to plastic deformation. Thus the analysis can be done by two-dimensional plane strain model leading to same accuracy with less computation effort. Analyses were performed as large displacement and large strain analyses by the PASULA computing package, which has been developed at VTT for a variety of structural analysis and for heat conduction calculations. A special routine was developed for automatic mesh generation. Before the analysis of the VVER 440-, BWR- and EPR-type canisters the calculation methodology was validated with test results, which were received from pressure tests performed with a short BWR canister in Germany

  2. 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.

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

    International Nuclear Information System (INIS)

    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 105 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 105 years

  4. Design basis for the copper/steel canister. Stage three. Final report

    International Nuclear Information System (INIS)

    The development of the copper/iron canister proposed for the containment of high-level waste in the Swedish disposal programme has been studied from the points of view of choice of materials, manufacturing technology and Q A. This report describes the observations on progress which has been made between March 1995 and February 1996 and the results of further literature studies. A first trial canister has been produced by SKB using a fabricated steel liner and an extruded copper tubular, a second one using a fabricated tubular is at an advanced stage. A change from a fabricated steel inner canister to a proposed cast canister has been justified by a criticality argument but the technology for producing a cast canister is at present untried. It is considered that such a change will require a significant development programme. The microstructure achieved in the extruded copper tubular for the first canister is unacceptable. An improved microstructure may be achieved by extruding at a lower temperature but this remains to be demonstrated. Similar problems exist with plate used for the fabricated tubular but some more favourable structures have been achieved already by this route. Seam welding of the first tubular failed through a suspected material problem. The second fabricated tubular welded without difficulty. However it was necessary to constrain it during welding and it subsequently distorted during machining. There was some evidence of hot tearing close to the weld. The distortion problem may be overcome by a stress relieving anneal but this could cause further grain size problems. 19 refs

  5. Analyses of atmospheric radon 222 / canisters exposed by Greenpeace in Niger (Arlit / Akokan sector)

    International Nuclear Information System (INIS)

    The companies SOMAIR and COMINAK, subsidiaries of the AREVA group, are mining uranium deposits in northern Niger. In the course of a field mission carried out in November 2009, a Greenpeace International team deposited detectors (canisters of activated charcoal) to measure radon 222, a radioactive gas formed by the decay of the radium 226 present in the uranium ore. This report includes the results of the analysis of the activated charcoal canisters conducted in CRIIRAD's laboratory, and a brief commentary on the interpretation of the results. (authors)

  6. 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

  7. 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.

  8. 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.

  9. STS-105 ICC is moved to the payload canister for transport to pad 39A

    Science.gov (United States)

    2001-01-01

    KENNEDY SPACE CENTER, Fla. -- An overhead crane in the Space Station Processing Facility lifts the Integrated Cargo Carrier from its workstand to move it to the payload canister. The ICC holds several payloads for mission STS-105, the Early Ammonia Servicer and two experiment containers. The ICC will join the Multi-Purpose Logistics Module Leonardo in the payload canister for transport to Launch Pad 39A where they will be placed in the payload bay of Space Shuttle Discovery. Launch of STS-105 is scheduled for 5:38 p.m. EDT Aug. 9

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

    International Nuclear Information System (INIS)

    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

  11. Integrity of copper/steel canisters under crystalline bedrock repository conditions

    International Nuclear Information System (INIS)

    In the Swedish nuclear waste disposal programme, the need to store the spent nuclear fuel safely for very long times has prompted a strategy which includes a long life canister. Technical as well as economical considerations related to design, choice of materials and manufacturing technology have lead to the selection of a reference design to be used for the continued development work. The canisters are cylindrical with a diameter close to 1 meter and a height of about 5 meters. In order to meet the need for an appropriate combination of mechanical strength, toughness, durability and corrosion resistance, the canisters comprise an inner vessel made of steel or cast iron to cope with mechanical stresses and an outer vessel made of almost pure copper to provide corrosion resistance. The Swedish nuclear industry has recently extended its development work to full-scale tests. Such experience is needed not least for the evaluation of the long-term integrity of the canister. This work has been closely followed by the Swedish Nuclear Power Inspectorate (SKI) who have also carried out independent investigations and analyses. It should be emphasized that the findings relate to a canister which is under development and cannot, in general, be expected to be relevant for the fully developed canister. Significant results of the analyses include the identification of conceivable modes of canister failures. Such failures may be related to defects, segregation, limitations in inspectability, long term creep properties, adverse mechanical load situations, etc. It is assessed that the distribution functions of these failures might have their largest uncertainties at the tails extending to comparatively short times. Specific issues related to canister manufacture, scaling and non destructive testing which have been found to warrant further investigation are: defects in the copper ingot which may transfer to the rolled copper plate; the amount of work applied during the rolling or

  12. 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

  13. Cost analysis for application of solidified waste fission product canisters in U.S. Army steam plants

    International Nuclear Information System (INIS)

    The main objectives of the present study are to design steam plants using projected waste fission product canister characteristics, to analyze the overall impact and cost/benefit to the nuclear fuel cycle associated with these plants, and to develop plans for this application if the cost analysis so warrants it. The construction and operation of a steam plant fueled with waste fission product canisters would require the involvement and cooperation of various government agencies and private industry; thus the philosophies of these groups were studied. These philosophies are discussed, followed by a forecast of canister supply, canister characteristics, and strategies for Army canister use. Another section describes the safety and licensing of these steam plants since this affects design and capital costs. The discussion of steam plant design includes boiler concepts, boiler heat transfer, canister temperature distributions, steam plant size, and steam plant operation. Also, canister transportation is discussed since this influences operating costs. Details of economics of Army steam plants are provided including steam plant capital costs, operating costs, fuel reprocessor savings due to Army canister storage, and overall economics. Recommendations are made in the final section

  14. An Assessment of Using Vibrational Compaction of Calcined HLW and LLW in DWPF Canisters

    International Nuclear Information System (INIS)

    Since 1963, the INEL has calcined almost 8 million gallons of liquid mixed waste and liquid high-level waste, converting it to some 1.1 million gallons of dry calcine (about 4275.0 m3), which consists of alumina-and zirconia-based calcine and zirconia-sodium blend calcine. In addition, if all existing and projected future liquid wastes are solidified, approximately 2,000 m3 of additional calcine will be produced primarily from sodium-bearing waste. Calcine is a more desirable material to store than liquid radioactive waste because it reduces volume, is much less corrosive, less chemically reactive, less mobile under most conditions, easier to monitor and more protective of human health and the environment. This paper describes the technical issue involved in the development of a feasible solution for further volume reduction of calcined nuclear waste for transportation and long term storage, using a standard DWPF canister. This will be accomplished by developing a process wherein the canisters are transported into a vibrational machine, for further volume reduction by about 35%. The random compaction experiments show that this volume reduction is achievable. The main goal of this paper is to demonstrate through computer modeling that it is feasible to use volume reduction vibrational machine without developing stress/strain forces that will weaken the canister integrity. Specifically, the paper presents preliminary results of the stress/strain analysis of the DWPF canister as a function of granular calcined height during the compaction and verifying that the integrity of the canister is not compromised. This preliminary study will lead to the development of better technology for safe compactions of nuclear waste that will have significant economical impact on nuclear waste storage and treatment. The preliminary results will guide us to find better solutions to the following questions: 1) What are the optimum locations and directions (vertical versus horizontal or

  15. 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.

  16. Development of single tubing-type canister for cryo-storage of bull semen and their effect on sperm motility and viability

    Directory of Open Access Journals (Sweden)

    Mohd Iswadi Ismail

    2014-04-01

    Full Text Available The objective of this study was to evaluate the potential of using single tubing-type canister on sperm quality. Semen was collected from the Bali cattle bull by electroejaculation technique and was cryopreserved in liquid nitrogen using slow freezing cryopreservation method. Two type of canister volume was used in this study; commercial canister (342.25π x 278 mm² and single tubing-type canister (4π x 90 mm². Makler counting chamber and computer assisted sperm analyzer (CASA were used to evaluate the sperm motility and viability of post-thaw sperm. Results showed that the bull sperm motility and viability at the bottom of tubing-type canister was statistically higher and significant as compared to the commercial canister (p<0.05. Significant changes were found in sperm kinetics (VCL, VAP, VSL of tubing-type canister compared to commercial canister. No significant changes in the motility and viability of the bull sperm at the top of tubing-type canister and commercial canister. There were no significant changes in sperm progression (LIN, WOB, PROG in both the canisters. Developed tubing-type canister in this study showed potential as an alternative to be used in bull sperm cryo-storage.

  17. 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.

  18. Spent Nuclear Fuel (SNF) Project Multi Canister Overpack (MCO) Process Flow Diagram Mass Balance Calculations

    Energy Technology Data Exchange (ETDEWEB)

    KLEM, M.J.

    2000-09-08

    The purpose of this calculation document is to develop the bases for the material balances of the Multi-Canister Overpack (MCO) Level 1 Process Flow Diagram (PFD). The attached mass balances support revision two of the PFD for the MCO and provide future reference.

  19. Corrosion issues in relation to copper canisters for disposal of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Saario, T. (VTT Technical Research Centre of Finland, Espoo (Finland))

    2010-05-15

    Spent nuclear fuel in Finland and Sweden will be disposed in deposition holes excavated in granitic bedrock at a depth of about 400 to 500 m. The release of activated products is prevented by a multi-barrier concept. A copper canister with 50 mm wall thickness will be the main corrosion shield, expected to last un-perforated for over 100 000 years. There are several different corrosion mechanisms that in principle may threaten the integrity of the copper canister. Likewise, there are several different scenarios which may expose the copper canister to these corrosion phenomena. During the years, most of the corrosion related concerns have been satisfactorily dealt with, leaving a few to be further researched. This paper gives an overview of the corrosion related issues with regard to the integrity of the copper canister, as well as the main approaches developed by the share holders to ensure the integrity for the duration needed. A more detailed description of the few remaining ongoing corrosion related research items is also included. (orig.)

  20. Stress analysis of glass-canister interaction: a study of residual stresses and fracturing

    International Nuclear Information System (INIS)

    Residual stresses and cracking in canisters filled with vitrified nuclear waste are simulated using finite element computer calculations. Cooling rates, internal heat generation, and thermal expansion coefficients significantly affect stress levels. Glass behavior within the softening temperature range is taken to follow the instant freezing concept of Bartenev

  1. 40 CFR 86.153-98 - Vehicle and canister preconditioning; refueling test.

    Science.gov (United States)

    2010-07-01

    ... controlled to 50±25 grains of water vapor per pound of dry air) maintained at a nominal flow rate of 0.8 cfm... at least 1200 canister bed volumes of ambient air (with humidity controlled to 50±25 grains of water...'s air conditioner (if so equipped) shall be turned off. Ambient temperature shall be controlled...

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

    International Nuclear Information System (INIS)

    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/

  3. 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.

  4. 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.

  5. Gap Analysis to Support Modeling the Long-Term Degradation of Used Nuclear Fuel Canisters

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, Philip J.; Sunderland, Dion J.; Ross, Steven B.; Montgomery, Robert O.; Hanson, Brady D.; Devanathan, Ram

    2015-04-01

    Welded stainless steel canisters are being used worldwide for dry storage of used nuclear fuel (UNF) assemblies, and the number of canisters in use is steadily increasing. In support of work currently being pursued at Pacific Northwest National Laboratory to understand the atmospheric corrosion behavior of spent fuel dry storage systems, a gap analysis is underway to assess the state of knowledge for modeling of the long-term degradation of a UNF canister. The fundamental aim of this work is to inform research and development (R&D) efforts to establish a sound technical basis to support the extended dry storage of UNF for 100+ years. The analysis is considering all major components of the atmosphere corrosion degradation processes, ranging from contaminant sources and climatic interactions to regional conditions of particle transport and deposition, to microscale effects leading to stress corrosion cracking. The results of this gap analysis will be used to define the R&D pathway to develop an integrated multi-scale atmospheric corrosion modeling capability for UNF in dry storage canisters that can support the safe and reliable performance of these structures for more than 100 years.

  6. Reliability in NDT of canister for the Swedish spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ronneteg, Ulf [SKB, Oskarshamn (Sweden); Mueller, Christina; Pavlovic, Mato [BAM, Berlin (Germany)

    2009-07-01

    The Swedish KBS-3 design for the disposal of spent fuel is based on encapsulation of the fuel in canisters consisting of cast iron inserts and an outer 5 cm thick shield of copper. The canisters are embedded in bentonite clay and will be disposed in crystalline bedrock at a depth of about 500 m. To verify that the canisters fulfil the requirements, an extensive program for quality control is developed. In this program the use of NDT is vital and therefore it is very important to determine the reliability of the developed NDT methods. The reliability of the NDT methods has been studied for six years and the results are used in several ways. One task is to identify areas where further optimizations of the NDT methods are necessary and another is to study the effect of influencing parameters prior to the upcoming qualifications. Furthermore, the reliability analyses of the NDT are combined with analyses of the manufacturing and welding processes to give input to the overall safety analyses of the long-term properties of the canisters. (orig.)

  7. Interim Storage of RH-TRU 72B Canisters at the DOE Oak Ridge Reservation

    International Nuclear Information System (INIS)

    This paper describes an evaluation performed by the Department of Energy (DOE) Oak Ridge Operations (ORO) office for potential interim storage of remote-handled (RH) transuranic (TRU) 72B waste canisters at the Oak Ridge National Laboratory (ORNL). The evaluation included the conceptual design of a devoted canister storage facility and an assessment of the existing RHTRU waste storage facilities for storage of canisters. The concept for the devoted facility used modular concrete silos located on an above-grade storage pad. The assessment of the existing facilities considered the potential methods, facility modifications, and conceptual equipment that might be used for storage of 400 millisievert per hour (mSv/hr) canisters. The results of the evaluation indicated that the initial investment into a devoted facility was relatively high as compared to the certainty that significant storage capacity was necessary prior to the Waste Isolation Pilot Plant (WIPP) accepting RH-TRU waste for disposal. As an alternative, the use of individual concrete overpacks provided an incremental method that could be used with the existing storage facilities and outside storage pads. For the concrete overpack concepts considered, the cylindrical design stored in a vertical orientation was determined to be the most effective

  8. Plutonium Immobilization Project - Can-In-Canister Hardware Development/Selection

    International Nuclear Information System (INIS)

    The Plutonium Immobilization Project (PIP) is a program funded by the U.S. Department of Energy to develop technology to disposition excess weapons grade plutonium. This program introduces the ''Can-in-Canister'' (CIC) technology that immobilizes the plutonium by encapsulating it in ceramic forms (or pucks) and ultimately surrounding it with high-level waste glass to provide a deterrent to recovery. Since there are significant radiation, contamination and security concerns, the project team is developing unique technologies to remotely perform plutonium immobilization tasks. This paper covers the design, development and testing of the magazines (cylinders containing cans of ceramic pucks) and the rack that holds them in place inside the waste glass canister. Several magazine and rack concepts were evaluated to produce a design that gives the optimal balance between resistance to thermal degradation and facilitation of remote handling. This paper also reviews the effort to develop a join ted arm robot that can remotely load seven magazines into defined locations inside a stationary canister working only through the 4 inch (102 mm) diameter canister throat

  9. Plutonium Immobilization Project - Can-In-Canister Hardware Development/Selection

    International Nuclear Information System (INIS)

    The Plutonium Immobilization Project (PIP) is a program funded by the U.S. Department of Energy to develop technology to disposition excess weapons grade plutonium. This program introduces the ''Can-in-Canister'' (CIC) technology that immobilizes the plutonium by encapsulating it in ceramic forms (or pucks) and ultimately surrounding it with high-level waste glass to provide a deterrent to recovery. Since there are significant radiation, contamination and security concerns, the project team is developing unique technologies to remotely perform plutonium immobilization tasks. This paper covers the design, development and testing of the magazines (cylinders containing cans of ceramic pucks) and the rack that holds them in place inside the waste glass canister. Several magazine and rack concepts were evaluated to produce a design that gives the optimal balance between resistance to thermal degradation and facilitation of remote handling. This paper also reviews the effort to develop a jointed arm robot that can remotely load seven magazines into defined locations inside a stationary canister working only through the 4 inch (102 mm) diameter canister throat

  10. 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/

  11. Development of fabrication technology for copper canisters with cast inserts. Status report in August 2001

    International Nuclear Information System (INIS)

    This report contains an account of the results of trial fabrication of copper canisters with cast inserts carried out during the period 1998 - 2001. The work of testing of fabrication methods is being focused on a copper thickness of 50 mm. Occasional canisters with 30 mm copper thickness are being fabricated for the purpose of gaining experience and evaluating fabrication and inspection methods for such canisters. For the fabrication of copper tubes, SKB has concentrated its efforts on seamless tubes made by extrusion and pierce and draw processing. Five tubes have been extruded and two have been pierced and drawn during the period. Materials testing has shown that the resultant structure and mechanical properties of these tubes are good. Despite certain problems with dimensional accuracy, it can be concluded that both of these methods can be developed for use in the serial production of SKB' copper tubes. No new trial fabrication with roll forming of copper plate and longitudinal welding has been done. This method is nevertheless regarded as a potential alternative. Copper lids and bottoms are made by forging of continuous-cast bars. The forged blanks are machined to the desired dimensions. Due to the Canister Laboratory's need for lids to develop the technique for sealing welding, a relatively large number of forged blanks have been fabricated. It is noted in the report that the grain size obtained in lids and bottoms is much coarser than in fabricated copper tubes. Development work has been commenced for the purpose of optimizing the forging process. Nine cast inserts have been cast during the three-year period. The results of completed material testing of test pieces taken at different places along the length of the inserts have in several cases shown an unacceptable range of variation in strength properties and structure. In the continued work, insert fabrication will be developed in terms of both casting technique and iron composition. Development work on

  12. 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

  13. 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

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

    International Nuclear Information System (INIS)

    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 obtained in

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

    International Nuclear Information System (INIS)

    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 obtained in

  16. Thermal analysis of dry concrete canister storage system for CANDU spent fuel

    International Nuclear Information System (INIS)

    This paper presents the results of a thermal analysis of the concrete canisters for interim dry storage of spent, irradiated Canadian Deuterium Uranium(CANDU) fuel. The canisters are designed to contain 6-year-old fuel safely for periods of 50 years in stainless steel baskets sealed inside a steel-lined concrete shield. In order to assure fuel integrity during the storage, fuel rod temperature shall not exceed the temperature limit. The contents of thermal analysis include the following : 1) Steady state temperature distributions under the conservative ambient temperature and insolation load. 2) Transient temperature distributions under the changes in ambient temperature and insolation load. Accounting for the coupled heat transfer modes of conduction, convection, and radiation, the computer code HEATING5 was used to predict the thermal response of the canister storage system. As HEATING5 does not have the modeling capability to compute radiation heat transfer on a rod-to-rod basis, a separate calculating routine was developed and applied to predict temperature distribution in a fuel bundle. Thermal behavior of the canister is characterized by the large thermal mass of the concrete and radiative heat transfer within the basket. The calculated results for the worst case (steady state with maximum ambient temperature and design insolation load) indicated that the maximum temperature of the 6 year cooled fuel reached to 182.4 .deg. C, slightly above the temperature limit of 180 .deg. C. However,the thermal inertia of the thick concrete wall moderates the internal changes and prevents a rise in fuel temperature in response to ambient changes. The maximum extent of the transient zone was less than 75% of the concrete wall thickness for cyclic insolation changes. When transient nature of ambient temperature and insolation load are considered, the fuel temperature will be a function of the long term ambient temperature as opposed to daily extremes. The worst design

  17. 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.

  18. 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

  19. Nonlinear dynamic impact analysis for installing a dry storage canister into a vertical concrete cask

    International Nuclear Information System (INIS)

    In this paper, a series of dynamic impact analysis for installing a dry storage canister into a vertical concrete cask (VCC) is performed. The dry storage system considered herein is called HCDSS-69, recently developed by INER and being capable of accommodating 69 bundles of BWR spent nuclear fuels. The impact accident is stemming from a conservative consideration of accidental movement when the canister is being hoisted into a VCC. According to NUREG-0554, the accidental movement is conservatively simulated by 80 mm- and 160 mm-height free-drop motions and then with straight and 2°-oblique impact to a pedestal in VCC. A symmetric fully 3-D finite element model is built and analyzed using the explicit finite element code, LS-DYNA. Geometrical, contact, and material nonlinearities are all taken into account. The analysis result concludes that the permanent deformations of the canister are not severe to affect fuel retrieve after the impact accident and the maximum stress intensity in the canister shell can meet the ASME code appendix F F-1340, preventing the leakage of radioactive materials. The study also found that with properly reducing the wall thickness of the pedestal cylinder, the maximum acceleration and permanent deformation of the canister can be much alleviated, even though the drop height is increased to the double of the required brake distance specified in NUREG-0554. The damages of the pedestal in each analysis are moderate so that the heat transfer condition after the impact accident can be bounded by the off-normal event for half-blockage of air inlets

  20. 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.

  1. 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

  2. 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.

  3. 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

  4. The gas-cooled Li2O moderator/breeder canister blanket for fusion-synfuels

    International Nuclear Information System (INIS)

    A new integrated power and breeding blanket is described. The blanket incorporates features that make it suitable for synthetic fuel production. It is matched to the thermal and electrical requirements of the General Atomic water-splitting process for producing hydrogen. The fusion reaction is the Tandem Mirror Reactor (TMR) using Mirror Advanced Reactor Study (MARS) physics. The canister blanket is a high temperature, pressure balanced, crossflow heat exchanger contained within a low activity, independently cooled, moderate temperature, first wall structural envelope. The canister uses Li2O as the moderator/breeder and helium as the coolant. ''In situ'' tritium control, combined with slip stream processing and self-healing permeation barriers, assures a hydrogen product essentially free of tritium. The blanket is particularly adapted to synfuels production but is equally useful for electricity production or co-generation

  5. 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...

  6. Automated waste canister docking and emplacement using a sensor-based intelligent controller

    International Nuclear Information System (INIS)

    A sensor-based intelligent control system is described that utilizes a multiple degree-of-freedom robotic system for the automated remote manipulation and precision docking of large payloads such as waste canisters. Computer vision and ultrasonic proximity sensing are used to control the automated precision docking of a large object with a passive target cavity. Real-time sensor processing and model-based analysis are used to control payload position to a precision of ± 0.5 millimeter

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

    International Nuclear Information System (INIS)

    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

  8. 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

  9. 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.

  10. Acceptance of canisters of consolidated spent nuclear fuel by the Federal Waste Management System

    International Nuclear Information System (INIS)

    This report is one of a series of eight prepared by E. R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. The eight reports are concerned with the conditions under which spent fuel and high level waste will be accepted in the following categories: failed fuel; consolidated fuel and associated structural parts; non-fuel-assembly hardware; fuel in metal storage casks; fuel in multi-element sealed canisters; inspection and testing requirements for wastes; canister criteria; spent fuel selection for deliver; and defense and commercial high-level waste packages. This document discusses canister standards and criteria. 12 refs., 7 figs., 28 tabs

  11. Localization of plastic deformation in copper canisters for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Savolainen, K.; Saukkonen, T.; Haenninen, H. (Aalto Univ. School of Science and Technology, Espoo (Finland))

    2010-05-15

    In Finland and Sweden the spent nuclear fuel will be deposited in a deep repository in copper corrosion barrier canisters surrounding cast iron inserts. The 50 mm thick copper canisters will we sealed using either electron beam welding (EB) or friction stir welding (FSW) to join the tubes and the lids/bottoms. The canisters will deform in the repository conditions, e.g., due to the hydrostatic pressure. The deformation will localize to different discontinuities, such as defects and microstructural heterogeneity. This study compared the localization of plastic deformation in EB and FSW welds as well as in the base materials (both forged and extruded) using optical strain measurement methods. The results show that in the base materials the deformation occurs very uniformly over the entire gauge length. In FSW welds the deformation localizes in the weld at a line of entrapped oxide particles. In EB welds the deformation localizes to the large grains in the middle of the weld or at the steep grain size gradient between the weld and the tube material. Tensile strength is lowest in the EB welds (175 MPa as compared to 200 MPa or higher for the other samples). Elongation to fracture of the FSW welds is similar to those of the base materials, but for the EB welds it is significantly lower (appr. 40 % compared to 65 %). (orig.)

  12. Inner material requirements and candidates screening for spent fuel disposal canister

    International Nuclear Information System (INIS)

    In the context of the present Spanish 'once-through' nuclear fuel cycle, the need arises to complete the geological repository reference concept with a spent fuel canister final design. One of the main issues in its design is selecting the inner material to be placed inside the canister, between the steel walls and the spent fuel assemblies. The primary purpose of this material will be to avoid the possibility of a criticality event once the canister walls have been finally breached by corrosion and the spent fuel is flooded with groundwater. That is an important role because the increase in heat generation from such an event would act against spent fuel stability and compromise bentonite barrier functions, negatively affecting overall repository performance. To prevent this possibility a detailed set of requirements for a material to fulfil this role in the repository environment have been devised and presented in this paper. With these requirements in view, eight potentially interesting candidates were selected and evaluated: cast iron or steel, borosilicate glass, spinel, depleted uranium, dehydrated zeolites, haematite, phosphates, and olivine. Among these, the first four materials or material families are found promising for this application. In addition, other relevant non-performance-related aspects of candidate materials, which could help on decision making, are also considered and evaluated. (authors)

  13. Thermal analysis of heat storage canisters for a solar dynamic, space power system

    Science.gov (United States)

    Wichner, R. P.; Solomon, A. D.; Drake, J. B.; Williams, P. T.

    1988-01-01

    A thermal analysis was performed of a thermal energy storage canister of a type suggested for use in a solar receiver for an orbiting Brayton cycle power system. Energy storage for the eclipse portion of the cycle is provided by the latent heat of a eutectic mixture of LiF and CaF2 contained in the canister. The chief motivation for the study is the prediction of vapor void effects on temperature profiles and the identification of possible differences between ground test data and projected behavior in microgravity. The first phase of this study is based on a two-dimensional, cylindrical coordinates model using an interim procedure for describing void behavor in 1-g and microgravity. The thermal analysis includes the effects of solidification front behavior, conduction in liquid/solid salt and canister materials, void growth and shrinkage, radiant heat transfer across the void, and convection in the melt due to Marangoni-induced flow and, in 1-g, flow due to density gradients. A number of significant differences between 1-g and o-g behavior were found. This resulted from differences in void location relative to the maximum heat flux and a significantly smaller effective conductance in 0-g due to the absence of gravity-induced convection.

  14. 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.

  15. Conceptual design study of a concrete canister spent-fuel storage facility

    International Nuclear Information System (INIS)

    This report presents a conceptual design study for the interim storage of CANDU spent fuel in concrete canisters. The canisters will be concrete flasks, which contain fuel prepackaged in double steel containment, and will be cooled by natural air convection. This is one of the methods proposed as a potential alternative to water pool storage. A preliminary study of this concept was done by CAFS (Committee Assessing Fuel Storage), and WNRE (Whiteshell Nuclear Research Establishment) is currently conducting a development and demonstration program. This study of a central facility for the storage of all Canadian spent fuel arisings to the year 2000 was completed in 1975. A brief description of the facilities required and the operations involved, a summary of costs, a survey of the monitoring requirements and a prediction of the personnel exposures associated with this method of storing spent fuel are reported here. The estimated total cost of interim storage in cylindrical canisters at a central site is $6.02/kg U (1975 dollars). Approximately half of this cost is incurred in the shipment of fuel from the reactors to the storage facility. (author)

  16. Stress corrosion cracking of stainless-steel canister for concrete cask storage of spent fuel

    Science.gov (United States)

    Tani, Jun-ichi; Mayuzumi, Masami; Hara, Nobuyoshi

    2008-09-01

    Resistance to external stress corrosion cracking (ESCC) and crevice corrosion were examined for various candidate canister materials in the spent fuel dry storage condition using concrete casks. A constant load ESCC test was conducted on the candidate materials in air after deposition of simulated sea salt particles on the specimen gage section. Highly corrosion resistant stainless steels (SS), S31260 and S31254, did not fail for more than 46 000 h at 353 K with relative humidity of 35%, although the normal stainless steel, S30403 SS failed within 500 h by ESCC. Crevice corrosion potentials of S31260 and S31254 SS became larger than 0.9 V (SCE) in synthetic sea water at temperatures below 298 K, while those of S30403 and S31603 SS were less than 0 V (SCE) at the same temperature range. No rust was found on S31260 and S31254 SS specimens at temperatures below 298 K in the atmospheric corrosion test, which is consistent with the temperature dependency of crevice corrosion potential. From the test result, the critical temperature of atmospheric corrosion was estimated to be 293 K for both S31260 and S31254 SS. Utilizing the ESCC test result and the critical temperature, together with the weather station data and the estimated canister wall temperature, the integrity of canister was assessed from the view point of ESCC.

  17. Analysis of burns caused by pre-filled gas canisters used for lamps or portable camping stoves.

    Science.gov (United States)

    Desouches, C; Salazard, B; Romain, F; Karra, C; Lavie, A; Volpe, C Della; Manelli, J C; Magalon, G

    2006-12-01

    The use of pre-filled valveless gas canisters for lamps or camping stoves has caused a number of serious burn incidents. We performed a retrospective analysis of all of the patients who were victims of such incidents admitted to the Marseille Burn Centre between January 1990 and March 2004. There were a total of 21 patients burned in such conditions. Adult males made up the majority of the victims of this sort. Lesions were often extensive (60% of the patients were burned over more than 10% of their body surface) and systematically deep. In order of frequency, burn locations were: the lower limbs, the upper limbs, the hands and the face. The incidents principally occurred during replacement of the canister near an open flame. The marketing of a canister with a valve in order to avoid gas leaks did not cause the old canisters to be taken off the market. On the contrary, European Safety Standard EN417, updated in October 2003, validated the use of these valveless canisters. The severity of the lesions caused and the existence of safe equivalent products requires the passage of a law that forbids valveless canisters. PMID:16982156

  18. 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.

  19. Challenge to Overcome the Concern of SCC in Canister During Long-Term Storage of Spent Fuel

    International Nuclear Information System (INIS)

    In order to put the concrete cask in practical use in Japan (an island country), stress corrosion cracking (SCC) of canister must be coped with. It is required to take measures for one or two of the three factors, i.e. welding residual stress, material, and environment, to cope with the SCC that may result in loss of the containment function of the canister. Prevention of loss of containment due to SCC of a canister was evaluated either by a method of comparing the amount of salt on the canister surface during storage with the minimum amount of salt to initiate rust and SCC or by a method of comparing the wetting time of the canister surface under salty-air field environment with the lifetime of the SCC fracture of the canister material. Although the use of highly corrosion-resistance stainless steel is one solution, it brings about a cost rise of the concrete cask storage. In order to suppress the cost rise, it should be evaluated whether the measure against SCC of the normal stainless steel is possible by reducing welding residual stress. In addition, technology should be developed to reduce salt particles in the air flowing into the storage facility and concrete cask. (author)

  20. 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.

  1. Performance Assessment and Sensitivity Analyses of Disposal of Plutonium as Can-in-Canister Ceramic

    Energy Technology Data Exchange (ETDEWEB)

    Rainer Senger

    2001-09-25

    The purpose of this analysis is to examine whether there is a justification for using high-level waste (HLW) as a surrogate for plutonium disposal in can-in-canister ceramic in the total-system performance assessment (TSPA) model for the Site Recommendation (SR). In the TSPA-SR model, the immobilized plutonium waste form is not explicitly represented, but is implicitly represented as an equal number of canisters of HLW. There are about 50 metric tons of plutonium in the U. S. Department of Energy inventory of surplus fissile material that could be disposed. Approximately 17 tons of this material contain significant quantities of impurities and are considered unsuitable for mixed-oxide (MOX) reactor fuel. This material has been designated for direct disposal by immobilization in a ceramic waste form and encapsulating this waste form in high-level waste (HLW). The remaining plutonium is suitable for incorporation into MOX fuel assemblies for commercial reactors (Shaw 1999, Section 2). In this analysis, two cases of immobilized plutonium disposal are analyzed, the 17-ton case and the 13-ton case (Shaw et al. 2001, Section 2.2). The MOX spent-fuel disposal is not analyzed in this report. In the TSPA-VA (CRWMS M&O 1998a, Appendix B, Section B-4), the calculated dose release from immobilized plutonium waste form (can-in-canister ceramic) did not exceed that from an equivalent amount of HLW glass. This indicates that the HLW could be used as a surrogate for the plutonium can-in-canister ceramic. Representation of can-in-canister ceramic as a surrogate is necessary to reduce the number of waste forms in the TSPA model. This reduction reduces the complexity and running time of the TSPA model and makes the analyses tractable. This document was developed under a Technical Work Plan (CRWMS M&O 2000a), and is compliant with that plan. The application of the Quality Assurance (QA) program to the development of that plan (CRWMS M&O 2000a) and of this Analysis is described in

  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. Development of single tubing-type canister for cryo-storage of bull semen and their effect on sperm motility and viability

    OpenAIRE

    Mohd Iswadi Ismail; Khairul Osman; Siti Fatimah Ibrahim; Farah Hanan Fatihah Jaafar; Nur Azianie Abd Ghani; Fazly Ann Zainalabidin; Abas Mazni Othman

    2014-01-01

    The objective of this study was to evaluate the potential of using single tubing-type canister on sperm quality. Semen was collected from the Bali cattle bull by electroejaculation technique and was cryopreserved in liquid nitrogen using slow freezing cryopreservation method. Two type of canister volume was used in this study; commercial canister (342.25π x 278 mm²) and single tubing-type canister (4π x 90 mm²). Makler counting chamber and computer assisted sperm analyzer (CASA) were used to ...

  4. Analysis of factors influencing the reliability of retrievable storage canisters for containment of solid high-level radioactive waste

    International Nuclear Information System (INIS)

    The reliability of stainless steel type 304L canisters for the containment of solidified high-level radioactive wastes in the glass and calcine forms was studied. A reference system, drawn largely from information furnished by Battelle Northwest Laboratories and Atlantic Richfield Hanford Company is described. Operations include filling the canister with the appropriate waste form, interim storage at a reprocessing plant, shipment in water to a Retrievable Surface Storage Facility (RSSF), interim storage at the RSSF, and shipment to a final disposal facility. The properties of stainless steel type 304L, fission product oxides, calcine, and glass were reviewed, and mechanisms of corrosion were identified and studied. The modes of corrosion important for reliability were stress-corrosion cracking, internal pressurization of the canister by residual impurities present, intergranular attack at the waste-canister interface, and potential local effects due to migration of fission products. The key role of temperature control throughout canister lifetime is considered together with interactive effects. Methods of ameliorating adverse effects and ensuring high reliability are identified and described. Conclusions and recommendations are presented

  5. Design basis for the copper/steel canister. Stage four. Final report

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-06-01

    The development of the copper/iron canister which has been proposed by SKB for the containment of high level nuclear waste has been studied from the points of view of choice of materials, manufacturing technology and quality assurance. Cast steel has been rejected in favour of cast iron as a candidate material for the load bearing liner. Nodular (or ductile) iron is selected and this is capable of providing mechanical properties which are equally suitable as those of the originally selected high strength low alloy steel. The material specified for the overpack is Oxygen free copper with 50 ppm of phosphorus added. Corrosion studies supported by SKB indicate that in the absence of mechanical failure or accelerated localised corrosion the overpack should provide corrosion shielding of the canister for its full design life. Published work claiming that the nodular iron liner would have corrosion characteristics similar to the carbon steel which had been examined in depth is flawed since the microstructures of the iron and carbon steel specimens used were not investigated. It is highly unlikely that nodular irons in the form used for the experiments would have similar structures to nodular iron in the canisters by chance. If the overpack were breached during the aerobic period of the repository life then very rapid penetration of the inner liner could occur. It has been recognised that the roll forming method is not suitable for serial production and alternatives are being sought. The electron beam welding process has been explored with tenacity but has so far failed to produce a satisfactory lid weld. A new welder is being developed for supply to the SKB pilot plant where development will be continued. An alternative welding process, friction stir welding, is being examined as a candidate for attaching lids. Surface breaking defects may be detected using eddy current methods but there is currently no reliable way of detecting small sub surface defects in the overpack

  6. Design basis for the copper/steel canister. Stage four. Final report

    International Nuclear Information System (INIS)

    The development of the copper/iron canister which has been proposed by SKB for the containment of high level nuclear waste has been studied from the points of view of choice of materials, manufacturing technology and quality assurance. Cast steel has been rejected in favour of cast iron as a candidate material for the load bearing liner. Nodular (or ductile) iron is selected and this is capable of providing mechanical properties which are equally suitable as those of the originally selected high strength low alloy steel. The material specified for the overpack is Oxygen free copper with 50 ppm of phosphorus added. Corrosion studies supported by SKB indicate that in the absence of mechanical failure or accelerated localised corrosion the overpack should provide corrosion shielding of the canister for its full design life. Published work claiming that the nodular iron liner would have corrosion characteristics similar to the carbon steel which had been examined in depth is flawed since the microstructures of the iron and carbon steel specimens used were not investigated. It is highly unlikely that nodular irons in the form used for the experiments would have similar structures to nodular iron in the canisters by chance. If the overpack were breached during the aerobic period of the repository life then very rapid penetration of the inner liner could occur. It has been recognised that the roll forming method is not suitable for serial production and alternatives are being sought. The electron beam welding process has been explored with tenacity but has so far failed to produce a satisfactory lid weld. A new welder is being developed for supply to the SKB pilot plant where development will be continued. An alternative welding process, friction stir welding, is being examined as a candidate for attaching lids. Surface breaking defects may be detected using eddy current methods but there is currently no reliable way of detecting small sub surface defects in the overpack

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  9. 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

  10. 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

  11. Processes, Techniques, and Successes in Welding the Dry Shielded Canisters of the TMI-2 Reactor Core Debris

    International Nuclear Information System (INIS)

    The Idaho National Engineering and Environmental Laboratory (INEEL) is operated by Bechtel-BWXT Idaho LLC (BBWI), which recently completed a very successful $100 million Three-Mile Island-2 (TMI-2) program for the Department of Energy (DOE). This complex and challenging program used an integrated multidisciplinary team approach that loaded, welded, and transported an unprecedented 25 dry shielded canisters (DSC) in seven months, and did so ahead of schedule. The program moved over 340 canisters of TMI-2 core debris that had been in wet storage into a dry storage facility at the INEEL. The main thrust of this paper is relating the innovations, techniques, approaches, and lessons learned associated to welding of the DSC's. This paper shows the synergism of elements to meet program success and shares these lessons learned that will facilitate success with welding of dry shielded canisters in other DOE complex dry storage programs

  12. Modeling of thermal evolution of near field area around single pit mode nuclear waste canister disposal in soft rocks

    International Nuclear Information System (INIS)

    Soft rocks like argillites/shales are under consideration worldwide as host rock for geological disposal of vitrified as well as spent fuel nuclear waste. The near field around disposed waste canister at 400-500m depth witnesses a complex heat field evolution due to varying thermal characteristics of rocks, coupling with hydraulic processes and varying intensity of heat flux from the canister. Smooth heat dissipation across the rock is desirable to avoid buildup of temperature beyond design limit (100 °C) and resultant micro fracturing due to thermal stresses in the rocks and intervening buffer clay layers. This also causes enhancement of hydraulic conductivity of the rocks, radionuclide transport and greater groundwater ingress towards the canister. Hence heat evolution modeling constitutes an important part of safety assessment of geological disposal facilities

  13. Selection of candidate canister materials for high-level nuclear waste containment in a tuff repository

    International Nuclear Information System (INIS)

    A repository located at Yucca Mountain at the Nevada Test Site is a potential site for permanent geological disposal of high-level nuclear waste. The repository can be located in a horizon in welded tuff, a volcanic rock, which is above the static water level at this site. The environmental conditions in this unsaturated zone are expected to be air and water vapor dominated for much of the containment period. Type 304L stainless steel is the reference material for fabricating canisters to contain the solid high-level wastes. Alternative stainless alloys are considered because of possible susceptibility of 304L to localized and stress forms of corrosion. For the reprocessed glass wastes, the canisters serve as the recipient for pouring the glass with the result that a sensitized microstructure may develop because of the times at elevated temperatures. Corrosion testing of the reference and alternative materials has begun in tuff-conditioned water and steam environments. 21 references, 8 figures, 8 tables

  14. 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.

  15. Canister Storage Building (CSB) safety analysis report, phase 3: Safety analysis documentation supporting CSB construction

    International Nuclear Information System (INIS)

    The US Department of Energy established the K Basins Spent Nuclear Fuel Project to address safety and environmental concerns associated with deteriorating spent nuclear fuel presently stored under water in the Hanford Site's K Basins, which are located near the Columbia River. Recommendations for a series of aggressive projects to construct and operate systems and facilities to manage the safe removal of K Basins fuel were made in WHC-EP-0830, Hanford Spent Nuclear Fuel Recommended Path Forward, and its subsequent update, WHC-SD-SNF-SP-005, Hanford Spent Nuclear Fuel Project Integrated Process Strategy for K Basins Fuel. The integrated process strategy recommendations include the following steps: Fuel preparation activities at the K Basins, including removing the fuel elements from their K Basin canisters, separating fuel particulate from fuel elements and fuel fragments greater than 0.6 cm (0.25 in.) in any dimension, removing excess sludge from the fuel and fuel fragments by means of flushing, as necessary, and packaging the fuel into multicanister overpacks (MCOs); Removal of free water by draining and vacuum drying at a cold vacuum drying facility ES-122; Dry shipment of fuel from the Cold Vacuum Drying to the Canister Storage Building (CSB), a new facility in the 200 East Area of the Hanford Site

  16. 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.

  17. 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.

  18. Critical review of welding technology for canisters for disposal of spent fuel and high level waste

    International Nuclear Information System (INIS)

    Nagra is the Swiss national cooperative for the disposal of radioactive waste and is responsible for final disposal of all types of waste produced in Switzerland, which are partitioned into two repository types, one for spent fuel (SF), vitrified high-level waste (HLW) and long-lived intermediate level waste and one for low and intermediate level waste. In the general licences applied for these repositories, documentation has to show that long-term safety can be ensured and that factors for the construction, operation, and closure of the facility have been considered. Nagra has commissioned TWI to carry out a critical review of welding technologies for the sealing of HLW and SF canisters made of carbon steel. In conjunction with a material selection report, the information gained will be used as a preliminary step to provide input to developing design concepts for the canisters. The features to be considered are: a) Suitability of techniques for thickness of weld required; b) Suitability for remote operation, maintenance and set-up; c) Welding speed, weld quality, tolerances and cost; d) Effect of welding process on parent materials properties including microstructure corrosion resistance, distortion and residual stress; e) Potential post-weld treatments to reduce residual stress and enhance corrosion resistance; f) Suitability of inspection techniques for the weld thickness required; g) Impact of welding techniques on the canister design and material selection; h) Critique of emerging technologies which may be suitable in the future. The review of potential welding technologies began with a feasibility study carried out by TWI experts, where the unsuitable processes were rejected. For the remaining processes attention was focused on previous applications for the material and thickness suggested, and especially on safety critical applications such as applied in the nuclear and pressure vessel industry. Once the relevant information was gathered each process was

  19. Creep of OFHC and silver copper at simulated final repository canister-service conditions

    International Nuclear Information System (INIS)

    Results of high-resolution creep rate measurements are described for estimating very long term creep life of copper and silver alloyed copper at room temperature and at stresses approaching the expected service conditions of final repository canisters. The aim was to assess the limiting service stress levels for potential canister wall materials. The 0.1% silver alloyed copper showed minimum creep rates of 10-9 to 10-10 l/h, corresponding to 1 % strain in about 1000 to 10000 years, at room temperature and uniaxial stress level of 50 to 75 MPa. The predicted time to 1 % strain, when extrapolated from literature data, was at least one order of magnitude shorter. From the results of the present work, the 1 % creep life for OFHC copper was at most a few hundreds of years at 50 MPa stress level. The technique developed and used in this work for measuring very low strain rates appears useful for assessing low temperature creep life of practical structures essentially without accelerating the test from the service conditions. (au)

  20. Creep of OFHC and silver copper at simulated final repository canister-service conditions

    International Nuclear Information System (INIS)

    Result of high-resolution creep rate measurements are described for estimating very long term creep life of copper and silver alloyed copper at room temperature and at stresses approaching the expected service conditions of final repository canisters. The aim was to assess the limiting service stress levels for potential canister wall materials. The 0.1 % silver alloyed copper showed minimum creep rates of 10-9 to 10-10 l/h, corresponding to 1 % strain in about 1000 to 10000 years, at room temperature and uniaxial stress level of 50 to 75 MPa. The predicted time to 1 % strain, when extrapolated from literature data, was at least one order of magnitude shorter. From the results of the present work, the 1 % creep life for OFHC copper was at most a few hundreds of years at 50 MPa stress level. The technique developed and used in this work for measuring very low strain rates appears useful for assessing low temperature creep life of practical structures essentially without accelerating the test from the service conditions

  1. Acceptance of spent nuclear fuel in multiple element sealed canisters by the Federal Waste Management System

    International Nuclear Information System (INIS)

    This report is one of a series of eight prepared by E.R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. The eight reports are concerned with the conditions under which spent fuel and high level waste will be accepted in the following categories: (1) failed fuel; (2) consolidated fuel and associated structural parts; (3) non-fuel-assembly hardware; (4) fuel in metal storage casks; (5) fuel in multi-element sealed canisters; (6) inspection and testing requirements for wastes; (7) canister criteria; (8) spent fuel selection for delivery; and (9) defense and commercial high-level waste packages. 14 refs., 27 figs

  2. The Swedish Concept for Disposal of Spent Nuclear Fuel: Differences Between Vertical and Horizontal Waste Canister Emplacement

    International Nuclear Information System (INIS)

    The Swedish Nuclear Power Inspectorate (SKI) is preparing for the review of licence applications related to the disposal of spent nuclear fuel. The Swedish Nuclear Fuel and Waste Management Company (SKB) refers to its proposals for the disposal of spent nuclear fuel as the KBS-3 concept. In the KBS-3 concept, SKB plans that, after 30 to 40 years of interim storage, spent fuel will be disposed of at a depth of about 500 m in crystalline bedrock, surrounded by a system of engineered barriers. The principle barrier to radionuclide release is a cylindrical copper canister. Within the copper canister, the spent fuel is supported by a cast iron insert. Outside the copper canister is a layer of bentonite clay, known as the buffer, which is designed to provide mechanical protection for the canisters and to limit the access of groundwater and corrosive substances to their surfaces. The bentonite buffer is also designed to sorb radionuclides released from the canisters, and to filter any colloids that may form within the waste. SKB is expected to base its forthcoming licence applications on a repository design in which the waste canisters are emplaced in vertical boreholes (KBS-3V). However, SKB has also indicated that it might be possible and, in some respects, beneficial to dispose of the waste canisters in horizontal tunnels (KBS-3H). There are many similarities between the KBS-3V and KBS-3H designs. There are, however, uncertainties associated with both of the designs and, when compared, both possess relative advantages and disadvantages. SKB has identified many of the key factors that will determine the evolution of a KBS-3H repository and has plans for research and development work in many of the areas where the differences between the KBS-3V and KBS-3H designs mean that they could be significant in terms of repository performance. With respect to the KBS-3H design, key technical issues are associated with: 1. The accuracy of deposition drift construction. 2. Water

  3. 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.

  4. Finite element analysis of stresses and deformations occurring in the spent nuclear fuel (SNF) disposal canister deposited in a deep geological repository

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Bo [Department of Mechanical and Aerospace Engineering, Rutgers, The State University of New Jersey, 98 Brett Road, Piscataway, NJ 08854-8058 (United States); School of Aeronautics, Northwestern Polytechnical University of China, Xi’an 710072 (China); Gea, Hae Chang [Department of Mechanical and Aerospace Engineering, Rutgers, The State University of New Jersey, 98 Brett Road, Piscataway, NJ 08854-8058 (United States); Kwon, Young Joo, E-mail: yjkwon@hongik.ac.kr [Department of Mechanical and Design Engineering, Hongik University, 2639 Sejong-ro, Jochiwon, Sejong 339-701 (Korea, Republic of)

    2014-01-15

    Highlights: • The weight reduction design of the PWR SNF disposal canister is presented. • The optimal SNF basket rotation angle to minimize canister weight is sought. • Conventional structural analysis and Kriging method are developed for this purpose. • The optimal SNF basket rotation angle is determined to be 45°. • The canister weight is correspondingly reduced by 16% from 25 tons to 21 tons. -- Abstract: Numerical computer experimental methodologies are investigated for the weight reduction of a spent nuclear fuel (SNF) disposal canister designed to be deposited in a Korean deep geological repository from a pressurized water reactor (PWR). Finite element analyses of stresses and deformations occurring inside the cylindrical canister under the deposited conditions are performed to assess its structural strength at various rotation angles (φ) of the SNF basket. Specifically, the cross sections of four square tube shaped SNF baskets (assemblies) contained in the canister are rotated. Using a conventional structural analysis and a Kriging method, an optimal rotation angle is determined in relation to canister diameter and weight. Both sets of results are in agreement. It was also determined that the computed deformation changes slightly in relation to variances in rotation angle, while the stress incurred inside the cast iron insert of the canister noticeably changes reaching its highest value at φ = 45° while still maintaining safe structural integrity. It is concluded that the diameter of the canister can be reduced from its original design value (102 cm) to 95.8463 cm resulting in a ∼16.0% reduction in canister weight for an optimal rotation angle of 45°.

  5. 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

  6. 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

  7. 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

  8. 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

  9. Test Plan to Determine the Maximum Surface Temperatures for a Plutonium Storage Cubicle with Horizontal 3013 Canisters

    International Nuclear Information System (INIS)

    A simulated full-scale plutonium storage cubicle with 22 horizontally positioned and heated 3013 canisters is proposed to confirm the effectiveness of natural circulation. Temperature and airflow measurements will be made for different heat generation and cubicle door configurations. Comparisons will be made to computer based thermal Hydraulic models

  10. Copper canister with cast inner component. Amendment to project on Alternative Systems Study (PASS), SKB TR 93-04

    International Nuclear Information System (INIS)

    The Project on Alternative Systems Study, PASS, was described in a report dated october 1992. In the report, the reference repository concept KBS-3 is described together with three other alternatives. In the report several designs for fuel storage canisters are presented. This report describes a recently developed design for the inner component of the composite, steel and copper, canister which is the main alternative in the KBS-3-model. The new design will be manufactured by casting. A cast insert with inner walls eliminates the need for a stabilizing filler in the canister and guarantees that the fuel remains sub-critical during sufficient time in the repository. The cast insert is judged, to, in comparison with the steel tube alternative, lead to a considerably simplified process in the encapsulation plant and lower development and investment cost. Positive side effects of the design are that the mechanical strength is improved by a factor 2-3 and that the difficult filling operation is avoided in the encapsulation process. The drawbacks are higher weight and probably higher unit price for the empty canister

  11. 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.

  12. Feasibility of direct reactivity measurement in multi-canister overpacks at the Cold Vacuum Drying Facility

    International Nuclear Information System (INIS)

    A proposed method for measuring the chemical reaction rate (power) of breached N-Reactor fuel elements with water in a Multi-canister overpack (MCO) based on hydrogen release rate is evaluated. The reaction rate is measured at 50 C in an oxygen free water by applying a vacuum to boil the water and adding a low, measured flow of helium. The ratio of helium to hydrogen is used to infer the reaction rate. A test duration of less than 8 hours was found to provide sufficient accuracy for confidence in the measurement results. A more rigorous treatment of system measurement accuracy, which may yield shorter test durations, should be performed if this reactivity measurement is to be employed

  13. Acceptance of canisters of high-level waste by the Federal Waste Management System

    International Nuclear Information System (INIS)

    This report is one of a series of eight prepared by E. R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high-priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. The eight reports are concerned with the conditions under which spent fuel and high-level waste will be accepted. This document discusses the acceptance of canisters of high-level waste by the Federal Waste Management System. 16 refs., 7 figs., 11 tabs

  14. 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

  15. Multi-canister overpack project -- verification and validation, MCNP 4A

    Energy Technology Data Exchange (ETDEWEB)

    Goldmann, L.H.

    1997-11-10

    This supporting document contains the software verification and validation (V and V) package used for Phase 2 design of the Spent Nuclear Fuel Multi-Canister Overpack. V and V packages for both ANSYS and MCNP are included. Description of Verification Run(s): This software requires that it be compiled specifically for the machine it is to be used on. Therefore to facilitate ease in the verification process the software automatically runs 25 sample problems to ensure proper installation and compilation. Once the runs are completed the software checks for verification by performing a file comparison on the new output file and the old output file. Any differences between any of the files will cause a verification error. Due to the manner in which the verification is completed a verification error does not necessarily indicate a problem. This indicates that a closer look at the output files is needed to determine the cause of the error.

  16. 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.

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

    International Nuclear Information System (INIS)

    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

  18. Project JADE. Method and machinery description of equipment for deposition of a canister in a vertical deposition hole

    International Nuclear Information System (INIS)

    A systematic evaluation of different disposal methods has been carried out. The study is named Comparison of Disposal Methods. The evaluation has included a comparison of the technical aspects, safety aspects and costs of alternatives proposed within the so-called KBS-3 method. Three alternatives have been studied and compared: vertical emplacement (KBS-3V), horizontal emplacement (KBS-3H) and emplacement in medium long horizontal holes (MLH). KBS-3V is the reference method adopted in SKB's development and planning work. This report describes eight alternative disposal methods, with variations, and forms a technical basis for the assessment of methods involving vertical disposal (KBS-3V). The alternative of emplacement behind a radiation-shielding screen has been rejected by SKB, as it has been decided that disposal will be carried out with complete radiation shielding around the canister. However, the alternative is considered in the report for the sake of comparison. Based on the applicable technical specifications, the results of fault-effect analyses, radiation protection assessments and flexibility and complexity analyses for the entire disposal process, two methods for vertical emplacement have been identified as the best from a technical point of view: Transport of a horizontally-lying canister which is raised to a vertical position during emplacement. The canister is shielded during transport and the raising movement. Radiation protection can be complete or partial. Transport with a standing canister. Under transport and disposal, the canister is surrounded by a complete radiation shield, which has a telescopic lower part. This principle involves only a few, simple mechanical movements

  19. A review of materials and corrosion issues regarding canisters for disposal of spent fuel and high-level waste in Opalinus clay

    International Nuclear Information System (INIS)

    The project 'Entsorgungsnachweis' presented by NAGRA to the Swiss Federal Government in December 2002 assessed the feasibility of disposal of spent fuel (SF), vitrified high level waste (HLW) from reprocessing and long-lived intermediate level waste in an Opalinus Clay repository site in Northern Switzerland. NAGRA proposed the use of carbon steel canisters for disposal of SF/HLW and it also put forward an alternative concept of copper canisters with cast iron insert. In its reply the Federal Government acknowledged that NAGRA had successfully demonstrated the technical feasibility of disposal of SF/HLW. However, some of its experts raised a number of questions related to the choice of steel as canister material. Among others, it was questioned whether hydrogen formed by corrosion of steel in contact with saturated bentonite might adversely affect the barrier function of the Opalinus clay. It was also recommended that alternative canister materials and/or design concepts should be evaluated. To deal with these concerns NAGRA convened an international group of experts, the Canister Materials Review Board (CMRB), who were to review the existing information on canister materials that could be suitable for the proposed repository environment. Based on present knowledge of materials science, the CMRB was to recommend to NAGRA the most suitable material(s) for meeting the performance requirements for SF/HLW canisters. Specifically, the CMRB was to consider corrosion, including hydrogen generation, and stress-assisted failure processes that could affect the integrity and projected life time of SF/HLW canisters or impede the functioning of geological barriers while keeping in mind the overall feasibility of manufacturing, sealing and inspecting the canisters. The CMRB was further asked to identify the needs and provide advice for further studies by NAGRA on the long term performance and safety of SF/HLW canisters in the Swiss repository concept. For the assessment of the

  20. 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

  1. 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

  2. Factors Affecting the Estimation of Indoor Radon Using Passive Activated Charcoal Canisters

    Science.gov (United States)

    Scarpitta, Salvatore Charles

    1990-01-01

    Adsorption and desorption studies of 20 activated charcoals were conducted in a monolayer and a packed bed utilizing tracer gases. Kinetic studies, using xenon-133, demonstrate the existence of a two-compartment micropore volume with entrance capillaries which together determine the response characteristics of the charcoal to external concentration gradients of tracer gases. This new two -compartment model adequately describes the adsorption and desorption dynamics of radon in the presence of water vapor. Measurements with charcoal exposed to water vapor and Rn-222 in a monolayer and packed bed for exposure intervals of 1-7 days demonstrate that the uptake rate and total quantity of adsorbed Rn-222 are highly dependent upon the amount of water adsorbed. The effect of CO_2 on radon adsorption is small in any charcoal. The measured effective diffusion coefficient of radon in a packed bed of a peat based charcoal at 15% humidity and 25^circC is 7.97 times 10^{-6} cm^2/s. Condensed water vapor in the entrance capillaries reduces the effective pore radius, increasing the diffusion half-time, both into and out of the charcoal. The amount of adsorbed water per gram of charcoal required to block the entrance capillaries varies with the charcoal type. The proposed term for this quantity is the "break-point". A two-stage diffusion barrier charcoal monitor with a long diffusion path length was developed. This design inhibits passive airflow while maintaining the amount of adsorbed water vapor in the primary charcoal adsorbent below the break-point. Water removal at the entry port allows for longer exposure times improving the integrating capability necessary for indoor exposure assessment. The long diffusion path length increases the integration time -constant for radon adsorption normally 24 hours for conventional open-faced canisters to 50 hours for the improved canister. The increased integration time-constant allows for a 7 day sample to be measured at 70% humidity and 23

  3. Study on commercial realization of concrete cask for interim storage of spent nuclear fuel. Proposal of examination methodology for stainless steel canister lid weldment

    International Nuclear Information System (INIS)

    To realize the early utilization of economical and module storage methodology, the Japan Society of Mechanical Engineers (JSME) had already developed and issued a Code for Construction of SNF Facilities - Rules on Concrete Cask - (JSME S FB1-2003) in 2003. On the other hand, the former competent authority (Nuclear and Industrial Safety Agency) issued the safety requirements for concrete cask design in 2004. For confinement requirements, Ultrasonic Test (UT) should be used to inspect the lid weldment in addition to Dye Penetrant Test (PT). To contribute to the JSME codification activities for adjustment between the safety requirements and the JSME design code, we performed mock-up canister weldment tests considering the vapor from the hot water in the spent fuel canister, the UT tests with canister lid weldment at high temperature and the structural analysis of the canister under hypothetical drop conditions. As a result, the draft amendment of the JSME design code was proposed. (author)

  4. Hanford Spent Nuclear Fuel Project evaluation of multi-canister overpack venting and monitoring options during staging of K basins fuel

    International Nuclear Information System (INIS)

    This engineering study recommends whether multi-canister overpacks containing spent nuclear fuel from the Hanford K Basins should be staged in vented or a sealed, but ventable, condition during staging at the Canister Storage Building prior to hot vacuum conditioning and interim storage. The integrally related issues of MCO monitoring, end point criteria, and assessing the practicality of avoiding venting and Hot Vacuum Conditioning for a portion of the spent fuel are also considered

  5. 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.

  6. Brine: a computer program to compute brine migration adjacent to a nuclear waste canister in a salt repository

    International Nuclear Information System (INIS)

    This report presents a mathematical model used to predict brine migration toward a nuclear waste canister in a bedded salt repository. The mathematical model is implemented in a computer program called BRINE. The program is written in FORTRAN and executes in the batch mode on a CDC 7600. A description of the program input requirements and output available is included. Samples of input and output are given

  7. The leading of Titanium on corrosion resistance of AISI 321 stainless steel as material for nuclear waste canister

    International Nuclear Information System (INIS)

    Ultimate disposal of spent fuel or high level waste on underground, needs canister which has high corrosion resistance. The AISI 321 stainless steel which contains titanium as spent fuel or high level waste canister has been studied. The titanium content in alloy would retard the formation of Cr23C6 precipitate, so the corrosion attack will be avoided. The experiment was conducted by heating specimens at temperature of 700oC for 2 hours in which the analysis was performed by optical microscope, SEM and EDS. The analysis results showed that at the experiment temperature was undetected the Cr23C6 precipitate, although the TiC precipitate was formed with average diameter of 4.70 μm with the Ti content on the TiC and matrix were 96.20% and 0.925% of weight respectively. The fact that this material has a high corrosion resistance, so the use as spent fuel or high level waste canister will be sufficiently safe

  8. Demonstrating compliance with the waste acceptance preliminary specifications on foreign materials within DWPF canistered waste forms (U)

    International Nuclear Information System (INIS)

    The Defense Waste Processing Facility (DWPF) will employ a waste acceptance program based on the Waste Acceptance Preliminary Specifications (WAPS). These specifications require, among other criteria, that the canistered waste form contain no free liquids, free gases, organics, or explosives. Of particular importance is the absence of liquid water. This paper summarizes efforts and discusses experiments at the Savannah River Site for demonstrating compliance with the foreign materials specifications of the WAPS. Existing data, already in the literature, is being combined with the results of new experiments. For the volatility of the waste glass, documented work is combined with new results of thermogravimetric analysis experiments on simulated waste glass samples produced during scale glass melter campaigns. The volatility of these glass samples provides evidence that no free liquids, free gases, organics, or explosives are released upon heating the waste glass to its glass transition temperature. To show compliance of the absence of liquid water, documented work is being combined with the results of new experiments involving measurement of the internal gas pressure, the composition of the gas within the canisters, and the relative humidity of sealed, canistered waste forms produced during large-scale glass melter runs and the upcoming cold runs of the DWPF. (orig.)

  9. Experimental and Computational Investigations of Phase Change Thermal Energy Storage Canisters

    Science.gov (United States)

    Ibrahim, Mounir; Kerslake, Thomas; Sokolov, Pavel; Tolbert, Carol

    1996-01-01

    Two sets of experimental data are examined in this paper, ground and space experiments, for cylindrical canisters with thermal energy storage applications. A 2-D computational model was developed for unsteady heat transfer (conduction and radiation) with phase-change. The radiation heat transfer employed a finite volume method. The following was found in this study: (1) Ground Experiments: the convection heat transfer is equally important to that of the radiation heat transfer; radiation heat transfer in the liquid is found to be more significant than that in the void; including the radiation heat transfer in the liquid resulted in lower temperatures (about 15 K) and increased the melting time (about 10 min.); generally, most of the heat flow takes place in the radial direction. (2) Space Experiments: radiation heat transfer in the void is found to be more significant than that in the liquid (exactly the opposite to the Ground Experiments); accordingly, the location and size of the void affects the performance considerably; including the radiation heat transfer in the void resulted in lower temperatures (about 40 K).

  10. 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.

  11. 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).

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

    International Nuclear Information System (INIS)

    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)

  13. 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

  14. 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).

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

    International Nuclear Information System (INIS)

    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. 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.

  17. ANALYSIS OF SLUDGE BATCH 4 (MACROBATCH 5) FOR CANISTER S02902 AND SLUDGE BATCH 5 (MACROBATCH 6) FOR CANISTER S03317 DWPF POUR STREAM GLASS SAMPLES

    Energy Technology Data Exchange (ETDEWEB)

    Reigel, M.; Bibler, N.

    2010-10-04

    The Defense Waste Processing Facility (DWPF) began processing Sludge Batch 4 (SB4), Macrobatch 5 (MB5) on May 29, 2007. Sludge Batch 4 was a blend of the heel of Tank 40 from Sludge Batch 3 (SB3) and SB4 material qualified in Tank 51. On November 28, 2008, DWPF began processing Sludge Batch 5 (SB5) from Tank 40 which is a blend of the heel of Tank 40 from SB4, SB5 material qualified in Tank 51 and H-Canyon Pu and Np transfers. SB4 was processed using Frit 510 and SB5 used Frit 418. During processing of each sludge batch, the DWPF is required to take at least one glass sample to meet the objectives of the Glass Product Control Program and to complete the necessary Production Records so that the final glass product may be disposed of at a Federal Repository. During the processing of SB4 and SB5, glass samples were obtained during the pouring of canisters S02902 and S03317, respectively. The samples were transferred to the Savannah River National Laboratory (SRNL) where they were analyzed (durability, chemical and radionuclide composition). The following observations and conclusions are drawn from the analytical results provided in this report: (1) The sum of the oxides for the chemical composition of both the SB4 and SB5 pour stream glasses is within the Product Composition Control System (PCCS) acceptance limits (95 {le} sum of oxides {le} 105). (2) The calculated Sludge Dilution Factor (SDF) for SB4 is 2.52. The measured radionuclide content is in good agreement with the calculated values from the dried sludge results from the SB4 Waste Acceptance Production Specification (WAPS) sample (References 1 and 19). (3) The calculated SDF for SB5 is 2.60. The measured radionuclide content is in good agreement with the calculated values from the dried sludge results from the SB5 WAPS sample (References 2 and 20). (4) Scanning Electron Microscopy (SEM) analysis shows there are noble metal inclusions, primarily ruthenium, present in both pour stream samples. (5) The Product

  18. 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

  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. 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

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

    International Nuclear Information System (INIS)

    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. This estimate can

  2. 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

  3. Development of measurement technology of chlorine attached on canister using laser. Remote measurement in narrow space for the application during storage of spent fuel

    International Nuclear Information System (INIS)

    Stress corrosion cracking (SCC) at the surface of a stainless steel canister is a current issue in the interim storage of spent fuel by a concrete cask. The concentration measurement of salt attached on a canister is one of the important methods for inspection of environmental condition for SCC occurrence. Laser-induced breakdown spectroscopy (LIBS) is an attractive method for the concentration measurement of salt attached on the candidate material of a canister. In order to adopt the LIBS to the actual equipment, the remote measurement is needed because the surface of the canister is under radioactive and hot environmental condition. In this study, the remote measurement by using open pass LIBS was performed, and the prototype of compact devices for remote LIBS has been developed. The developed device for laser focusing can be inserted in the narrow space simulating the space between the concrete body and the canister. The results of open pass LIBS showed that the chlorine emission spectrum was measured in the narrow space when the distance from laser device to the measurement points was over 20 m. (author)

  4. Heat transfer and thermal storage performance of an open thermosyphon type thermal storage unit with tubular phase change material canisters

    International Nuclear Information System (INIS)

    Highlights: • A novel open heat pipe thermal storage unit is design to improve its performance. • Mechanism of its operation is phase-change heat transfer. • Tubular canisters with phase change material were placed in thermal storage unit. • Experiment and analysis are carried out to investigate its operation properties. - Abstract: A novel open thermosyphon-type thermal storage unit is presented to improve design and performance of heat pipe type thermal storage unit. In the present study, tubular canisters filled with a solid–liquid phase change material are vertically placed in the middle of the thermal storage unit. The phase change material melts at 100 °C. Water is presented as the phase-change heat transfer medium of the thermal storage unit. The tubular canister is wrapped tightly with a layer of stainless steel mesh to increase the surface wettability. The heat transfer mechanism of charging/discharging is similar to that of the thermosyphon. Heat transfer between the heat resource or cold resource and the phase change material in this device occurs in the form of a cyclic phase change of the heat-transfer medium, which occurs on the surface of the copper tubes and has an extremely high heat-transfer coefficient. A series of experiments and theoretical analyses are carried out to investigate the properties of the thermal storage unit, including power distribution, start-up performance, and temperature difference between the phase change material and the surrounding vapor. The results show that the whole system has excellent heat-storage/heat-release performance

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

    International Nuclear Information System (INIS)

    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 ω-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

  6. 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.

  7. Sampling and analysis plan for sludge located in fuel storage canisters of the 105-K West basin

    Energy Technology Data Exchange (ETDEWEB)

    Baker, R.B.

    1997-04-30

    This Sampling and Analysis Plan (SAP) provides direction for the first sampling of sludge from the K West Basin spent fuel canisters. The specially developed sampling equipment removes representative samples of sludge while maintaining the radioactive sample underwater in the basin pool (equipment is described in WHC-SD-SNF-SDD-004). Included are the basic background logic for sample selection, the overall laboratory analyses required and the laboratory reporting required. These are based on requirements put forth in the data quality objectives (WHC-SD-SNF-DQO-012) established for this sampling and characterization activity.

  8. Performance of CASTOR{sup R} HAW Cask Cold Trials for Loading, Transport and Storage of HAW canisters

    Energy Technology Data Exchange (ETDEWEB)

    Wilmsmeier, Marco; Vossnacke, Andre [GNS Gesellschaft fuer Nuklear-Service mbH, Hollestrasse 7A, D-45127 Essen (Germany)

    2008-07-01

    On the basis of reprocessing contracts, concluded between the German Nuclear Utilities (GNUs) and the reprocessing companies in France (AREVA NC) and the UK (Nuclear Decommissioning Authority), GNS has the task to return the resulting residues to Germany. The high active waste (HAW) residuals from nuclear fuel reprocessing are vitrified and filled into steel cans, the HAW canisters. According to reprocessing contracts the equivalent number of HAW canisters to heavy metals delivered has to be returned to the country of origin and stored at an interim storage facility where applicable. The GNS' CASTOR{sup R} HAW casks are designed and licensed to fulfil the requirements for transport and long-term storage of HAW canisters. The new cask type CASTOR{sup R} HAW28M is capable of storing 28 HAW canisters with a maximum thermal power of 56 kW in total. Prior to the first active cask loading at a reprocessing facility it is required to demonstrate all important handling steps with the CASTOR{sup R} HAW28M cask according to a specific and approved sequence plan (MAP). These cold trials have to be carried out at the cask loading plant and at the reception area of an interim storage facility in Gorleben (TBL-G), witnessed by the licensing authorities and their independent experts. At transhipment stations GNS performs internal trials to demonstrate safe handling. A brand-new, empty CASTOR{sup R} HAW28M cask has been shipped from the GNS cask assembly facility in Muelheim to the TBL-G for cold trials. With this cask, GNS has to demonstrate the transhipment of casks at the Dannenberg transfer station from rail to road, transport to and reception at the TBL-G as well as incoming dose rate and contamination measurements and preparation for storage. After removal of all shock absorbers with a cask specific handling frame, tilting operation and assembly of the secondary lid with a pressure sensor, the helium leak tightness and 'Block-mass' tests have to be carried out

  9. Sampling and analysis plan for sludge located in fuel storage canisters of the 105-K West basin

    International Nuclear Information System (INIS)

    This Sampling and Analysis Plan (SAP) provides direction for the first sampling of sludge from the K West Basin spent fuel canisters. The specially developed sampling equipment removes representative samples of sludge while maintaining the radioactive sample underwater in the basin pool (equipment is described in WHC-SD-SNF-SDD-004). Included are the basic background logic for sample selection, the overall laboratory analyses required and the laboratory reporting required. These are based on requirements put forth in the data quality objectives (WHC-SD-SNF-DQO-012) established for this sampling and characterization activity

  10. Multi Canister Overpack (MCO) Handling Machine Trolley Seismic Uplift Constraint Design Loads

    Energy Technology Data Exchange (ETDEWEB)

    SWENSON, C.E.

    2000-03-09

    The MCO Handling Machine (MHM) trolley moves along the top of the MHM bridge girders on east-west oriented rails. To prevent trolley wheel uplift during a seismic event, passive uplift constraints are provided as shown in Figure 1-1. North-south trolley wheel movement is prevented by flanges on the trolley wheels. When the MHM is positioned over a Multi-Canister Overpack (MCO) storage tube, east-west seismic restraints are activated to prevent trolley movement during MCO handling. The active seismic constraints consist of a plunger, which is inserted into slots positioned along the tracks as shown in Figure 1-1. When the MHM trolley is moving between storage tube positions, the active seismic restraints are not engaged. The MHM has been designed and analyzed in accordance with ASME NOG-1-1995. The ALSTHOM seismic analysis (Reference 3) reported seismic uplift restraint loading and EDERER performed corresponding structural calculations. The ALSTHOM and EDERER calculations were performed with the east-west seismic restraints activated and the uplift restraints experiencing only vertical loading. 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 east-west trolley restraints are not engaged. For this case, the associated trolley movements would result in east-west lateral loads on the uplift constraints due to friction, as shown in Figure 1-2. During preliminary evaluations, questions were raised as to whether the EDERER calculations considered the latest ALSTHOM seismic analysis loads (See NCR No. 00-SNFP-0008, Reference 5). Further evaluation led to the conclusion that the EDERER calculations used appropriate vertical loading, but the uplift restraints would need to be re-analyzed and modified to account for lateral loading. The disposition of NCR 00-SNFP-0008 will track the redesign and modification effort. The purpose of this calculation is to establish bounding seismic

  11. 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

  12. 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.

  13. 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

  14. Thermomechanical room and canister region benchmark analyses between STEALTH-WI and SPECTROM-32: Draft final report

    International Nuclear Information System (INIS)

    This report documents the benchmarking of the two-dimensional waste isolation version of STEALTH (designated STEALTH-WI) against the thermomechanical performance assessment calculations performed by RE/SPEC using SPECTROM-32. An axisymmetric, canister-scale (very-near-field) analysis was performed to compute the peak stress exerted by the salt on the waste package. A plane strain, room-scale (near-field) analysis was also performed to predict disposal room roof-to-floor closure and the temperatures at key locations in the vicinity of the disposal room. Comparisons between the STEALTH and SPECTROM-32 results showed that the temperature predictions agreed to within 5/degree/C, peak canister stresses better than 10%, and the average roof-to-floor closures within 30%. The stress and displacement differences were attributed to differences in the treatment of plasticity in the constitutive laws for salt employed in STEALTH and SPECTROM-32. The temperature differences were due to minor differences in the thermal models employed in STEALTH and SPECTROM- 41, the thermal analysis code which supplies temperatures for SPECTROM-32. 9 refs., 21 figs., 6 tabs

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

    International Nuclear Information System (INIS)

    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

  16. 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

  17. 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

  18. State of the art of the welding method for sealing spent nuclear fuel canister made of copper. Part 2 - EBW

    International Nuclear Information System (INIS)

    This report consist the results of the development of the electron beam welding (EBW) method for sealing spent nuclear fuel (SNF) disposal canister. This report has been used as background material for selection of the sealing method for the SNF canister. Report contains the state of the art knowledge of the EBW method and research and development (R and D) results done by Posiva. Relevant R and D results of EB-welds done by SKB are also reviewed in this report. Requirements set for the welding and weld are present. These requirements are based on the long term safety and also some part of requirements are set by other processes like non-destructive testing (NDT) and manufacturing processes of components. Initial state of the weld is described in this report. Initial state has significant effect on the long term safety issues like corrosion resistance and creep ductility. Also short and long term mechanical properties as well as corrosion properties are described. Microstructure and residual stresses of the weld is represented in this report. Report consists also imperfections of the weld and statistical analysis of the evaluation of the probability of the largest defect size on the weld. Results of corrosion and creep tests of EB-welds are reviewed in this report. EBW process and machine are described. Preliminary designing of the EBW-machine has been done including component handling equipments. Preliminary welding procedure specification (pWPS) has drawn up and qualification of the personnel is described briefly. In-line process and quality control system including seam tracking system is implemented in modern EBW machine. Also NDT methods for inspection of the weld are described in this report. Concerning the results from the research and development work it can be concluded that EB welding method is suitable method for sealing SNF canister. Weld material fulfils requirements set by the long term safety. The welding system is robust and reliable and it is based

  19. US NRC-Sponsored Research on Stress Corrosion Cracking Susceptibility of Dry Storage Canister Materials in Marine Environments - 13344

    International Nuclear Information System (INIS)

    At a number of locations in the U.S., spent nuclear fuel (SNF) is maintained at independent spent fuel storage installations (ISFSIs). These ISFSIs, which include operating and decommissioned reactor sites, Department of Energy facilities in Idaho, and others, are licensed by the U.S. Nuclear Regulatory Commission (NRC) under Title 10 of the Code of Federal Regulations, Part 72. The SNF is stored in dry cask storage systems, which most commonly consist of a welded austenitic stainless steel canister within a larger concrete vault or overpack vented to the external atmosphere to allow airflow for cooling. Some ISFSIs are located in marine environments where there may be high concentrations of airborne chloride salts. If salts were to deposit on the canisters via the external vents, a chloride-rich brine could form by deliquescence. Austenitic stainless steels are susceptible to chloride-induced stress corrosion cracking (SCC), particularly in the presence of residual tensile stresses from welding or other fabrication processes. SCC could allow helium to leak out of a canister if the wall is breached or otherwise compromise its structural integrity. There is currently limited understanding of the conditions that will affect the SCC susceptibility of austenitic stainless steel exposed to marine salts. NRC previously conducted a scoping study of this phenomenon, reported in NUREG/CR-7030 in 2010. Given apparent conservatisms and limitations in this study, NRC has sponsored a follow-on research program to more systematically investigate various factors that may affect SCC including temperature, humidity, salt concentration, and stress level. The activities within this research program include: (1) measurement of relative humidity (RH) for deliquescence of sea salt, (2) SCC testing within the range of natural absolute humidity, (3) SCC testing at elevated temperatures, (4) SCC testing at high humidity conditions, and (5) SCC testing with various applied stresses. Results

  20. 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.

  1. A unique tungsten-based tagging approach for maintaining of continuity of knowledge of nuclear waste copper canisters

    International Nuclear Information System (INIS)

    A new approach to the unique tagging of nuclear waste copper canisters is suggested. In this new method a combination of two different techniques, radiation and ultrasonic measurements, is used in order to get the same unique identifier of the cask. The necessary component of the method is a tungsten/lead insert marked with a binary or bar code and placed inside the container. The paper discusses results of the radiation measurements performed in lab as the first proof of the concept, as well as results of Monte-Carlo evaluation of the feasibility of proposed approach. The method makes it possible to maintain continuity of knowledge of nuclear waste for a time scale up to a few hundred years without comprising the environmental safety of casks. (author)

  2. Preselection of Ni-Cr(-Mo) alloys as potential canister materials for vitrified high active nuclear waste by electrochemical testing

    Science.gov (United States)

    Bort, H.; Wolf, I.; Leistikow, S.

    1987-07-01

    Several Ni-Cr(-Mo) alloys (Hastelloy C4, Inconel 625, Sanicro 28, Incoloy 825, Inconel 690) were tested by electrochemical methods to characterize their corrosion behavior in chloride containing solutions at various temperatures and pH-values in respect to their application as canister materials for final radioactive waste storage. Especially, Hastelloy C4 was tested by potentiodynamic, potentiostatic and galvanostic measurements. As electrolytes H 2SO 4 solutions were used, as parameters temperature, chloride content and pH-value were varied. All tested alloys showed a clearly limited resistance against pitting corrosion phenomena; under severe conditions even crevice corrosion phenomena were observed. The best corrosion behavior, however, is shown by Hastelloy C4, which has the lowest passivation current density of all tested alloys and the largest potential region with protection against local corrosion phenomena.

  3. Preselection of Ni-Cr(-Mo) alloys as potential canister materials for vitrified high active nuclear waste by electrochemical testing

    Energy Technology Data Exchange (ETDEWEB)

    Bort, H.; Wolf, I.; Leistikow, S.

    1987-07-01

    Several Ni-Cr(-Mo) alloys (Hastelloy C4, Inconel 625, Sanicro 28, Incoloy 825, Inconel 690) were tested by electrochemical methods to characterize their corrosion behavior in chloride containing solutions at various temperatures and pH-values in respect to their application as canister materials for final radioactive waste storage. Especially, Hastelloy C4 was tested by potentiodynamic, potentiostatic and galvanostic measurements. As electrolytes H/sub 2/SO/sub 4/ solutions were used, as parameters temperature, chloride content and pH-value were varied. All tested alloys showed a clearly limited resistance against pitting corrosion phenomena; under severe conditions even crevice corrosion phenomena were observed. The best corrosion behavior, however, is shown by Hastelloy C4, which has the lowest passivation current density of all tested alloys and the largest potential region with protection against local corrosion phenomena.

  4. STS-102 MPLM Leonardo is moved to the payload canister for transfer to Launch Pad 39B

    Science.gov (United States)

    2001-01-01

    KENNEDY SPACE CENTER, Fla. -- In the Space Station Processing Facility, an overhead crane begins lifting the Multi-Purpose Logistics Module Leonardo. The MPLM is being moved to the payload canister for transfer to Launch Pad 39B and installation in Space Shuttle Discovery. The Leonardo, one of Italy'''s major contributions to the International Space Station program, is a reusable logistics carrier. It is the primary delivery system used to resupply and return Station cargo requiring a pressurized environment. Leonardo is the primary payload on mission STS-102 and will deliver up to 10 tons of laboratory racks filled with equipment, experiments and supplies for outfitting the newly installed U.S. Laboratory Destiny. STS-102 is scheduled to launch March 8 at 6:45 a.m. EST.

  5. 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.

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

    International Nuclear Information System (INIS)

    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.)

  7. 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.)

  8. Mechanical Analysis of an SM 2 Blk IV restrained firing within a concentric canister launcher test unit

    Energy Technology Data Exchange (ETDEWEB)

    Kassner, M C; Kennedy, T C; Puttapitukporn, T; Rosen, R S

    1999-03-01

    The Office of Naval Research (ONR) and PMS512 have undertaken a program to develop a new Vertical Launching System (VLS) for future generation ships, such as the DD-21 Destroyer. The Naval Sea Systems Command Combat Weapons Program (NAVSEA 05K) and Naval Surface Warfare Center Dahlgren Division (NSWCDD) are working jointly with industry and universities to develop one such launcher design, the Concentric Canister Launcher (CCL). The basic CCL design consists of a tube made of two concentric cylinders; one end is open, the other is sealed with a hemispherical end cap. During firing, the missile exhaust gas is turned 180 degrees by the hemispherical end cap and flows through the annular space between inner and outer cylinders. Depending on the missile utilized and the particular service environment of the CCL, maximum temperatures within the cylinder material have been calculated to exceed 2000 F. In an earlier study [1], the authors determined the high temperature mechanical properties of several candidate alloys being considered for fabrication of the CCL. This study [1] found that, of these candidate materials, titanium alloys exhibit higher yield stresses than that of 316L stainless steel at temperatures up to about 1000 F; above 1500 F, the yield stress of 316L stainless steel is comparable to those of the titanium alloys. The 316L stainless steel was found to strain harden (increase its flow stress with increasing strain) at temperatures up to about 1800 F. The ability of the 316L stainless steel to strain harden at high temperatures may provide an added margin of safety for engineering design of the CCL. The objective of the current study was to perform a computer simulation of the structural response of a CCL during a restrained firing, one in which a SM-2 Blk IV missile would fail to exit the canister. A finite element model of the inner cylinder, outer cylinder, end rings (mounting brackets), and lateral restraints in the uptake was constructed. An elastic

  9. Mechanical Analysis of an SM 2 Blk IV restrained firing within a concentric canister launcher test unit; TOPICAL

    International Nuclear Information System (INIS)

    The Office of Naval Research (ONR) and PMS512 have undertaken a program to develop a new Vertical Launching System (VLS) for future generation ships, such as the DD-21 Destroyer. The Naval Sea Systems Command Combat Weapons Program (NAVSEA 05K) and Naval Surface Warfare Center Dahlgren Division (NSWCDD) are working jointly with industry and universities to develop one such launcher design, the Concentric Canister Launcher (CCL). The basic CCL design consists of a tube made of two concentric cylinders; one end is open, the other is sealed with a hemispherical end cap. During firing, the missile exhaust gas is turned 180 degrees by the hemispherical end cap and flows through the annular space between inner and outer cylinders. Depending on the missile utilized and the particular service environment of the CCL, maximum temperatures within the cylinder material have been calculated to exceed 2000 F. In an earlier study[1], the authors determined the high temperature mechanical properties of several candidate alloys being considered for fabrication of the CCL. This study[1] found that, of these candidate materials, titanium alloys exhibit higher yield stresses than that of 316L stainless steel at temperatures up to about 1000 F; above 1500 F, the yield stress of 316L stainless steel is comparable to those of the titanium alloys. The 316L stainless steel was found to strain harden (increase its flow stress with increasing strain) at temperatures up to about 1800 F. The ability of the 316L stainless steel to strain harden at high temperatures may provide an added margin of safety for engineering design of the CCL. The objective of the current study was to perform a computer simulation of the structural response of a CCL during a restrained firing, one in which a SM-2 Blk IV missile would fail to exit the canister. A finite element model of the inner cylinder, outer cylinder, end rings (mounting brackets), and lateral restraints in the uptake was constructed. An elastic

  10. Temperature history for canistered fuel lag storage areas during the loss of cooling air at the receiving and handling building of the MRS Facility

    International Nuclear Information System (INIS)

    The Pacific Northwest Laboratory has analyzed the temperature history at a canistered fuel lag storage area during a postulated failure of the cooling system. A two-dimensional analysis was performed using the GE2D computer code, which accounts for thermal radiation, conduction, and convective heat transfer. The results indicate that the system may not reach a steady-state condition because heat transfer through the top and the bottom of the system is not enough to remove the energy generated in the canistered fuel. Although limits for abnormal operation have not been set, the temperatures do not reach limiting conditions for normal operation for 32 hours, which should be enough time to repair the cooling system

  11. Temperature history for canistered fuel lag storage areas during the loss of cooling air at the receiving and handling building of the MRS Facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D.

    1986-04-01

    The Pacific Northwest Laboratory has analyzed the temperature history at a canistered fuel lag storage area during a postulated failure of the cooling system. A two-dimensional analysis was performed using the GE2D computer code, which accounts for thermal radiation, conduction, and convective heat transfer. The results indicate that the system may not reach a steady-state condition because heat transfer through the top and the bottom of the system is not enough to remove the energy generated in the canistered fuel. Although limits for abnormal operation have not been set, the temperatures do not reach limiting conditions for normal operation for 32 hours, which should be enough time to repair the cooling system.

  12. SCOPING EVALUATION TO EXPLORE - ROCK FALL ACCIDENT CONDITION ANALYSIS ON MULTI-PURPOSE CANISTER WASTE PACKAGES CORRELATED FROM INTERLOCKING BASKET WASTE PACKAGE DESIGN ANALYSIS (SCPB: N/A)

    Energy Technology Data Exchange (ETDEWEB)

    Z, Ceylan

    1995-12-08

    The objective of this analysis is to correlate the results of a rock fall analysis performed for the 12 Pressurized Water Reactor (PWR) Fuel Assembly Interlocking Basket waste package (WP) in order to determine the size of rock that can strike the Multi-Purpose Canister (MPC) waste packages without breaching the containment barriers. The purpose of this analysis is to document the models and methods used in the calculations.

  13. 32-Week Holding-Time Study of SUMMA Polished Canisters and Triple Sorbent Traps Used To Sample Organic Constituents in Radioactive Waste Tank Vapor Headspace

    International Nuclear Information System (INIS)

    Two sampling methods[SUMMA polished canisters and triple sorbent traps (TSTs)] were compared for long-term storage of trace organic vapor samples collected from the headspaces of high-level radioactive waste tanks at the U.S. Department of Energy's Hanford Site in Washington State. Because safety, quality assurance, radiological controls, the long-term stability of the sampling media during storage needed to be addressed. Samples were analyzed with a gas chromatograph/mass spectrometer (GC/MS) using cryogenic reconcentration or thermal desorption sample introduction techniques. SUMMA canister samples were also analyzed for total non-methane organic compounds (TNMOC) by GC/flame ionization detector (FID) using EPA Compendium Method TO-12 . To verify the long-term stability of the sampling media, multiple samples were collected in parallel from a typical passively ventilated radioactive waste tank known to contain moderately high concentrations of both polar and nonpolar organic compounds. Analyses for organic analytes and TNMOC were conducted at increasing intervals over a 32-week period to determine whether any systematic degradation of sample integrity occurred. Analytes collected in the SUMMA polished canisters generally showed good stability over the full 32 weeks with recoveries at the 80% level or better for all compounds studied. The TST data showed some loss (50-80% recovery) for a few high-volatility compounds even in the refrigerated samples; losses for unrefrigerated samples were far more pronounced with recoveries as low as 20% observed in a few cases

  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. Measurements of the equilibrium factor and radon dose in some houses in Cairo, Egypt using activated charcoal canister

    International Nuclear Information System (INIS)

    Radon concentration was measured in some houses in Cairo, Egypt using activated charcoal canisters. Equilibrium factor and radon annual equivalent dose are measured as well . The measurements were performed in closed and open ventilation rooms. The average radon concentration was found to be 11.55 Bq/m3 and 24.93 Bq/m3 in open rooms and closed rooms, respectively. The deduced values of equilibrium factor were ranging from 0.171 to 0.180 with an average of 0.175 in open rooms and ranging from 0.190 to 0.205 with an average of 0.194 in closed rooms. The average radon annual equivalent dose was found to be o.111 mSv/y and 0.267 mSv/y in open rooms and closed ones, respectively. These values are much lower than the maximum permissible dose of 1.0 mSv/y recommended by ICRP-60-1990.

  16. 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.

  17. 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.

  18. Reliable evaluation of acceptability of weld for final disposal based on the canister copper weld inspection using different NDT methods

    International Nuclear Information System (INIS)

    The inspection of the sealing weld is an important phase for the evaluation of the acceptability of final disposal canister, but the weld is only a part of the 3D shielding of copper shell. The main tasks for reliable NDT evaluation requires an extensive evaluation of the parameters - which contains typical inspection related items like repeatability, S/N ratio, POD, setting up the equipment for inspection, and all practices for inspections. The other parameters are material parameters, their variation must be taken into account in the evaluation of NDT reliability. Further parameters include human factors, i. e. human inspectors and their interaction with technical systems; their effects were studied on an example of the evaluation of eddy current data. Final parameters are related to evaluation of detected defects, which means sizing and base for acceptance and this can be done in different ways. Some examples are given and results are compared with different methods for instance between radiographic testing and ultrasonic testing by raw data analysis and PA-SAFT results. Also, preliminary curves for the evaluation of metallographic results of 55 defects will be shown by EB weld measurements. Some practical items concerning copper inspections will be also discussed related to acceptability.

  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. Analyses of atmospheric radon 222 / canisters exposed by Greenpeace in Niger (Arlit / Akokan sector); Analyses du radon 222 dans l'air ambiant / Capteurs exposes par Greenpeace au Niger (secteur Arlit et Akokan)

    Energy Technology Data Exchange (ETDEWEB)

    Chareyron, B.

    2010-07-01

    The companies SOMAIR and COMINAK, subsidiaries of the AREVA group, are mining uranium deposits in northern Niger. In the course of a field mission carried out in November 2009, a Greenpeace International team deposited detectors (canisters of activated charcoal) to measure radon 222, a radioactive gas formed by the decay of the radium 226 present in the uranium ore. This report includes the results of the analysis of the activated charcoal canisters conducted in CRIIRAD's laboratory, and a brief commentary on the interpretation of the results. (authors)

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

    International Nuclear Information System (INIS)

    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 of what-if calculations could

  3. 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

  4. 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

  5. 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.

  6. 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.

  7. 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

  8. 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

  9. Inspection of copper canister for spent nuclear fuel by means of ultrasound. FSW monitoring with emission, copper characterization 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))

    2008-09-15

    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 2007. In the first part of the report we further develop the concept of monitoring of the friction stir welding (FSW) process by means of acoustic emission (AE) technique implemented using multiple sensors formed into a circular array. After a brief introduction into the field of arrays and beamforming we focus on the features of uniform circular arrays (UCA). Results obtained from the simulations of UCA beamformer based on phase mode concept are presented for the continuous wave as well as for the pulse, noise-free input signals. The influence of white noise corrupting the input pulse is also considered and a simple regularization technique proposed as a solution to this problem. The second part of the report is concerned with aspects related to ultrasonic attenuation of copper material used for canisters. We compare resonant ultrasound spectroscopy (RUS) with other methods used for characterization of the copper material. RUS is a non-destructive technique based on sensing mechanical resonances present in a tested sample in the ultrasonic frequency range. Resonance frequencies observed in a material sample (with given geometry) are directly related to the vibration modes occurring in the inspected volume defined by the material parameters (elastic constants). We solve the inverse problem that consists in using the information about resonance frequencies acquired in physical measurements for estimating material parameters. Our aim in this project is to investigate the feasibility of RUS for the grain size estimation in copper using copper specimens that were provided by SKB. In the final part we consider the design of input signals for ultrasonic arrays. The Bayesian linear minimum mean squared error (LMMSE) estimator discussed in our former reports is studied. We show that it

  10. 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.

  11. 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.

  12. 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

  13. 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.

  14. 新型复合滤料层个人防护氨气滤毒罐%The development of new pattern composite filter canister for ammonia individual safety

    Institute of Scientific and Technical Information of China (English)

    李建华

    2012-01-01

    以改性椰壳活性炭和强酸离子交换纤维为复合滤料层制备出针对氨气的新型个人防护滤毒罐.按照国标GB2890-2009对制备的滤毒罐动态吸附性能及呼吸阻力性能进行了测试.扫描电镜照片显示改性处理的椰壳活性炭具有孔套孔结构,使改性活性炭的比表面积增大很多.呼吸防护时间超过2级5 min.通气阻力还需要改进处理.%The new type personal protection canisters of compos-ite filtering layer for ammonia were prepared using modified co-conut shell activated carbon and acidic ion exchange fiber. The dynamic adsorption and respiratory resistance were measured according to the national standard GB 2890-2009. The results of surface electronic microscopy revealed that the modified coco-nut shell activated carbon has a hole sets of holes structure which increases the specific surface area of the modified activa-ted carbon. The respiratory protection time exceeded the second class by 5 min. Airway resistance need to be improvement.

  15. Inspection of copper canisters for spent nuclear fuel by means of ultrasound. Synthetic aperture imaging, evaluation of ultrasonic attenuation in copper

    Energy Technology Data Exchange (ETDEWEB)

    Stepinski, Tadeusz (ed.); Lingvall, Fredrik; Wennerstroem, Erik; Ping Wu [Uppsala Univ. (Sweden). Department of Engineering Sciences

    2006-01-15

    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 2004/2005. After a short introduction a new super-resolution synthetic aperture imaging (SRSAI) technique is proposed. The proposed SRSAI is characterized by an excellent lateral resolution and much higher signal to noise ratio compared with ESAFT which was proposed in our previous reports. The ESAFT is based on the assumption that probability density of the imaged targets (so called prior) is Gaussian, which is the simplest case for the analysis. The increased performance of SRSAI is due to more realistic assumption concerning probability density function of targets in the region of interest. It is shown, both using simulations and in an experiment that if the target reflectivity is assumed to be positive a substantial increase of resolution and signal to noise ratio in the reconstructed image can be obtained. In the second chapter the result of the evaluation of ultrasonic attenuation in copper blocks with different grain size is presented. A short presentation of basic theory of the buffer-rod and the immersion methods is given in the beginning (the detailed derivation is provided in the appendix). Then the measurement setup and copper specimens are specified and the results of measurements made on copper specimens are presented. Correlation between the attenuation and grain size for the inspected specimens is discussed.

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

    International Nuclear Information System (INIS)

    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 simulated

  17. 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

  18. 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)

  19. 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

  20. 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.

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

    International Nuclear Information System (INIS)

    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

  2. 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

  3. 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

  4. Neutron-sensitive ZnS/{sup 10}B{sub 2}O{sub 3} ceramic scintillator detector as an alternative to a {sup 3}He-gas-based detector for a plutonium canister assay system

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, T., E-mail: nakamura.tatsuya@jaea.go.jp [J-PARC, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan); Ohzu, A. [Nuclear Science and Engineering Directorate, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan); Toh, K.; Sakasai, K.; Suzuki, H.; Honda, K. [J-PARC, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan); Birumachi, A.; Ebine, M. [Nuclear Science Institute, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan); Yamagishi, H. [J-PARC, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan); Takase, M.; Haruyama, M.; Kureta, M. [Nuclear Science and Engineering Directorate, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan); Soyama, K. [J-PARC, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan); Nakamura, H. [Tokai Reprocessing Development Center, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan); Seya, M. [Integrated Support Center for Nuclear Nonproliferation and Nuclear Security, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan)

    2014-11-01

    A neutron-sensitive ZnS/{sup 10}B{sub 2}O{sub 3} ceramic scintillator detector was developed as an alternative to a {sup 3}He-gas-based detector for use in a plutonium canister assay system. The detector has a modular structure, with a flat ZnS/{sup 10}B{sub 2}O{sub 3} ceramic scintillator strip that is installed diagonally inside a light-reflecting aluminium case with a square cross-section, and where the scintillation light is detected using two photomultiplier tubes attached at both ends of the case. The prototype detectors, which have a neutron-sensitive area of 30 mm×250 mm, exhibited a sensitivity of 21.7–23.4±0.1 cps/nv (mean±SD) for thermal neutrons, a {sup 137}Cs gamma-ray sensitivity of 1.1–1.9±0.2×10{sup −7} and a count variation of less than 6% over the detector length. A trial experiment revealed a temperature coefficient of less than −0.24±0.05%/°C over the temperature range of 20–50 °C. The detector design and the experimental results are presented.

  5. Neutron-sensitive ZnS/10B2O3 ceramic scintillator detector as an alternative to a 3He-gas-based detector for a plutonium canister assay system

    Science.gov (United States)

    Nakamura, T.; Ohzu, A.; Toh, K.; Sakasai, K.; Suzuki, H.; Honda, K.; Birumachi, A.; Ebine, M.; Yamagishi, H.; Takase, M.; Haruyama, M.; Kureta, M.; Soyama, K.; Nakamura, H.; Seya, M.

    2014-11-01

    A neutron-sensitive ZnS/10B2O3 ceramic scintillator detector was developed as an alternative to a 3He-gas-based detector for use in a plutonium canister assay system. The detector has a modular structure, with a flat ZnS/10B2O3 ceramic scintillator strip that is installed diagonally inside a light-reflecting aluminium case with a square cross-section, and where the scintillation light is detected using two photomultiplier tubes attached at both ends of the case. The prototype detectors, which have a neutron-sensitive area of 30 mm×250 mm, exhibited a sensitivity of 21.7-23.4±0.1 cps/nv (mean±SD) for thermal neutrons, a 137Cs gamma-ray sensitivity of 1.1-1.9±0.2×10-7 and a count variation of less than 6% over the detector length. A trial experiment revealed a temperature coefficient of less than -0.24±0.05%/°C over the temperature range of 20-50 °C. The detector design and the experimental results are presented.

  6. 同心筒水下发射筒口气泡变化的数值模拟%Numerical Simulation of Change of Bubble in Tube Opening Underwater Launch Using Concentric Canister Structure

    Institute of Scientific and Technical Information of China (English)

    邓佳; 毕世华; 李景须

    2015-01-01

    为了研究同心筒水下热发射过程中筒口气泡变化规律,采用三维多相流模型对发射过程进行了模拟。研究表明,同心筒结构应用于水下发射时筒口气泡受到弹体表面黏性力、气流附壁效应以及两相互相作用过程影响,筒口气泡的形态会经历3个典型阶段,筒口附近的压强和速度受气泡运动和发展影响而振荡变化。研究的结果在水下发射技术的发展上具有一定的理论意义和工程应用价值。%In order to study the development change rule of bubble in tube opening during underwater launching process using concentric canister structure,three-dimensional multiphase flow model was used to simulate the process of launching. The results indicate that when concentric tube structure is applied to the underwater launch,bubbles at concentric tube edge are affected by projectile surface viscous force,airflow coanda effect and the above two phase’s process between each other,and the shape of jet bubble experi-enced three typical forms. Pressure and velocity inside the bubbles oscillated and changed for the motion and development of bubble. Results of this study have theoretical significance and engineering value in the development of underwater launch technology.

  7. Multi-canister overpack design report

    International Nuclear Information System (INIS)

    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

  8. 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,...

  9. 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 .

  10. Study on the Corrosion of Shang and Zhou Dynasty Bronze Relics---A Case Study of Artificial Analogue of HLW Disposal System Canister%商周青铜文物的腐蚀研究--高放废物处置系统人为类似物研究实例

    Institute of Scientific and Technical Information of China (English)

    陈璋如; 刘月妙; 范光; 温志坚; 孙淑云; 李延祥; 张展适

    2015-01-01

    This paper presents compositional analyses and structure observations of unearthed Shang Zhou dynasty bronze relict samples and environmental conditions of the sites where the samples were collected.Research results revealed that the bronze relics had been continuously corroded with time.The corrosion layer consisted of at least two sub layers ,such as oxide and carbonate.There was a mechani‐cal corrosion sub layer found in few samples (a loose sub layer) . A SnO2 sub layer also occurred in a few samples. The thickness of bronze corrosion layers varied with the different climate environments in which the samples were taken. Their thickness was 50~260μm and 320~800μm and humid climate re‐gions respectively. In the view of natural corrosion resistance , bronze is recommendable as a material for HLW repository system canisters.%文章论述了出土商周时代青铜文物的化学成分和结构特征以及样品采集地区的环境条件。研究表明青铜文物随时间而不断腐蚀。腐蚀层至少由两个亚层组成,即氧化物层和碳酸盐亚层。有些样品出现物理腐蚀亚层,即疏松层;有些样品则产出SnO2亚层。青铜腐蚀层的厚度在不同环境条件下发生变化,干旱条件下为50~260μm ,潮湿条件下为320~800μm。从天然抗腐蚀观点看,青铜可推荐作为高放废物处置系统废物罐的材料。

  11. Mechanical performance and life prediction for canister copper

    Energy Technology Data Exchange (ETDEWEB)

    Rantala, J.; Auerkari, P.; Salonen, J.; Holmstroem, S.; Laukkanen, A. (VTT Technical Research Centre of Finland, Espoo (Finland)); Saukkonen, T. (Aalto Univ. School of Science and Technology, Espoo (Finland))

    2010-05-15

    The creep response of the OFP copper at 175 deg C has been studied by uniaxial, multiaxial creep testing, metallography and surface inspections. Surface cracks have been observed in a uniaxial test which has been running for more than 70 000 h (eight years). In a CT-specimen creep cavities have been observed after an exposure of 25 000 h. A purpose-built creep strain model has been implemented into FE analysis to characterize the stress multiaxiality and the stress state of the CT specimen. The results have been applied for life prediction of copper at foreseen levels of stress and temperature at the repository. (orig.)

  12. Modeling void growth and movement with phase change in thermal energy storage canisters

    Science.gov (United States)

    Darling, Douglas; Namkoong, David; Skarda, J. R. L.

    1993-01-01

    A scheme was developed to model the thermal hydrodynamic behavior of thermal energy storage salts. The model included buoyancy, surface tension, viscosity, phases change with density difference, and void growth and movement. The energy, momentum, and continuity equations were solved using a finite volume formulation. The momentum equation was divided into two pieces. The void growth and void movement are modeled between the two pieces of the momentum equations. Results showed this scheme was able to predict the behavior of thermal energy storage salts.

  13. Fluid Phase Separation (FPS) experiment for flight on a space shuttle Get Away Special (GAS) canister

    Science.gov (United States)

    Peters, Bruce; Wingo, Dennis; Bower, Mark; Amborski, Robert; Blount, Laura; Daniel, Alan; Hagood, Bob; Handley, James; Hediger, Donald; Jimmerson, Lisa

    1990-01-01

    The separation of fluid phases in microgravity environments is of importance to environmental control and life support systems (ECLSS) and materials processing in space. A successful fluid phase separation experiment will demonstrate a proof of concept for the separation technique and add to the knowledge base of material behavior. The phase separation experiment will contain a premixed fluid which will be exposed to a microgravity environment. After the phase separation of the compound has occurred, small samples of each of the species will be taken for analysis on the Earth. By correlating the time of separation and the temperature history of the fluid, it will be possible to characterize the process. The experiment has been integrated into space available on a manifested Get Away Special (GAS) experiment, CONCAP 2, part of the Consortium for Materials Complex Autonomous Payload (CAP) Program, scheduled for STS-42. The design and the production of a fluid phase separation experiment for rapid implementation at low cost is presented.

  14. Feasibility study of electron beam welding of spent nuclear fuel canisters

    International Nuclear Information System (INIS)

    A thick walled copper container is presently the prime Swedish alternative for encapsulation of spent nuclear fuel. In order to demonstrate the feasibility of encapsulating high-level nuclear waste in copper containers, a study of electron beam welding of thick copper has been performed. Two copper qualities have been investigated, oxygen free high conductivity (OFHC) copper and phosphorous desoxydized high conductivity copper (PDO). The findings in this study are summarized below. In 100 mm thick copper penetration can be achived at power level of about 75 kW (typically 150 kV x 500 mA) at welding speed of 100 mm/min. The welds in OFHC copper made under these conditions are free from major defects during constant welding conditions. The welds in PDO copper show a microporosity level considerably higher than those in OFHC copper, but no major defects are produced in the welds in PDO copper. In the ending of the weld (ie the fade out) it is still not possible to completely eliminate root and cold-shut defects. A semi-full-scale lid weld has been performed successfully. Automatic ultrasonic C-scan has been shown to be useful in detecting and displaying defects, but some problems still remain with defect sizing. The different speciments of OFHS copper had different attenuation of the ultrasonic signal, forged copper showing a far lower attenuation than hot extruded copper, indicating that attention must be paid in choosing copper that allows accurate ultrasonic testing. Resiudal stresses in the welded zone has been measured and are found to lie in the range -32N/mm2 to +36N/mm2. The peak stress was less than half the assumed value of the proof stress of the fused metal. (authors)

  15. Impact of Aluminum on Anticipated Corrosion in a Flooded spent nuclear fuel Multi -Canister Overpack

    International Nuclear Information System (INIS)

    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

  16. 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.

  17. 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.

  18. 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.

  19. Radiation effects on polymers for coatings on copper canisters used for the containment of radioactive materials

    International Nuclear Information System (INIS)

    The present work proposes applying polyurethane coatings as an additional barrier in the design of Canadian nuclear waste disposal containers. The goal of the present research is to investigate the physico-mechanical integrity of a natural castor oil-based polyurethane (COPU) to be used as a coating material in pH-radiation-temperature environments. As the first part to these inquiries, the present paper investigates the effect of a mixed radiation field supplied by a SLOWPOKE-2 nuclear research reactor on COPUs that differ only by their isocyanate structure. FTIR, DSC, DMA, WAXS, and MALDI are used to characterize the changes that occur as a result of radiation and to relate these changes to polymer structure and composition. The COPUs used in the present work have demonstrated sustained physico-mechanical properties up to accumulated doses of 2.0 MGy and are therefore suitable for end-uses in radiation environments such as those expected in the deep geological repository

  20. Radiation effects on polymers for coatings on copper canisters used for the containment of radioactive materials

    Science.gov (United States)

    Mortley, Aba; Bonin, H. W.; Bui, V. T.

    2008-05-01

    The present work proposes applying polyurethane coatings as an additional barrier in the design of Canadian nuclear waste disposal containers. The goal of the present research is to investigate the physico-mechanical integrity of a natural castor oil-based polyurethane (COPU) to be used as a coating material in pH-radiation-temperature environments. As the first part to these inquiries, the present paper investigates the effect of a mixed radiation field supplied by a SLOWPOKE-2 nuclear research reactor on COPUs that differ only by their isocyanate structure. FTIR, DSC, DMA, WAXS, and MALDI are used to characterize the changes that occur as a result of radiation and to relate these changes to polymer structure and composition. The COPUs used in the present work have demonstrated sustained physico-mechanical properties up to accumulated doses of 2.0 MGy and are therefore suitable for end-uses in radiation environments such as those expected in the deep geological repository.

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

    International Nuclear Information System (INIS)

    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

  2. Topical safety analysis report for the transportation of the NUHOMS reg-sign dry shielded canister

    International Nuclear Information System (INIS)

    A thermal analyses for the 10CFR71 normal conditions of . transport and hypothetical accident conditions is presented for the conceptual NUHOMS reg-sign Transportation Cask and NUHOMS reg-sign-24P DSC system. The purpose of the thermal analyses presented herein is to demonstrate that the conceptual NUHOMS reg-sign Transportation Cask with the NUHOMS reg-sign-24P DSC provides suitable heat dissipation to maintain the heat removal capacity of the loaded NUHOMS reg-sign Transportation Cask. The thermal analyses results show that the maximum temperatures and pressures of the NUHOMS reg-sign Transportation Cask and the NUHOMS reg-sign-24P DSC are within their allowable material temperature and pressure limits

  3. Fluid Phase Separation (FPS) experiment for flight on the shuttle in a Get Away Special (GAS) canister: Design and fabrication

    Science.gov (United States)

    1990-01-01

    The separation of fluid phases in microgravity environments is of importance to environmental control and life support systems (ECLSS) and materials processing in space. A successful fluid phase separation experiment will demonstrate a proof of concept for the separation technique and add to the knowledge base of material behavior. The phase separation experiment will contain a premixed fluid that will be exposed to a microgravity environment. After the phase separation of the compound has occurred, small samples of each of the species will be taken for analysis on Earth. By correlating the time of separation and the temperature history of the fluid, it will be possible to characterize the process. The phase separation experiment is totally self-contained, with three levels of containment on all fluids, and provides all necessary electrical power and control. The controller regulates the temperature of the fluid and controls data logging and sampling. An astronaut-activated switch will initiate the experiment and an unmaskable interrupt is provided for shutdown. The experiment has been integrated into space available on a manifested Get Away Special (GAS) experiment, CONCAP 2, part of the Consortium for Materials Complex Autonomous Payload (CAP) Program, scheduled for STS 42 in April 1991. Presented here are the design and the production of a fluid phase separation experiment for rapid implementation at low cost.

  4. 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.

  5. Additional guidance for including nuclear safety equivalency in the Canister Storage Building and Cold Vacuum Drying Facility final safety analysis report

    Energy Technology Data Exchange (ETDEWEB)

    Garvin, L.J.

    1997-05-20

    This document provides guidance for the production of safety analysis reports that must meet both DOE Order 5480.23 and STD 3009, and be in compliance with the DOE regulatory policy that imposes certain NRC requirements.

  6. Feasibility of disposal of high-level radioactive waste into the seabed. Volume 8: Review of processes near a buried waste canister

    International Nuclear Information System (INIS)

    One of the options suggested for disposal of high-level radioactive waste resulting from the generation of nuclear power is burial beneath the deep ocean floor in geologically stable sediment formations which have no economic value. The 8-volume series provides an assessment of the technical feasibility and radiological safety of this disposal concept based on the results obtained by ten years of co-operation and infomation exchange among the Member countries participating in the NEA Seabed Working Group. This report investigates the phenomena arriving in the proximity of the waste package immersed in the sea sediments

  7. 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.

  8. 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.

  9. TVO's new encapsulation method for spent nuclear fuel

    International Nuclear Information System (INIS)

    Teollisuuden Voima Oy has developed a new encapsulation method for spent nuclear fuel s.c. cold process. Instead of casting (400 deg C) molten lead the new canister is filled with cold granulated material like quartz sand, lead shots or glassa beads. The new canister concept ACPC (Advanced Cold Process Canister) consists of the outer oxygen free copper canister and of the inner steel canister. The function of the steel canister is merely to give mechanical strength. The copper canister acts as corrosion protection guaranteeing practically lifetime of millions of years for the ACPC concept

  10. 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.

  11. 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.

  12. 碎裂开盖鱼雷发射箱内流场变化研究%Research on Flow Field Change Inside of the Fragmenting Opening Torpedo Launch Canister

    Institute of Scientific and Technical Information of China (English)

    潘登; 赵修平; 姜守辉

    2015-01-01

    利用计算流体力学的仿真软件FLUENT对鱼雷发射箱内流场变化情况进行仿真计算,通过仿真计算分析发射箱内的压强分布规律及其对易碎裂盖体的影响.仿真结果表明,后盖在助推器射流直接冲击下会受到极不均匀的压强作用,可能使后盖中心破裂而不碎;前盖受后盖反射的射流影响而开盖.研究结果对发射箱的碎裂开盖方案设计有一定的指导意义.%The change of the flow field inside a torpedo launch canster was simulated and calculated by the CFD software FLUENT, and the pressure distribution regular pattern inside the launch canster and its influence over the fragile lids were analyszed through simulation and calculation. The simulation result indicated that the back lid got an extremely inhomoge?neous pressure effect under the direct impact of the jet flow of the roll booster, it's possible to fracture the back lid at the center but did not fragment it;the front lid opened under the influence of the jet flow which reverberated by the back lid. The result of the study has particular guidance significance over the design of the scheme of fragmenting to open the lids.

  13. 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

    International Nuclear Information System (INIS)

    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

  14. 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

    International Nuclear Information System (INIS)

    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

  15. Les enjeux de la professionnalisation des agents de développement. L'ingénierie territoriale prise en étau entre les conceptions organique et mécaniste du développement territorial

    OpenAIRE

    Dany Lapostolle

    2011-01-01

    This article connects the regulation mechanisms of territorial public action, to issues related to the professionalization of development agents, who are pivotal actors of territorial engineering. By entering into epistemic communities, they pave the way for a pragmatic regulation of public action. However, the competitive register of territorial excellence, coupled with new forms of bureaucratic controls, transform the “project territories” into segments of policies implementation set by com...

  16. 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.

  17. 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.

  18. 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.

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

    International Nuclear Information System (INIS)

    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

  20. Evaluation of special safety issues associated with handling the Three Mile Island Unit 2 core debris

    International Nuclear Information System (INIS)

    This document reports the results of recent tests and analyses evaluating safety concerns relating to Three Mile Island Unit 2 (TMI-2) core debris pyrophoricity, radiolytic hydrogen and oxygen, and the potential for steam generation in shipping canisters during a fire. Recommendations drawn from these results include the following: (1) hydrogen-oxygen recombiners should be installed in each core debris canister, (2) water should be removed from each canister by drip drying (no vacuum pumping is required), (3) the maximum weight of the loaded, dewatered canisters and the minimum volume of gas/vapor in each canister should be controlled and measured by weighting before and after dewatering, (4) a cover gas of approximately two atmospheres of argon should be added to each canister, (5) each canister should be weighed and pressure checked prior to shipping, (6) the shipping cask should be designed to limit the temperature of the canister contents after the standard hypothetical accident (fire) such that the design pressure of the canister/cask will not be exceeded, (7) provisions should be made for canister venting during long-term storage and for cask venting in the event of an overpressure condition resulting from an ''extended'' fire, and (8) some pyrophoricity testing of samples taken during defueling should be conducted to assure adequate safety-related information during canister opening

  1. Dose Rate Calucaltion for the DHL W/DOE SNF Codisposal Waste Package

    Energy Technology Data Exchange (ETDEWEB)

    G. Radulescu

    2000-02-12

    The purpose of this calculation is to determine the surface dose rates of the short codisposal waste package (WP) of defense high-level waste (DHLW) and TRIGA (Training, Research, Isotopes, General Atomics) spent nuclear fuel (SNF). The WP contains the TRIGA SNF, in a standardized 18-in. DOE (U.S. Department of Energy) SNF canister, and five 3-m-long Savannah River Site (SRS) DHLW pour glass canisters, which surround the DOE SNF canister.

  2. 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.

  3. Thermal analysis of VWSB-IP1 at Tarapur

    Energy Technology Data Exchange (ETDEWEB)

    Verma, V.; Singh, R.K.; Vaze, K.K. [Bhabha Atomic Research Centre, Trombay, Mumbai (India). Reactor Safety Div.

    2013-11-15

    High Level Liquid Radioactive Waste (HLLRW) produced during reprocessing of spent fuel from nuclear reactors is encased in the canisters after vitrification. The vitrified waste has high heat generation rate due to decay heat and needs interim storage under surveillance. The waste needs to be cooled continuously until major portion of the decay heat is dissipated. Natural circulation air cooling has been considered to cool the canisters. Canisters are placed in a storage vault and cooled by induced axial flow of air with the help of stack. The capacity of storage vault for Vitrified Waste Storage Block (VWSB) Facility proposed at Integrated Plant-1, Tarapur is designed for interim storage of waste generated of 30 yrs of IP1 plant operation. Canister and concrete temperature should be within the prescribed limit. Parametric studies have been carried out for the relevant parameters such as stack and duct dimensions, plenum height etc. Details canister temperature have been obtained using CFD code CFD-ACE+. Axial and radial temperature variation in the canisters, thimble and ventilation pipe have been evaluated in a location. Effect of natural convection (in air) within the canister and between thimble and canister is also studied. It was found that canister centerline temperature reduces by 20 C. (orig.)

  4. Test report for K Basin MK I lid removal and replacement system

    Energy Technology Data Exchange (ETDEWEB)

    Omberg, R.P.; Roe, N.R. [Westinghouse Hanford Co., Richland, WA (United States)

    1996-08-21

    This report provides the results of acceptance testing of sampling equipment for use in the Hanford K Basin. The equipment, MK I Lid Removal/Replacement Tools, were designed to remove/replace MK I Spent Fuel Canister lids so that other equipment may be used to sample the canister contents. The tools were determined to be acceptable for their intended use.

  5. Commercial Nuclear Waste Research and Development Program. Annual report, fiscal year 1984

    International Nuclear Information System (INIS)

    This document is a report of fiscal year 1984 activities performed to meet task objectives of the Nevada Nuclear Waste Storage Investigations (NNWSI) Project Plan. Noteworthy accomplishments were: operation of a fuel temperature test which evaluates fuel performance in an environment expected for metal cask storage. This test is being performed for the DOE Richland Operations Dry Storage Fuel Integrity Demonstration Program; characterization of fuel assemblies and storage canisters, which includes fuel assembly visual inspection, videotape and still photography, and contamination swipe sampling; and canister gas and full volume filtration sampling, contamination swipe sampling, and residue collection; removal of fuel assemblies from seal welded storage canisters, using the E-MAD canister cutter; integrity monitoring of fuel assemblies in seal welded canisters, by sampling and analysis of canister cover gas; and fuel assemblies in unwelded canisters, by collection and analysis of fuel surface contamination swipe samples and canister residue; emplacement/removal of canisterized fuel assemblies in the E-MAD drywells; and preparation of a summary storage history of all fuel assemblies at the E-MAD facility, including FY 1983 activities involving those fuel assemblies

  6. 75 FR 43225 - Culturally Significant Objects Imported for Exhibition Determinations: “The Holocaust-Uniforms...

    Science.gov (United States)

    2010-07-23

    ... Culturally Significant Objects Imported for Exhibition Determinations: ``The Holocaust--Uniforms, Canisters... ``The Holocaust--Uniforms, Canisters, and Shoes,'' imported from abroad for temporary exhibition within... objects as part of the permanent exhibit at the U.S. Holocaust Memorial Museum, Washington, DC, from on...

  7. 25 CFR 542.41 - What are the minimum internal control standards for drop and count for Tier C gaming operations?

    Science.gov (United States)

    2010-04-01

    ...) The count of each box shall be recorded in ink or other permanent form of recordation. (ii) A second... canister has been recorded. (i) The count of each canister shall be recorded in ink or other permanent form... weigh/count unless a weigh scale with a printer is used. (6) The gaming machine drop shall be...

  8. 25 CFR 542.21 - What are the minimum internal control standards for drop and count for Tier A gaming operations?

    Science.gov (United States)

    2010-04-01

    ... of each box shall be recorded in ink or other permanent form of recordation. (ii) A second count... canister has been recorded. (i) The count of each canister shall be recorded in ink or other permanent form... perform the initial recording of the weigh/count unless a weigh scale with a printer is used. (6)...

  9. 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

  10. Degradation and Failure of Stored Radiological Material Containers and Packages, Idaho National Engineering and Environmental Laboratory, United States of America

    International Nuclear Information System (INIS)

    On 3 December 2003, at the Naval Reactors Facility, Idaho National Engineering and Environmental Laboratory, in the United States of America, a canister containing an irradiated non-fuel test specimen failed catastrophically while stored in a water pool. The failure made a large noise, dislodged the stainless steel canister made from 10 cm diameter schedule 40 pipe 45 cm long, ruptured its brass cap, and projected part of the cap 3 m away underwater. No injuries or other damage occurred and there was no measurable release of radioactivity to the environment. The brass cap screwed onto the canister, with two nitrile rubber O-rings providing a watertight seal. Investigators found evidence of water leakage inside the canister. Their preliminary conclusion is that during the 14 years the canister was stored in the water pool, the nitrile rubber seals degraded from exposure to high flux gamma radiation emitted from the test specimen. Water leaked into the canister and the canister subsequently resealed tightly as a result of the brass cap’s corrosion. Radiolysis caused the captured water to break down into hydrogen and oxygen gas, pressurizing the canister. (Decomposition of the nitrile rubber could also generate flammable gases.) The investigators concluded that the hydrogen detonated and caused the failure. Although the ignition source is still not clear, it could have been thermal energy from the specimen, reactions from radicals produced by the radiolysis, sparking from interaction of metallic components, or static electricity discharge

  11. Foreign encapsulation concepts of spent fuel

    International Nuclear Information System (INIS)

    The foreign encapsulation concepts of spent fuel in countries, which have geologies similar to that in Finland are reviewed. The main interests in this report were alternative concepts of canister materials as well as the design, fabrication and testing programs planned to evaluate the canister performance. Also present concepts of mined geological repositories and facilities for waste handling before final disposal are reviewed. (author)

  12. Defense Waste Processing Facility -- Radioactive operations -- Part 3 -- Remote operations

    International Nuclear Information System (INIS)

    The Savannah River Site's Defense Waste Processing Facility (DWPF) near Aiken, South Carolina is the nation's first and world's largest vitrification facility. Following a ten year construction period and nearly three years of non-radioactive testing, the DWPF began radioactive operations in March 1996. Radioactive glass is poured from the joule heated melter into the stainless steel canisters. The canisters are then temporarily sealed, decontaminated, resistance welded for final closure, and transported to an interim storage facility. All of these operations are conducted remotely with equipment specially designed for these processes. This paper reviews canister processing during the first nine months of radioactive operations at DWPF. The fundamental design consideration for DWPF remote canister processing and handling equipment are discussed as well as interim canister storage

  13. Interaction of DOE SNF and Packaging Materials

    International Nuclear Information System (INIS)

    A sensitivity analysis was conducted to identify and evaluate potential destructive interactions between the materials in US Department of Energy (USDOE) spent nuclear fuels (SNFs) and their storage/disposal canisters. The technical assessment was based on the thermodynamic properties as well as the chemical and physical characteristics of the materials expected inside the canisters. No chemical reactions were disclosed that could feasibly corrode stainless steel canisters to the point of failure. However, the possibility of embrittlement (loss of ductility) of the stainless steel through contact with liquid metal fission products or hydrogen inside the canisters cannot be dismissed. Higher-than-currently-permitted internal gas pressures must also be considered. These results, based on the assessment of two representative 90-year-cooled fuels that are stored at 200C in stainless steel canisters with internal blankets of helium, may be applied to most of the fuels in the USDOE's SNF inventory

  14. Interaction of DOE SNF and Packaging Materials

    Energy Technology Data Exchange (ETDEWEB)

    P. A. Anderson

    1998-09-01

    A sensitivity analysis was conducted to identify and evaluate potential destructive interactions between the materials in US Department of Energy (USDOE) spent nuclear fuels (SNFs) and their storage/disposal canisters. The technical assessment was based on the thermodynamic properties as well as the chemical and physical characteristics of the materials expected inside the canisters. No chemical reactions were disclosed that could feasibly corrode stainless steel canisters to the point of failure. However, the possibility of embrittlement (loss of ductility) of the stainless steel through contact with liquid metal fission products or hydrogen inside the canisters cannot be dismissed. Higher-than-currently-permitted internal gas pressures must also be considered. These results, based on the assessment of two representative 90-year-cooled fuels that are stored at 200°C in stainless steel canisters with internal blankets of helium, may be applied to most of the fuels in the USDOE's SNF inventory.

  15. Spent fuel dry storage technology development: electrically heated drywell storage test (1kW and 2kW operation)

    International Nuclear Information System (INIS)

    The simulated drywell cell consists of a representative stainless steel spent fuel canister containing an electrical heater assembly, a concrete-filled shield plug to which the canister is attached, and a carbon steel liner that encloses the canister and shield plug. The entire test drywell is grouted into a hole drilled in the soil adjacent to the Engine Maintenance Assembly and Disassembly. Temperature instrumentation is provided on the canister and drywell liner, in the grout around the liner, and at a number of radial locations in the soil surrounding the drywell. Peak measured canister and liner temperatures are 276 and 2320F for 1.0 kW and 510 and 4580F for 2.0kW, respectively. A computer model was developed to predict the thermal response of the test configuration. Computer predictions of the transient and steady-state temperatures of the drywell components and surrounding soil show good agreement with the test data

  16. Estimation of the corrosion resistance of materials intended for enclosure of nuclear fuel waste. State of the art report 1977-09-27 and supplementary remarks

    International Nuclear Information System (INIS)

    Within the KBS project the Swedish Corrosion Institute has got the task to evaluate the corrosion resistance of different materials proposed to be used in canisters for nuclear waste. For this purpose the Institute appointed a reference group of specialists, mainly within the fields of corrosion and materisl. KBS has proposed three different alternatives of canisters: 1. Titanium lined with lead as a canister material for reprocessed and vitrified waste. 2. Copper as a canister material for direct disposal of spent fuel. 3. Alumina as a canister material for direct disposal of spent fuel. For further evaluations the Institute points out the need of complementary investigations, e.g. on variations in ground water composition at a depth of 500 m, especially the content of oxygen, chloride, nitrite, sulphate and organic matter. (author)

  17. Concept of BN-350 reactor spent fuel handling during its storage after shutdown

    International Nuclear Information System (INIS)

    Full text: According to the Kazakhstan Government Decree (456; Apr 22, 1999), the fast BN-350 reactor has been shutdown in 1999 and the plan of its decommissioning has been started. By a decision of the Kazakhstan Government is defined that its spent fuel must be placed into 50 year's long-term safe storage with following dismantling and final disposal. This plan most important part is the concept of the spent fuel handling during its storage after shutdown. Next main stages of this concept are discussed here: discharge of the fuel, spent fuel packaging into special canisters, temporary storage of these canisters in the rector pool, to choose the site for long-term spent fuel storage and transport cask, making a choice of safety technique for canisters dry storage during 50 years. At first stage the fuel has been removed from the core and placing into reactor pool. The second stage lasted practically about two years. During this period all spent fuel has been packaged into special sealed canisters filled with inert gas. On next stage the site outside of the reactor one for long-term storage of these canisters and the way to ship them into the casks were chosen. The selected site is placed within the territory of former Semipalatinsk nuclear site. But the casks for spent fuel ought to be shipping as by rail and trailer vehicles too. Until its shipping will be started the canisters have been temporary stored under the water into BN-350 reactor pool. The following step is to define the transport cask design and the way to store the canisters at the site during long-term period about 50 years. There are two projects, which have been under consideration. The first is to use single-canister iron cask designed for shipping only. In this case the canisters with spent fuel ought to be transporting to the site into these casks and then transshipped into special 'silo'-type storage. Each canister has to be placed into individual near surface silo. The goal of this way is to

  18. Notice of construction for the 105 KE encapsulation activity

    International Nuclear Information System (INIS)

    This is a Notice of Construction for the 105-KE Basin Fuel Encapsulation Activity. This Notice of Construction is being submitted to the State of Washington Department of Health pursuant to the Washington Administrative Code, 246-247, ''Radiation Protection--Air Emissions,'' as amended. The encapsulation activity will consist of all of the activities described in Section 1.3 as necessary to complete encapsulation of the fuel elements and sludge contained in the 105-KE Basin. The encapsulation activity will be carried out on the 105-KE Basin floor under a minimum of 3 m (10 ft) of water. The encapsulation activity.includes emptying irradiated fuel elements from the existing canisters stored in the 105-KE Basin, packaging.these fuel elements in new canisters, packaging sludge from previous fuel inventories and the current inventory in new canisters, capping these canisters of fuel elements and sludge, and returning the new canisters to the storage racks in the basin. The empty canisters will be cleaned and crushed underwater. Packaging and disposal of the crushed canisters will take place above the water

  19. Non-LWR and special LWR spent fuels: Characteristics and criticality aspects of packaging and disposal

    International Nuclear Information System (INIS)

    Two important categories of potential repository wastes, non-LWR spent fuels and special LWR spent fuels, were evaluated regarding their future quantities and packaging requirements, including criticality aspects. Assuming that canisters similar to those planned for vitrified high-level waste and conventional LWR fuel assemblies prove to be acceptable for these fuels, it is estimated that about 1400 such canisters will be needed for the non-LWR and special LWR spent fuels on hand and projected through the year 2020. Approximately half of these canisters are for HTGR spent fuel, a quarter are for TMI-2 spent fuel and core debris, and the balance are for fuels from a score of diverse sources. About two thirds of these spent fuels can be accommodated in extended vitrified HLW-size canisters and the remainder will fit into canisters proposed for use in a tuff repository. Since many of these fuels are of higher enrichment and/or lower burnup than standard LWR spent fuel, estimation of the number of fuel assemblies that could be safely placed in a canister required preliminary calculations of neutron multiplication factors to ensure non-criticality. These estimates showed that the canister configurations used would provide more than adequate volume for the neutron poisons needed for this purpose. 36 refs., 19 figs., 20 tabs

  20. 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.

  1. Effects of brine migration on waste storage systems. Final report

    International Nuclear Information System (INIS)

    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 10-7 m s-1 (13 m y-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 10-4 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

  2. Effects of fuel burn-up and cooling periods on thermal responses in a repository for spent nuclear fuels

    International Nuclear Information System (INIS)

    Flexible methods and codes have been developed to calculate thermal responses in a spent nuclear fuel repository located in granitic bedrock. The thermal resistance of the backfill material makes an essential contribution to the temperature at the canister surface. The backfill is however so thin compared to the rock masses of the repository that a stationary solution for the temperature drop across the backfill can be used. Combining analytical solutions of thermal diffusion in the rock and the stationary temperature following the released heat power across the backfill gives a very good description of the thermal behaviour of a repository. Temperatures at different points in a repository were calculated for encapsuled spent fuel with different cooling times (10, 20, 30 and 40 years) and having different burn-up values (33, 35 and 45 MWd/kgU). The amount of spent fuel in each canister was supposed to originate from 1.4 tons of fresh Uranium in the calculated examples. A more crucial variable in a thermal analysis of a repository configuration is however the heat power of canisters as a function of time and not the content of canisters. The study shows that the largest temperature rise at the boundary of a centrally located canister and buffer mass will be at most 85 K with assumed thermal data, when the heat power of 900 canisters does not exceed 1200-1400 W each at the time of emplacement depending on burn-up and cooling time. Assuming a higher value for the heat conductivity of the buffer material corresponding to that of the water saturated bentonite gives 15% higher limits respectively. The canisters were supposed to be placed in 15 rows 25 m apart from each other and 60 canisters in each row. The distance between canisters was taken to be 6 m

  3. K East basin sludge volume estimates for integrated water treatment system

    International Nuclear Information System (INIS)

    This document provides estimates of the volume of sludge expected from Integrated Process Strategy (IPS) processing of the fuel elements and in the fuel storage canisters in K East Basin. The original estimates were based on visual observations of fuel element condition in the basin and laboratory measurements of canister sludge density. Revision 1 revised the volume estimates of sludge from processing of the fuel elements based on additional data from evaluations of material from the KE Basin fuel subsurface examinations. A nominal Working Estimate and an upper level Working Bound is developed for the canister sludge and the fuel wash sludge components in the KE Basin

  4. 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.

  5. Criticality safety evaluation report for spent nuclear fuelprocessing and storage facilities

    Energy Technology Data Exchange (ETDEWEB)

    Schwinkendorf, K.N., Fluor Daniel Hanford

    1997-03-24

    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 either 54 Mark IV or 48 Mark IA fuel assemblies. The criticality evaluations include loading baskets into the MCO/Cask, operations at the Cold Vacuum Drying Facility (CVDF), and storage in the Canister Storage Building (CSB). Many conservatisms have been built into this analysis, the primary one being the selection of the k{sub eff} @ 0.95 criticality safety limit.

  6. Possible effects of external electrical fields on the corrosion of copper in bentonite

    Energy Technology Data Exchange (ETDEWEB)

    Taxen, Claes (Swerea KIMAB (Sweden))

    2011-12-15

    External potentials that develop across a repository may interact with the copper canister. A study was undertaken to investigate the potential corrosion effects of voltage differences in a repository. A set of experiments was performed to study the tendency of copper in bentonite to corrode under influence of an externally applied electrical field. A model study was made to estimate possible corrosion effects of an external electrical field on a full-scale canister in the KBS-3 concept. The interaction between the repository represented by a copper canister in bentonite, and an external electrical field is illustrated with an example

  7. Dose Calculations for the Co-Disposal WP-of HLW-Glass and the Triga SNF

    Energy Technology Data Exchange (ETDEWEB)

    G. Radulescu

    1999-08-02

    This calculation is prepared by the Monitored Geologic Repository (MGR) Waste Package Operations (WPO). The purpose of this calculation is to determine the surface dose rates of a codisposal waste package (WP) containing a centrally located Department of Energy (DOE) standardized 18-in. spent nuclear fuel (SNF) canister, loaded with the TRIGA (Training, Research, Isotopes, General Atomics) SNF. This canister is surrounded by five 3-m long canisters, loaded with Savannah River Site (SRS) high-level waste (HLW) glass. The results are to support the WP design and radiological analyses.

  8. 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...

  9. 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...

  10. 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...

  11. Retrieval system for emplaced spent unreprocessed fuel (SURF) in salt bed depository: accident event analysis and mechanical failure probabilities. Final report

    International Nuclear Information System (INIS)

    This report provides support in developing an accident prediction event tree diagram, with an analysis of the baseline design concept for the retrieval of emplaced spent unreprocessed fuel (SURF) contained in a degraded Canister. The report contains an evaluation check list, accident logic diagrams, accident event tables, fault trees/event trees and discussions of failure probabilities for the following subsystems as potential contributors to a failure: (a) Canister extraction, including the core and ram units; (b) Canister transfer at the hoist area; and (c) Canister hoisting. This report is the second volume of a series. It continues and expands upon the report Retrieval System for Emplaced Spent Unreprocessed Fuel (SURF) in Salt Bed Depository: Baseline Concept Criteria Specifications and Mechanical Failure Probabilities. This report draws upon the baseline conceptual specifications contained in the first report

  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. 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)

  14. 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.

  15. 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.

  16. A passive air-cooled dry storage facility for vitrified high-level wastes

    International Nuclear Information System (INIS)

    A conceptual design of air-cooled dry storage vault facility for vitrified high-level waste (HLW) canisters is developed for a site in northern Japan. The facility is designed for the reception and unloading of shielded seagoing transportation casks of vitrified HLW canisters, for the inspection of these canisters, and for their temporary storage for a period of up to 50 years. The waste is to be at least 9 years old when received, and the facility will be capable of storing up to 2,500 canisters. This paper provides a conceptual design to identify construction requirements, materials, and space requirements that are unique to the vitrified HLW storage facility. It also identifies the types of special systems and equipment needed in such a facility

  17. Defense Waste Processing Facility (DWPF) startup test program: Glass characterization

    International Nuclear Information System (INIS)

    Liquid high-level nuclear waste will be immobilized at the Savannah River Site (SRS) by vitrification in borosilicate glass. The glass will be processed in the Defense Waste Processing Facility (DWPF) and poured into stainless steel canisters for eventual geologic disposal. Six simulated glass compositions will be processed in the DWPF during initial startup. The glass in 86 of the first 106 full sized canisters will be sampled and characterized. Extensive glass characterization will determine the following: (1) sampling frequency for radioactive operation, (2) verification of the compositionally dependent process-product models, (3) verification of melter mixing, (4) representativeness of the glass from the canister throat sampler, and (5) homogeneity of the canister glass

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

    International Nuclear Information System (INIS)

    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

  19. 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.

  20. Research program to study the gamma radiation effects in Spanish bentonites

    International Nuclear Information System (INIS)

    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)

  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. Treatment and disposal of radioactive wastes from nuclear power plants. Program for encapsulation, deep geologic deposition and research, development and demonstration

    International Nuclear Information System (INIS)

    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

  3. 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.

  4. 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)

  5. Development of in-can melting process and equipment, 1979 and 1980

    Energy Technology Data Exchange (ETDEWEB)

    Petkus, L.L.; Larson, D.E.; Bjorklund, W.J.; Holton, L.K.

    1981-09-01

    Nonradioactive process testing continued with the in-can melter as part of an investigation into the applicability of this vitrification process to various calcined high-level and incinerator ash radioactive wastes. The investigation in this report concentrated on how waste composition and canister fins affect in-can melter capacity and how waste composition affects glass quality. Process performance proved to be generally satisfactory. Pilot-scale in-can melter runs were performed with synthetic, nonradioactive, high-level wastes to produce eight canisters of glass. The synthetic wastes processed included high-level wastes from Savannah River, West Valley, and ICPP, as well as transuranic ash waste. Full-scale in-can melter runs using nonradioactive materials were also conducted, producing ten canisters of glass. Of the ten canisters, nine contained Savannah River Plant glass and one canister contained glass from synthetic zirconia calcine waste from the ICPP. 11.4 tons of glass was produced in test runs. In the full-scale in-can melter furnace, the baffles separating the six heating zones were removed because of baffle warping. A remotely operated section connecting the spray calciner to the canister was tested. Some problems were encountered with calcine plugging.

  6. Development of in-can melting process and equipment, 1979 and 1980

    International Nuclear Information System (INIS)

    Nonradioactive process testing continued with the in-can melter as part of an investigation into the applicability of this vitrification process to various calcined high-level and incinerator ash radioactive wastes. The investigation in this report concentrated on how waste composition and canister fins affect in-can melter capacity and how waste composition affects glass quality. Process performance proved to be generally satisfactory. Pilot-scale in-can melter runs were performed with synthetic, nonradioactive, high-level wastes to produce eight canisters of glass. The synthetic wastes processed included high-level wastes from Savannah River, West Valley, and ICPP, as well as transuranic ash waste. Full-scale in-can melter runs using nonradioactive materials were also conducted, producing ten canisters of glass. Of the ten canisters, nine contained Savannah River Plant glass and one canister contained glass from synthetic zirconia calcine waste from the ICPP. 11.4 tons of glass was produced in test runs. In the full-scale in-can melter furnace, the baffles separating the six heating zones were removed because of baffle warping. A remotely operated section connecting the spray calciner to the canister was tested. Some problems were encountered with calcine plugging

  7. Feasibility of dry cask-to-cask and pool-to-cask spent fuel transfer based on single-element transfer cask experience

    International Nuclear Information System (INIS)

    Spent fuel transportation casks and canister-based storage systems are generally loaded underwater in a nuclear plant's spent fuel pool/cask loading pit. Several reasons exist for exploring the feasibility of dry cask-to-cask and pool-to-cask spent fuel transfer. These include: the accommodation of plants which do not have sufficient crane capacity to handle large 90 tonne (100 ton) storage canisters or shipping casks, and construction of an MRS without the need for extensive hot cell facilities. In the case of DOE's ''Multi-Purpose Canister'' (MPC) scenario, use of such a transfer system would allow all plants with adequate transport routes to use large canisters at-reactor, and those without adequate transport routes to use the MRS for loading of large canisters without the need for hot cell facilities. The dry transfer option would not only allow the use of large canisters for all fuel, but would assist DOE in meeting MRS deadlines since licensing and construction of hot-cell facilities significantly affect schedule. This paper reviews the regulatory issues and technical design considerations for a single-element dry transfer system. Also summarized are lessons learned from the TMI-2 fuel transfer system which are directly applicable to the design, testing, startup, and use of a future dry cask-to-cask or pool-to-cask transfer system

  8. Analysis of SKB MiniCan-Experiment 3

    International Nuclear Information System (INIS)

    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 insert. The

  9. 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

  10. 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

  11. Spent fuel dry storage technology development: thermal evaluation of three adjacent drywells (each containing a 0.6 kW PWR spent fuel assembly)

    International Nuclear Information System (INIS)

    A spent fuel Adjacent Drywell Test was conducted at the Engine-Maintenance, Assembly and Disassembly (E-MAD) facility on the Nevada Test Site utilizing three nearly identical pressurized water reactor spent fuel assemblies each having a decay heat level of approximately 0.6 kW. Each fuel assembly was encapsulated inside the E-MAD Hot Bay and placed in an instrumented near-surface drywell storage cell for thermal testing. Each fuel assembly was sealed inside a 14-in. diam, 168-in.-long stainless steel canister and attached to a concrete-filled, 20-in.-diam, 34-in.-long, shield plug. The canister assembly was then placed in a carbon steel drywell liner which had been grouted into a hole drilled in the soil adjacent to E-MAD. The three drywells were located 25 feet apart in a linear array. Thermocouples, provided to measure canister, liner and soil temperatures, were inserted into tubes on the outside of the canister and drywell liner and were attached to plastic pipes which were grouted into holes in the soil. Temperatures from the three drywells and the adjacent soil were recorded throughout the Adjacent Drywell Test. Drywell thermal data showed virtually no thermal interaction between adjacent drywells. However, peak temperatures reached by the three drywells did show a fairly significant difference. Peak canister and drywell liner temperatures were reached in August 1981 for all three drywells. The two previously unused drywells responded similarly with peak canister and liner temperatures reaching 1990F and 1580F, respectively. Comparable peak temperatures for the third drywell which had previously contained spent fuel for nearly 21 months prior to the Adjacent Drywell Test reached 2100F for the canister and 1690F for the drywell liner. This difference is attributed to a decrease in soil thermal conductivity caused by the dryout of soil around the drywell used for previous spent fuel testing

  12. Spent fuel dry storage technology development: electrically heated drywell storage test (3kW operation)

    International Nuclear Information System (INIS)

    An electrically heated drywell storage cell test has been in operation since March 1978 at the Engine-Maintenance, Assembly and Disassembly (E-MAD) facility on the Nevada Test Site in support of spent fuel dry storage technology development. This document presents the test data obtained at electric heater power output of 3.0 kW and compares the data with that for heater power outputs of 1.0 kW and 2.0 kW. The simulated drywell storage cell consists of a stainless steel canister (representative of the spent fuel canisters being tested at E-MAD) containing an electrical heater assembly, a concrete-filled shield plug to which the canister is attached, and a carbon steel liner that encloses the canister and shield plug. The entire test drywell is grouted into a hole drilled in the soil adjacent to E-MAD. Temperature instrumentation is provided on the exterior of the canister and liner, in the grout around the liner, and at six radial locations in the soil surrounding the drywell. Peak measured canister and liner temperatures are 7850F and 7470F, respectively. Previous testing showed peak measured canister and liner temperatures of 2760F and 2320F for 1.0 kW and 5060F and 4580F for 2.0 kW, respectively. A previously developed computer model was utilized to predict the thermal response of the test configuration. Computer predictions of the transient and steady-state temperatures of the drywell components and surrounding soil are presented and are compared with the test data

  13. 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.

  14. Encapsulation and handling of spent nuclear fuel for final disposal

    International Nuclear Information System (INIS)

    The handling and embedding of those metal parts which arrive to the encapsulation station with the fuel is described. For the encapsulation of fuel two alternatives are presented, both with copper canisters but with filling of lead and copper powder respectively. The sealing method in the first case is electron beam welding, in the second case hot isostatic pressing. This has given the headline of the two chapters describing the methods: Welded copper canister and Pressed copper canister. Chapter 1, Welded copper canister, presents the handling of the fuel when it arrives to the encapsulation station, where it is first placed in a buffer pool. From this pool the fuel is transferred to the encapsulation process and thereby separated from fuel boxes and boron glass rod bundles, which are transported together with the fuel. The encapsulation process comprises charging into a copper canister, filling with molten lead, electron beam welding of the lid and final inspection. The transport to and handling in the final repository are described up to the deposition and sealing in the deposition hole. Handling of fuel residues is treated in one of the sections. In chapter 2, Pressed copper canister, only those parts of the handling, which differ from chapter 1 are described. The hot isostatic pressing process is given in the first sections. The handling includes drying, charging into the canister, filling with copper powder, seal lid application and hot isostatic pressing before the final inspection and deposition. In the third chapter, BWR boxes in concrete moulds, the handling of the metal parts, separated from the fuel, are dealt with. After being lifted from the buffer pool they are inserted in a concrete mould, the mould is filled with concrete, covered with a lid and after hardening transferred to its own repository. The deposition in this repository is described. (author)

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

    International Nuclear Information System (INIS)

    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 10-4 m/yr and a median F-ratio of 3.3 x 105 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

  16. Aluminium oxide as an encapsulation material for unreprocessed nuclear fuel waste - evaluation from the viewpoint of corrosion

    International Nuclear Information System (INIS)

    The Nuclear Fuel Safety Project (KBS) has proposed that spent unreprocessed nuclear fuel shall be disposed of by encapsulation in canisters of high-purity alumina sintered under isostatic pressure. The canisters will have a wall thickness of 100 mm and are to be placed in vertical boreholes extending from horizontal tunnels 500 m below ground in igneous rock. In each borehole one canister is deposited embedded in a quartz sand/bentonite buffer. An expert group of 10 Swedish specialists has arrived at the following conclusions. The alumina is not thermodynamically stable in water. In pure water hydration will occur, below 100degC leading to the formation of either Al(OH)3 in the amorphous state or crystalline gibbsite (Al2O3 x 3H2O). Corrosion may take place by slow dissolution or flaking off of a surface layer. Various immersion tests showed that the corrosion rate will be less than 0.1 μm/year, probably one or two powers of ten lower. If the alumina canister in the storage has sufficiently large surface defects and is under sufficiently high mechanical tension the defects may grow slowly into propagating cracks, ultimately leading to fracture, so-called delayed fracture. On the basis of results from fracture mechanical studies and after introduction of safety factors with respect to possible unknown features of the delayed fracture it was judged possible to eliminate the risk of delayed fracture if the canisters pass the following production control: - Proof testing at 150 MN/m2, using acoustic emission technique to ensure that crack growth does not occur during the unstressing cycle. - Surface acoustic wave examination with respect to surface inclusions, canisters with inclusions larger than 100 μm within a 100 μm deep surface zone being rejected. Canisters which pass the production control mentioned are estimated to have a life of hundreds of thousands of years. (author)

  17. Defense High Level Waste Disposal Container System Description

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-10-12

    The Defense High Level Waste Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the accesses using a rail mounted transporter, and emplaced in emplacement drifts. The defense high level waste (HLW) disposal container provides long-term confinement of the commercial HLW and defense HLW (including immobilized plutonium waste forms (IPWF)) placed within disposable canisters, and withstands the loading, transfer, emplacement, and retrieval loads and environments. U.S. Department of Energy (DOE)-owned spent nuclear fuel (SNF) in disposable canisters may also be placed in a defense HLW disposal container along with commercial HLW waste forms, which is known as 'co-disposal'. The Defense High Level Waste Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container/waste package maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual canister temperatures after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Defense HLW disposal containers for HLW disposal will hold up to five HLW canisters. Defense HLW disposal containers for co-disposal will hold up to five HLW canisters arranged in a ring and one DOE SNF canister in the ring. Defense HLW disposal containers also will hold two Multi-Canister Overpacks (MCOs) and two HLW canisters in one disposal container. The disposal container will include outer and inner cylinders, outer and inner cylinder lids, and may include a canister guide. An exterior label will provide a means by which to identify the disposal container and its contents. Different

  18. 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

  19. Improved drying rate diagnostics for saturated fuel debris at the INEEL

    Energy Technology Data Exchange (ETDEWEB)

    Childs, K.; Christensen, A. [Idaho National Engineering and Environmental Laboratory, Idaho Falls, ID (United States)

    1999-09-01

    A fuel canning station (FCS) has been operated for {approximately}2 yr to prepare for the dry storage of a variety of spent reactor fuels stored in pools at the Idaho National Engineering and Environmental Laboratory (INEEL). The FCS dewaters the fuel and then passivates possibly pyrophoric components in the fuel. Fuel-loaded canisters are placed into a heated insert, the canister is connected to a vacuum system, and the fuel is heated under a vacuum to remove the water. The dewatering system must also verify that the water was removed. The dryness criteria state that the canister pressure shall not exceed a defined pressure for a specified isolation time. Dewatering did not work well for defected TRIGA elements that had corroded in pool storage, leaving the intact fuel meat mixed with a bed of fines from metal oxides and from sludge that continuously accumulated within the pool. Dewatering these cans proved to be very time consuming. Fueled canisters were heated to 60 C and evacuated between 5 and 10 torr. At these conditions, intact fuels were rapidly dried (<10 h). TRIGA drying periods extended to 9 days. Dryness was qualitatively monitored using the canister pressure-control valve position. The valve closes as the gas flow rate declines, providing an indication that drying is complete. However, the valve remained open when drying TRIGA fuel, leaving no indication of dryness. In addition, dryness could not be verified because the canister pressure exceeded the defined pressure during isolation. Air leakage into the evacuated canister prevented the dryness from being verified. Air in-leakage and water vapor cannot easily be discriminated by the aforementioned procedures. Because the canister design does not seal above atmospheric pressure, a drying temperature that yielded a vapor pressure less than atmospheric pressure was chosen. A sufficiently long isolation test could then determine if air was accumulating in the canister; however, the low temperature reduced

  20. 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.

  1. The HAW-project: Demonstration facility for the disposal of high-level waste in salt

    International Nuclear Information System (INIS)

    The HAW-project plants the testwise emplacement of 30 vitrified highly radioactive canisters containing Cs-137 and Sr-90 at the 800 m level of the Asse salt mine for a testing period of approximately five years. The major objective of this project is the pilot testing and demonstration of safe methods for the final disposal of high-level radioactive waste (HAW) in geological salt formations. During the years 1985 to 1989 the underground test field was excavated, the measuring equipment installed, and two preceedings inactive electrical tests taken into operation. Furthermore, the components of a system for transportation and emplacement of highly radioactive canisters was fabricated, installed, and preliminarily tested. After some delays in the licensing procedure the emplacement of the 30 radioactive canisters is now envisaged for early 1991. For handling of the radioactive canisters and their emplacement into the boreholes a system consisting of a transport cask, a transport vehicle, a disposal machine, and of a borehole slider has been developed and will be tested. The actual scientific investigation programme is based on the estimation and observation of the interaction between the radioactive canisters and the rock salt. This programme includes measurement of thermally and radiolytically induced water and gas release from the rock salt and the radiolytical decomposition of salt minerals. Also the thermally induced stress and deformation fields in the surrounding rock mass will be investigated carefully. (orig./HP)

  2. 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)

  3. 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.

  4. Vacuum-insulated catalytic converter

    Energy Technology Data Exchange (ETDEWEB)

    Benson, David K. (Golden, CO)

    2001-01-01

    A catalytic converter has an inner canister that contains catalyst-coated substrates and an outer canister that encloses an annular, variable vacuum insulation chamber surrounding the inner canister. An annular tank containing phase-change material for heat storage and release is positioned in the variable vacuum insulation chamber a distance spaced part from the inner canister. A reversible hydrogen getter in the variable vacuum insulation chamber, preferably on a surface of the heat storage tank, releases hydrogen into the variable vacuum insulation chamber to conduct heat when the phase-change material is hot and absorbs the hydrogen to limit heat transfer to radiation when the phase-change material is cool. A porous zeolite trap in the inner canister absorbs and retains hydrocarbons from the exhaust gases when the catalyst-coated substrates and zeolite trap are cold and releases the hydrocarbons for reaction on the catalyst-coated substrate when the zeolite trap and catalyst-coated substrate get hot.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  7. New Strategies for Licensing the Storage and Transportation of High Burn-up Spent Nuclear Fuel in the United States - 12546

    International Nuclear Information System (INIS)

    An alternative approach may be needed to the licensing of high-burnup fuel for storage and transportation based on the assumption that spent fuel cladding may not always remain intact. The approach would permit spent fuel to be retrieved on a canister basis and could lessen the need for repackaging of spent fuel. This approach is being presented as a possible engineering solution to address the uncertainties and lack of data availability for cladding properties for high burnup fuel and extended storage time frames. The proposed approach does not involve relaxing current safety standards for criticality safety, containment, or permissible external dose rates. Packaging strategies and regulations should be developed to reduce the potential for requiring fuel to be repackaged unnecessarily. This would lessen the chance of accidents and mishaps during loading and unloading of casks, and decrease dose to workers. A packaging approach that shifts the safety basis from reliance upon the fuel condition to reliance upon an inner canister could eliminate or lessen the need for repackaging. In addition, the condition of canisters can be more readily monitored and inspected than the condition of fuel cladding. Canisters can also be repaired and/or replaced when deemed necessary. In contrast, once a fuel assembly is loaded into a canister and placed in a storage overpack, there is little opportunity to monitor its condition or take mitigating measures if cladding degradation is suspected or proven to occur. (authors)

  8. TVO-92 safety analysis of spent fuel disposal

    Energy Technology Data Exchange (ETDEWEB)

    Vieno, T.; Hautojaervi, A.; Koskinen, L.; Nordman, H. [Technical Research Centre of Finland, Espoo (Finland). Nuclear Engineering Lab.

    1993-08-01

    The spent fuel from the TVO I and TVO II reactors at the Olkiluoto nuclear power plant is planned to be disposed in a repository constructed at a depth of about 500 meters in crystalline bedrock. Teollisuuden Voima Oy (TVO) has carried out preliminary site investigations for spent fuel disposal between 1987 and 1992 at five areas in Finland (Olkiluoto, Kivetty, Romuvaara, Syyry and Veitsivaara). The Safety analysis of the disposal system is presented in the report. Spent fuel will be encapsulated in composite copper-steel canisters. The canister design (ACP canister) consists of an inner container of steel as a load-bearing element and an outer container of oxygen-free copper to provide a shield against corrosion. In the repository the canisters will be emplaced in vertical holes drilled in the floors of horizontal deposition tunnels. The annulus between the canister and the rock is filled with compacted bentonite. The results of the safety analysis attest that the planned disposal system fulfils the safety requirements. Suitable places for the repository can be found at each of the five investigation sites.

  9. Phased Startup Initiative Phase 3 Test Procedure (OCRWM)

    Energy Technology Data Exchange (ETDEWEB)

    PAJUNEN, A.L.

    2000-05-10

    The purpose of this test procedure is to safely operate the Fuel Retrieval System (FRS) and Integrated Water Treatment System (IWTS) with specific fuel canisters, and show that canisters containing fuel can be retrieved from the canister queue, decapped in the Canister Decapper, loaded into the Primary Clean Machine (PCM) for fuel cleaning, fuel sorted on the Process Table, then loaded back into fuel canisters and relocated in Basin Storage. Additional Data are collected during this test, beyond that collected during production operations. These data support qualifying the cleaning performance of the PCM, assessing the quantity of scrap generated during the cleaning, and evaluating the impact of fuel retrieval operations on the Basin water quality. The additional data collected primarily consist of weighing fuel and scrap at selected points in the operation, as well as photographing fuel and scrap as it is processed. The time to perform operations is also monitored for comparison with design predictions. Water quality data are collected to establish a base line to predict the effectiveness of equipment design for control of contamination and visibility during production operation.

  10. High-level waste long-term management technology development : Development of a geological disposal system

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. Y.; Choi, J. W.; Lee, M. S.; and others

    2012-04-15

    For the purpose of developing the geological disposal system proposed in this research project successfully, the analysis of source terms of various waste forms, conceptual design and performance assessment of the engineered barrier system including a disposal canister, and the cost analysis based on the domestic unit cost should be carried out. To this end the following research items will be developed at this stage: {Omicron} Development of a technology for assessing the source-term of advanced fuel cycle - Verification and enhancement of a source-term assessment program - Radiological performance assessment of A-Kfs - Assessment of national inventories of major radionuclides {Omicron} Development of disposal canisters for HLA from pyro-processing - Development of canister sealing technology - Estimation of corrosion life-time of canisters with Korean Ebs - Analysis of shear stress on the canister with Korean Ebs {Omicron} Development of a geological disposal system for HLA from advanced fuel cycle - Hydrological analysis for determining the layout of disposal tunnels - Preparation of database regarding the domestic unit disposal cost - Thermal analysis of disposal tunnels with ventilation during operation period - Feasibility assessment of A-Kfs from the technical, safety, and economical points of views.

  11. TVO-92 safety analysis of spent fuel disposal

    International Nuclear Information System (INIS)

    The spent fuel from the TVO I and TVO II reactors at the Olkiluoto nuclear power plant is planned to be disposed in a repository constructed at a depth of about 500 meters in crystalline bedrock. Teollisuuden Voima Oy (TVO) has carried out preliminary site investigations for spent fuel disposal between 1987 and 1992 at five areas in Finland (Olkiluoto, Kivetty, Romuvaara, Syyry and Veitsivaara). The Safety analysis of the disposal system is presented in the report. Spent fuel will be encapsulated in composite copper-steel canisters. The canister design (ACP canister) consists of an inner container of steel as a load-bearing element and an outer container of oxygen-free copper to provide a shield against corrosion. In the repository the canisters will be emplaced in vertical holes drilled in the floors of horizontal deposition tunnels. The annulus between the canister and the rock is filled with compacted bentonite. The results of the safety analysis attest that the planned disposal system fulfils the safety requirements. Suitable places for the repository can be found at each of the five investigation sites

  12. Preliminary study on corrosion layers of unearthed bronzes relics of Xizhou dynasty, China

    International Nuclear Information System (INIS)

    Engineering barrier for HLW repository is composed of vitrified waste, canister and buffer/backfill material. Assessment of the applicability of metal as candidate materials of canister for HLW has been conducted in some countries.. Many bronze relics in Xizhou Dynasty, China, dated from more than 3000 years ago, have been preserved perfectly. The study on the corrosion of the bronze relics would contribute to the material selection and design of canister for HLW. The corrosion products of ancient bronzes consist of copper carbonate hydrate, lead carbonate, copper oxides etc. The corrosion mechanism of the substrate were mainly characterized by electrochemical corrosion, whereas the corrosion mechanism of surface layers was characterized by direct chemical corrosion and electrochemical corrosion. (author)

  13. Taking it all back home

    International Nuclear Information System (INIS)

    Reprocessing contracts stipulate that Cogema's and BNFL's foreign customers will take back their vitrified residues to ensure subsequent storage themselves. National policies have been defined by those customers for the interim storage on return. Belgium and Japan have chosen to store them in glass canisters in air-cooled pits - at Mol and at Rokkasho-mura, respectively (similar to their current stores at the reprocessing plants) -while Germany and Switzerland have opted to use storage flasks. Aware of the need for vitrified residue return, almost 10 years ago Transnucleaire began developing a new model of flask to suit the various needs of the utilities concerned. Named TN 28 V in view of its basic payload of 28 vitrified waste canisters, this flask is currently being manufactured in two versions: one for the routine transport of glass-containing canisters and another for their transport followed by a long period of interim storage. (author)

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. Gas Generation from K East Basin Sludges - Series II Testing

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Samuel A.; Delegard, Calvin H.; Schmidt, Andrew J.; Sell, Rachel L.; Silvers, Kurt L.; Gano, Susan R.; Thornton, Brenda M.

    2001-03-14

    This report describes work to examine the gas generation behavior of actual K East (KE) Basin floor, pit and canister sludge. Mixed and unmixed and fractionated KE canister sludge were tested, along with floor and pit sludges from areas in the KE Basin not previously sampled. The first report in this series focused on gas generation from KE floor and canister sludge collected using a consolidated sampling technique. The third report will present results of gas generation testing of irradiated uranium fuel fragments with and without sludge addition. The path forward for management of the K Basin Sludge is to retrieve, ship, and store the sludge at T Plant until final processing at some future date. Gas generation will impact the designs and costs of systems associated with retrieval, transportation and storage of sludge.

  19. In situ characterization of Hanford K Basins fuel

    International Nuclear Information System (INIS)

    Irradiated N Reactor uranium metal fuel is stored underwater in the Hanford K East and K West Basins. In K East Basin, fuel is stored in open canisters and defected fuel is free to react with the basin water. In K West Basin, the fuel is stored in sealed canisters filled with water containing a corrosion inhibitor (potassium nitrite). To gain a better understanding of the physical condition of the fuel in these basins, visual surveys using high resolution underwater cameras were conducted. The inspections included detailed lift and look examinations of a number of fuel assemblies from selected canisters in each basin. These examinations formed the bases for selecting specific fuel elements for laboratory testing and analyses as prescribed in the characterization plan for Hanford K Basin Spent Nuclear Fuel

  20. 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.

  1. 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.

  2. Conceptual design for remote handling methods using the HIP process in the Calcine Immobilization Program

    International Nuclear Information System (INIS)

    This report recommends the remote conceptual design philosophy for calcine immobilization using the hot isostatic press (HIP) process. Areas of remote handling operations discussed in this report include: (1) introducing the process can into the front end of the HIP process, (2) filling and compacting the calcine/frit mixture into the process can, (3) evacuating and sealing the process can, (4) non-destructive testing of the seal on the process can, (5) decontamination of the process can, (6) HIP furnace loading and unloading the process can for the HIPing operation, (7) loading an overpack canister with processed HIP cans, (8) sealing the canister, with associated non-destructive examination (NDE) and decontamination, and (9) handling canisters for interim storage at the Idaho Chemical Processing Plant (ICPP) located on the Idaho National Engineering Laboratory (INEL) site

  3. Feature test report for the Small Debris Collection and Packaging System

    Energy Technology Data Exchange (ETDEWEB)

    Brisbin, S.A.

    1995-03-17

    The Spent Nuclear Fuel Equipment Engineering group performed feature testing of the Small Debris Collection and Packaging System (SDCPS) in the 305 Cold Test Facility from January 30, 1995, to February 1, 1995. Feature testing of the Small Debris Collection and Packaging System (SDCPS) was performed for the following reasons: To assess the feasibility of using ``drop-out`` vessels to collect small debris (<2.5 cm) in MK-II fuel canisters while transferring sludge to the Weasel Pit. To evaluate system performance under conditions similar to those in the K-Basins (e.g. submerged under 4.9 meters of water and operated with long handled tools) while using a surrogate sludge mixed with debris. To determine if canister weight could be used to predict the volume of sludge and/or debris contained within the canisters during system operation.

  4. The HAW-project: Demonstration facility for the disposal of high-level waste in salt

    International Nuclear Information System (INIS)

    To satisfy the test objectives thirty highly radioactive canisters containing the radionuclides Cs-137 and Sr-90 will be emplaced in six boreholes located in two test galleries at the 800 m-level in the Asse salt mine. For handling of the radioactive canisters and their emplacement into the boreholes a system consisting of a transport cask, a transport vehicle, a disposal machine, and of a borehole slider has been developed. The actual scientific investigation programme is based on the estimation and observation of the interaction between the radioactive canisters and the rock salt. This programme includes measurement of thermally and radiolytically induced water and gas release from the rock salt and the radiolytical decomposition of salt minerals. Also the thermally induced stress and deformation fields in the surrounding rock mass will be investigated carefully. (orig./DG)

  5. 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

    International Nuclear Information System (INIS)

    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

  6. TRIGA FUEL PHASE I AND II CRITICALITY CALCULATION

    Energy Technology Data Exchange (ETDEWEB)

    L. Angers

    1999-11-23

    The purpose of this calculation is to characterize the criticality aspect of the codisposal of TRIGA (Training, Research, Isotopes, General Atomic) reactor spent nuclear fuel (SNF) with Savannah River Site (SRS) high-level waste (HLW). The TRIGA SNF is loaded into a Department of Energy (DOE) standardized SNF canister which is centrally positioned inside a five-canister defense SRS HLW waste package (WP). The objective of the calculation is to investigate the criticality issues for the WP containing the five SRS HLW and DOE SNF canisters in various stages of degradation. This calculation will support the analysis that will be performed to demonstrate the viability of the codisposal concept for the Monitored Geologic Repository (MGR).

  7. Measurement of volatile organic compounds during start-up of bioremediation of French limited superfund site in Crosby Texas using wind dependent whole-air sampling

    International Nuclear Information System (INIS)

    Whole-air sampling was performed before and after the start-up of the bioremediation of an industrial (primarily petrochemical) waste lagoon in Crosby Texas, near Houston. Four 'Sector Samplers' were deployed at the four corners of the French Limited Superfund Site. These samplers collect air into one of two SUMMA polished canisters depending upon wind direction and speed. When the wind blows at the sampler from across the waste lagoon, air is routed to the 'IN' sector canister, otherwise sample is collected in the 'OUT' sector canister. As such, each sampler provides its own background sample, and, upon gas chromatographic analysis, individual compounds can be associated with the waste lagoon. Five sets of 24-hour sector samples were taken; the first set was collected prior to the start of the bioremediation effort and the remaining four sets were taken sequentially for four 24-hour periods after the start-up of the procedure

  8. 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.

  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 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.

  11. 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.

  12. 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.

  13. Considerations for Disposition of Dry Cask Storage System Materials at End of Storage System Life

    International Nuclear Information System (INIS)

    Dry cask storage systems are deployed at nuclear power plants for used nuclear fuel (UNF) storage when spent fuel pools reach their storage capacity and/or the plants are decommissioned. An important waste and materials disposition consideration arising from the increasing use of these systems is the management of the dry cask storage systems' materials after the UNF proceeds to disposition. Thermal analyses of repository design concepts currently under consideration internationally indicate that waste package sizes for the geologic media under consideration may be significantly smaller than the canisters being used for on-site dry storage by the nuclear utilities. Therefore, at some point along the UNF disposition pathway, there could be a need to repackage fuel assemblies already loaded into the dry storage canisters currently in use. In the United States, there are already over 1650 of these dry storage canisters deployed and approximately 200 canisters per year are being loaded at the current fleet of commercial nuclear power plants. There is about 10 cubic meters of material from each dry storage canister system that will need to be dispositioned. The concrete horizontal storage modules or vertical storage overpacks will need to be reused, re-purposed, recycled, or disposed of in some manner. The empty metal storage canister/cask would also have to be cleaned, and decontaminated for possible reuse or recycling or disposed of, likely as low-level radioactive waste. These material disposition options can have impacts of the overall used fuel management system costs. This paper will identify and explore some of the technical and interface considerations associated with managing the dry cask storage system materials. (authors)

  14. Survey of waste package designs for disposal of high-level waste/spent fuel in selected foreign countries

    International Nuclear Information System (INIS)

    This report presents the results of a survey of the waste package strategies for seven western countries with active nuclear power programs that are pursuing disposal of spent nuclear fuel or high-level wastes in deep geologic rock formations. Information, current as of January 1989, is given on the leading waste package concepts for Belgium, Canada, France, Federal Republic of Germany, Sweden, Switzerland, and the United Kingdom. All but two of the countries surveyed (France and the UK) have developed design concepts for their repositories, but none of the countries has developed its final waste repository or package concept. Waste package concepts are under study in all the countries surveyed, except the UK. Most of the countries have not yet developed a reference concept and are considering several concepts. Most of the information presented in this report is for the current reference or leading concepts. All canisters for the wastes are cylindrical, and are made of metal (stainless steel, mild steel, titanium, or copper). The canister concepts have relatively thin walls, except those for spent fuel in Sweden and Germany. Diagrams are presented for the reference or leading concepts for canisters for the countries surveyed. The expected lifetimes of the conceptual canisters in their respective disposal environment are typically 500 to 1,000 years, with Sweden's copper canister expected to last as long as one million years. Overpack containers that would contain the canisters are being considered in some of the countries. All of the countries surveyed, except one (Germany) are currently planning to utilize a buffer material (typically bentonite) surrounding the disposal package in the repository. Most of the countries surveyed plan to limit the maximum temperature in the buffer material to about 100 degree C. 52 refs., 9 figs

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

    International Nuclear Information System (INIS)

    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. 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.)

  17. Survey of waste package designs for disposal of high-level waste/spent fuel in selected foreign countries

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, K.J.; Lakey, L.T.; Silviera, D.J.

    1989-09-01

    This report presents the results of a survey of the waste package strategies for seven western countries with active nuclear power programs that are pursuing disposal of spent nuclear fuel or high-level wastes in deep geologic rock formations. Information, current as of January 1989, is given on the leading waste package concepts for Belgium, Canada, France, Federal Republic of Germany, Sweden, Switzerland, and the United Kingdom. All but two of the countries surveyed (France and the UK) have developed design concepts for their repositories, but none of the countries has developed its final waste repository or package concept. Waste package concepts are under study in all the countries surveyed, except the UK. Most of the countries have not yet developed a reference concept and are considering several concepts. Most of the information presented in this report is for the current reference or leading concepts. All canisters for the wastes are cylindrical, and are made of metal (stainless steel, mild steel, titanium, or copper). The canister concepts have relatively thin walls, except those for spent fuel in Sweden and Germany. Diagrams are presented for the reference or leading concepts for canisters for the countries surveyed. The expected lifetimes of the conceptual canisters in their respective disposal environment are typically 500 to 1,000 years, with Sweden's copper canister expected to last as long as one million years. Overpack containers that would contain the canisters are being considered in some of the countries. All of the countries surveyed, except one (Germany) are currently planning to utilize a buffer material (typically bentonite) surrounding the disposal package in the repository. Most of the countries surveyed plan to limit the maximum temperature in the buffer material to about 100{degree}C. 52 refs., 9 figs.

  18. 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.

  19. Materials considerations relative to multibarrier waste isolation

    International Nuclear Information System (INIS)

    The environmental conditions associated with the storage of radioactive wastes are reviewed, and the corrosion of potential waste containment materials under these conditions is evaluated. The desired service life of about 1000 years is beyond the time period for which existing corrosion data can be extrapolated with certainty; however, titanium alloys seem to offer the most promise. The mechanical requirements for canisters and overpacks are considered and several candidate materials are selected. Designs for a canister and an overpack have been developed, and these are used to estimate the costs for three possible materials of construction

  20. Cold vacuum drying facility site evaluation report

    International Nuclear Information System (INIS)

    In order to transport Multi-Canister Overpacks to the Canister Storage Building they must first undergo the Cold Vacuum Drying process. This puts the design, construction and start-up of the Cold Vacuum Drying facility on the critical path of the K Basin fuel removal schedule. This schedule is driven by a Tri-Party Agreement (TPA) milestone requiring all of the spent nuclear fuel to be removed from the K Basins by December, 1999. This site evaluation is an integral part of the Cold Vacuum Drying design process and must be completed expeditiously in order to stay on track for meeting the milestone

  1. 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.

  2. Depression Augments Activity-Related Pain in Women but not in Men with Chronic Musculoskeletal Conditions

    OpenAIRE

    Heather Adams; Pascal Thibault; Nicole Davidson; Maureen Simmonds; Ana Velly; Michael JL Sullivan

    2008-01-01

    OBJECTIVES: The primary objective of the present study was to examine the role of sex as a moderator of the relation between depression and activity-related pain.METHODS: The study sample consisted of 83 participants (42 women, 41 men) with musculoskeletal conditions. Participants were asked to lift a series of 18 canisters that varied in weight (2.9 kg, 3.4 kg and 3.9 kg) and distance from the body. Participants were asked to rate their pain while they lifted each canister and estimate the w...

  3. Safety analysis report for packaging (onsite) Castor GSF cask

    International Nuclear Information System (INIS)

    The CASTOR GSF packaging was designed and fabricated to be a certified Type B(U) packaging and comply with the requirements of the International Atomic Energy Agency (IAEA) for transport of up to five sealed canisters of vitrified radioactive materials. This onsite Safety Analysis Report for Packaging (SARP) provides the analysis and evaluations necessary to demonstrate that the casks, with the canister payload, meet the intent of the Type B packaging regulations set forth in 10 CFR 71 and therefore meet the onsite transportation safety requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping

  4. 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.

  5. Method and design for externally applied laser welding of internal connections in a high power electrochemical cell

    Energy Technology Data Exchange (ETDEWEB)

    Martin, Charles E; Fontaine, Lucien; Gardner, William H

    2014-01-21

    An electrochemical cell includes components that are welded from an external source after the components are assembled in a cell canister. The cell canister houses electrode tabs and a core insert. An end cap insert is disposed opposite the core insert. An external weld source, such as a laser beam, is applied to the end cap insert, such that the end cap insert, the electrode tabs, and the core insert are electrically coupled by a weld which extends from the end cap insert to the core insert.

  6. Management of Damaged SNF Handling Operations At Paks NPP

    International Nuclear Information System (INIS)

    The issue of handling leaky fuel is one of the crucial issues of nuclear energy. It is directly connected with transportation of leaky spent nuclear fuel (SNF) that is mostly stored in the station's cooling pools. At present damaged spent nuclear fuel (SNF) of thirty VVER-440 spent fuel assemblies (SFA) is loaded into the ventilated canisters of types 28 and 29 and temporary stored in the cooling pool of Paks NPP. The report presents the milestones of preparation and safety justification of the technology for preparing canisters with damaged SNF of Paks NPP for transport to FSUE “PA “Mayak” (Russia) for reprocessing. (author)

  7. 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

  8. 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

  9. Comments on a paper tilted `The sea transport of vitrified high-level radioactive wastes: Unresolved safety issues`

    Energy Technology Data Exchange (ETDEWEB)

    Sprung, J.L.; McConnell, P.E.; Nigrey, P.J.; Ammerman, D.J. [and others

    1997-05-01

    The cited paper estimates the consequences that might occur should a purpose-built ship transporting Vitrified High Level Waste (VHLW) be involved in a severe collision that causes the VHLW canisters in one Type-B package to spill onto the floor of a major ocean fishing region. Release of radioactivity from VHLW glass logs, failure of elastomer cask seals, failure of VHLW canisters due to stress corrosion cracking (SCC), and the probabilities of the hypothesized accident scenario, of catastrophic cask failure, and of cask recovery from the sea are all discussed.

  10. Estimation of unsaturated zone liquid water flux at boreholes UZ No. 4, UZ No. 5, UZ No. 7, and UZ No. 13, Yucca Mountain, Nevada, from saturation and water potential profiles

    International Nuclear Information System (INIS)

    The unsaturated zone at Yucca Mountain, Nevada, is being investigated as a potential location for a high-level nuclear waste repository. Characterization of the liquid water flux and its spatial distribution under natural conditions provides estimates of the amount of water that could potentially contact the waste canisters and transport soluble radionuclides to the accessible environment. Estimates of ambient water flux may affect design requirements by indicating the degree to which reliance must be placed on engineered barriers or the waste-generated heat to keep the waste canisters dry

  11. Research-, development- and demonstrationprogram for a KBS-3 repository with horizontal deposition; Forsknings-, utvecklings- och demonstrationsprogram foer ett KBS-3-foervar med horisontell deponering

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-12-01

    This report gives an account of what RD and D efforts that would be necessary in order to make a full comparison between a KBS-3 repository with horizontal deposition of several canisters in each deposition hole, and the reference design with vertical deposition. A horizontal repository has no deposition tunnels, since the canisters are deposited directly from the transport tunnels. This means that the excavated rock mass is much reduced, compared to vertical deposition. Economically, it is calculated that horizontal deposition can be about 1 billion SEK (about 100 million USD) cheaper. The costs for performing the RD and D program are estimated to 150 million SEK.

  12. DWPF glass transition temperatures: What they are and why they are important

    Energy Technology Data Exchange (ETDEWEB)

    Marra, S.L.; Jantzen, C.M.; Ramsey, A.A.

    1991-12-31

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site will immobilize high-level radioactive liquid waste in borosilicate glass. The glass will be poured into stainless steel canisters for eventual disposal in a geologic repository. The Department of Energy has defined a set of requirements for the DWPF canistered waste form which must be met in order to assure compatibility with, and acceptance by, the repository. These requirements are the Waste Acceptance Preliminary Specifications (WAPS). The WAPS require DWPF to report glass transition temperatures for the projected range of compositions. This information will be used by the repository to establish waste package design limits.

  13. DWPF glass transition temperatures: What they are and why they are important

    Energy Technology Data Exchange (ETDEWEB)

    Marra, S.L.; Jantzen, C.M.; Ramsey, A.A.

    1991-01-01

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site will immobilize high-level radioactive liquid waste in borosilicate glass. The glass will be poured into stainless steel canisters for eventual disposal in a geologic repository. The Department of Energy has defined a set of requirements for the DWPF canistered waste form which must be met in order to assure compatibility with, and acceptance by, the repository. These requirements are the Waste Acceptance Preliminary Specifications (WAPS). The WAPS require DWPF to report glass transition temperatures for the projected range of compositions. This information will be used by the repository to establish waste package design limits.

  14. 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.

  15. Comments on a paper tilted 'The sea transport of vitrified high-level radioactive wastes: Unresolved safety issues'

    International Nuclear Information System (INIS)

    The cited paper estimates the consequences that might occur should a purpose-built ship transporting Vitrified High Level Waste (VHLW) be involved in a severe collision that causes the VHLW canisters in one Type-B package to spill onto the floor of a major ocean fishing region. Release of radioactivity from VHLW glass logs, failure of elastomer cask seals, failure of VHLW canisters due to stress corrosion cracking (SCC), and the probabilities of the hypothesized accident scenario, of catastrophic cask failure, and of cask recovery from the sea are all discussed

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

    International Nuclear Information System (INIS)

    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

  17. 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

  18. Czech safety concept: 2013 state of the art

    International Nuclear Information System (INIS)

    The Czech Republic operates four WWER 440 reactors (Dukovany) and two WWER 1000 reactors (Temelin). The four 440 MW Dukovany units were installed and began operation during the period 1985-1988. The two WWER 1000 reactors at Temelin started operation in 2002 and 2003. Currently, more than 8000 SF assemblies from WWER 440 reactors and 900 spent assemblies from WWER 1000 reactors spent fuel assemblies are stored in dry storage facilities located in the area of both NPP in approved casks or in pools at reactor sites. More than 4 000 assemblies are expected to be spent by 2025 at Dukovany reactors and 4 600 assemblies by 2042 at Temelin reactors. The multi-billion Euro contract to build two new nuclear reactors at the current site of Temelin with the option for an additional one in Dukovany has recently been launched in the Czech Republic. It is expected that more than 8 000 fuel assemblies would be spent in the three new nuclear reactors in the Czech Republic during their 60 years of electricity production. The basic reference plan is to directly dispose of all of the spent fuel assemblies in a deep geological repository (DGR), starting operation not earlier than in 2065. The DGR is planned to be located in granite host rock, because no other type of host rock in sufficient volume is available in the Czech Republic. Currently seven candidate sites for DGR suitable for geological disposal of SF assemblies have been selected, but due to negative community attitudes at the notion of have a repository in their backyard, they are still awaiting a detailed geological survey. According to proposed reference designs, SF assemblies should be in steel-based canisters emplaced in vertical or horizontal boreholes in granite host rock at approximately 500 m under the surface and surrounded by compacted bentonite. The Czech safety concept is based on the KBS-3 concept developed in Sweden. The Swedish concept is primarily based on almost thermodynamic stability of copper overpack in

  19. Analysis of thermal energy storage material with change-of-phase volumetric effects

    Science.gov (United States)

    Kerslake, Thomas W.; Ibrahim, Mounir B.

    1990-01-01

    NASA's Space Station Freedom proposed hybrid power system includes photovoltaic arrays with nickel hydrogen batteries for energy storage and solar dynamic collectors driving Brayton heat engines with change-of-phase Thermal Energy Storage (TES) devices. A TES device is comprised of multiple metallic, annular canisters which contain a eutectic composition LiF-CaF2 Phase Change Material (PCM) that melts at 1040 K. A moderately sophisticated LiF-CaF2 PCM computer model is being developed in three stages considering 1-D, 2-D, and 3-D canister geometries, respectively. The 1-D model results indicate that the void has a marked effect on the phase change process due to PCM displacement and dynamic void heat transfer resistance. Equally influential are the effects of different boundary conditions and liquid PCM natural convection. For the second stage, successful numerical techniques used in the 1-D phase change model are extended to a 2-D (r,z) PCM containment canister model. A prototypical PCM containment canister is analyzed and the results are discussed.

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

    International Nuclear Information System (INIS)

    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.)