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Sample records for canisters

  1. Shielded Canister Transporter

    International Nuclear Information System (INIS)

    Eidem, G.G. Jr.; Fages, R.

    1993-01-01

    The Hanford Waste Vitrification Plant (HWVP) will produce canisters filled with high-level radioactive waste immobilized in borosilicate glass. This report discusses a Shielded Canister Transporter (SCT) which will provide the means for safe transportation and handling of the canisters from the Vitrification Building to the Canister Storage Building (CSB). The stainless steel canisters are 0.61 meters in diameter, 3.0 meters tall, and weigh approximately 2,135 kilograms, with a maximum exterior surface dose rate of 90,000 R/hr. The canisters are placed into storage tubes to a maximum of three tall (two for overpack canisters) with an impact limiter placed at the tube bottom and between each canister. A floor plug seals the top of the storage tube at the operating floor level of the CSB

  2. SRL canister impact tests

    International Nuclear Information System (INIS)

    Kelker, J.W. Jr.

    1986-05-01

    The Defense Waste Processing Facility (DWPF) is being constructed at the SRP for the containerization of high-level nuclear waste as a waste form for eventual permanent disposal. The waste will be incorporated in molten glass and solidified in Type 304L stainless steel canisters 2 feet in diameter x 9 feet 10 inches long. The canisters have a minimum wall thickness of 3/8 inch. Over a three-year period, nineteen drop-tests of nine canisters, filled with simulated waste glass, were made in support of the DWPF containerization program. Eight of the canister evaluation tests were of Type 304L stainless steel material and one was of commercially pure titanium. Three different length (9.44, 5.06, and 7.88 inch) nozzle configurations containing final closure upset welds were evaluated for the stainless steel canisters. All impact tests of the stainless steel canisters, which included bottom-, side-, and top-drops, were acceptable. The bottom-drop test of the titanium canister, which contained a final closure upset weld, was acceptable; however, the top-drop resulted in a breaching of the top head where it joins the nozzle. The final closure titanium upset weld was acceptable. The titanium canister wall thickness was 1/4 inch

  3. Mechanical integrity of canisters

    International Nuclear Information System (INIS)

    Nilsson, Fred

    1992-12-01

    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

  4. The concrete canister program

    International Nuclear Information System (INIS)

    Ohta, M.M.

    1978-02-01

    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)

  5. Manufacture of disposal canisters

    International Nuclear Information System (INIS)

    Nolvi, L.

    2009-12-01

    The report summarizes the development work carried out in the manufacturing of disposal canister components, and present status, in readiness for manufacturing, of the components for use in assembly of spent nuclear fuel disposal canister. The disposal canister consist of two major components: the nodular graphite cast iron insert and overpack of oxygen-free copper. The manufacturing process for copper components begins with a cylindrical cast copper billet. Three different manufacturing processes i.e. pierce and draw, extrusion and forging are being developed, which produce a seamless copper tube or a tube with an integrated bottom. The pierce and draw process, Posiva's reference method, makes an integrated bottom possible and only the lid requires welding. Inserts for BWR-element are cast with 12 square channels and inserts for VVER 440-element with 12 round channels. Inserts for EPR-elements have four square channels. Casting of BWR insert type has been studied so far. Experience of casting inserts for PWR, which is similar to the EPR-type, has been got in co-operation with SKB. The report describes the processes being developed for manufacture of disposal canister components and some results of the manufacturing experiments are presented. Quality assurance and quality control in manufacture of canister component is described. (orig.)

  6. CANISTER TRANSFER SYSTEM DESCRIPTION DOCUMENT

    International Nuclear Information System (INIS)

    B. Gorpani

    2000-01-01

    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

  7. Canister Transfer System Description Document

    International Nuclear Information System (INIS)

    2000-01-01

    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

  8. A Film Canister Colorimeter.

    Science.gov (United States)

    Gordon, James; James, Alan; Harman, Stephanie; Weiss, Kristen

    2002-01-01

    A low-cost, low-tech colorimeter was constructed from a film canister. The student-constructed colorimeter was used to show the Beer-Lambert relationship between absorbance and concentration and to calculate the value of the molar absorptivity for permanganate at the wavelength emission maximum for an LED. Makes comparisons between this instrument…

  9. 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

  10. Status report, canister fabrication

    International Nuclear Information System (INIS)

    Andersson, Claes-Goeran; Eriksson, Peter; Westman, Marika; Emilsson, Goeran

    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

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

    International Nuclear Information System (INIS)

    H. Marr

    1999-01-01

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

  12. K West Basin canister survey

    International Nuclear Information System (INIS)

    Pitner, A.L.

    1998-01-01

    A survey was conducted of the K West Basin to determine the distribution of canister types that contain the irradiated N Reactor fuel. An underwater camera was used to conduct the survey during June 1998, and the results were recorded on videotape. A full row-by-row survey of the entire basin was performed, with the distinction between aluminum and stainless steel Mark 1 canisters made by the presence or absence of steel rings on the canister trunions (aluminum canisters have the steel rings). The results of the survey are presented in tables and figures. Grid maps of the three bays show the canister lid ID number and the canister type in each location that contained fuel. The following abbreviations are used in the grid maps for canister type designation: IA = Mark 1 aluminum, IS = Mark 1 stainless steel, and 2 = Mark 2 stainless steel. An overall summary of the canister distribution survey is presented in Table 1. The total number of canisters found to contain fuel was 3842, with 20% being Mark 1 Al, 25% being Mark 1 SS, and 55% being Mark 2 SS. The aluminum canisters were predominantly located in the East and West bays of the basin

  13. DISPOSABLE CANISTER WASTE ACCEPTANCE CRITERIA

    Energy Technology Data Exchange (ETDEWEB)

    R.J. Garrett

    2001-07-30

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

  14. Transport of multiassembly sealed canisters

    International Nuclear Information System (INIS)

    Quinn, R.D.; Lehnert, R.A.; Rosa, J.M.

    1992-01-01

    A significant portion of the commercial spent nuclear fuel in dry storage in the US will be stored in multiassembly sealed canisters before the DOE begins accepting fuel from utilities in 1998. This paper reports that it is desirable from economic and ALARA perspectives to transfer these canisters directly from the plant to the MRS. To this end, it is necessary that the multiassembly sealed canisters, which have been licensed for storage under 10CFR72, be qualified for shipment within a suitable shipping cask under the rules of 10CFR71. Preliminary work performed to date indicates that it is feasible to license a current canister design for transportation, and work is proceeding on obtaining NRC approval

  15. Evaluation of a molybdenum assay canister

    International Nuclear Information System (INIS)

    Yoshizumi, T.T.; Keener, S.J.

    1988-01-01

    The performance characteristics of a commercial molybdenum assay canister were evaluated. The geometrical variation of the technetium-99m (/sup 99m/Tc) activity reading was studied as a function of the elution volume for the standard vials. It was found that the /sup 99m/Tc canister activity reading was ∼ 5% lower than that of the standard method. This is due to attenuation by the canister wall. However, the effect of the geometric variation on the clinical dose preparation was found to be insignificant. The molybdenum-99 ( 99 Mo) contamination level was compared by two methods: (1) the commercial canister and (2) the standard assay kit. The 99 Mo contamination measurements with the canister indicated consistently lower readings than those with the standard 99 Mo assay kit. The authors conclude that the canister may be used in the clinical settings. However, the user must be aware of the problems and the limitations associated with this canister

  16. Spent fuel canister docking station

    International Nuclear Information System (INIS)

    Suikki, M.

    2006-01-01

    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

  17. Design premises for canister for spent nuclear fuel

    International Nuclear Information System (INIS)

    Werme, L.

    1998-09-01

    The purpose of this report is to establish the basic premises for designing canisters for the disposal of spent nuclear fuel, the requirements for canister characteristics, and the design criteria, and to present alternative canister designs that satisfy these premises. The point of departure for canister design has been that the canister must be able to be used for both BWR and PWR fuel

  18. Inspection of disposal canisters components

    International Nuclear Information System (INIS)

    Pitkaenen, J.

    2013-12-01

    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

  19. Canister arrangement for storing radioactive waste

    Science.gov (United States)

    Lorenzo, D.K.; Van Cleve, J.E. Jr.

    1980-04-23

    The subject invention relates to a canister arrangement for jointly storing high level radioactive chemical waste and metallic waste resulting from the reprocessing of nuclear reactor fuel elements. A cylindrical steel canister is provided with an elongated centrally disposed billet of the metallic waste and the chemical waste in vitreous form is disposed in the annulus surrounding the billet.

  20. Waste canister for storage of nuclear wastes

    Science.gov (United States)

    Duffy, James B.

    1977-01-01

    A waste canister for storage of nuclear wastes in the form of a solidified glass includes fins supported from the center with the tips of the fins spaced away from the wall to conduct heat away from the center without producing unacceptable hot spots in the canister wall.

  1. Waste canister for storage of nuclear wastes

    International Nuclear Information System (INIS)

    Duffy, J.B.

    1977-01-01

    A waste canister for storage of nuclear wastes in the form of a solidified glass includes fins supported from the center with the tips of the fins spaced away from the wall to conduct heat away from the center without producing unacceptable hot spots in the canister wall. 4 claims, 4 figures

  2. Effect of Canister Movement on Water Turbidity

    International Nuclear Information System (INIS)

    TRIMBLE, D.J.

    2000-01-01

    Requirements for evaluating the adherence characteristics of sludge on the fuel stored in the K East Basin and the effect of canister movement on basin water turbidity are documented in Briggs (1996). The results of the sludge adherence testing have been documented (Bergmann 1996). This report documents the results of the canister movement tests. The purpose of the canister movement tests was to characterize water turbidity under controlled canister movements (Briggs 1996). The tests were designed to evaluate methods for minimizing the plumes and controlling water turbidity during fuel movements leading to multi-canister overpack (MCO) loading. It was expected that the test data would provide qualitative visual information for use in the design of the fuel retrieval and water treatment systems. Video recordings of the tests were to be the only information collected

  3. 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.

  4. CANISTER HANDLING FACILITY DESCRIPTION DOCUMENT

    International Nuclear Information System (INIS)

    Beesley. J.F.

    2005-01-01

    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

  5. Canister storage building natural phenomena design loads

    International Nuclear Information System (INIS)

    Tallman, A.M.

    1996-02-01

    This document presents natural phenomena hazard (NPH) loads for use in the design and construction of the Canister Storage Building (CSB), which will be located in the 200 East Area of the Hanford Site

  6. Canister transfer into repository in shaft alternative

    International Nuclear Information System (INIS)

    Raiko, H.; Kukkola, T.; Autio, J.

    2005-09-01

    In this report, a study of lift transportation of a massive canister for spent nuclear fuel is considered. The canister is transferred from ground level to repository, which lies in the depth of 400 to 500 m in the bedrock. The canister is a massive metal vessel, whose weight is 19 to 29 tons, and which is strongly irradiant (gamma and neutrons), and which contains 1.4 to 2.2 tons of very strongly radio-active material, the activity of the fuel should not be spread in the environment even during postulated accidents. The study observes that the lift alternative is possible to be built and through good design practices and good maintenance procedures its safety, reliability and usability can be kept on such high level that canister transport is estimated to be licensable. (orig.)

  7. Conceptual designs of radioactive canister transporters

    International Nuclear Information System (INIS)

    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

  8. Design premises for canister for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Werme, L

    1998-09-01

    The purpose of this report is to establish the basic premises for designing canisters for the disposal of spent nuclear fuel, the requirements for canister characteristics, and the design criteria, and to present alternative canister designs that satisfy these premises. The point of departure for canister design has been that the canister must be able to be used for both BWR and PWR fuel 43 refs, 4 figs, 6 tabs

  9. Grain boundary corrosion of copper canister material

    International Nuclear Information System (INIS)

    Fennell, P.A.H.; Graham, A.J.; Smart, N.R.; Sofield, C.J.

    2001-03-01

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

  10. Am/Cm canister temperature evaluation in CIM5

    International Nuclear Information System (INIS)

    Baich, M.A.

    2000-01-01

    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

  11. Preliminary design for spent fuel canister handling systems in a canister transfer and installation vehicle

    International Nuclear Information System (INIS)

    Wendelin, T.; Suikki, M.

    2008-12-01

    The report presents a spent fuel canister transfer and installation vehicle. The vehicle is used for carrying the fuel canister into a disposal tunnel and installing it into a deposition hole. The report outlines basic requirements and a design for canister handling equipment used in a canister transfer and installation vehicle, a description regarding the operation and maintenance of the equipment, as well as a cost estimate. Specific vehicles will be manufactured for all canister types in order to minimize the height of the disposal tunnels. This report is only focused on a transfer and installation vehicle for OL1-2 fuel canisters. Detailed designing and selection of final components have not yet been carried out. The report also describes the vehicle's requirements for the structures of a repository system, as well as actions in possible malfunction or fault situations. The spent fuel canister is brought from an encapsulation plant by a canister lift down to the repository level. The fuel canister is driven from the canister lift by an automated guided vehicle onto a canister hoist at a canister loading station. The canister transfer and installation vehicle is waiting for the canister with its radiation shield in an upright position above the canister hoist. The hoist carries the canister upward until the vehicle's own lifting means grab hold of the canister and raise it up into the vehicle's radiation shield. This is followed by turning the radiation shield to a transport position and by closing it in a radiation-proof manner against a rear radiation shield. The vehicle is driven along the central tunnel into the disposal tunnel and parked on top of the deposition hole. The vehicle's radiation shield is turned to the upright position and the canister is lowered with the vehicle's hydraulic winches into a bentonite-lined deposition hole. The radiation shield is turned back to the transport position and the vehicle can be driven out of the disposal tunnel

  12. CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    Energy Technology Data Exchange (ETDEWEB)

    C.E. Sanders

    2005-04-07

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

  13. CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    International Nuclear Information System (INIS)

    C.E. Sanders

    2005-01-01

    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

  14. 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.)

  15. 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.

  16. Remote controlled mover for disposal canister transfer

    International Nuclear Information System (INIS)

    Suikki, M.

    2013-10-01

    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.)

  17. Friction welded closures of waste canisters

    International Nuclear Information System (INIS)

    Klein, R.F.

    1987-01-01

    Liquid radioactive waste presently stored in underground tanks is to undergo a vitrifying process which will immobilize it into a solid form. This solid waste will be contained in a stainless steel canister. The canister opening requires a positive-seal weld, the properties and thickness of which must be at least equal to those of the canister material. All studies and tests performed in the work discussed in this paper have the inertia friction welding concept to be highly feasible in this application. This paper describes the decision to investigate the inertia friction welding process, the inertia friction welding process itself, and a proposed equipment design concept. This system would provide a positive, reliable, inspectable, and full-thickness seal weld while utilizing easily maintainable equipment. This high-quality weld can be achieved even in highly contaminated hot cell

  18. Chemical compatibility of DWPF canistered waste forms

    International Nuclear Information System (INIS)

    Harbour, J.R.

    1993-01-01

    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

  19. Corrosion resistance of copper canister weld material

    International Nuclear Information System (INIS)

    Gubner, Rolf; Andersson, Urban

    2007-03-01

    The proposed design for a final repository for spent fuel and other long-lived residues is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will be placed in granite bedrock and surrounded by compacted bentonite clay. The canister design is based on a thick cast iron insert fitted inside a copper canister. SKB has since several years developed manufacturing processes for the canister components using a network of manufacturers. For the encapsulation process SKB has built the Canister Laboratory to demonstrate and develop the encapsulation technique in full scale. The critical part of the encapsulation of spent fuel is the sealing of the canister which is done by welding the copper lid to the cylindrical part of the canister. Two welding techniques have been developed in parallel, Electron Beam Welding (EBW) and Friction Stir Welding (FSW). During the past two decades, SKB has developed the technology EBW at The Welding Institute (TWI) in Cambridge, UK. The development work at the Canister Laboratory began in 1999. In electron beam welding, a gun is used to generate the electron beam which is aimed at the joint. The beam heats up the material to the melting point allowing a fusion weld to be formed. The gun was developed by TWI and has a unique design for use at reduced pressure. The system has gone through a number of improvements under the last couple of years including implementation of a beam oscillation system. However, during fabrication of the outer copper canisters there will be some unavoidable grain growth in the welded areas. As grains grow they will tend to concentrate impurities at the new grain boundaries that might pose adverse effects on the corrosion resistance of welds. As a new method for joining, SKB has been developing friction stir welding (FSW) for sealing copper canisters for spent nuclear fuel in cooperation with TWI since 1997. FSW was invented in 1991 at TWI and is a thermo

  20. Corrosion resistance of copper canister weld material

    Energy Technology Data Exchange (ETDEWEB)

    Gubner, Rolf; Andersson, Urban [Corrosion and Metals Research Institute, Sto ckholm (Sweden)

    2007-03-15

    The proposed design for a final repository for spent fuel and other long-lived residues is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will be placed in granite bedrock and surrounded by compacted bentonite clay. The canister design is based on a thick cast iron insert fitted inside a copper canister. SKB has since several years developed manufacturing processes for the canister components using a network of manufacturers. For the encapsulation process SKB has built the Canister Laboratory to demonstrate and develop the encapsulation technique in full scale. The critical part of the encapsulation of spent fuel is the sealing of the canister which is done by welding the copper lid to the cylindrical part of the canister. Two welding techniques have been developed in parallel, Electron Beam Welding (EBW) and Friction Stir Welding (FSW). During the past two decades, SKB has developed the technology EBW at The Welding Institute (TWI) in Cambridge, UK. The development work at the Canister Laboratory began in 1999. In electron beam welding, a gun is used to generate the electron beam which is aimed at the joint. The beam heats up the material to the melting point allowing a fusion weld to be formed. The gun was developed by TWI and has a unique design for use at reduced pressure. The system has gone through a number of improvements under the last couple of years including implementation of a beam oscillation system. However, during fabrication of the outer copper canisters there will be some unavoidable grain growth in the welded areas. As grains grow they will tend to concentrate impurities at the new grain boundaries that might pose adverse effects on the corrosion resistance of welds. As a new method for joining, SKB has been developing friction stir welding (FSW) for sealing copper canisters for spent nuclear fuel in cooperation with TWI since 1997. FSW was invented in 1991 at TWI and is a thermo

  1. Copper canisters for nuclear high level waste disposal. Corrosion aspects

    International Nuclear Information System (INIS)

    Werme, L.; Sellin, P.; Kjellbert, N.

    1992-10-01

    A corrosion analysis of a thick-walled copper canister for spent fuel disposal is discussed. The analysis has shown that there are no rapid mechanisms that may lead to canister failure, indicating an anticipated corrosion service life of several millions years. If further analysis of the copper canister is considered, it should be concentrated on identifying and evaluating processes other than corrosion, which may have a potential for leading to canister failure. (au)

  2. Near-field performance of the advanced cold process canister

    International Nuclear Information System (INIS)

    Werme, L.

    1990-09-01

    A near-field performance evaluation of an Advanced Cold Process Canister for spent fuel disposal has been performed jointly by TVO, Finland and SKB, Sweden. The canister consists of a steel canister as a load bearing element, with an outer corrosion shield of copper. The canister design was originally proposed by TVO. In the analysis, as well internal (ie corrosion processes from the inside of the canister) as external processes (mechanical and chemical) have been considered both prior to and after canister breach. Throughout the analysis, present day underground conditions has been assumed to persist during the service life of the canister. The major conclusions for the evaluation are: Internal processes cannot cause the canister breach under foreseen conditions, ie localized corrosion for the steel or copper canisters can be dismissed as a failure mechanism. The evaluation of the effects of processes outside the canister indicate that there is no rapid mechanism to endanger the integrity of the canister. Consequently the service life of the canister will be several million years. This factor will ensure the safety of the concept. (orig.)

  3. Studies of waste-canister compatibility

    International Nuclear Information System (INIS)

    McCoy, H.E.

    1983-01-01

    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 300 0 C, 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 300 0 C, but the extent of chemical reaction was not determined. Testing for 0.5 h at 800 0 C 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

  4. Plutonium Immobilization Project - Robotic canister loading

    International Nuclear Information System (INIS)

    Hamilton, R.L.

    2000-01-01

    The Plutonium Immobilization Program (PIP) is a joint venture between the Savannah River Site (SRS), Lawrence Livermore National Laboratory (LLNL), Argonne National Laboratory (ANL), and Pacific Northwest National Laboratory (PNNL). When operational in 2008, the PIP will fulfill the nation's nonproliferation commitment by placing surplus weapons-grade plutonium in a permanently stable ceramic form and making it unattractive for reuse. Since there are significant radiation and security concerns, the program team is developing novel and unique technology to remotely perform plutonium immobilization tasks. The remote task covered in this paper employs a jointed arm robot to load seven 3.5 inch diameter, 135-pound cylinders (magazines) through the 4 inch diameter neck of a stainless steel canister. Working through the narrow canister neck, the robot secures the magazines into a specially designed rack pre-installed in the canister. To provide the deterrent effect, the canisters are filled with a mixture of high-level waste and glass at the Defense Waste Processing Facility (DWPF)

  5. Techniques for freeing deposited canisters. Final report

    International Nuclear Information System (INIS)

    Kalbantner, P.; Sjoeblom, R.

    2000-06-01

    Four different techniques for removal of the bentonite buffer around a deposited canister have been identified, studied and evaluated: mechanical, hydrodynamical, thermal, and electrical techniques. Different techniques to determine the position of the canister in the buffer have also been studied: mechanical, electromagnetic, thermal and acoustic techniques. The mechanical techniques studied are full-face boring, milling and core-drilling. It is expected that the bentonite can be machined relatively easily. It is assessed that cooling by means of flushing water over the outer surfaces of the tools is not feasible in view of the tendency of bentonite to form a gel. The mechanical techniques are characterized by the potential of damaging the canister, a high degree of complexity, and high requirements of energy/power input. The generated byproduct is solid and cannot be removed by means of flushing. Removal is assessed to be simplest in conjunction with full-face boring and most difficult when coredrilling is applied. The hydrodynamical techniques comprise high-pressure hydrodynamic techniques, where pressures above and below 100 bar, and low pressure hydrodynamical techniques (< 10 bar) are separated. At pressures above 100 bar, a water jet with a diameter of approximately a millimetre cuts through the material. If desired, sand can be added to the jet. At pressures below 100 bar the jet has a diameter of one or a few centimetres. The liquid contains a few percent of salt, which is essential for the efficiency of the process. The flushing is important not only because it removes the modified bentonite but also because it frees previously unaffected bentonite and thereby makes it accessible to chemical modification. All of the hydrodynamical techniques are applicable for freeing the end surface as well as the mantle surface. The degree of complexity and the requirement on energy/power decrease with a decrease in pressure. A significant potential for damaging the

  6. Decontamination of high-level waste canisters

    International Nuclear Information System (INIS)

    Nesbitt, J.F.; Slate, S.C.; Fetrow, L.K.

    1980-12-01

    This report presents evaluations of several methods for the in-process decontamination of metallic canisters containing any one of a number of solidified high-level waste (HLW) forms. The use of steam-water, steam, abrasive blasting, electropolishing, liquid honing, vibratory finishing and soaking have been tested or evaluated as potential techniques to decontaminate the outer surfaces of HLW canisters. Either these techniques have been tested or available literature has been examined to assess their applicability to the decontamination of HLW canisters. Electropolishing has been found to be the most thorough method to remove radionuclides and other foreign material that may be deposited on or in the outer surface of a canister during any of the HLW processes. Steam or steam-water spraying techniques may be adequate for some applications but fail to remove all contaminated forms that could be present in some of the HLW processes. Liquid honing and abrasive blasting remove contamination and foreign material very quickly and effectively from small areas and components although these blasting techniques tend to disperse the material removed from the cleaned surfaces. Vibratory finishing is very capable of removing the bulk of contamination and foreign matter from a variety of materials. However, special vibratory finishing equipment would have to be designed and adapted for a remote process. Soaking techniques take long periods of time and may not remove all of the smearable contamination. If soaking involves pickling baths that use corrosive agents, these agents may cause erosion of grain boundaries that results in rough surfaces

  7. Design of double containment canister cask storage system

    International Nuclear Information System (INIS)

    Asami, M.; Matsumoto, T.; Oohama, T.; Kuriyama, K.; Kawakami, K.

    2004-01-01

    Spent fuels discharged from Japanese LWR will be stored as recycled-fuel-resources in interim storage facilities. The concrete cask storage system is one of important forms for the spent fuel interim storage. In Japan, the interim storage facility will be located near the coast, therefore it is important to prevent SCC (Stress Corrosion Cracking) caused by sea salt particles and to assure the containment integrity of the canister which contains spent fuels. KEPCO, NFT and OCL have designed the double containment canister cask storage system that can assure the long-term containment integrity and monitor the containment performance without storage capacity decrease. Major features of the combined canister cask system are shown as follows: This system can survey containment integrity of dual canisters by monitoring the pressure of the gap between canisters. The primary canister has dual lids sealed by welding. The secondary canister has single lid tightened by bolts and sealed by metallic gaskets. The primary canister is contained in the transport cask during transportation, and the gap between the primary canister and the transport cask is filled with He gas. Under storage condition in the concrete cask, the primary canister is contained in the secondary canister, and the gap between these canisters is filled with helium gas. Hence this system can prevent the primary canister to contact sea salt particle in the air and from SCC. Decrease of cooling performance because of the double canister is compensated by fins fitted on the secondary canister surface. Then, this system can prevent the decrease of storage capacity determined by the fuel temperature limit. This system can assure that the primary canister will keep intact for long term storage. Therefore, in the case of pressure down of the gap between canisters, it can be considered that the secondary canister containment is damaged, and the primary canister will be transferred to another secondary canister at the

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

    International Nuclear Information System (INIS)

    Sreten Mastilovic

    2001-01-01

    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

  9. Design report of the canister for nuclear fuel disposal

    International Nuclear Information System (INIS)

    Raiko, H.; Salo, J.P.

    1996-12-01

    The report provides a summary of the design of the canister for final disposal of 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 11 assemblies of BWR or VVER 440 fuel. The canister shall be tight with a high probability for about 100 000 years. The design basis of the canister is set, the performed analyses are summarised and the results are assessed and discussed in the report. (26 refs.)

  10. Criticality safety calculations for the nuclear waste disposal canisters

    International Nuclear Information System (INIS)

    Anttila, M.

    1996-12-01

    The criticality safety of the copper/iron canisters developed for the final disposal of the Finnish spent fuel has been studied with the MCNP4A code based on the Monte Carlo technique and with the fuel assembly burnup programs CASMO-HEX and CASMO-4. Two rather similar types of spent fuel disposal canisters have been studied. One canister type has been designed for hexagonal VVER-440 fuel assemblies used at the Loviisa nuclear power plant (IVO canister) and the other one for square BWR fuel bundles used at the Olkiluoto nuclear power plant (TVO canister). (10 refs.)

  11. Canister disposition plan for the DWPF Startup Test Program

    International Nuclear Information System (INIS)

    Harbour, J.R.; Payne, C.H.

    1990-01-01

    This report details the disposition of canisters and the canistered waste forms produced during the DWPF Startup Test Program. The six melter campaigns (DWPF Startup Tests FA-13, WP-14, WP-15, WP-16, WP-17, and FA-18) will produce 126 canistered waste forms. In addition, up to 20 additional canistered waste forms may be produced from glass poured during the transition between campaigns. In particular, this canister disposition plan (1) assigns (by alpha-numeric code) a specific canister to each location in the six campaign sequences, (2) describes the method of access for glass sampling on each canistered waste form, (3) describes the nature of the specific tests which will be carried out, (4) details which tests will be carried out on each canistered waste form, (5) provides the sequence of these tests for each canistered waste form, and (6) assigns a storage location for each canistered waste form. The tests are designed to provide evidence, as detailed in the Waste Form Compliance Plan (WCP 1 ), that the DWPF product will comply with the Waste Acceptance Product Specifications (WAPS 2 ). The WAPS must be met before the canistered waste form is accepted by DOE for ultimate disposal at the Federal Repository. The results of these tests will be included in the Waste Form Qualification Report (WQR)

  12. Near-field performance of the advanced cold process canister

    International Nuclear Information System (INIS)

    Werme, L.

    1991-12-01

    A near-field performance evaluation of an advanced cold process canister for spent fuel disposal has been performed jointly by TVO, Finland and SKB, Sweden. The canister consists of a steel canister as a load bearing element, with an outer corrosion shield of copper. In the analysis, as well internal (ie corrosion processes from the inside of the canister) as external processes (mechanical and chemical) have been considered both prior to and after canister breach. The major conclusions for the evaluation are: Internal processes cannot cause the canister breach under foreseen conditions, ie local-iced corrosion for the steel or copper canisters can be dismissed as a failure mechanism; The evaluation of the effects of processed outside the canister indicate that there is no rapid mechanism to endanger the integrity of the canister. Consequently the service life of the canister will be several million years. For completeness also evaluation of post-failure behaviour was carried out. Analyses were focussed on low probability phenomena from faults in canisters. Some items were identified where further research is justified in order to increase knowledge of the phenomena and thus strengthen the confidence of safety margins. However, it can be concluded that the risks of these scenarios can be judged to be acceptable. This is due to the fact that firstly, the probability of occurrence of most of these scenarios can be controlled to a large extent through technical measures. Secondly, these analyses indicated that the consequences would not be severe

  13. 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.

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

    International Nuclear Information System (INIS)

    Raiko, H.; Salo, J.P.

    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.)

  15. 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

  16. Decontamination processes for waste glass canisters

    International Nuclear Information System (INIS)

    Rankin, W.N.

    1981-06-01

    The process which will be used to decontaminate waste glass canisters at the Savannah River Plant consists of: decontamination (slurry blasting); rinse (high-pressure water); and spot decontamination (high-pressure water plus slurry). No additional waste will be produced by this process because glass frit used in decontamination will be mixed with the radioactive waste and fed into the glass melter. Decontamination of waste glass canisters with chemical and abrasive blasting techniques was investigated. The ability of a chemical technique with HNO 3 -HF and H 2 C 2 O 4 to remove baked-on contamination was demonstrated. A correlation between oxide removal and decontamination was observed. Oxide removal and, thus, decontamination by abrasive blasting techniques with glass frit as the abrasive was proposed and demonstrated

  17. SOURCE TERMS FOR HLW GLASS CANISTERS

    International Nuclear Information System (INIS)

    J.S. Tang

    2000-01-01

    This calculation is prepared by the Monitored Geologic Repository (MGR) Waste Package Design Section. The objective of this calculation is to determine the source terms that include radionuclide inventory, decay heat, and radiation sources due to gamma rays and neutrons for the high-level radioactive waste (HLW) from the, West Valley Demonstration Project (WVDP), Savannah River Site (SRS), Hanford Site (HS), and Idaho National Engineering and Environmental Laboratory (INEEL). This calculation also determines the source terms of the canister containing the SRS HLW glass and immobilized plutonium. The scope of this calculation is limited to source terms for a time period out to one million years. The results of this calculation may be used to carry out performance assessment of the potential repository and to evaluate radiation environments surrounding the waste packages (WPs). This calculation was performed in accordance with the Development Plan ''Source Terms for HLW Glass Canisters'' (Ref. 7.24)

  18. Decontamination processes for waste glass canisters

    International Nuclear Information System (INIS)

    Rankin, W.N.

    1982-01-01

    A Defense Waste Processing Facility (DWPF) is currently being designed to convert Savannah River Plant liquid, high-level radioactive waste into a solid form, such as borosilicate glass. To prevent the spread of radioactivity, the outside of the canisters of waste glass must have very low levels of smearable radioactive contamination before they are removed from the DWPF. Several techniques were considered for canister decontamination: high-pressure water spray, electropolishing, chemical dissolution, and abrasive blasting. An abrasive blasting technique using a glass frit slurry has been selected for use in the DWPF. No additional equipment is needed to process waste generated from decontamination. Frit used as the abrasive will be mixed with the waste and fed to the glass melter. In contrast, chemical and electrochemical techniques require more space in the DWPF, and produce large amounts of contaminated by-products, which are difficult to immobilize by vitrification

  19. Decontamination processes for waste glass canisters

    International Nuclear Information System (INIS)

    Rankin, W.N.

    1981-01-01

    The process which will be used to decontaminate waste glass canisters at the Savannah River Plant consists of: decontamination (slurry blasting); rinse (high-pressure water); and spot decontamination (high-pressure water plus slurry). No additional waste will be produced by this process because glass frit used in decontamination will be mixed with the radioactive waste and fed into the glass melter. Decontamination of waste glass canisters with chemical and abrasive blasting techniques was investigated. The ability of a chemical technique with HNO 3 -HF and H 2 C 2 O 4 to remove baked-on contamination was demonstrated. A correlation between oxide removal and decontamination was observed. Oxide removal and, thus, decontamination by abrasive blasting techniques with glass frit as the abrasive was proposed and demonstrated

  20. Multi-purpose canister project overview

    International Nuclear Information System (INIS)

    Williams, J.

    1995-01-01

    In this presentation, the author lists the approved and proposed dry storage technologies. He discusses the compatibility of dry storage systems with waste management systems. Historical aspects, recent history, key features of the program approach, benefits, specifications, acquisition and potential utility use of the multi-purpose canister (MPC) are covered. The MPCs provide standardization in the waste management system and a cost savings to utilities and government. MPC will be developed to the same level as existing dry storage systems

  1. Canister storage building hazard analysis report

    International Nuclear Information System (INIS)

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

    1997-01-01

    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

  2. Stress corrosion cracking of copper canisters

    International Nuclear Information System (INIS)

    King, Fraser; Newman, Roger

    2010-12-01

    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

  3. 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

  4. Fracturing of simulated high-level waste glass in canisters

    International Nuclear Information System (INIS)

    Peters, R.D.; Slate, S.C.

    1981-09-01

    Waste-glass castings generated from engineering-scale developmental processes at the Pacific Northwest Laboratory are generally found to have significant levels of cracks. The causes and extent of fracturing in full-scale canisters of waste glass as a result of cooling and accidental impact are discussed. Although the effects of cracking on waste-form performance in a repository are not well understood, cracks in waste forms can potentially increase leaching surface area. If cracks are minimized or absent in the waste-glass canisters, the potential for radionuclide release from the canister package can be reduced. Additional work on the effects of cracks on leaching of glass is needed. In addition to investigating the extent of fracturing of glass in waste-glass canisters, methods to reduce cracking by controlling cooling conditions were explored. Overall, the study shows that the extent of glass cracking in full-scale, passively-cooled, continuous melting-produced canisters is strongly dependent on the cooling rate. This observation agrees with results of previously reported Pacific Northwest Laboratory experiments on bench-scale annealed canisters. Thus, the cause of cracking is principally bulk thermal stresses. Fracture damage resulting from shearing at the glass/metal interface also contributes to cracking, more so in stainless steel canisters than in carbon steel canisters. This effect can be reduced or eliminated with a graphite coating applied to the inside of the canister. Thermal fracturing can be controlled by using a fixed amount of insulation for filling and cooling of canisters. In order to maintain production rates, a small amount of additional facility space is needed to accomodate slow-cooling canisters. Alternatively, faster cooling can be achieved using the multi-staged approach. Additional development is needed before this approach can be used on full-scale (60-cm) canisters

  5. Can-in-canister cold demonstration in DWPF (U)

    International Nuclear Information System (INIS)

    Kuehn, N.H.

    1996-07-01

    The Department of Energy Fissile Materials Disposition Program is evaluating a number of options for disposition of weapons-usable plutonium surplus to national defense needs. One of the immobilization options is the Can-In-Canister approach. In this option small cans of a plutonium glass, which contains a neutron absorber, are placed on a support structure in a large Savannah River Site Defense Waste Processing Facility (DWPF) canister. The top is then welded onto the canister. This canister is filled with High Level Waste (HLW) glass at the DWPF. The HLW glass provides the radiation source for proliferation resistance. These canisters are to be placed in a Federal Repository. To provide information on the technical feasibility of this option prior to the Record of Decision on plutonium disposition, the Department of Energy Fissile Materials Disposition Program funded a demonstration in the DWPF. This demonstration was conducted before the start of radioactive operations. Two test canisters containing cans of surrogate (non- radioactive) plutonium glass were successfully filled with simulated HLW glass at the DWPF using standard pouring procedures. One canister had twenty cans of surrogate plutonium glass. The other had eight cans of surrogate plutonium glass. After the canisters were filled, the contents of the canisters were examined to provide data on the effect of the rack and cans on the filling of the DWPF canister, the effect of the pour on the surrogate plutonium glass and the effect of the rack and cans on the simulated HLW glass. There was no deformation of the support racks during the pour. The simulated HLW glass filled all the regions around the rack and cans and the regions between the cans and the wall of the canister. This report discusses the design of the racks and cans, the modification of the DWPF canisters to accommodate the rack and cans, the conditions during the pours and the results of the post pour analysis

  6. Further assessment studies of the Advanced Cold Process Canister

    International Nuclear Information System (INIS)

    Henshaw, J.; Hoch, A.; Sharland, S.M.

    1990-08-01

    A preliminary assessment of the performance of the Advanced Cold Process Canister (ACPC) was carried out recently by Marsh. The aim of the study presented in this report is to re-examine the validity of some of the assumptions made, and re-evaluate the canister performance as appropriate. Two areas were highlighted in the preliminary study as requiring more detailed quantitative evaluation. 1) Assessment of the risk of internal stress-corrosion cracking induced by irradiation of moist air inside the canister if, under fault conditions, significant water was carried into the canister before sealing. 2) Evaluation of the corrosion behaviour subsequent to first breach of outer container. (author)

  7. Shaft shock absorber tests for a spent fuel canister

    International Nuclear Information System (INIS)

    Kukkola, T.; Toermaelae, V.P.

    2005-06-01

    The disposal canister for spent nuclear fuel will be transferred by a lift to the repository, which is 500 m deep in the bedrock. Model tests were carried out with the objective to estimate weather feasible shock absorber can be developed against the design accident case where the canister should survive a free fall to the lift shaft. If the velocity of the canister is not controlled by air drag or by any other deceleration means, the impact velocity may reach ultimate speed of 100m/s. The canister would retain its integrity in impact on water when the bottom pit of the lift well is filled with groundwater. However, the canister would hit the pit bottom with high velocity since the water hardly slows down the canister. The impact to the bottom of the pit should be dampened mechanically. The tests demonstrated that 20 m high filling to the bottom pit of the lift well by Light Expanded Clay Aggregate (LECA), gives fair impact absorption to protect the fuel canister. Presence of ground water is not harmful for impact absorption system provided that the ceramic gravel is not floating too high from the pit bottom. Almost ideal impact absorption conditions are met if the water high level does not exceed two thirds of the height of the gravel. Shaping of the bottom head of the cylindrical canister does not give meaningful advantages to the impact absorption system. The flat nose bottom head of the fuel canister gives adequate deceleration properties. (orig.)

  8. 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.

  9. 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.

  10. Canister storage building trade study. Final report

    International Nuclear Information System (INIS)

    Swenson, C.E.

    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

  11. Pitting corrosion on a copper canister

    International Nuclear Information System (INIS)

    Hermansson, H.P.; Beverskog, B.

    1996-02-01

    It is demonstrated that normal pitting can occur during oxidizing conditions in the repository. It is also concluded that a new theory for pitting corrosion has to be developed, as the present theory is not in accordance with all practical and experimental observations. A special variant of pitting, based on the growth of sulfide whiskers, is suggested to occur during reducing conditions. However, such a mechanism needs to be demonstrated experimentally. A simple calculational model of canister corrosion was developed based on the results of this study. 69 refs, 3 figs

  12. Multi-Canister overpack pressure testing

    International Nuclear Information System (INIS)

    SMITH, K.E.

    1998-01-01

    The Multi-Canister Overpack (MCO) shield plug closure assembly will be hydrostatically tested at the fabricator's shop to the 150 psig design test requirement in accordance with the ASME Code. Additionally, the MCO shell and collar will be hydrostatically tested at the fabricator's shop to the 450 psig design test requirement. Commercial practice has not required a pressure test of the closure weld after spent fuel is loaded in the containers. Based on this precedent and Code Case N-595-I, the MCO closure weld will not be pressure tested in the field

  13. Canister storage building hazard analysis report

    International Nuclear Information System (INIS)

    POWERS, T.B.

    1999-01-01

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

  14. Status of the Multipurpose Canister (MPC) Project

    International Nuclear Information System (INIS)

    Hopper, J.P.

    1996-01-01

    The multipurpose canister (MPC) project represents a cornerstone of the current U.S. Department of Energy's Office of Civilian Radioactive Waste Management (OCRWM) program for handling spent nuclear fuel. The MPC and associated support equipment is being designed to accommodate the requirements for not only storage and transport but also for the specified disposal requirements of the mined geologic repository system. The phase 1 design effort for the MPC system, being performed by the Westinghouse team on behalf of TRW Environmental Safety Systems (TESS), the OCRWM management ampersand operating (M ampersand O) contractor, is on schedule for delivery of completed safety analysis reports (SARs) in April 1996

  15. Choices of canisters and elements for the first fuel and canister sludge shipment from K East Basin

    International Nuclear Information System (INIS)

    Makenas, B.J.

    1996-01-01

    The K East Basin contains open-top canisters with up to fourteen N Reactor fuel assemblies distributed between the two barrels of each canister. Each fuel assembly generally consists of inner and outer concentric elements fabricated from uranium metal with zirconium alloy cladding. The canisters also contain varying amounts of accumulated sludge. Retrieval of sample fuel elements and associated sludge for examination is scheduled to occur in the near future. The purpose of this document is to specify particular canisters and elements of interest as candidate sources of fuel and sludge to be shipped to laboratories

  16. Test manufacture of the canister insert 135

    International Nuclear Information System (INIS)

    Raiko, H.

    2005-10-01

    This report describes the insert-manufacturing test of a disposal canister for spent nuclear fuel that was made by Metso Foundries Jyvaeskylae Oy, in June 2004 on contract for Posiva Oy. The test manufacture was a part of the co-operation development programme of encapsulation technology between SKB AB and Posiva Oy. Insert casting was specified according to the current manufacturing specifications of SKB. The canister insert was of BWR-type with integral bottom. This was the second trial manufacture of this type of insert in Finland and, in total, the third test manufacture of insert by Metso Foundries Jyvaeskylae Oy. The result fulfilled all the requirements but the material mechanical properties of the cast material. The measured ultimate strength and elongation at rupture were lower than specified in the upper part of the cast. The reason for this was revealed in the metallurgical investigation of the cast material. The cast contained slag (dross). Avoiding the dross formation will be the most demanding challenge of the forthcoming development of the cast procedure. (orig.)

  17. Multi-Canister Overpack (MCO) Topical Report

    International Nuclear Information System (INIS)

    LORENZ, B.D.

    2000-01-01

    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

  18. Design analysis report for the canister

    International Nuclear Information System (INIS)

    Raiko, Heikki; Sandstroem, Rolf; Ryden, Haakan; Johansson, Magnus

    2010-04-01

    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

  19. Test manufacture of a canister insert

    International Nuclear Information System (INIS)

    Raiko, H.

    2004-11-01

    This report describes the insert-manufacturing test of a disposal canister for spent nuclear fuel that was made by Metso Paper Oy, Jyvaeskylae Foundry, in 2003 on contract for Posiva Oy. The test manufacture was a part of the co-operation development programme of encapsulation technology between SKB AB and Posiva Oy. Insert casting was specified according to the current manufacturing specifications of SKB. The canister insert was of BWR-type with integral bottom. This was the first trial manufacture of this type of insert in Finland and, in total, the second test manufacture of insert by Metso Paper. The result fulfilled all the requirements but the material mechanical properties and metallurgical structure of the cast material. The measured tensile strength, ultimate strength and elongation at rupture were lower than specified. The reason for this was revealed in the metallurgical investigation of the cast material. The nodulizing of the graphite was not occurred during the casting process according to the requirements. (orig.)

  20. Canister storage building design basis accident analysis documentation

    International Nuclear Information System (INIS)

    KOPELIC, S.D.

    1999-01-01

    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 design basis accident analysis documentation

    Energy Technology Data Exchange (ETDEWEB)

    KOPELIC, S.D.

    1999-02-25

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

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

    International Nuclear Information System (INIS)

    CROWE, R.D.

    1999-01-01

    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

  3. Interim transfer canister for consolidating nuclear fuel rods

    International Nuclear Information System (INIS)

    Formanek, F.J.

    1987-01-01

    This patent describes a canister for receiving and consolidating a group of uniformly spaced apart nuclear fuel rods, comprising: a rectangular, vertically oriented straight back panel; a pair of oppositely disposed side panels connected perpendicularly to the back panel, having a vertical straight upper portion and an inwardly tapered lower portion; a front panel opposite the back panel and connected to the side panels, having a straight vertical upper portion and inwardly tapered lower portion; whereby the back, side and front panels define a rectangular upper opening at the upper end of the canister and a generally rectangular lower opening at the other end, the lower opening having a cross-sectional area less than one-half that of the upper opening; parallel plate members spanning the canister from the front panel to the back panel, each plate spaced from the other the same uniform distance, the plates extending downwardly into the tapered portion of the canister while remaining spaced above the tapered sidewalls; first base means at the lower end of the canister, removably mounted and having an oblique orientation generally downward from the front panel to the back panel, for guiding the fuel rods to be inserted preferentially toward the lower portion of the back panel; and second base means removably mounted within the canister below first base means and oriented transversely to the longitudinal extent of the canister, for supporting the fuel rods when the first base means is removed from the canister

  4. BRIC-100VC Biological Research in Canisters (BRIC)-100VC

    Science.gov (United States)

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

    2016-01-01

    The Biological Research in Canisters (BRIC) is an anodized-aluminum cylinder used to provide passive stowage for investigations of the effects of space flight on small specimens. The BRIC 100 mm petri dish vacuum containment unit (BRIC-100VC) has supported Dugesia japonica (flatworm) within spring under normal atmospheric conditions for 29 days in space and Hemerocallis lilioasphodelus L. (daylily) somatic embryo development within a 5% CO2 gaseous environment for 4.5 months in space. BRIC-100VC is a completely sealed, anodized-aluminum cylinder (Fig. 1) providing containment and structural support of the experimental specimens. The top and bottom lids of the canister include rapid disconnect valves for filling the canister with selected gases. These specialized valves allow for specific atmospheric containment within the canister, providing a gaseous environment defined by the investigator. Additionally, the top lid has been designed with a toggle latch and O-ring assembly allowing for prompt sealing and removal of the lid. The outside dimensions of the BRIC-100VC canisters are 16.0 cm (height) x 11.4 cm (outside diameter). The lower portion of the canister has been equipped with sufficient storage space for passive temperature and relative humidity data loggers. The BRIC- 100VC canister has been optimized to accommodate standard 100 mm laboratory petri dishes or 50 mL conical tubes. Depending on storage orientation, up to 6 or 9 canisters have been flown within an International Space Station (ISS) stowage locker.

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

    International Nuclear Information System (INIS)

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

    2000-01-01

    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

  6. Design basis for the copper/steel canister

    International Nuclear Information System (INIS)

    Bowyer, W.H.

    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

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

    International Nuclear Information System (INIS)

    JOHNSON, D.M.

    2000-01-01

    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

  8. Evaluation of canister weld flaw depth for concrete storage cask

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Tae Chul; Cho, Chun Hyung [Korea Radioactive Waste Agency, Daejeon (Korea, Republic of); Jung, Sung Hun; Lee, Young Oh; Jung, In Su [Korea Nuclear Engineering and Service Corp, Daejeon (Korea, Republic of)

    2017-03-15

    Domestically developed concrete storage casks include an internal canister to maintain the confinement integrity of radioactive materials. In this study, we analyzed the depth of flaws caused by loads that propagate canister weld cracks under normal, off-normal and accident conditions, and evaluated the maximum allowable weld flaw depth needed to secure the structural integrity of the canister weld and to reduce the welding time of the internal canister lid of the concrete storage cask. Structural analyses for normal, off-normal and accident conditions were performed using the general-purpose finite element analysis program ABAQUS; the allowable flaw depth was assessed according to ASME B and PV Code Section XI. Evaluation results revealed an allowable canister weld flaw depth of 18.75 mm for the concrete storage cask, which satisfies the critical flaw depth recommended in NUREG-1536.

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

    International Nuclear Information System (INIS)

    Bond, A.E.; Hoch, A.R.; Jones, G.D.; Tomczyk, A.J.; Wiggin, R.M.; Worraker, W.J.

    1997-05-01

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

  10. 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

  11. Multi-Canister overpack sealing configuration

    International Nuclear Information System (INIS)

    SMITH, K.E.

    1998-01-01

    The Spent Nuclear Fuel (SNF) position regarding the Multi-Canister Overpack (MCO) sealing configuration is to initially rely on an American Society of Mechanical Engineers (ASME) Section III Subsection NB code compliant mechanical closure/sealing system to quickly and safely establish and maintain full confinement of radioactive materials prior to and during MCO fuel drying activities. Previous studies have shown the mechanical seal to be the preferred closure method, based on dose, cost, and schedule considerations. The cost and schedule impacts of redesigning the mechanical closure to a welded shield plug do not support changing the closure system. The SNF Project has determined that the combined mechanical/welded closure system meets or exceeds the regulatory requirements to provide redundant seals while accommodating key safety and schedule limitations that are unique to K Basins fuel removal effort

  12. Multi-Canister overpack internal HEPA filters

    International Nuclear Information System (INIS)

    SMITH, K.E.

    1998-01-01

    The rationale for locating a filter assembly inside each Multi-Canister Overpack (MCO) rather than include the filter in the Cold Vacuum Drying (CVD) process piping system was to eliminate the potential for contamination to the operators, processing equipment, and the MCO. The internal HEPA filters provide essential protection to facility workers from alpha contamination, both external skin contamination and potential internal depositions. Filters installed in the CVD process piping cannot mitigate potential contamination when breaking the process piping connections. Experience with K-Basin material has shown that even an extremely small release can result in personnel contamination and costly schedule disruptions to perform equipment and facility decontamination. Incorporating the filter function internal to the MCO rather than external is consistent with ALARA requirements of 10 CFR 835. Based on the above, the SNF Project position is to retain the internal HEPA filters in the MCO design

  13. Shippingport Spent Fuel Canister System Description

    International Nuclear Information System (INIS)

    JOHNSON, D.M.

    2000-01-01

    In 1978 and 1979, a total of 72 blanket fuel assemblies (BFAs), irradiated during the operating cycles of the Shippingport Atomic Power Station's Pressurized Water Reactor (PWR) Core 2 from April 1965 to February 1974, were transferred to the Hanford Site and stored in underwater storage racks in Cell 2R at the 221-T Canyon (T-Plant). The initial objective was to recover the produced plutonium in the BFAs, but this never occurred and the fuel assemblies have remained within the water storage pool to the present time. The Shippingport Spent Fuel Canister (SSFC) is a confinement system that provides safe transport functions (in conjunction with the TN-WHC cask) and storage for the BFAs at the Canister Storage Building (CSB). The current plan is for these BFAs to be retrieved from wet storage and loaded into SSFCs for dry storage. The sealed SSFCs containing BFAs will be vacuum dried, internally backfilled with helium, and leak tested to provide suitable confinement for the BFAs during transport and storage. Following completion of the drying and inerting process, the SSFCs are to be delivered to the CSB for closure welding and long-term interim storage. The CSB will provide safe handling and dry storage for the SSFCs containing the BFAs. The purpose of this document is to describe the SSFC system and interface equipment, including the technical basis for the system, design descriptions, and operations requirements. It is intended that this document will be periodically updated as more equipment design and performance specification information becomes available

  14. 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

  15. Development of the DWPF canister temporary shrink-fit seal

    International Nuclear Information System (INIS)

    Kelker, J.W. Jr.

    1986-04-01

    The Defense Waste Processing Facility is being constructed at The Savannah River Plant for the containerization of high-level nuclear waste in a wasteform for eventual permanent disposal. The waste will be incorporated in molten glass and solidified in type 304L stainless steel canisters, 2-feet in diameter x 9-feet 10-inches long, containing a flanged 6-in.-diam pipe fill-nozzle. The canisters have a minimum wall thickness of 3/8 in. Utilizing the heat from the glass filling operation, a shrink-fit seal for a plug in the end of the canister fill nozzle was developed that: will withstand the radioactive environment; will prevent the spread of contamination, and will keep moisture and water from entering the canister during storage and decontamination of the canister by wet-frit blasting to remove smearable and oxide-film fixed radioactive nuclides; is removable and can be replaced by a new oversize plug in the event the seal fails the pressure decay leakage test ( -4 atm cc/sec helium); will keep the final weld closure clean and free of nuclear contamination; will withstand being pressed into the nozzle without exposing external contamination or completely breaking the seal; is reliable; and is easily installed. The seal consists of: a removable sleeve (with a tapered bore) which is shrink-fitted into the nozzle bore during canister fabrication; and a tapered plug which is placed into the sleeved nozzle after the canister is filled with radioactive molten glass. A leak-tight shrink-fit seal is formed between the nozzle, sleeve, and plug upon temperature equilibrium. The temporarily sealed canister is transferred from the Melt cell to the Decon cell, and the surface is decontaminated. Next it is transferred to the Weld/Test cell where the temporary seal is pressed down into the nozzle, revealing a clean cavity where the canister final closure weld is made

  16. Shaft shock absorber tests for a spent fuel canister

    International Nuclear Information System (INIS)

    Kukkola, T.; Toermaelae, V.P.

    2003-01-01

    The holding canister for spent nuclear fuel will be transferred by a lift to the final disposal tunnels 500m deep in the bedrock. Model tests were carried out with an objective to estimate weather feasible shock absorbing properties can be met in a design accident case where the canister should survive a free fall due to e.g. sabotage. If the velocity of the canister is not controlled by air drag or any other deceleration means, the impact velocity may reach ultimate speed of 100m/s. The canister would retain its integrity when stricken by the surface penetration impact if the bottom pit of the lift well would be filled with groundwater. However the canister would hit the pit bottom with high velocity since the water hardly slows down the canister. The impact to the bottom of the pit should be dampened mechanically. The tests demonstrated that 20m high filling to the bottom pit of the lift well by ceramic gravel, trade mark LECA-sora, gives a fair impact absorption to protect the spent fuel canister. Presence of ground water is not harmful for impact absorption system provided that the ceramic gravel is not floating too high from the pit bottom. Almost ideal impact absorption conditions are met if the water high level does not exceed two thirds of the height of the gravel. Shaping of the bottom head of the cylindrical canister does not give meaningful advantages to the impact absorption system. The flat nose bottom head of the fuel canister gives adequate deceleration properties. (orig.)

  17. DOE requests waiver on double containment for HLW canisters

    International Nuclear Information System (INIS)

    Lobsenz, G.

    1994-01-01

    The Energy Department has asked the Nuclear Regulatory Commission to waive double containment requirements for vitrified high-level radioactive waste canisters, saying the additional protection is not necessary and too costly. NRC said it had received a petition from DOE contending that the vitrified waste canisters were durable enough without double containment to prevent any potential plutonium release during handling and shipping. DOE said testing had shown that the vitrified waste canisters were similar - even superior - in durability to spent reactor fuel shipments, which NRC specifically exempted from the double containment requirement

  18. 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

  19. Zero-Headspace Coal-Core Gas Desorption Canister, Revised Desorption Data Analysis Spreadsheets and a Dry Canister Heating System

    Science.gov (United States)

    Barker, Charles E.; Dallegge, Todd A.

    2005-01-01

    Coal desorption techniques typically use the U.S. Bureau of Mines (USBM) canister-desorption method as described by Diamond and Levine (1981), Close and Erwin (1989), Ryan and Dawson (1993), McLennan and others (1994), Mavor and Nelson (1997) and Diamond and Schatzel (1998). However, the coal desorption canister designs historically used with this method have an inherent flaw that allows a significant gas-filled headspace bubble to remain in the canister that later has to be compensated for by correcting the measured desorbed gas volume with a mathematical headspace volume correction (McLennan and others, 1994; Mavor and Nelson, 1997).

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

    International Nuclear Information System (INIS)

    Hoekmark, Harald; Faelth, Billy

    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

  1. MCC-15: waste/canister accident testing and analysis method

    International Nuclear Information System (INIS)

    Slate, S.C.; Pulsipher, B.A.; Scott, P.A.

    1985-02-01

    The Materials Characterization Center (MCC) at the Pacific Northwest Laboratory (PNL) is developing standard tests to characterize the performance of nuclear waste forms under normal and accident conditions. As part of this effort, the MCC is developing MCC-15, Waste/Canister Accident Testing and Analysis. MCC-15 is used to test canisters containing simulated waste forms to provide data on the effects of accidental impacts on the waste form particle size and on canister integrity. The data is used to support the design of transportation and handling equipment and to demonstrate compliance with repository waste acceptance specifications. This paper reviews the requirements that led to the development of MCC-15, describes the test method itself, and presents some early results from tests on canisters representative of those proposed for the Defense Waste Processing Facility (DWPF). 13 references, 6 figures

  2. Multi-canister overpack operations and maintenance manual

    International Nuclear Information System (INIS)

    PIERCE, S.R.

    1999-01-01

    This manual provides general operating and maintenance instructions for the Multi-Canister Overpack. Procedure outlines included are conceptual in nature and will be modified, expanded, and refined during preparation of detailed operating procedures

  3. TMI-2 fuel canister interface requirements for INEL. Revision 1

    International Nuclear Information System (INIS)

    Wilkins, D.E.; Martz, D.E.; Reno, H.W.

    1984-06-01

    This report focuses on fuel canister interface requirements at INEL which should be incorporated into the canister design criteria. The requirements will ensure compatibility with existing INEL structures and equipment to be used for receipt, unloading, and storage of fuel canisters. INEL can and does receive and store radioactive materials in many different forms, including reactor fuel. INEL requires detailed descriptions of canisters and casks. Therefore, requirements listed represent engineering design features which will simplify the handling and storage operations; consequently, they are not to be viewed as absolute or non-negotiable. However, the core acquisition contract was negotiated with certain storage assumptions which effect costs of storage. Deviations from those assumptions which significantly effect costs would require approval by DOE-Idaho. If some stated requirements are too restrictive, modifications based on sound engineering principles may be negotiated with INEL. 11 figures

  4. 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.

  5. 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.

  6. Grain boundary corrosion of copper canister weld material

    International Nuclear Information System (INIS)

    Gubner, Rolf; Andersson, Urban; Linder, Mats; Nazarov, Andrej; Taxen, Claes

    2006-01-01

    The proposed design for a final repository for spent fuel and other long-lived residues in Sweden is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will then be placed in granite bedrock and surrounded by compacted bentonite clay. The canister design is based on a thick cast inner container fitted inside a corrosion-resistant copper canister. During fabrication of the outer copper canisters there will be some unavoidable grain growth in the welded areas. As grains grow, they will tend to concentrate impurities within the copper at the new grain boundaries. The work described in this report was undertaken to determine whether there is any possibility of enhanced corrosion at grain boundaries within the copper canister, based on the recommendations of the report SKB-TR--01-09 (INIS ref. 32025363). Grain boundary corrosion of copper is not expected to be a problem for the copper canisters in a repository. However, as one step in the experimental verification it is necessary to study grain boundary corrosion of copper in an environment where it may occur. A literature study aimed to find one or several solutions that are aggressive with respect to grain boundary corrosion of copper. Copper specimens cut from welds of real copper canisters where exposed to aerated ammonium hydroxide solution for a period of 14 days at 80 degrees C and 10 bar pressure. The samples were investigated prior to exposure using the scanning Kelvin probe technique to characterize anodic and cathodic areas on the samples. The degree of corrosion was determined by optical microscopy. No grain boundary corrosion could be observed in the autoclave experiments, however, a higher rate of corrosion was observed for the weld material compared to the base material. The work suggests that grain boundary corrosion of copper weld material is most unlikely to adversely affect SKB's copper canisters under the conditions in the repository

  7. Spent nuclear fuel canister storage building conceptual design report

    International Nuclear Information System (INIS)

    Swenson, C.E.

    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

  8. Characterization of materials for waste-canister compatibility studies

    International Nuclear Information System (INIS)

    McCoy, H.E.; Mack, J.E.

    1981-10-01

    Sample materials of 7 waste forms and 15 potential canister materials were procured for compatibility tests. These materials were characterized before being placed in test, and the results are the main topic of this report. A test capsule was designed for the tests in which disks of a single waste form were contacted with duplicate samples of canister materials. The capsules are undergoing short-term tests at 800 0 C and long-term tests at 100 and 300 0 C

  9. Grain boundary corrosion of copper canister weld material

    Energy Technology Data Exchange (ETDEWEB)

    Gubner, Rolf; Andersson, Urban; Linder, Mats; Nazarov, Andrej; Taxen, Claes [Corrosion and Metals Research Inst. (KIMAB), Stockholm (Sweden)

    2006-01-15

    The proposed design for a final repository for spent fuel and other long-lived residues in Sweden is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will then be placed in granite bedrock and surrounded by compacted bentonite clay. The canister design is based on a thick cast inner container fitted inside a corrosion-resistant copper canister. During fabrication of the outer copper canisters there will be some unavoidable grain growth in the welded areas. As grains grow, they will tend to concentrate impurities within the copper at the new grain boundaries. The work described in this report was undertaken to determine whether there is any possibility of enhanced corrosion at grain boundaries within the copper canister, based on the recommendations of the report SKB-TR--01-09 (INIS ref. 32025363). Grain boundary corrosion of copper is not expected to be a problem for the copper canisters in a repository. However, as one step in the experimental verification it is necessary to study grain boundary corrosion of copper in an environment where it may occur. A literature study aimed to find one or several solutions that are aggressive with respect to grain boundary corrosion of copper. Copper specimens cut from welds of real copper canisters where exposed to aerated ammonium hydroxide solution for a period of 14 days at 80 degrees C and 10 bar pressure. The samples were investigated prior to exposure using the scanning Kelvin probe technique to characterize anodic and cathodic areas on the samples. The degree of corrosion was determined by optical microscopy. No grain boundary corrosion could be observed in the autoclave experiments, however, a higher rate of corrosion was observed for the weld material compared to the base material. The work suggests that grain boundary corrosion of copper weld material is most unlikely to adversely affect SKB's copper canisters under the conditions in the repository.

  10. QA/QC For Radon Concentration Measurement With Charcoal Canister

    International Nuclear Information System (INIS)

    Pantelic, G.; Zivanovic, M.; Rajacic, M.; Krneta Nikolic, J.; Todorovic, D.

    2015-01-01

    The primary concern of any measuring of radon or radon progeny must be the quality of the results. A good quality assurance program, when properly designed and diligently followed, ensures that laboratory staff will be able to produce the type and quality of measurement results which is needed and expected. Active charcoal detectors are used for testing the concentration of radon in dwellings. The method of measurement is based on radon adsorption on coal and measurement of gamma radiation of radon daughters. Upon closing the detectors, the measurement was carried out after achieving the equilibrium between radon and its daughters (at least 3 hours) using NaI or HPGe detector. Radon concentrations as well as measurement uncertainties were calculated according to US EPA protocol 520/5-87-005. Detectors used for the measurements were calibrated by 226Ra standard of known activity in the same geometry. Standard and background canisters are used for QA and QC, as well as for the calibration of the measurement equipment. Standard canister is a sealed canister with the same matrix and geometry as the canisters used for measurements, but with the known activity of radon. Background canister is a regular radon measurement canister, which has never been exposed. The detector background and detector efficiency are measured to ascertain whether they are within the warning and acceptance limits. (author).

  11. Drop tests of the Three Mile Island knockout canister

    International Nuclear Information System (INIS)

    Box, W.D.; Aaron, W.S.; Shappert, L.B.; Childress, P.C.; Quinn, G.J.; Smith, J.V.

    1986-09-01

    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

  12. Research on corrosion aspects of the advanced cold process canister

    International Nuclear Information System (INIS)

    Blackwood, D.J.; Hoch, A.R.; Naish, C.C.; Rance, A.

    1994-01-01

    The Advanced Cold Process Canister (ACPC) is a waste canister being developed jointly by SKB and TVO for the disposal of spent nuclear fuel. It comprises an outer copper canister, with a carbon steel canister inside. A concern regarding the use of the ACPC is that, in the unlikely event that the outer copper canister is penetrated, the anaerobic corrosion of the carbon steel container may result in the formation of hydrogen gas bubbles. These bubbles could disrupt the backfill, and thus increase water flow through the near field and the flux of radionuclides to the host geology. A number of factors that influence the rate at which hydrogen evolves as a result of the anaerobic corrosion of carbon steel in artificial granitic groundwaters have been investigated. A previously observed, time-dependent decline in the hydrogen evolution rate has been confirmed as being due to the production of magnetite film. Once the magnetite film is about 0.7-1.0 μm thick, the rate of hydrogen evolution reaches a steady state value. The pH and the ionic strength of the groundwater were both found to influence the long-term hydrogen evolution rate. The results of the experimental programme were used to update a model of the corrosion behaviour and hydrogen production from the Advanced Cold Process Canister. 36 figs, 5 tabs, 13 refs

  13. Physical properties of encapsulate spent fuel in canisters

    International Nuclear Information System (INIS)

    1999-01-01

    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)

  14. 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.

  15. 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.

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

    International Nuclear Information System (INIS)

    TU, K.C.

    1999-01-01

    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 and transmittal record

  17. Mechanical Integrity of Canisters Using a Fracture Mechanics Approach

    Energy Technology Data Exchange (ETDEWEB)

    Koyama, Tomofumi; Guoxiang Zhang; Lanru Jing [Royal Inst. of Technology, Stockholm (Sweden). Dept. of Land and Water Resources Engineering

    2006-07-15

    This report presents the methods and results of a research project about numerical modeling of mechanical integrity of cast-iron canisters for the final disposal of spent nuclear fuel in Sweden, using combined boundary element (BEM) and finite element (FEM) methods. The objectives of the project are: 1) to investigate the possibility of initiation and growth of fractures in the cast-iron canisters under the mechanical loading conditions defined in the premises of canister design by Swedish Nuclear Fuel and Waste Management Co. (SKB); 2) to investigate the maximum bearing capacity of the cast iron canisters under uniformly distributed and gradually increasing boundary pressure until plastic failure. Achievement of the two objectives may provide some quantitative evidence for the mechanical integrity and overall safety of the cast-iron canisters that are needed for the final safety assessment of the geological repository of the radioactive waste repository in Sweden. The geometrical dimension, distribution and magnitudes of loads and Material properties of the canisters and possible fractures were provided by the latest investigations of SKB. The results of the BEM simulations, using the commercial code BEASY, indicate that under the currently defined loading conditions the possibility of initiation of new fractures or growth of existing fractures (defects) are very small, due to the reasons that: 1) the canisters are under mainly compressive stresses; 2) the induced tensile stress regions are too small in both dimension and magnitude to create new fractures or to induce growth of existing fractures, besides the fact that the toughness of the fractures in the cast iron canisters are much higher that the stress intensity factors in the fracture tips. The results of the FEM simulation show a approximately 75 MPa maximum pressure beyond which plastic collapse of the cast-iron canisters may occur, using an elastoplastic Material model. This figure is smaller compared

  18. 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

  19. Design, production and initial state of the canister

    International Nuclear Information System (INIS)

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

    2010-12-01

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

  20. Burst Test Qualification Analysis of DWPF Canister-Plug Weld

    International Nuclear Information System (INIS)

    Gupta, N.K.; Gong, Chung.

    1995-02-01

    The DWPF canister closure system uses resistance welding for sealing the canister nozzle and plug to ensure leak tightness. The welding group at SRTC is using the burst test to qualify this seal weld in lieu of the shear test in ASME B ampersand PV Code, Section IX, paragraph QW-196. The burst test is considered simpler and more appropriate than the shear test for this application. Although the geometry, loading and boundary conditions are quite different in the two tests, structural analyses show similarity in the failure mode of the shear test in paragraph QW-196 and the burst test on the DWPF canister nozzle Non-linear structural analyses are performed using finite element techniques to study the failure mode of the two tests. Actual test geometry and realistic stress strain data for the 304L stainless steel and the weld material are used in the analyses. The finite element models are loaded until failure strains are reached. The failure modes in both tests are shear at the failure points. Based on these observations, it is concluded that the use of a burst test in lieu of the shear test for qualifying the canister-plug weld is acceptable. The burst test analysis for the canister-plug also yields the burst pressures which compare favorably with the actual pressure found during burst tests. Thus, the analysis also provides an estimate of the safety margins in the design of these vessels

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

    International Nuclear Information System (INIS)

    C.A Kouts

    2006-01-01

    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

  2. Canisters and nonfuel components at commercial nuclear reactors

    International Nuclear Information System (INIS)

    Gibbard, K.; Disbrow, J.

    1994-01-01

    This paper discusses detailed data on canisters and nonfuel components (NFC) at US commercial nuclear power reactors. A wide variety of NFC have been reported on the Form RW-859, open-quotes Nuclear Fuel Dataclose quotes survey. They may have been integral with an assembly, noncanistered in baskets, destined for disposal as low-level radioactive waste, or stored in canisters. Similarly, data on the family of canistered spent nuclear fuel (SNF) in storage pools was compiled. Approximately 85 percent of the 40,194 pieces of nonfuel assembly (NFA) hardware reported were integral with an assembly. This represents data submitted by 95 of the 107 reactors in 10 generic assembly classes. In addition, a total of 286 canisters have been reported as being in storage pools as of December 31, 1992. However, an additional 264 open baskets were also reported to contain miscellaneous SNF and nonfuel materials, garbage and debris. All of these 286 canisters meet the dimensional envelope requirements specified for disposal for open-quotes standard fuelclose quotes under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (10 CFR 961); most of the baskets do not

  3. SNF Interim Storage Canister Corrosion and Surface Environment Investigations

    International Nuclear Information System (INIS)

    Bryan, Charles R.; Enos, David G.

    2015-01-01

    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.

  4. Electropolishing decontamination system for high-level waste canisters

    International Nuclear Information System (INIS)

    Larson, D.E.; Berger, D.N.; Allen, R.P.; Bryan, G.H.; Place, B.G.

    1988-10-01

    As part of a US Department of Energy (DOE) project agreement with the Federal Ministry for Research and Technology (BMFT) in the Federal Republic of Germany (FRG). The Nuclear Waste Treatment Program at the Pacific Northwest Laboratory (PNL) is preparing 30 radioactive canisters containing borosilicate glass for use in high-level waste repository related tests at the Asse Salt Mine. After filling, the canisters will be welded closed and decontaminated in preparation for shipping to the FRG. Electropolishing was selected as the primary decontamination approach, and an electropolishing system with associated canister inspection equipment has been designed and fabricated for installation in a large hot cell. This remote electropolishing system, which is currently undergoing preliminary testing, is described in this report. 3 refs., 3 figs., 1 tab

  5. Drop tests of the Three Mile Island knockout canister

    International Nuclear Information System (INIS)

    Box, W.D.; Aaron, W.S.; Shappert, L.B.; Childress, P.C.; Quinn, G.J.; Smith, J.V.

    1987-01-01

    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 confirmed 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. This report presents the data generated and the results obtained from a series of four drop tests that included two drops with the test assembly in the vertical position and two drops with the assembly in the horizontal position

  6. A welding system for spent fuel canister lid

    International Nuclear Information System (INIS)

    Suikki, M.; Wendelin, T.

    2008-06-01

    The report presents a proposed welding system for spent fuel canister lids. The system is used for welding the copper lid to the copper overpack. The apparatus will be installed in the encapsulation plant. The report presents basic requirements for and implementation of the welding system, operation, service and maintenance of the equipment, as well as a cost estimate. Some aspects of the apparatus design are quite specified, but the actual detailed planning and final selection of components is not included. The report also describes actions for possible malfunction and fault conditions. Closing of the copper cylinder's lid is carried out by electron beam welding, which must be performed in vacuum. The welding system for spent fuel canister lid consists of two welding chambers, a canister docking system, an EB-welding machine with its accessories, a vacuum apparatus, as well as necessary auxiliary equipment. The system's equipment is housed in a welding room, an auxiliary system room, an operation control room, as well as mounted on the ceiling of a transfer corridor. One of the welding chambers is intended for carrying out test welding procedures and for calibration of welding parameters. The actual spent fuel canister lid welding chamber has a weldingready canister docked thereto in an airtight manner. The chamber is pumped for a vacuum, followed by closing the canister's copper lid and carrying out the lid welding process. The lid is brought into the chamber prior to docking the canister by means of a canister transfer trolley lifting gear. Lifting of the canister and rotating it during a welding process are also handled by means of the transfer trolley. The lid welding chamber houses equipment for the alignment and installation of the lid, as well as heating means for the top side of a copper overpack for ensuring a sufficient installation clearance between the lid and the overpack. The equipment not needed in the immediate vicinity of welding chambers, is

  7. Development of cold sprayed Cu coating for canister

    International Nuclear Information System (INIS)

    Kim, Hyung Jun; Kang, Yoon Ha

    2010-01-01

    Cold sprayed Cu deposition was studied for the application of outer part of canister for high level nuclear waste. Five commercially available pure Cu powders were analyzed and sprayed by high pressure cold spray system. Electrochemical corrosion test using potentiostat in 3.5% NaCl solution was conducted as well as microstructural analysis including hardness and oxygen content measurements. Overall evaluation of corrosion performance of cold sprayed Cu deposition is inferior to forged and extruded Cu plates, but some of Cu depositions are comparable to Cu plates. The simulated corrosion test in 200m underground cave is still in progress. The effect of cold spray process parameters was also studied and the results show that the type of nozzle is the most important other than powder feed rate, spray distance, and scan speed. 1/10 scale miniature of canister was manufactured confirming that the production of full scale canister is possible

  8. Prototype spent-fuel canister design, analysis, and test

    International Nuclear Information System (INIS)

    Leisher, W.B.; Eakes, R.G.; Duffey, T.A.

    1982-03-01

    Sandia National Laboratories was asked by the US Energy Research and Development Administration (now US Department of Energy) to design the spent fuel shipping cask system for the Clinch River Breeder Reactor Plant (CRBRP). As a part of this task, a canister which holds liquid sodium and the spent fuel assembly was designed, analyzed, and tested. The canister body survived the regulatory Type-B 9.1-m (30-ft) drop test with no apparent leakage. However, the commercially available metal seal used in this design leaked after the tests. This report describes the design approach, analysis, and prototype canister testing. Recommended work for completing the design, when funding is available, is included

  9. High-level radioactive waste glass and storage canister design

    International Nuclear Information System (INIS)

    Slate, S.C.; Ross, W.A.

    1979-01-01

    Management of high-level radioactive wastes is a primary concern in nuclear operations today. The main objective in managing these wastes is to convert them into a solid, durable form which is then isolated from man. A description is given of the design and evaluation of this waste form. The waste form has two main components: the solidified waste and the storage canister. The solid waste form discussed in this study is glass. Waste glasses have been designed to be inert to water attack, physically rugged, low in volatility, and stable over time. Two glass-making processes are under development at PNL. The storage canister is being designed to provide high-integrity containment for solidified wastes from processing to terminal storage. An outline is given of the steps in canister design: material selection, stress and thermal analyses, quality verification, and postfill processing. Examples are given of results obtained from actual nonradioactive demonstration tests. 14 refs

  10. Test plan for the Sample Transfer Canister system

    International Nuclear Information System (INIS)

    Flanagan, B.D.

    1998-01-01

    The Sample Transfer Canister will be used by the Waste Receiving and Processing Facility (WRAP) for the transport of small quantity liquid samples that meet the definition of a limited quantity radioactive material, and may also be corrosive and/or flammable. These samples will be packaged and shipped in accordance with the US Department of Transportation (DOT) regulation 49 CFR 173.4, ''Exceptions for small quantities.'' The Sample Transfer Canister is of a ''French Can'' design, intended to be mated with a glove box for loading/unloading. Transport will typically take place north of the Wye Barricade between WRAP and the 222-S Laboratory. The Sample Transfer Canister will be shipped in an insulated ice chest, but the ice chest will not be a part of the small quantity package during prototype testing

  11. Waste canister closure welding using the inertia friction welding process

    International Nuclear Information System (INIS)

    Klein, R.F.; Siemens, D.H.; Kuruzar, D.L.

    1986-02-01

    Liquid radioactive waste presently stored in underground tanks is to undergo a vitrifying process which will immobilize it in a solid form. This solid waste will be contained in a stainless steel canister. The canister opening requires a positive seal weld, the properties and thickness of which are at least equal to those of the canister material. This paper describes the inertia friction welding process and a proposed equipment design concept that will provide a positive, reliable, inspectable, and full thickness seal weld while providing easily maintainable equipment, even though the weld is made in a highly contaminated hot cell. All studies and tests performed have shown the concept to be highly feasible. 2 refs., 6 figs

  12. 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.

  13. Canisters and nonfuel components at commercial nuclear reactors

    International Nuclear Information System (INIS)

    Gibbard, K.; Thorpe, J.; Moore, R.S.

    1995-01-01

    The Energy Information Administration of the U.S. Department of Energy (DOE) collects data annually from commercial nuclear power reactors via the Nuclear Fuel Data survey, Form RW-859. Over the past three years, the survey has collected data on the quantities and types of nonfuel components and on the quantities and contents of canisters in storage at reactor sites. This paper focuses on the annual changes in the data, specific implications of these changes, and lessons that should be applied to future revisions of the study. The total number of canisters reported by utilities for each year from 1986 to 1993 is listed. Changes in the quantities of nonfuel components report by General Reactors from 1992 to 1993 are also provided. Comparisons of canister and nonfuel components components data from year to year and from reactor to reactor point out that survey questions on these topics have been interpreted differently by reactor personnel

  14. Development of Copper Canister through Cold Sprayed Coating Method

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Min Soo; Choi, Jong Won; Choi, Heui Joo; Lee, Jong Youl; Jeong, Jong Tae; Kim, Sung Ki; Cho, Dong Keun

    2007-12-15

    General thickness of a copper canister is 5 cm for a underground disposal application. The lower limit of a thickness is determined by a forging technology. But many experts in this area agrees that the thickness 1 cm is enough at the underground disposal for the life time of 1,000,000 years. Thus new technology is suggested for the making 1 cm thickness copper canister, that is a cold spray coating method(CSC). In this report, the CSC is examined and the technical possibility for making copper canister is measured. The overview of CSC and its characteristics are discussed. Various copper particles for the CSC are analyzed and the formed coating layers are examined to find their porosity and uniformity. A Tafa copper particle and Chang-sung copper particle are selected for making 1 cm thick test specimen. Using the CSC specimens, tensile test and XRD analysis are performed. As a corrosion evaluation, a electrochemical test such as a polarization test is done, together with humid corrosion test and chloric acid immersion test. Through the corrosion tests, it is tried to confirm that the CSC is valuable method for making a copper canister. Consequently, it is confirmed that the CSC method is very usful for making 1 cm thick copper canister. the porosity of CSC layer is very low at 0.3 in case of Tafa copper layer. In corrosion tests, the CSC layers are very stable in active environments. It is hard to say that the difference of processing method but the purity of copper is important for the corrosion rate evaluation. The CSC method is very effective method for making 1 cm thick copper canister, It is hoped that the CSC method is applied in a HLW underground disposal system in the future.

  15. Development of Copper Canister through Cold Sprayed Coating Method

    International Nuclear Information System (INIS)

    Lee, Min Soo; Choi, Jong Won; Choi, Heui Joo; Lee, Jong Youl; Jeong, Jong Tae; Kim, Sung Ki; Cho, Dong Keun

    2007-12-01

    General thickness of a copper canister is 5 cm for a underground disposal application. The lower limit of a thickness is determined by a forging technology. But many experts in this area agrees that the thickness 1 cm is enough at the underground disposal for the life time of 1,000,000 years. Thus new technology is suggested for the making 1 cm thickness copper canister, that is a cold spray coating method(CSC). In this report, the CSC is examined and the technical possibility for making copper canister is measured. The overview of CSC and its characteristics are discussed. Various copper particles for the CSC are analyzed and the formed coating layers are examined to find their porosity and uniformity. A Tafa copper particle and Chang-sung copper particle are selected for making 1 cm thick test specimen. Using the CSC specimens, tensile test and XRD analysis are performed. As a corrosion evaluation, a electrochemical test such as a polarization test is done, together with humid corrosion test and chloric acid immersion test. Through the corrosion tests, it is tried to confirm that the CSC is valuable method for making a copper canister. Consequently, it is confirmed that the CSC method is very usful for making 1 cm thick copper canister. the porosity of CSC layer is very low at 0.3 in case of Tafa copper layer. In corrosion tests, the CSC layers are very stable in active environments. It is hard to say that the difference of processing method but the purity of copper is important for the corrosion rate evaluation. The CSC method is very effective method for making 1 cm thick copper canister, It is hoped that the CSC method is applied in a HLW underground disposal system in the future

  16. Criticality safety for TMI-2 canister storage at INEL

    International Nuclear Information System (INIS)

    Jones, R.R.; Briggs, J.B.; Ayers, A.L. Jr.

    1986-01-01

    Canisters containing Three Mile Island Unit 2 (TMI-2) core debris will be researched, stored, and prepared for final disposition at the Idaho National Engineering Laboratory (INEL). The canisters will be placed into storage modules and assembled into a storage rack, which will be located in the Test Area North (TAN) storage pool. Criticality safety calculations were made (a) to ensure that the storage rack is safe for both normal and accident conditions and (b) to determine the effects of degradation of construction materials (Boraflex and polyethylene) on criticality safety

  17. 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.

  18. Chemical stability of copper-canisters in deep repository

    International Nuclear Information System (INIS)

    Ahonen, L.

    1995-12-01

    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.)

  19. High-level waste canister envelope study: structural analysis

    International Nuclear Information System (INIS)

    1977-11-01

    The structural integrity of waste canisters, fabricated from standard weight Type 304L stainless steel pipe, was analyzed for sizes ranging from 8 to 24 in. diameter and 10 to 16 feet long under normal, abnormal, and improbable life cycle loading conditions. The canisters are assumed to be filled with vitrified high-level nuclear waste, stored temporarily at a fuel reprocessing plant, and then transported for storage in an underground salt bed or other geologic storage. In each of the three impact conditions studies, the resulting impact force is far greater than the elastic limit capacity of the material. Recommendations are made for further study

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

    International Nuclear Information System (INIS)

    FARWICK, C.C.

    1999-01-01

    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. Corrosion resistance of metal materials for HLW canister

    International Nuclear Information System (INIS)

    Furuya, Takashi; Muraoka, Susumu; Tashiro, Shingo

    1982-02-01

    In order to verify the materials as an important artificial barrier for canister of vitrified high-level waste from spent fuel reprocessing, data and reports were researched on corrosion resistance of the materials under conditions from glass form production to final disposal. Then, in this report, investigated subjects, improvement methods and future subjects are reviewed. It has become clear that there would be no problem on the inside and outside corrosion of the canister during glass production, but long term corrosion and radiation effect tests and the vitrification methods would be subjects in future on interim storage and final disposal conditions. (author)

  2. 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

  3. Mechanical Integrity of Copper Canister Lid and Cylinder. Sensitivity study

    International Nuclear Information System (INIS)

    Karlsson, Marianne

    2002-08-01

    This report is part of a study of the mechanical integrity of canisters used for disposal of nuclear fuel waste. The overall objective is to determine and ensure the static and long-term strength of the copper canister lid and cylinder casing. The canisters used for disposal nuclear fuel waste of type BWR consists of an inner part (insert) of ductile cast iron and an outer part of copper. The copper canister is to provide a sealed barrier between the contents of the canister and the surroundings. The study in this report complements the finite element analyses performed in an earlier study. The analyses aim to evaluate the sensitivity of the canister to tolerances regarding the gap between the copper cylinder and the cast iron insert. Since great uncertainties regarding the material's long term creep properties prevail, analyses are also performed to evaluate the effect of different creep data on the resulting strain and stress state. The report analyses the mechanical response of the lid and flange of the copper canister when subjected to loads caused by pressure from swelling bentonite and from groundwater at a depth of 500 meter. The loads acting on the canister are somewhat uncertain and the cases investigated in this report are possible cases. Load cases analysed are: Pressure 15 MPa uniformly distributed on lid and 5 MPa uniformly distributed on cylinder; Pressure 5 MPa uniformly distributed on lid and 15 MPa uniformly distributed on cylinder; Pressure 20 MPa uniformly distributed on lid and cylinder; and Side pressures 10 MPa and 20 MPa uniformly distributed on part of the cylinder. Creep analyses are performed for two of the load cases. For all considered designs high principal stresses appear on the outside of the copper cylinder in the region from the weld down to the level of the lid lower edge. Altering the gap between lid and cylinder and/or between cylinder and insert only marginally affects the resulting stress state. Fitting the lid in the cylinder

  4. Interaction between rock, bentonite buffer and canister. FEM calculations of some mechanical effects on the canister in different disposal concepts

    International Nuclear Information System (INIS)

    Boergesson, L.

    1992-07-01

    An important task of the buffer of highly compacted bentonite is to offer a mechanical protection to the canister. This role has been investigated by a number of finite element calculations using the complex elasto plastic material models for the bentonite that have been developed on the basis of laboratory tests and adapted to the code ABAQUS. The following main functions and scenarios have been investigated for some different canister types and repository concepts: - The effect of the water and swelling pressure, - The effect of a rock shear perpendicular to the canister axis, - The effect of creep in the copper after a rock shear displacement, - The thermomechanical effects when an initially saturated buffer is used

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

    International Nuclear Information System (INIS)

    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

  6. 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.

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

    International Nuclear Information System (INIS)

    1995-01-01

    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

  8. Examination of sludge from the Hanford K Basins fuel canisters

    International Nuclear Information System (INIS)

    Makenas, B.J.

    1998-01-01

    Samples of sludges with a high uranium content have been retrieved from the fuel canisters in the Hanford K West and K East basins. The composition of these samples contrasts markedly with the previously reported content of sludge samples taken from the K East basin floor. Chemical composition, chemical reactivity, and particle size of sludge are summarized in this paper

  9. High level waste canister emplacement and retrieval concepts study

    International Nuclear Information System (INIS)

    1975-09-01

    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

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

    International Nuclear Information System (INIS)

    FARWICK, C.C.

    2000-01-01

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

  11. Effects of glacial meltwater on corrosion of copper canisters

    International Nuclear Information System (INIS)

    Ahonen, L.; Vieno, T.

    1994-08-01

    The composition of glacial meltwater and its reactions in the bedrock are examined. The evidences that there are or should be from past intrusions of glacial meltwater and oxygen deep in the bedrock are also considered. The study is concluded with an evaluation of the potential effects of oxygenated meltwater on the corrosion of copper canisters. (46 refs., 3 figs., 2 tabs.)

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

    International Nuclear Information System (INIS)

    1983-04-01

    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.)

  13. Chemical durability of copper canisters under crystalline bedrock repository conditions

    International Nuclear Information System (INIS)

    Sjoeblom, R.; Hermansson, H.P.; Amcoff, Oe.

    1995-01-01

    In the Swedish waste management programme, the copper canister is expected to provide containment of the radionuclides for a very long time, perhaps million of years. The purpose of the present paper is to analyze prerequisites for assessments of corrosion lifetimes for copper canisters. The analysis is based on compilations of literature from the following areas: chemical literature on copper and copper corrosion, mineralogical literature with emphasis on the stability of copper in near surface environments, and chemical and mineralogical literature with emphasis on the stabilities and thermodynamics of species and phases that may exist in a repository environment. Three main types of situations are identified: (1) under oxidizing and low chloride conditions, passivating oxide type of layers may form on the copper surface; (2) under oxidizing and high chloride conditions, the species formed may all be dissolved; and (3) under reducing conditions, non-passivating sulfide type layers may form on the copper surface. Considerable variability and uncertainty exists regarding the chemical environment for the canister, especially in certain scenarios. Thus, the mechanisms for corrosion can be expected to differ greatly for different situations. The lifetime of a thick-walled copper canister subjected to general corrosion appears to be long for most reasonable chemistries. Localized corrosion may appear for types (1) and (3) above but the mechanisms are widely different in character. The penetration caused by localized corrosion can be expected to be very sensitive to details in the chemistry. 20 refs, 3 figs, 1 tab

  14. The development of a Martian atmospheric Sample collection canister

    Science.gov (United States)

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

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

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

    International Nuclear Information System (INIS)

    Bowyer, W.H.

    2000-06-01

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

  16. Structural Integrity Evaluation for Damaged Fuel Canister of a Research Reactor

    International Nuclear Information System (INIS)

    Oh, Jinho; Kwak, Jinsung; Lee, Sangjin; Lee, Jongmin; Ryu, Jeong-Soo

    2016-01-01

    The purpose of this document is to confirm the structural integrity of damaged fuel canister through the numerical simulation. The analysis results of canister including damaged fuel are evaluated with design limits of the ASME Sec. III NF Codes and Standards. The main function of canister is to store and protect the damaged fuel assembly generated from the operation of the research reactor. The canister is classified into safety class NNS (Non-nuclear Safety) and seismic category II. The shape of the canister is designed into commercialized circular tube due to economic benefit and easy manufacturing. The damaged fuel assembly is loaded in a dedicated canister by using special tool and supported by lower block in the canister. Then it is move into the damaged fuel storage rack under safeguards arrangements. The canister is securely supported at guide plate and base plate of rack. The structural integrity evaluation for the canister is performed by using response spectrum analysis. The analysis results show that the stress intensity of the canister under the seismic loads is within the ASME Code limits. Thus, the validity of the present design of the canister has been demonstrated

  17. Feasibility study for a DOE research and production fuel multipurpose canister

    International Nuclear Information System (INIS)

    Lopez, D.A.; Abbott, D.G.

    1994-02-01

    This is a report of the feasibility of multipurpose canisters for transporting, storing, and sing of Department of Energy research and production spent nuclear fuel. Six representative Department of Energy fuel assemblies were selected, and preconceptual canister designs were developed to accommodate these assemblies. The study considered physical interface, structural adequacy, criticality safety, shielding capability, thermal performance of the canisters, and fuel storage site infrastructure. The external envelope of the canisters was designed to fit within the overpack casks for commercial canisters being developed for the Department of Energy Office of Civilian Radioactive Waste Management. The budgetary cost of canisters to handle all fuel considered is estimated at $170.8M. One large conceptual boiling water reactor canister design, developed for the Office of Civilian Radioactive Waste Management, and two new canister designs can accommodate at least 85% of the volume of the Department of Energy fuel considered. Canister use minimizes public radiation exposure and is cost effective compared with bare fuel handling. Results suggest the need for additional study of issues affecting canister use and for conceptual design development of the three canisters

  18. Structural Integrity Evaluation for Damaged Fuel Canister of a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Jinho; Kwak, Jinsung; Lee, Sangjin; Lee, Jongmin; Ryu, Jeong-Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The purpose of this document is to confirm the structural integrity of damaged fuel canister through the numerical simulation. The analysis results of canister including damaged fuel are evaluated with design limits of the ASME Sec. III NF Codes and Standards. The main function of canister is to store and protect the damaged fuel assembly generated from the operation of the research reactor. The canister is classified into safety class NNS (Non-nuclear Safety) and seismic category II. The shape of the canister is designed into commercialized circular tube due to economic benefit and easy manufacturing. The damaged fuel assembly is loaded in a dedicated canister by using special tool and supported by lower block in the canister. Then it is move into the damaged fuel storage rack under safeguards arrangements. The canister is securely supported at guide plate and base plate of rack. The structural integrity evaluation for the canister is performed by using response spectrum analysis. The analysis results show that the stress intensity of the canister under the seismic loads is within the ASME Code limits. Thus, the validity of the present design of the canister has been demonstrated.

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

    International Nuclear Information System (INIS)

    Palmer, A.J.

    1995-01-01

    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

  20. Structural assessment of a space station solar dynamic heat receiver thermal energy storage canister

    Science.gov (United States)

    Thompson, R. L.; Kerslake, T. W.; Tong, M. T.

    1988-01-01

    The structural performance of a space station thermal energy storage (TES) canister subject to orbital solar flux variation and engine cold start up operating conditions was assessed. The impact of working fluid temperature and salt-void distribution on the canister structure are assessed. Both analytical and experimental studies were conducted to determine the temperature distribution of the canister. Subsequent finite element structural analyses of the canister were performed using both analytically and experimentally obtained temperatures. The Arrhenius creep law was incorporated into the procedure, using secondary creep data for the canister material, Haynes 188 alloy. The predicted cyclic creep strain accumulations at the hot spot were used to assess the structural performance of the canister. In addition, the structural performance of the canister based on the analytically determined temperature was compared with that based on the experimentally measured temperature data.

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

    International Nuclear Information System (INIS)

    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

  2. Safety evaluation for the inner canister closure station

    International Nuclear Information System (INIS)

    Glasscock, J.R.

    1987-01-01

    The Inner Canister Closure Station (ICCS), built by Remote Technology Corporation, will be operability tested. The ICCS is used to remotely leak test Inner Canister Closures (ICC's) and replace ICC's that are not water tight. After operability testing, the ICCS will be inspected and sent to the 717-F mock-up shop for remotability demonstration and dimensional checks, then installed in the Vitrification Building, 221-S. An analysis of potential safety hazards, equipment safety features, and procedural controls indicates that the ICCS can be operated without undue hazard to employees or to the public. A safety inspection and a new equipment inspection will be held before operation to verify that the ICCS meets Savannah River Site safety requirements. 4 refs., 6 figs

  3. Test design requirements: Canister-scale heater test

    International Nuclear Information System (INIS)

    Schauer, M.I.; Craig, P.A.; Stickney, R.G.

    1986-03-01

    This document establishes the Test Design Requirements for the design of a canister scale heater test to be performed in the Exploratory Shaft test facility. The purpose of the test is to obtain thermomechanical rock mass response data for use in validation of the numerical models. The canister scale heater test is a full scale simulation of a high-level nuclear waste container in a prototypic emplacement borehole. Electric heaters are used to simulate the heat loads expected in an actual waste container. This document presents an overview of the test including objectives and justification for the test. A description of the test as it is presently envisioned is included. Discussions on Quality Assurance and Safety are also included in the document. 12 refs., 1 fig

  4. Sediment mechanical response due to emplacement of a waste canister

    International Nuclear Information System (INIS)

    Karnes, C.H.; Dawson, P.R.; Silva, A.J.; Brown, W.T.

    1980-01-01

    Preliminary studies have been conducted to determine the interaction between a waste canister and seabed sediment during and after emplacement. Empirical and approximate methods for determining the depth reached by a freefall penetrator indicate that a boosted penetrator emplacement method may be necessary. Hole closure is necessary, but has not been verified because calculations and laboratory experiments show sensitivity to boundary conditions which control the degree of dynamic hole closure. Laboratory studies show that closure will take place by creep deformation but closure times in seabed environments are uncertain. For assumed thermomechanical properties of sediments, it is shown that a heat generating waste canister will probably not move a significant distancce during the heat generation period

  5. Multi-purpose canister system evaluation: A systems engineering approach

    International Nuclear Information System (INIS)

    1994-09-01

    This report summarizes Department of Energy (DOE) efforts to investigate various container systems for handling, transporting, storing, and disposing of spent nuclear fuel (SNF) assemblies in the Civilian Radioactive Waste Management System (CRWMS). The primary goal of DOE's investigations was to select a container technology that could handle the vast majority of commercial SNF at a reasonable cost, while ensuring the safety of the public and protecting the environment. Several alternative cask and canister concepts were evaluated for SNF assembly packaging to determine the most suitable concept. Of these alternatives, the multi-purpose canister (MPC) system was determined to be the most suitable. Based on the results of these evaluations, the decision was made to proceed with design and certification of the MPC system. A decision to fabricate and deploy MPCs will be made after further studies and preparation of an environmental impact statement

  6. Theoretical predictions for glass flow into an evacuated canister

    International Nuclear Information System (INIS)

    Routt, K.R.; Crow, K.R.

    1983-01-01

    Radioactive waste currently stored at the Savannah River Plant in liquid form is to be immobilized by incorporating it into a borosilicate glass. The glass melter for this process will consist of a refractory lined, steel vessel operated at a glass temperature of 1150 0 C. At the end of a two-year projected melter lifetime, the glass inside the melter is to be drained prior to disposition of the melter vessel. One proposed technique for accomplishing this drainage is by sucking the glass into an evacuated canister. The theoretical bases for design of an evacuated canister for draining a glass melter have been developed and tested. The theoretical equations governing transient and steady-state flow were substantiated with both a silicone glass simulant and molten glass

  7. Decontamination of Savannah River Plant waste glass canisters

    International Nuclear Information System (INIS)

    Rankin, W.N.

    1982-01-01

    A Defense Waste Processing Facility (DWPF) is currently being designed to convert Savannah River Plant (SRP) liquid, high-level radioactive waste into a solid form, such as borosilicate glass. The outside of the canisters of waste glass must have very low levels of smearable radioactive contamination before they are removed from the DWPF to prevent the spread of radioactivity. Several techniques were considered for canister decontamination: high-pressure water spray, electropolishing, chemical dissolution, and abrasive blasting. An abrasive blasting technique using a glass frit slurry has been selected for use in the DWPF. No additional equipment is needed to process waste generated from decontamination. Frit used as the abrasive will be mixed with the waste and fed to the glass melter. In contrast, chemical and electrochemical techniques require more space in the DWPF, and produce large amounts of contaminated byproducts which are difficult to immobilize by vitrification

  8. Settlement of Canisters with smectite clay envelopes in deposition holes

    International Nuclear Information System (INIS)

    Pusch, R.

    1986-12-01

    Settlement of canisters containing radioactive waste and being surrounded by dense smectite clay is caused by the stresses and heat induced in the clay. Consolidation by water expulsion of the clay underlying a model canister with 5 cm diameter and 30 cm length would theoretically account for a maximum finite settlement of about 70 my m in a few weeks, while shear-induced creep would yield a settlement of only a few microns in the same time period. These predictions were checked by running a laboratory test in which a dead load of 80 kg was applied to a small cylindrical copper canister embedded in Na bentonite. The settlement, which increased in proportion to log time, turned out to be about 6 my m in the first 2.5 months. After the first loading period at room temperature, heating to 50 degrees C and, after a 4 months long 'room temperature' period, to 70 degrees C took place. This cycling gave strong, instant settlement and upheaval because of the different thermal expansion of the interacting components of the system. After the development of constant temperature conditions in the entire system and completion of the consolidation or expansion that followed from the thermo-mechanical interactions, the settlement proceeded at a rather high rate at 70 degrees C, still following a log time creep law, but with somewhat stronger retardation. At room temperature, i.e. in the post-heating periods, the settlement seemed to cease, on the other hand. The conclusion from the study is that the canister movements under isothermal conditions were in accordance with the log t-type creep settlement that was predicted in theoretical grounds. Pre-heating and low stresses may account for extraordinary retardation of the settlement. (author)

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

    International Nuclear Information System (INIS)

    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

  10. 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).

  11. 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.

  12. Development of ultrasonic immersion inspection technique for spent fuel canisters

    International Nuclear Information System (INIS)

    Schankula, J.J.

    1982-07-01

    This report summarizes ultrasonic nondestructive testing development for metal matrix supported spent fuel disposal canisters. The work has concentated in two areas: inspection for lack of bond at the shell/matrix interface and inspection for voids in the matrix. The capabilities and limitations of these techniques have been fully established. Unbonded areas as small as 4 mm in diameter and voids 6 mm in diameter, 25 mm deep in the matrix, can readily be detected

  13. Analysis of probability of defects in the disposal canisters

    International Nuclear Information System (INIS)

    Holmberg, J.-E.; Kuusela, P.

    2011-06-01

    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.)

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

    International Nuclear Information System (INIS)

    KLEM, M.J.

    2000-01-01

    In 1998, a major change in the technical strategy for managing Multi Canister Overpacks (MCO) while stored within the Canister Storage Building (CSB) occurred. The technical strategy is documented in Baseline Change Request (BCR) No. SNF-98-006, Simplified SNF Project Baseline (MCO Sealing) (FDH 1998). This BCR deleted the hot conditioning process initially adopted for the Spent Nuclear Fuel Project (SNF Project) as documented in WHC-SD-SNF-SP-005, Integrated Process Strategy for K Basins Spent Nuclear Fuel (WHC 199.5). In summary, MCOs containing Spent Nuclear Fuel (SNF) from K Basins would be placed in interim storage following processing through the Cold Vacuum Drying (CVD) facility. With this change, the needs for the Hot Conditioning System (HCS) and inerting/pressure retaining capabilities of the CSB storage tubes and the MCO Handling Machine (MHM) were eliminated. Mechanical seals will be used on the MCOs prior to transport to the CSB. Covers will be welded on the MCOs for the final seal at the CSB. Approval of BCR No. SNF-98-006, imposed the need to review and update the CSB functions and requirements baseline documented herein including changing the document title to ''Spent Nuclear Fuel Project Canister Storage Building Functions and Requirements.'' This revision aligns the functions and requirements baseline with the CSB Simplified SNF Project Baseline (MCO Sealing). This document represents the Canister Storage Building (CSB) Subproject technical baseline. It establishes the functions and requirements baseline for the implementation of the CSB Subproject. The document is organized in eight sections. Sections 1.0 Introduction and 2.0 Overview provide brief introductions to the document and the CSB Subproject. Sections 3.0 Functions, 4.0 Requirements, 5.0 Architecture, and 6.0 Interfaces provide the data described by their titles. Section 7.0 Glossary lists the acronyms and defines the terms used in this document. Section 8.0 References lists the

  15. Stakeholder involvement in the evaluation of a multipurpose canister system

    International Nuclear Information System (INIS)

    Williams, J.R.; Kane, D.; Smith, T.B. Jr.

    1994-01-01

    The U.S. Department of Energy (DOE), Office of Civilian Radioactive Waste Management (OCRWM), began evaluating a multipurpose canister (MPC) concept in October of 1992. This followed recommendations by the Nuclear Waste Technical Review Board (NWTRB) and the U.S. Nuclear Regulatory Commission (NRC) that DOE develop a nuclear waste management system that achieves system integration, standardization, and reduced fuel-handling operations. Industry organizations such as Edison Electric Institute (EEI) and Electric Power Research Institute (EPRI) had conducted earlier studies that concluded advantages to the nuclear waste management system may be offered by such a concept. The MPC concept involves a metal canister which would contain multiple spent nuclear fuel assemblies. The canister would be sealed at the nuclear power plant and would not be reopened. The MPC would then be placed inside separate casks or overpacks for storage, transportation, and disposal. An important factor in DOE's evaluation of the MPC concept was the involvement of external parties. This paper describes that involvement process for the OCRWM's MPC implementation program. External parties who have an interest or stake in the program are referred to as stakeholders

  16. Canister displacement in KBS-3V. A theoretical study

    International Nuclear Information System (INIS)

    Boergesson, Lennart; Hernelind, Jan

    2006-02-01

    The vertical displacement of the canister in the KBS-3V concept has been studied in a number of consolidation and creep calculations using the FE-program ABAQUS. The creep model used for the calculations is based on Singh-Mitchell's creep theory, which has been adapted to and verified for the buffer material MX-80 in earlier tests. A porous elastic model with Drucker-Prager plasticity has been used for the consolidation calculations. For simplicity the buffer has been assumed to be water saturated from start. In one set of calculations only the consolidation and creep in the buffer without considering the interaction with the backfill was studied. In the other set of calculations the interaction with the backfill was included for a backfill consisting of an in situ compacted mixture of 30% bentonite and 70% crushed rock. The motivation to also study the behaviour of the buffer alone was that the final choice of backfill material and backfilling technique is not made yet so that set of calculations simulates a backfill that has identical properties with the buffer. The two cases represent two extreme cases, one with a backfill that has a low stiffness and the lowest allowable swelling pressure and one that has the highest possible swelling pressure and stiffness. The base cases in the calculations correspond to the final average density at saturation of 2,000 kg/m 3 with the expected swelling pressure of 7 MPa in a buffer. In order to study the sensitivity of the system to loss in bentonite mass and swelling pressure seven additional calculations were done with reduced swelling pressure down to 80 kPa corresponding to a density at water saturation of about 1,500 kg/m 3 . The calculations included two stages, where the first stage models the swelling and consolidation that takes place in order for the buffer to reach force equilibrium. This stage takes place during the saturation phase and the subsequent consolidation/swelling phase. The second stage models the

  17. Canister displacement in KBS-3V. A theoretical study

    Energy Technology Data Exchange (ETDEWEB)

    Boergesson, Lennart [Clay Technology AB, Lund (Sweden); Hernelind, Jan [FEMTech AB, Vaesteraas (Sweden)

    2006-02-15

    The vertical displacement of the canister in the KBS-3V concept has been studied in a number of consolidation and creep calculations using the FE-program ABAQUS. The creep model used for the calculations is based on Singh-Mitchell's creep theory, which has been adapted to and verified for the buffer material MX-80 in earlier tests. A porous elastic model with Drucker-Prager plasticity has been used for the consolidation calculations. For simplicity the buffer has been assumed to be water saturated from start. In one set of calculations only the consolidation and creep in the buffer without considering the interaction with the backfill was studied. In the other set of calculations the interaction with the backfill was included for a backfill consisting of an in situ compacted mixture of 30% bentonite and 70% crushed rock. The motivation to also study the behaviour of the buffer alone was that the final choice of backfill material and backfilling technique is not made yet so that set of calculations simulates a backfill that has identical properties with the buffer. The two cases represent two extreme cases, one with a backfill that has a low stiffness and the lowest allowable swelling pressure and one that has the highest possible swelling pressure and stiffness. The base cases in the calculations correspond to the final average density at saturation of 2,000 kg/m{sup 3} with the expected swelling pressure of 7 MPa in a buffer. In order to study the sensitivity of the system to loss in bentonite mass and swelling pressure seven additional calculations were done with reduced swelling pressure down to 80 kPa corresponding to a density at water saturation of about 1,500 kg/m{sup 3}. The calculations included two stages, where the first stage models the swelling and consolidation that takes place in order for the buffer to reach force equilibrium. This stage takes place during the saturation phase and the subsequent consolidation/swelling phase. The second stage

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

    International Nuclear Information System (INIS)

    Enos, David; Bryan, Charles R.

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-01

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

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

    International Nuclear Information System (INIS)

    HEARD, F.J.

    1999-01-01

    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

  1. 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.

  2. Thermohydraulic analysis of BWR and PWR spent fuel assemblies contained within square canisters

    International Nuclear Information System (INIS)

    Wiles, L.E.; McCann, R.A.

    1981-09-01

    This report presents the results of several thermohydraulic simulations of spent fuel assembly/canister configurations performed in support of a program investigating the feasibility of storing spent nuclear fuel assemblies in canisters that would be stored in an air environment. Eleven thermohydraulic simulations were performed. Five simulations were performed using a single BWR fuel assembly/canister design. The various cases were defined by changing the canister spacing and the heat generation rate of the fuel assembly. For each simulation a steady-state thermohydraulic solution was achieved for the region inside the canister. Similarly, six simulations were performed for a single PWR fuel assembly/canister design. The square fuel rod arrays were contained in square canisters which would permit closer packing of the canisters in a storage facility. However, closer packing of the canisters would result in higher fuel temperatures which would possibly have an adverse impact on fuel integrity. Thus, the most important aspect of the analysis was to define the peak fuel assembly temperatures for each case. These results are presented along with various temperature profiles, heat flux distributions, and air velocity profiles within the canister. 48 figures, 4 tables

  3. 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

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

    International Nuclear Information System (INIS)

    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

  5. Kinetic modelling of bentonite-canister interaction. Long-term predictions of copper canister corrosion under oxic and anoxic conditions

    International Nuclear Information System (INIS)

    Wersin, P.; Spahiu, K.; Bruno, J.

    1994-09-01

    A new modelling approach for canister corrosion which emphasises chemical processes and diffusion at the bentonite-canister interface is presented. From the geochemical boundary conditions corrosion rates for both an anoxic case and an oxic case are derived and uncertainties thereof are estimated via sensitivity analyses. Time scales of corrosion are assessed by including calculations of the evolution of redox potential in the near field and pitting corrosion. This indicates realistic corrosion depths in the range of 10 -7 and 4*10 -5 mm/yr, respectively for anoxic and oxic corrosion. Taking conservative estimates, depths are increased by a factor of about 200 for both cases. From these predictions it is suggested that copper canister corrosion does not constitute a problem for repository safety, although certain factors such as temperature and radiolysis have not been explicitly included. The possible effect of bacterial processes on corrosion should be further investigated as it might enhance locally the described redox process. 35 refs, 11 figs, 6 tabs

  6. Kinetic modelling of bentonite-canister interaction. Long-term predictions of copper canister corrosion under oxic and anoxic conditions

    Energy Technology Data Exchange (ETDEWEB)

    Wersin, P; Spahiu, K; Bruno, J [MBT Tecnologia Ambiental, Cerdanyola (Spain)

    1994-09-01

    A new modelling approach for canister corrosion which emphasises chemical processes and diffusion at the bentonite-canister interface is presented. From the geochemical boundary conditions corrosion rates for both an anoxic case and an oxic case are derived and uncertainties thereof are estimated via sensitivity analyses. Time scales of corrosion are assessed by including calculations of the evolution of redox potential in the near field and pitting corrosion. This indicates realistic corrosion depths in the range of 10{sup -7} and 4*10{sup -5} mm/yr, respectively for anoxic and oxic corrosion. Taking conservative estimates, depths are increased by a factor of about 200 for both cases. From these predictions it is suggested that copper canister corrosion does not constitute a problem for repository safety, although certain factors such as temperature and radiolysis have not been explicitly included. The possible effect of bacterial processes on corrosion should be further investigated as it might enhance locally the described redox process. 35 refs, 11 figs, 6 tabs.

  7. Thermo-hydro-mechanical mode of canister retrieval test

    International Nuclear Information System (INIS)

    Zandarin, M.T.; Olivella, S.; Gens', A.; Alonso, E.E.

    2010-01-01

    Document available in extended abstract form only. The Canister Retrieval Tests (CRT) is a full scale in situ experiment performed by SKB at Aespoe Laboratory. The experiment involves placing a canister equipped with electrical heaters inside of a deposition hole bored in Aespoe diorite. The deposition hole is 8.55 metres deep and has a diameter of 1.76 metres. The space between canister and the hole is filled with a MX-80 bentonite buffer. The bentonite buffer was installed in form of blocks and rings of bentonite. At the top of the canister bentonite bricks occupy the volume between the canister top surface and the bottom surface of the plug. Due to the bentonite ring size there are two gaps; once between canister and buffer which was left empty and another one between buffer and rock that was filled with bentonite pellets. The top of the hole was sealed with a retaining plug composed of concrete and a steel plate. The plug was secured against heave caused by the swelling clay with nine cables anchored in the rock. An artificial pressurised saturation system was used because the supply of water from the rock was judged to be insufficient for saturating the buffer in a feasible time. A large number of instruments were installed to monitor the test as follows: - Canister - temperature and strain. - Rock mass - temperature and stress. - Retaining system - force and displacement. - Buffer - temperature, relative humidity, pore pressure and total pressure. After dismantling the tests the final dry density and water content of bentonite and pellets were measured. The comprehensive record of the Thermo-Hydro-Mechanical (THM) processes in the buffer give the possibility to investigate theoretical formulations and models, since the results of THM analyses can be checked against experimental data. As part of the European project THERESA, a 2-D axisymmetric model simulation of CRT bas been carried out. Some of the main objectives of this simulation are the study of the

  8. An assessment of KW Basin radionuclide activity when opening SNF canisters

    International Nuclear Information System (INIS)

    Bergmann, D.W.; Mollerus, F.J.; Wray, J.L.

    1995-01-01

    N Reactor spent fuel is being stored in sealed canisters in the KW Basin. Some of the canisters contain damaged fuel elements. There is the potential for release of Cs 137, Kr 85, H3, and other fission products and transuranics (TRUs) when canisters are opened. Canister opening is required to select and transfer fuel elements to the 300 Area for examination as part of the Spent Nuclear Fuel (SNF) Characterization program. This report estimates the amount of radionuclides that can be released from Mark II spent nuclear fuel (SNF) canisters in KW Basin when canisters are opened for SNF fuel sampling as part of the SNF Characterization Program. The report also assesses the dose consequences of the releases and steps that can be taken to reduce the impacts of these releases

  9. Effect of HNO3-cerium(IV) decontamination on stainless steel canister materials

    International Nuclear Information System (INIS)

    Westerman, R.E.; Mackey, D.B.

    1991-01-01

    Stainless steel canisters will be filled with vitrified radioactive waste at the West Valley Demonstration Project (WVDP), West Valley, NY. After they are filled, the sealed canisters will be decontaminated by immersion in a HNO 3 -Ce(IV) solution, which will remove the oxide film and a small amount of metal from the surface of the canisters. Studies were undertaken in support of waste form qualification activities to determine the effect of this decontamination treatment on the legibility of the weld-bead canister identification label, and to determine whether this decontamination treatment could induce stress-corrosion cracking (SCC) in the AISI 304L stainless steel (SS) canister material. Neither the label legibility nor the canister integrity with regard to SCC were found to be prejudiced by the simulated decontamination treatment

  10. Proposal of a SiC disposal canister for very deep borehole disposal

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Heui-Joo; Lee, Minsoo; Lee, Jong-Youl; Kim, Kyungsu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    In this paper authors proposed a silicon carbide, SiC, disposal canister for the DBD concept in Korea. A. Kerber et al. first proposed the SiC canister for a geological disposal of HLW, CANDU or HTR spent nuclear fuels. SiC has some drawbacks in welding or manufacturing a large canister. Thus, we designed a double layered disposal canister consisting of a stainless steel outer layer and a SiC inner layer. KAERI has been interested in developing a very deep borehole disposal (DBD) of HLW generated from pyroprocessing of PWR spent nuclear fuel and supported the relevant R and D with very limited its own budget. KAERI team reviewed the DBD concept proposed by Sandia National Laboratories (SNL) and developed its own concept. The SNL concept was based on the steel disposal canister. The authors developed a new technology called cold spray coating method to manufacture a copper-cast iron disposal canister for a geological disposal of high level waste in Korea. With this method, 8 mm thin copper canister with 400 mm in diameter and 1200 mm in height was made. In general, they do not give any credit on the lifetime of a disposal canister in DBD concept unlike the geological disposal. In such case, the expensive copper canister should be replaced with another one. We designed a disposal canister using SiC for DBD. According to an experience in manufacturing a small size canister, the fabrication of a large-size one is a challenge. Also, welding of SiC canister is not easy. Several pathways are being paved to overcome it.

  11. Gas liquid sampling for closed canisters in KW Basin - test plan

    International Nuclear Information System (INIS)

    Pitkoff, C.C.

    1995-01-01

    Test procedures for the gas/liquid sampler. Characterization of the Spent Nuclear Fuel, SNF, sealed in canisters at KW-Basin is needed to determine the state of storing SNF wet. Samples of the liquid and the gas in the closed canisters will be taken to gain characterization information. Sampling equipment has been designed to retrieve gas and liquid from the closed canisters in KW basin. This plan is written to outline the test requirements for this developmental sampling equipment

  12. 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.

  13. Corrosion resistance of canisters for final disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    Mattsson, E.

    1979-01-01

    A group of Swedish scientists has evaluated from the corrosion point of view three alternative canister types for final disposal of waste from nuclear reactors in boreholes in rock 500 m below ground. Titanium canisters with a wall-thickness of 6 mm and 100 mm thick lead lining have been estimated to have a life of at least thousands of years, and probably tens of thousands of years. Copper canisters with 200-mm-thick walls would last for hundreds of thousands of years. The third type, α-alumina sintered under isostatic pressure, is a very promising canister material

  14. Mechanical design of the storage tubes in the HWVP canister storage building

    International Nuclear Information System (INIS)

    Divona, C.J.; Fages, R.; Janicek, G.P.; Mullally, J.A.

    1993-01-01

    Canisters of high-level waste from the Hanford Waste Vitrification Plant (HWVP) will be stored in an adjacent facility, the Canister Storage Building (CSB). The canisters are stored vertically in an array of tubes within the shielded vault area of the CSB. This paper describes the mechanical design of the storage tubes, the shield floor plugs that confine the waste within the tubes and the impact absorber system used to assure that the canisters are not breached in the event of an accidental drop. Installation and testing of the components is also discussed

  15. SITE-94. CAMEO: A model of mass-transport limited general corrosion of copper canisters

    International Nuclear Information System (INIS)

    Worgan, K.J.; Apted, M.J.

    1996-12-01

    This report describes the technical basis for the CAMEO code, which models the general, uniform corrosion of a copper canister either by transport of corrodants to the canister, or by transport of corrosion products away from the canister. According to the current Swedish concept for final disposal of spent nuclear fuels, extremely long containment times are achieved by thick (60-100 mm) copper canisters. Each canister is surrounded by a compacted bentonite buffer, located in a saturated, crystalline rock at a depth of around 500 m below ground level. Three diffusive transport-limited cases are identified for general, uniform corrosion of copper: General corrosion rate-limited by diffusive mass-transport of sulphide to the canister surface under reducing conditions; General corrosion rate-limited by diffusive mass-transport of oxygen to the canister surface under mildly oxidizing conditions; General corrosion rate-limited by diffusive mass-transport of copper chloride away from the canister surface under highly oxidizing conditions. The CAMEO code includes general corrosion models for each of the above three processes. CAMEO is based on the well-tested CALIBRE code previously developed as a finite-difference, mass-transfer analysis code for the SKI to evaluate long-term radionuclide release and transport in the near-field. A series of scoping calculations for the general, uniform corrosion of a reference copper canister are presented

  16. Production methods and costs of oxygen free copper canisters for nuclear waste disposal

    International Nuclear Information System (INIS)

    Rajainmaeki, H.; Nieminen, M.; Laakso, L.

    1991-08-01

    The fabrication technology and costs of various manufacturing alternatives to make large copper canisters for spent fuel repository are discussed. The capsule design is based on the TVO's new advanced cold process concept where a steel canister is surrounded by the oxygen free copper canister. This study shows that already at present there exist several possible manufacturing routes, which result in consistently high quality canisters. Hot rolling, bending and EB-welding the seam is the best way to assure the small grain size which is preferable for the best inspectability of the final EB-welded seam of the lid. The same route turns out also to be the most economical

  17. Production methods and costs of oxygen free copper canisters for nuclear waste disposal

    International Nuclear Information System (INIS)

    Rajainmaeki, H.; Nieminen, M.; Laakso, L.

    1991-06-01

    The fabrication technology and costs of various manufacturing alternatives to make large copper canisters for spent fuel repository are discussed. The capsule design is based on the TVO's new advanced cold process concept where a steel canister is surrounded by the oxygen free copper canister. This study shows that already at present there exist several possible manufacturing routes, which results in consistently high quality canisters. Hot rolling, bending and EB-welding the seam is the best way to assure the small grain size which is preferable for the best inspectability of the final EB-welded seam of the lid. The same route turns out also to be the most economical. (au)

  18. Mechanical Integrity of Copper Canister Lid and Cylinder

    International Nuclear Information System (INIS)

    Karlsson, Marianne

    2002-01-01

    This report compiles finite element analyses performed to ensure the structural integrity of canisters used for storing of nuclear fuel waste of type BWR. The report comprises analyses performed on the canister lid and cylinder casing in order to determine static and long-term strength of the structure. The report analyses the mechanical response of the lid and flange of the copper canister when subjected to loads caused by pressure from swelling bentonite and from ground water at a depth of 500 meter. The loads acting on the canister are somewhat uncertain and the cases investigated in this report are possible cases. Load cases analysed are: Pressure 15 MPa uniformly distributed on lid and 5 MPa uniformly distributed on cylinder; Pressure 5 MPa uniformly distributed on lid and 15 MPa uniformly distributed on cylinder; Pressure 20 MPa uniformly distributed on lid and cylinder; Side pressures 10 MPa and 20 MPa uniformly distributed on part of the cylinder. Creep analyses are also performed in order to estimate the stresses that will arise when the canister is placed in the repository. The analyses in this report are recreated from the original analyses but the models differ in geometry. Also, there is no information in the original reports on material data, time-independent as well as creep data, and analysis procedure. The data used in the recreated analyses are based on information from References 2, 3, 6 and 7. The results presented in this report are based on the supplementary analyses. These results differ from the original results. Most likely this is due to differences in model geometry. The original results are appended to the report and are summarised for comparison with results from the supplementary analyses. Otherwise, these results are not further discussed. For all load cases, high tensile stresses are found in the lid fillet between the planar part and the flange. High tensile stresses are also found in the weld surface and on the outer side of the

  19. Investigation and modelling of friction stir welded copper canisters

    Energy Technology Data Exchange (ETDEWEB)

    Kaellgren, Therese

    2010-02-15

    This work has been focused on characterisation of FSW joints, and modelling of the process, both analytically and numerically. The Swedish model for final deposit of nuclear fuel waste is based on copper canisters as a corrosion barrier with an inner pressure holding insert of cast iron. Friction Stir Welding (FSW) is the method to seal the copper canister, a technique invented by The Welding Institute (TWI). The first simulations were based on Rosenthal's analytical medium plate model. The model is simple to use, but has limitations. Finite element models (FEM) were developed, initially with a two-dimensional geometry. Due to the requirements of describing both the heat flow and the tool movement, three-dimensional models were developed. These models take into account heat transfer, material flow, and continuum mechanics. The geometries of the models are based on the simulation experiments carried out at TWI and at Swedish Nuclear Fuel Waste and Management Co (SKB). Temperature distribution, material flow and their effects on the thermal expansion were predicted for a full-scale canister and lid. The steady state solutions have been compared with temperature measurements, showing good agreement. In order to understand the material flow during welding a marker technique is used, which involves inserting dissimilar material into the weld zone before joining. Different materials are tested showing that brass rods are the most suitable material in these welds. After welding, the weld line is sliced, etched and examined by optical microscope. To understand the material flow further, and in the future predict the flow, a FEM is developed. This model and the etched samples are compared showing similar features. Furthermore, by using this model the area that is recrystallised can be predicted. The predicted area and the grain size and hardness profile agree well

  20. Friction Stir Welding of Copper Canisters for Nuclear Waste

    Energy Technology Data Exchange (ETDEWEB)

    Kaellgren, Therese

    2005-07-01

    The Swedish model for final disposal of nuclear fuel waste is based on copper canisters as a corrosion barrier with an inner pressure holding insert of cast iron. One of the methods to seal the copper canister is to use the Friction Stir Welding (FSW), a method invented by The Welding Institute (TWI). This work has been focused on characterisation of the FSW joints, and modelling of the process, both analytically and numerically. The first simulations were based on Rosenthal's analytical medium plate model. The model is simple to use, but has limitations. Finite element models were developed, initially with a two-dimensional geometry. Due to the requirements of describing both the heat flow and the tool movement, three-dimensional models were developed. These models take into account heat transfer, material flow, and continuum mechanics. The geometries of the models are based on the simulation experiments carried out at TWI and at Swedish Nuclear Fuel Waste and Management Co (SKB). Temperature distribution, material flow and their effects on the thermal expansion were predicted for a full-scale canister and lid. The steady state solutions have been compared with temperature measurements, showing good agreement. Microstructure and hardness profiles have been investigated by optical microscope, Scanning Electron Microscope (SEM), Electron Back Scatter Diffraction (EBSD) and Rockwell hardness measurements. EBSD visualisation has been used to determine the grain size distribution and the appearance of twins and misorientation within grains. The orientation maps show a fine uniform equiaxed grain structure. The root of the weld exhibits the smallest grains and many annealing twins. This may be due to deformation after recrystallisation. The appearance of the nugget and the grain size depends on the position of the weld. A large difference can be seen both in hardness and grain size between the start of the weld and when the steady state is reached.

  1. Heat transfer analysis of the geologic disposal of spent fuel and high-level waste storage canisters

    International Nuclear Information System (INIS)

    Allen, G.K.

    1980-08-01

    Near-field temperatures resulting from the storage of high-level waste canisters and spent unreprocessed fuel assembly canisters in geologic formations were determined. Preliminary design of the repository was modeled for a heat transfer computer code, HEATING5, which used the Crank-Nicolson finite difference method to evaluate transient heat transfer. The heat transfer system was evaluated with several two- and three-dimensional models which transfer heat by a combination of conduction, natural convention, and radiation. Physical properties of the materials in the model were based upon experimental values for the various geologic formations. The effects of canister spacing, fuel age, and use of an overpack were studied for the analysis of the spent fuel canisters; salt, granite, and basalt were considered as the storage media for spent fuel canisters. The effects of canister diameter and use of an overpack were studied for the analysis of the high-level waste canisters; salt was considered as the only storage media for high-level waste canisters. Results of the studies on spent fuel assembly canisters showed that the canisters could be stored in salt formations with a maximum heat loading of 134 kw/acre without exceeding the temperature limits set for salt stability. The use of an overpack had little effect on the peak canister temperatures. When the total heat load per acre decreased, the peak temperatures reached in the geologic formations decreased; however, the time to reach the peak temperatures increased. Results of the studies on high-level waste canisters showed that an increased canister diameter will increase the canister interior temperatures considerably; at a constant areal heat loading, a 381 mm diameter canister reached almost a 50 0 C higher temperature than a 305 mm diameter canister. An overpacked canister caused almost a 30 0 C temperature rise in either case

  2. 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.

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

    Science.gov (United States)

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

    2009-01-01

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

  4. 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.

  5. FPIN2 posttest analysis of cylindrical canisters in SLSF Experiment P4

    Energy Technology Data Exchange (ETDEWEB)

    Hughes, T H; Kramer, J M

    1984-12-01

    Results demonstrate that the clad deformation is dominated by the expansion of the fuel when it melts. In our analysis we moved the end space volume and some of the fuel-clad radial gap volume to an artificial central hole. This approximation may affect the details in the early parts of the transient, but clearly did not affect the major cladding deformation. It is also clear that the accuracy of the value of the fuel expansion upon melting is significant as is the dimensional accuracy of the fuel and canisters. The major conclusions from the FPIN2 posttest analysis of the cylindrical canisters in SLSF Experiment P4 are: The maximum melt fractions in the two canisters were about 75%. Both canisters experienced about the same diametral strains of 12% prior to failure. These strains were almost entirely due to the additional volume that must be created inside the canisters to accommodate the expansion of fuel on melting. The mode of cladding failure was plastic instability by necking of the canister walls. The failure time of the 20% CW canister and the nonmechanical failure of the 10% CW canister are consistent with the FPIN2 calculations using the plastic instability failure criteria.

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  7. Test plan for K Basin Sludge Canister and Floor Sampling Device

    International Nuclear Information System (INIS)

    Meling, T.A.

    1995-01-01

    This document provides the test plan and procedure forms for conducting the functional and operational acceptance testing of the K Basin Sludge Canister and Floor Sampling Device(s). These samplers samples sludge off the floor of the 100K Basins and out of 100K fuel storage canisters

  8. 42 CFR 84.1155 - Filters used with canisters and cartridges; location; replacement.

    Science.gov (United States)

    2010-10-01

    ...; location; replacement. 84.1155 Section 84.1155 Public Health PUBLIC HEALTH SERVICE, DEPARTMENT OF HEALTH... Efficiency Respirators and Combination Gas Masks § 84.1155 Filters used with canisters and cartridges; location; replacement. (a) Particulate matter filters used in conjunction with a canister or cartridge...

  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. The design analysis of ACP-canister for nuclear waste disposal

    International Nuclear Information System (INIS)

    Raiko, H.

    1992-05-01

    The design basis, dimensioning and some manufacturing aspects of the Advanced Cold Process Canister (ACPC) for the nuclear waste disposal is summarized in the report. The strength of the canister has been evaluated in normal design load condition and in extreme high hydrostatic pressure load condition possibly caused by ice age (orig.)

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

    International Nuclear Information System (INIS)

    Heard, F.J.

    1999-01-01

    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

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

    International Nuclear Information System (INIS)

    Callan, J.E.

    1986-01-01

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

  13. Performance Specification Shippinpark Pressurized Water Reactor Fuel Drying and Canister Inerting System for PWR Core 2 Blanket Fuel Assemblies Stored within Shippingport Spent Fuel Canisters

    International Nuclear Information System (INIS)

    JOHNSON, D.M.

    2000-01-01

    This specification establishes the performance requirements and basic design requirements imposed on the fuel drying and canister inerting system for Shippingport Pressurized Water Reactor (PWR) Core 2 blanket fuel assemblies (BFAs) stored within Shippingport spent fuel (SSFCs) canisters (fuel drying and canister inerting system). This fuel drying and canister inerting system is a component of the U.S. Department of Energy, Richland Operations Office (RL) Spent Nuclear Fuels Project at the Hanford Site. The fuel drying and canister inerting system provides for removing water and establishing an inert environment for Shippingport PWR Core 2 BFAs stored within SSFCs. A policy established by the U.S. Department of Energy (DOE) states that new SNF facilities (this is interpreted to include structures, systems and components) shall achieve nuclear safety equivalence to comparable U.S. Nuclear Regulatory Commission (NRC)-licensed facilities. This will be accomplished in part by applying appropriate NRC requirements for comparable NRC-licensed facilities to the fuel drying and canister inerting system, in addition to applicable DOE regulations and orders

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

    International Nuclear Information System (INIS)

    Callan, J.E.

    1986-01-01

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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

    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

  16. Thermal studies of the canister staging pit in a hypothetical Yucca Mountain canister handling facility using computational fluid dynamics

    International Nuclear Information System (INIS)

    Soltani, Mehdi; Barringer, Chris; Bues, Timothy T. de

    2007-01-01

    The proposed Yucca Mountain nuclear waste storage site will contain facilities for preparing the radioactive waste canisters for burial. A previous facility design considered was the Canister Handling Facility Staging Pit. This design is no longer used, but its thermal evaluation is typical of such facilities. Structural concrete can be adversely affected by the heat from radioactive decay. Consequently, facilities must have heating ventilation and air conditioning (HVAC) systems for cooling. Concrete temperatures are a function of conductive, convective and radiative heat transfer. The prediction of concrete temperatures under such complex conditions can only be adequately handled by computational fluid dynamics (CFD). The objective of the CFD analysis was to predict concrete temperatures under normal and off-normal conditions. Normal operation assumed steady state conditions with constant HVAC flow and temperatures. However, off-normal operation was an unsteady scenario which assumed a total HVAC failure for a period of 30 days. This scenario was particularly complex in that the concrete temperatures would gradually rise, and air flows would be buoyancy driven. The CFD analysis concluded that concrete wall temperatures would be at or below the maximum temperature limits in both the normal and off-normal scenarios. While this analysis was specific to a facility design that is no longer used, it demonstrates that such facilities are reasonably expected to have satisfactory thermal performance. (author)

  17. 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

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

    International Nuclear Information System (INIS)

    Lydmark, Sara; Pedersen, Karsten

    2011-03-01

    '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 10 0 -10 2 viable sulphate-reducing and acetogenic bacteria and 10 2 -10 4 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

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

    International Nuclear Information System (INIS)

    Jantzen, C.M.

    1992-01-01

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

  20. Reliability in sealing of canister for spent nuclear fuel

    International Nuclear Information System (INIS)

    Ronneteg, Ulf; Cederqvist, Lars; Ryden, Haakan; Oeberg, Tomas; Mueller, Christina

    2006-06-01

    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

  1. 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

  2. Corrosion of canister materials for radioactive waste disposal

    Energy Technology Data Exchange (ETDEWEB)

    Kienzler, Bernhard [KIT Karlsruhe (Germany). Institut fuer Nukleare Entsorgung (INE)

    2017-08-15

    In the period between 1980 and 2004, corrosion studies on various metallic materials have been performed at the Research Center Karlsruhe. The objectives of these experimental studies addressed mainly the performance of canister materials for heat producing, high-level wastes and spent nuclear fuels for a repository in a German salt dome. Additional studies covered the performance of steels for packaging wastes with negligible heat production under conditions to be expected in rocksalt and in the Konrad iron ore mine. The results of the investigations have been published in journals and conference proceedings but also in ''grey literature''. This paper presents a summary of the results of corrosion experiments with fine-grained steels and nodular cast steel.

  3. Multi-Canister overpack inservice inspection and maintenance

    International Nuclear Information System (INIS)

    SMITH, K.E.

    1998-01-01

    The factors to be considered in establishing inservice inspection and maintenance requirements for the Multi-Canister Overpack (MCO) include evaluating the likelihood of degradation to the MCO pressure boundary due to erosion and corrosion, reviewing commercial practice for NRC licensed spent nuclear fuel storage systems, and examining the individual MCO components for maintenance needs. Reviews of the potential for MCO erosion and corrosion conclude that neither will pose a threat to the MCO pressure boundary. Consistent with commercial practice for spent fuel storage systems, the MCO closure weld will be helium leak tested prior to placement in interim storage. Beyond the CSB facility related monitoring plans (radiological monitoring, emissions monitoring, vault cooling data, etc.), no inservice inspection or maintenance of the MCO is required during interim storage

  4. Life Prediction of Spent Fuel Storage Canister Material

    Energy Technology Data Exchange (ETDEWEB)

    Ballinger, Ronald

    2018-04-16

    The original purpose of this project was to develop a probabilistic model for SCC-induced failure of spent fuel storage canisters, exposed to a salt-air environment in the temperature range 30-70°C for periods up to and exceeding 100 years. The nature of this degradation process, which involves multiple degradation mechanisms, combined with variable and uncertain environmental conditions dictates a probabilistic approach to life prediction. A final report for the original portion of the project was submitted earlier. However, residual stress measurements for as-welded and repair welds could not be performed within the original time of the project. As a result of this, a no-cost extension was granted in order to complete these tests. In this report, we report on the results of residual stress measurements.

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

    International Nuclear Information System (INIS)

    Naish, C.C.; Buttle, D.; Wallace-Sims, R.; O'Brien, T.M.

    1991-01-01

    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)

  6. Inspection of bottom and lid welds for disposal canisters

    International Nuclear Information System (INIS)

    Pitkaenen, J.

    2010-09-01

    This report presents the inspection techniques of copper electron beam and friction stir welds. Both welding methods are described briefly and a more detailed description of the defects occurring in each welding methods is given. The defect types form a basis for the design of non-destructive testing. The inspection of copper material is challenging due to the anisotropic properties of the weld and local changes in the grain size of the base material. Four different methods are used for inspection. Ultrasonic and radiographic testing techniques are used for inspection of volume. Eddy current and visual testing techniques are used for inspection of the surface and near surface area. All these methods have some limitations which are related to the physics of the used method. All inspection methods need to be carried out remotely because of the radiation from the spent nuclear fuel. All methods have been described in detail and the use of the chosen inspection techniques has been justified. Phased array technology has been applied in ultrasonic testing. Ultrasonic phased array technology enables the electrical modification of the sound field during inspection so that the sound field can be adjusted dynamically for different situations and detection of different defect types. The frequency of the phased array probe has been chosen to be 3.5 MHz. It is a compromise between good sizing and defect detectability. It must be taken into account that ultrasonic testing is not suitable for detection of defect types which are in the direction of the beam. Ultrasonic and radiographic testing techniques complement each other in case of planar defects. Positioning of the indication in the radial direction is rather limited in radiographic testing. Surface inspection has been added to the inspection routine because indications from the outer surface of the canister cannot be distinguished from weld defects in the radiographic image. A 9 MeV linear accelerator has been used in the

  7. 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.

  8. Corrosion of canister materials for radioactive waste disposal

    International Nuclear Information System (INIS)

    Kienzler, Bernhard

    2017-01-01

    In the period between 1980 and 2004, corrosion studies on various metallic materials have been performed at the Research Center Karlsruhe. The objectives of these experimental studies addressed mainly the performance of canister materials for heat producing, high-level wastes and spent nuclear fuels for a repository in a German salt dome. Additional studies covered the performance of steels for packaging wastes with negligible heat production under conditions to be expected in rocksalt and in the Konrad iron ore mine. The results of the investigations have been published in journals and conference proceedings but also in ''grey literature''. This paper presents a summary of the results of corrosion experiments with fine-grained steels and nodular cast steel.

  9. Multi-Canister overpack necessity of the rupture disk

    International Nuclear Information System (INIS)

    SMITH, K.E.

    1998-01-01

    The Multi-Canister Overpack (MCO) rupture disk precludes the MCO from pressurization above the design limit during transport from the K Basins to the Cold Vacuum Drying (CVD) Facility and prior to connection of the CVD process piping. Removal of the rupture disk from the MCO design would: (a) result in unacceptable dose consequences in the event a thermal runaway accident occurred; (b) increase residual risk; and (c) remove a degree of specificity from the dose calculations. The potential cost savings of removing the rupture disk from the MCO design is offset by the cost of design modifications, changes to hazard analyses and safety analyses, and changes to existing documentation. Retaining the rupture disk mitigates the consequences of MCO overpressurization, and considering the overall economic impacts to the SNF Project, is the most cost effective approach

  10. Defense Waste Processing Facility Canister Closure Weld Current Validation Testing

    Energy Technology Data Exchange (ETDEWEB)

    Korinko, P. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Maxwell, D. N. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2018-01-29

    Two closure welds on filled Defense Waste Processing Facility (DWPF) canisters failed to be within the acceptance criteria in the DWPF operating procedure SW4-15.80-2.3 (1). In one case, the weld heat setting was inadvertently provided to the canister at the value used for test welds (i.e., 72%) and this oversight produced a weld at a current of nominally 210 kA compared to the operating procedure range (i.e., 82%) of 240 kA to 263 kA. The second weld appeared to experience an instrumentation and data acquisition upset. The current for this weld was reported as 191 kA. Review of the data from the Data Acquisition System (DAS) indicated that three of the four current legs were reading the expected values, approximately 62 kA each, and the fourth leg read zero current. Since there is no feasible way by further examination of the process data to ascertain if this weld was actually welded at either the target current or the lower current, a test plan was executed to provide assurance that these Nonconforming Welds (NCWs) meet the requirements for strength and leak tightness. Acceptance of the welds is based on evaluation of Test Nozzle Welds (TNW) made specifically for comparison. The TNW were nondestructively and destructively evaluated for plug height, heat tint, ultrasonic testing (UT) for bond length and ultrasonic volumetric examination for weld defects, burst pressure, fractography, and metallography. The testing was conducted in agreement with a Task Technical and Quality Assurance Plan (TTQAP) (2) and applicable procedures.

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

    International Nuclear Information System (INIS)

    Colleen Shelton-Davis

    2005-01-01

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

  12. 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.

  13. Production methods and costs of oxygen free copper canisters for nuclear waste disposal

    International Nuclear Information System (INIS)

    Aalto, H.; Rajainmaeki, H.; Laakso, L.

    1996-10-01

    The fabrication technology and costs of various manufacturing alternatives to make large copper canisters for disposal of spent nuclear fuel from reactors of Teollisuuden Voima Oy (TVO) and Imatran Voima Oy (IVO) are discussed. The canister design is based on the Posiva's concept where solid insert structure is surrounded by the copper mantle. During recent years Outokumpu Copper Products and Posiva have continued their work on development of the copper canisters. Outokumpu Copper Products has also increased capability to manufacture these canisters. In the study the most potential manufacturing methods and their costs are discussed. The cost estimates are based on the assumption that Outokumpu will supply complete copper mantles. At the moment there are at least two commercially available production methods for copper cylinder manufacturing. These routes are based on either hot extrusion of the copper tube or hot rolling, bending and EB-welding of the tube. Trial fabrications has been carried out with both methods for the full size canisters. These trials of the canisters has shown that both the forming from rolled plate and the extrusion are possible methods for fabricating copper canisters on a full scale. (orig.) (26 refs.)

  14. Finite element modelling of an evacuated canister for removal of molten radioactive glass

    International Nuclear Information System (INIS)

    Hatchell, B.K.; Deibler, J.E.; Ketner, G.L.

    1994-05-01

    Pacific Northwest Laboratory (PNL) has prepared a preliminary design for the West Valley Demonstration Project evacuated canister system. The function of the evacuated canister is to remove radioactive molten glass from a hot cell melter cavity during a planned melter shutdown. The proposed evacuated canister system consists of an L-shaped 4-inch 304L stainless steel (SS) schedule 40 pipe, sealed at one end with an aluminum plug and attached at the other end to a canister. While the canister is being filled, it is positioned and held above the melter at approximately 15 degree from horizontal by two turntable-mounted cranes. ANSYS finite element analyses were conducted to evaluate the heat transfer from the glass to the canister and establish a maximum canister temperature for material strength evaluation. Finite element structural analyses were conducted to identify areas that required reinforcement for high temperature use. Finite element results will be used to locate strain gauges at high stress locations during prototype testing

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

    Science.gov (United States)

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

    2005-01-01

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

  16. Equipment for deployment of canisters with spent nuclear fuel and bentonite buffer in horizontal holes

    International Nuclear Information System (INIS)

    Henttonen, V.; Suikki, M.

    1992-08-01

    The study presents the predesign of equipment for the deployment of canisters in long horizontal holes. The canisters are placed in the centre of the hole and are surrounded by a bentonite buffer. In thE study the canisters are assumed to have a diameter of 1.6 m and a length of 5.9 m, including the hemispherical ends. Their total weight is 60 tonnes. The bentonite buffer after homogenization is 400 mm thick, making a total package diameter of 2.4 m. The deployment system consists of four wagons for handling The canisters and the bentonite blocks. To ensure safe emplacement, every part is installed separately in its final position. This also makes it possible to use small clearances between the canisters and the bentonite blocks and between the blocks and the rock wall. With small clearances, backfilling can be avoided. Another basic design idea is that the wagons are equipped with wheels, which are in direct contact with the rock walls. Thus, rails, which have to be removed as the deployment progresses, are unnecessary. To minimize the time taken for deploying one canister, the wagons are designed so that only three trips from the service area to the deposit area are needed. Due to the radiation in the vicinity of the canisters, the wagons have to be teleoperated

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

    International Nuclear Information System (INIS)

    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

  18. Equipment for deployment of canisters with spent nuclear fuel and bentonite buffer in horisontal holes

    International Nuclear Information System (INIS)

    Henttonen, V.; Suikki, M.

    1992-06-01

    This study presents the predesign of equipment for the deployment of canisters in long horizontal holes. The canisters are placed in the centre of the hole and are surrounded by a bentonite buffer. In this study the canisters are assumed to have a diameter of 1.6 m and a length of 5.9 m, including the hemispherical ends. Their total weight is 60 tonnes. The bentonite buffer after homogenization is 400 mm thick, making a total package diameter of 2.4 m. The deployment system consists of four wagons for handling the canisters and the bentonite blocks. To ensure safe emplacement, every part is installed separately in its final position. This also makes it possible to use small clearances between the canisters and the bentonite blocks and between the blocks and the rock wall. With small clearances, backfilling can be avoided. Another basic design idea is that the wagons are equipped with wheels, which are in direct contact with the rock walls. Thus, rails, which have to be removed as the deployment progresses, are unnecessary. To minimize the time taken for deploying one canister, the wagons are designed so that only three trips from the service area to the deposit area are needed. Due to the radiation in the vicinity of the canisters, the wagons have to be teleoperated. (au)

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

    International Nuclear Information System (INIS)

    Hamilton, L.

    2001-01-01

    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

  20. Scoping calculations for canister-tunnel migration of corrodants, oxidants and radionuclides

    International Nuclear Information System (INIS)

    Shaw, W.; Worth, D.

    1992-03-01

    This report presents the mathematical models and results obtained for a set of scooping calculations which estimate the possible extent of the vertical migration of canister corrodants, oxidants (forming a redox front) and radionuclides between a copper canister containing spent nuclear fuel, and an overlying emplacement tunnel. The KBS-3 concept for the disposal of spent nuclear fuel is copper canisters, vertically emplaced in deposition holes bored in the floor of a tunnel, situated deep underground. The deposition holes are filled with a buffer of bentonite and the tunnel is backfilled with a mixture of sand and bentonite. (au)

  1. Volumes, Masses, and Surface Areas for Shippingport LWBR Spent Nuclear Fuel in a DOE SNF Canister

    International Nuclear Information System (INIS)

    J.W. Davis

    1999-01-01

    The purpose of this calculation is to estimate volumes, masses, and surface areas associated with (a) an empty Department of Energy (DOE) 18-inch diameter, 15-ft long spent nuclear fuel (SNF) canister, (b) an empty DOE 24-inch diameter, 15-ft long SNF canister, (c) Shippingport Light Water Breeder Reactor (LWBR) SNF, and (d) the internal basket structure for the 18-in. canister that has been designed specifically to accommodate Seed fuel from the Shippingport LWBR. Estimates of volumes, masses, and surface areas are needed as input to structural, thermal, geochemical, nuclear criticality, and radiation shielding calculations to ensure the viability of the proposed disposal configuration

  2. General aspects of the mechanical integrity of canisters

    International Nuclear Information System (INIS)

    Saario, Timo

    2007-01-01

    This paper attempts to introduce a new point of view to the mechanical integrity of the canisters, 'mechanical integrity evolutionary path'. The mechanical integrity evolutionary path is a description of development of the critical parameters involved in the prevailing degradation modes as a function of time. The degradation mechanisms considered are: mechanical overload; creep; and stress corrosion cracking. For each degradation mechanism one may consider two different states; initial state; critical state. The initial state considered will be different for different degradation mechanisms. For example stress corrosion cracking (SCC), which involves electrochemical steps is not possible without a surface covering aqueous phase. Thus, potentially, the initial state for SCC is that existing after saturation. On the other hand, the initial state for a possible mechanical overload can be different in different periods during the mechanical integrity evolutionary path. During the handling and transport stages the initial state is 'ex works', while during a glaciation the initial state has been altered due to creep, corrosion and possible SCC processes. The canister will go through mechanical overload during saturation and bentonite swelling phases and it will deform to fit the form of the insert. The initial state for this period is 'ex works', with e.g. manufacturing defects. The insert is designed to bear the load after closing the gap. In the 'ex works' state directionality of the mechanical properties has been raised lately as a new issue worth checking. Within the projected evolutionary path two events have been especially considered; seismic events and glaciation. A glacier 2 km thick would increase the hydrostatic pressure with 20 MPa if there were a mechanism transmitting the load into the aqueous phase. Remembering what makes ice skating possible such a mechanism seems plausible. For mechanical overload the critical state is relatively straightforward to

  3. General aspects of the mechanical integrity of canisters

    Energy Technology Data Exchange (ETDEWEB)

    Saario, Timo [VTT Materials and Building (Finland)

    2007-09-15

    This paper attempts to introduce a new point of view to the mechanical integrity of the canisters, 'mechanical integrity evolutionary path'. The mechanical integrity evolutionary path is a description of development of the critical parameters involved in the prevailing degradation modes as a function of time. The degradation mechanisms considered are: mechanical overload; creep; and stress corrosion cracking. For each degradation mechanism one may consider two different states; initial state; critical state. The initial state considered will be different for different degradation mechanisms. For example stress corrosion cracking (SCC), which involves electrochemical steps is not possible without a surface covering aqueous phase. Thus, potentially, the initial state for SCC is that existing after saturation. On the other hand, the initial state for a possible mechanical overload can be different in different periods during the mechanical integrity evolutionary path. During the handling and transport stages the initial state is 'ex works', while during a glaciation the initial state has been altered due to creep, corrosion and possible SCC processes. The canister will go through mechanical overload during saturation and bentonite swelling phases and it will deform to fit the form of the insert. The initial state for this period is 'ex works', with e.g. manufacturing defects. The insert is designed to bear the load after closing the gap. In the 'ex works' state directionality of the mechanical properties has been raised lately as a new issue worth checking. Within the projected evolutionary path two events have been especially considered; seismic events and glaciation. A glacier 2 km thick would increase the hydrostatic pressure with 20 MPa if there were a mechanism transmitting the load into the aqueous phase. Remembering what makes ice skating possible such a mechanism seems plausible. For mechanical overload the critical

  4. Multi Canister Overpack (MCO) Topical Report [SEC 1 THRU 6

    International Nuclear Information System (INIS)

    GARVIN, L.J.

    2002-01-01

    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 facilities to permit the safe retrieval, packaging, transport, conditioning, 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 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. A common thread that was identified among the structures was the MCO. Each structure exists for the specific purpose of treating or storing the MCO and its contents. Normally, an extensive amount of MCO-related documentation would be generated for each of the facility safety analysis reports. However, the expedited schedule for removing spent fuel from the K Basins requires that the documentation effort be minimized and repetitious activities be eliminated. Therefore, this topical report has been prepared to address those aspects of the MCO that will be common to the facilities. The MCO will be included in each facility's safety documentation by reference to this topical report. By capturing the design of the MCO and its safety evaluation in a single document, repetition, inconsistency, and duplication of

  5. 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.

  6. Modelling studies for the assessment of the Advanced Cold Process Canister

    International Nuclear Information System (INIS)

    Henshaw, J.; Hoch, A.R.; Sharland, S.M.

    1991-01-01

    The Advanced Cold Process Canister (ACPC) is a new concept for the encapsulation of spent nuclear fuel for geological disposal. It consists of steel canister encased in a copper overpack. In this paper, modelling studies to assess the performance of the ACPC under repository conditions are presented. The production of nitric acid and ammonia through radiolysis of any water remaining inside the canister under fault conditions has been examined in this study. However, results suggest that only low levels are possible, and the risk of stress-corrosion cracking is considered small. The corrosion behavior subsequent to a breach in the outer canister was also considered. A model was constructed to predict the hydrogen gas production due to corrosion reactions, and evolution of the corrosion behavior

  7. Gas and liquid sampling for closed canisters in KW Basin - Work Plan

    International Nuclear Information System (INIS)

    Pitkoff, C.C.

    1995-01-01

    Work Plan for the design and fabrication of gas/liquid sampler for closed canister sampling in KW Basin. This document defines the tasks associated with the design, fabrication, assembly, and acceptance testing equipment necessary for gas and liquid sampling of the Mark I and Mark II canisters in the K-West basin. The sampling of the gas space and the remaining liquid inside the closed canisters will be used to help understand any changes to the fuel elements and the canisters. Specifically, this work plan will define the scope of work and required task structure, list the technical requirements, describe design configuration control and verification methodologies, detail quality assurance requirements, and present a baseline estimate and schedule

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

    International Nuclear Information System (INIS)

    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

  9. 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.

  10. Study on radon concentration monitoring using activated charcoal canisters in high humidity environments

    International Nuclear Information System (INIS)

    Wang Yuexing; Wang Haijun; Yang Yifang; Qin Sichang; Wang Zhentao; Zhang Zhenjiang

    2009-01-01

    The effects of humidity on the sensitivity using activated charcoal canisters for measuring radon concentrations in high humidity environments were studied. Every canister filled with 80 g of activated charcoal, and they were exposed to 48 h or 72 h in the relative humidity of 68%, 80%, 88% and 96% (28 degree C), respectively. The amount of radon absorbed in the canisters was determined by counting the gamma rays from 214 Pb and 214 Bi (radon progeny). The results showed that counts decreased with the increase of relative humidity. There was a negative linear relationship between count and humidity. In the relative humidity range of 68%-96%, the sensitivity of radon absorption decreased about 2.4% for every 1% (degree)rise in humidity. The results also showed that the exposure time of the activated charcoal canisters should be less than 3 days. (authors)

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

    International Nuclear Information System (INIS)

    Cannell, G.R.

    1999-01-01

    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

  12. Fuel canister and blockage pin fabrication for SLSF Experiment P4

    International Nuclear Information System (INIS)

    Rhude, H.V.; Folkrod, J.R.; Noland, R.A.; Schaus, P.S.; Benecke, M.W.; Delucchi, T.A.

    1983-01-01

    As part of its fast breeder reactor safety research program, Argonne National Laboratory (ANL) has conducted an experiment (SLSF Experiment P4) to determine the extent of fuel-failure propagation resulting from the release of molten fuel from one or more heat-generating fuel canisters. The test conditions consisted of 37 full-length FTR fuel pins operating at FTR rated core nominal peak fuel/reduced coolant conditions. Thirty-four of the the fuel pins were prototypical FTR mixed-oxide fuel pins. The other three fuel pins were fabricated with a mid-core section having an enlarged canister containing fully enriched UO 2 . Two of the canisters were cylindrical and one was fluted. The cylindrical canisters were designed to fail and release molten fuel into the 37-pin fuel cluster at near full power

  13. Hazard categorization of 100 K West fuel canister gas and liquid sampling

    International Nuclear Information System (INIS)

    Alwardt, L.D.

    1994-01-01

    This report documents the determination that the activities associated with the 100 K West fuel canister gas and liquid sampling are classified as Hazard Category Other (consequences are below criteria for Category 3)

  14. Effect of canister size on costs of disposal of SRP high-level wastes

    International Nuclear Information System (INIS)

    McDonell, W.R.

    1982-01-01

    The current plan for managing the high-level nuclear wastes at the Savannah River Plant (SRP) calls for processing them into solid forms contained in stainless steel canisters for eventual disposal in a federal geologic repository. A new SRP facility called the Defense Waste Processing Facility (DWPF) is being designed for the onsite waste processing operations. Preliminary evaluations indicate that costs of the overall disposal operation will depend significantly on the size of the canisters, which determines the number of waste forms to be processed. The objective of this study was to evaluate the effects of canister size on costs of DWPF process operations, including canister procurement, waste solidification, and interim storage, on offsite transport, and on repository costs of disposal, including provision of suitable waste packages

  15. Multi-purpose canisters as an alternative for storage, transportation, and disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    Hollaway, W.R.; Rozier, R.; Nitti, D.A.; Williams, J.R.

    1993-01-01

    A study was conducted to assess the feasibility of using multi-purpose canisters to handle spent nuclear fuel throughout the Civilian Radioactive Waste Management System. Multi-purpose canisters would be sealed, metallic containers maintaining multiple spent fuel assemblies in a dry, inert environment and overpacked separately and uniquely for the various system elements of storage, transportation, and disposal. Using five implementation scenarios, the multi-purpose canister was evaluated with regard to several measures of effectiveness, including number of handlings, radiation exposure, cost, schedule and licensing considerations, and public perception. Advantages and disadvantages of the multi-purpose canister were identified relative to the current reference system within each scenario, and the scenarios were compared to determine the most effective method of implementation

  16. Multiple-canister flow and transport code in 2-dimensional space. MCFT2D: user's manual

    International Nuclear Information System (INIS)

    Lim, Doo-Hyun

    2006-03-01

    A two-dimensional numerical code, MCFT2D (Multiple-Canister Flow and Transport code in 2-Dimensional space), has been developed for groundwater flow and radionuclide transport analyses in a water-saturated high-level radioactive waste (HLW) repository with multiple canisters. A multiple-canister configuration and a non-uniform flow field of the host rock are incorporated in the MCFT2D code. Effects of heterogeneous flow field of the host rock on migration of nuclides can be investigated using MCFT2D. The MCFT2D enables to take into account the various degrees of the dependency of canister configuration for nuclide migration in a water-saturated HLW repository, while the dependency was assumed to be either independent or perfectly dependent in previous studies. This report presents features of the MCFT2D code, numerical simulation using MCFT2D code, and graphical representation of the numerical results. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Patel, R.; Punshon, C.; Nicholas, J.; Bastid, P.; Zhou, R.; Schneider, C.; Bagshaw, N.; Howse, D.; Hutchinson, E. [TWI Ltd, Cambridge, (United Kingdom); Asano, R. [Hitachi Zosen Corporation, Osaka (Japan); King, S. [Integrity Corrosion Consulting Ltd, Calgary, Alberta (Canada)

    2012-10-15

    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

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

    International Nuclear Information System (INIS)

    Patel, R.; Punshon, C.; Nicholas, J.; Bastid, P.; Zhou, R.; Schneider, C.; Bagshaw, N.; Howse, D.; Hutchinson, E.; Asano, R.; King, S.

    2012-10-01

    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

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

    International Nuclear Information System (INIS)

    Garvin, L.J.

    1996-01-01

    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

  20. Canister materials proposed for final disposal of high level nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    Mattson, E; Odoj, R; Merz, E [eds.

    1981-06-01

    The nuclear waste will be enclosed in corrosion resistant canisters. These will be deposited in repositories in geological formations, such as granite, basalt, clay, bedded or domed salt, or the sediments beneath the deep ocean floor. There the canisters will be exposed to groundwater, brine or seawater at an elevated temperature. Species formed by radiolysis may effect the corrosivity of the agent. The corrosion resistance of candidate canister materials is evaluated by corrosion tests and by thermodynamic and mass transport calculations. Examinations of ancient metal objects after long exposure in nature may give additional information. On the basis of the work carried out so far, the principal candidate canister materials are titanium materials, copper, and highpurity alumina.

  1. Process and machinery description of equipment for deposition of canisters in medium-long deposition holes

    International Nuclear Information System (INIS)

    Kalbantner, P.

    2001-08-01

    In this report twelve methods are presented to deposit a canister with spent nuclear fuel in a horizontal hole, several canisters per hole (MLH). These methods are part of the KBS-3 system. They have been developed successively, after an analysis of weak points and strong points in previously described methods. In conformance with the guidelines for Project JADE, a choices of system has been considered during the development work. This is whether canister and bentonite buffer should be deposited 'in parts', i.e. at different occasions, but shortly after each other or 'in a package', i.e. together in a single package. The other choice in the guidelines for the JADE project, whether the canister should be placed in a radiation shield or not during transport in the secondary tunnels, was not relevant to MLR. The basic technical problem is depositing heavy objects, the canister and the buffer components, in an horizontal hole which is approximately 200 m deep. Two methods for depositing of the bentonite barrier and the canisters in separate processes have been studied. For depositing of the bentonite barrier and the canister 'in a package', four alternative techniques have been studied: a metallic sleeve around the package, a loading scoop that is rotated, a fork carriage and rails. The repeated transports in a hole, a consequence of depositing several canisters in the same hole, could lead to the rock being crushed. The mutual impact of machines, load and rock wall has therefore been particularly considered. In several methods, the use of a gangway has been proposed (steel plates or layer of ice). A failure mode and effect analysis has been performed for one of the twelve methods. When comparing with a method to deposit one canister per hole using the same technique, the need for equipment and resources is far larger for this MLH method if incidents should occur during depositing. The development work reported here has not yet yielded a definitive method for placing

  2. Data compilation report: Gas and liquid samples from K West Basin fuel storage canisters

    International Nuclear Information System (INIS)

    Trimble, D.J.

    1995-01-01

    Forty-one gas and liquid samples were taken from spent fuel storage canisters in the K West Basin during a March 1995 sampling campaign. (Spent fuel from the N Reactor is stored in sealed canisters at the bottom of the K West Basin.) A description of the sampling process, gamma energy analysis data, and quantitative gas mass spectroscopy data are documented. This documentation does not include data analysis

  3. 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.

  4. Application of the TEMPEST computer code to canister-filling heat transfer problems

    International Nuclear Information System (INIS)

    Farnsworth, R.K.; Faletti, D.W.; Budden, M.J.

    1988-03-01

    Pacific Northwest Laboratory (PNL) researchers used the TEMPEST computer code to simulate thermal cooldown behavior of nuclear waste glass after it was poured into steel canisters for long-term storage. The objective of this work was to determine the accuracy and applicability of the TEMPEST code when used to compute canister thermal histories. First, experimental data were obtained to provide the basis for comparing TEMPEST-generated predictions. Five canisters were instrumented with appropriately located radial and axial thermocouples. The canister were filled using the pilot-scale ceramic melter (PSCM) at PNL. Each canister was filled in either a continous or a batch filling mode. One of the canisters was also filled within a turntable simulant (a group of cylindrical shells with heat transfer resistances similar to those in an actual melter turntable). This was necessary to provide a basis for assessing the ability of the TEMPEST code to also model the transient cooling of canisters in a melter turntable. The continous-fill model, Version M, was found to predict temperatures with more accuracy. The turntable simulant experiment demonstrated that TEMPEST can adequately model the asymmetric temperature field caused by the turntable geometry. Further, TEMPEST can acceptably predict the canister cooling history within a turntable, despite code limitations in computing simultaneous radiation and convection heat transfer between shells, along with uncertainty in stainless-steel surface emissivities. Based on the successful performance of TEMPEST Version M, development was initiated to incorporate 1) full viscous glass convection, 2) a dynamically adaptive grid that automatically follows the glass/air interface throughout the transient, and 3) a full enclosure radiation model to allow radiation heat transfer to non-nearest neighbor cells. 5 refs., 47 figs., 17 tabs

  5. Gas and liquid sampling for closed canisters in K-West basins - functional design criteria

    International Nuclear Information System (INIS)

    Pitkoff, C.C.

    1994-01-01

    The purpose of this document is to provide functions and requirements for the design and fabrication of equipment for sampling closed canisters in the K-West basin. The samples will be used to help determine the state of the fuel elements in closed canisters. The characterization information obtained will support evaluation and development of processes required for safe storage and disposition of Spent Nuclear Fuel (SNF) materials

  6. Corrosion studies on HGW-canister materials for marine disposal

    International Nuclear Information System (INIS)

    Taylor, K.J.; Bland, I.D.; Marsh, G.P.

    1984-07-01

    A combination of mathematical modelling and experimental studies has been used to investigate and assess the long term corrosion behaviour of heat generating waste canister/ overpack materials under conditions relevant to deep ocean disposal. Preliminary operation of the model, using improved electrochemical kinetic data from the experimental programme, has indicated that the general corrosion rate of carbon steel at 90 deg C will be 57 μm yr -1 which is equivalent to a metal loss of 57 mm in 1000 years. This prediction compares favourably with the results from long term tests, which are also in progress, for plain and electron beam welded carbon steel specimens embedded in marine sediment at 90 deg C under active dissolution conditions. Tests with γ-radiation at a dose rate of 1.5 x 10 5 R h -1 have shown that the pH of seawater falls to 3.7 after 5000 hours exposure causing a significant increase in the corrosion rate of carbon steel from 50 to 80 μm yr -1 . Further work is in progress to investigate the mechanism of this acidification and whether it also occurs at the more realistic lower radiation dose rates. (author)

  7. Native copper as a natural analogue for copper canisters

    International Nuclear Information System (INIS)

    Marcos, N.

    1989-12-01

    This paper discusses the occurrence of native copper as found in geological formations as a stability analogue of copper canisters that are planned to be used for the disposal of spent nuclear fuel in the Finnish bedrock. A summary of several publications on native copper occurrences is presented. The present geochemical and geohydrological conditions in which copper is met with in its metallic state show that metallic copper is stable in a wide range of temperatures. At low temperatures native copper is found to be stable where groundwater has moderate pH (about 7), low Eh (< +100 mV), and low total dissolved solids, especially chloride. Microscopical and microanalytical studies were carried out on a dozen of rock samples containing native copper. The results reveal that the metal shows no significant alteration. Only the surface of copper grains is locally coated. In the oldest samples there exist small corrosion cracks; the age of the oldest samples is over 1,000 million years. A review of several Finnish groundwater studies suggests that there are places in Finland where the geohydrological conditions are favourable for native copper stability. (orig.)

  8. 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.

  9. Probabilistic analysis of canister inserts for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Dillstroem, Peter [Det Norske Veritas, Stockholm (Sweden)

    2005-10-01

    In this study, probabilistic analysis of canister inserts for spent nuclear fuel has been performed. The main conclusions are: 1. For the baseline case, the probability of failure is insignificant ({approx} 2x10{sup -9}). This is the case even though several conservative assumptions have been made both in underlying deterministic analysis and in the probabilistic analysis. 2. The initiation event dominates (over the local collapse event) when the external pressure is below the baseline case (p = 44 MPa). The local collapse event dominates when the external pressure is above the baseline case. 3. The local collapse event is strongly dependent of the assumed external pressure. 4. The analysis of collapse only considers the first local collapse event, total collapse of the insert will occur at a much higher pressure. 5. The resulting probabilities are more dependent on the assumption regarding the eccentricity of the cassette than the assumption regarding outer corner radius of the profiles for steel section cassette. The results indicate that the maximum allowed eccentricity should not be larger than 5 mm. 6. The probability of initiation of crack growth is calculated using a defect distribution where one assumes the existence of one crack-like defect. A simple scaling argument can be applied to consider the number of defects through the thickness.

  10. Design basis for the copper canister. Stage one

    International Nuclear Information System (INIS)

    Bowyer, W. H.

    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

  11. Probabilistic analysis of canister inserts for spent nuclear fuel

    International Nuclear Information System (INIS)

    Dillstroem, Peter

    2005-10-01

    In this study, probabilistic analysis of canister inserts for spent nuclear fuel has been performed. The main conclusions are: 1. For the baseline case, the probability of failure is insignificant (∼ 2x10 -9 ). This is the case even though several conservative assumptions have been made both in underlying deterministic analysis and in the probabilistic analysis. 2. The initiation event dominates (over the local collapse event) when the external pressure is below the baseline case (p = 44 MPa). The local collapse event dominates when the external pressure is above the baseline case. 3. The local collapse event is strongly dependent of the assumed external pressure. 4. The analysis of collapse only considers the first local collapse event, total collapse of the insert will occur at a much higher pressure. 5. The resulting probabilities are more dependent on the assumption regarding the eccentricity of the cassette than the assumption regarding outer corner radius of the profiles for steel section cassette. The results indicate that the maximum allowed eccentricity should not be larger than 5 mm. 6. The probability of initiation of crack growth is calculated using a defect distribution where one assumes the existence of one crack-like defect. A simple scaling argument can be applied to consider the number of defects through the thickness

  12. NAC gets OK for waste canister, looks for buyers

    International Nuclear Information System (INIS)

    Newman, P.

    1994-01-01

    The Nuclear Regulatory Commission has given design approval to the first dual-purpose waste cansiter suitable for storing and transporting irradiated nuclear fuel. The cask could be commercially available by January 1995. NRC issued a transportation certificate for the canister, which was developed by Atlanta-based NAC Services Inc., a subsidiary of NAC Holding Inc. That certificate, which says the cask is a suitable vessel for transporting radioactive wastes by rail and truck, is the first credential of a two-part licensing process the design must acquire. Testing of the cask has been extensive, including drop tests and pin-puncture tests. Roughly 19 feet long and eight feet in diameter, the cask is designed to hold 26 pressurized water reactor fuel assemblies. NAC officials say the cask design will soon be adapted to accomodate larger boiling water reactor fuel assemblies. Utilities will need some convincing that the dual-purpose $1.5 million cask is worth the money, particularly since companies currently have no use for the cask's transportation capabilities

  13. The Meaning of the Sampling of the ZPPR Canisters And Proposed New Surveillance Operating Instructions

    Energy Technology Data Exchange (ETDEWEB)

    Charles W. Solbrig

    2007-01-01

    Analysis of the sample data taken from the ZPPR canisters containing Uranium plate fuel indicates that (as of February 2004) hydriding could be occurring in 35 of them. Since there appears to be no way of determining that a getter is functional, the getters in all the canisters should be replaced now (unless canister residence time can be determined) to prevent further hydriding. In addition, the surveillance procedure should be modified. Canisters to be inspected should be selected sequentially, 12 each quarter resulting in all being opened once every five years. Three of the 12 should be sampled and results reported before opening any of the canisters. Water vapor and pressure should be measured as well as the current hydrogen, oxygen, and nitrogen. Then all 12 canisters should be opened for physical evaluation of the plate conditions and correlation with the sample measurements. The getters should be replaced at each inspection ensuring that no getter is used more than five years. The data should be analyzed each year and a conclusion made on the adequacy of the surveillance procedure and modifications made if it is inadequate.

  14. Development of a facility for fabricating nuclear waste canisters from radioactively contaminated steel

    International Nuclear Information System (INIS)

    Logan, J.A.; Larsen, M.M.

    1986-01-01

    This paper describes design of a facility and processes capable of using radioactively contaminated waste steel as the principal raw material for fabricating stainless steel canisters to be used for disposal of nuclear high-level waste. By such action, expenditure (i.e., permanent loss to society) of thousands of tons of uncontaminated chromium and nickel to fabricate such canisters can be avoided. Moreover, the cost and risks involved in disposing of large accumulations of radioactively contaminated steel as low-level radioactive waste (LLRW), that would otherwise be necessary, can also be avoided. The canister fabrication processes (involving centrifugal casting) described herein have been tested and proven for this application. The performance characteristics of stainless steel canisters so fabricated have been tested and agreed to by the organizations that have been involved in this development work (Battelle Memorial Institute, DuPont, EGandG and the Savannah River Laboratory) as equivalent to the performance characteristics of canisters fabricated of uncontaminated wrought stainless steel. It is estimated that the production cost for fabricating canisters by the methods described will not differ greatly from the production cost using uncontaminated wrought steel, and the other costs avoided by not having to dispose of the contaminated steel as LLRW could cause this method to produce the lowest ultimate overall costs

  15. 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)

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

    International Nuclear Information System (INIS)

    1993-08-01

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

  17. Decontamination of stainless steel canisters that contain high-level waste

    International Nuclear Information System (INIS)

    Bray, L.A.

    1987-01-01

    At the West Valley Nuclear Services Company (WVNSC) in West Valley, New York, high-level radioactive waste (HLW) will be vitrified into a borosilicate glass form and poured into large, stainless steel canisters. During the filling process, volatile fission products, principally 137 Cs, condense on the exterior of the canisters. The smearable contamination that remains on the canisters after they are filled and partially cooled must be removed from the canisters' exterior surfaces prior to their storage and ultimate shipment to a US Department of Energy (DOE) repository for disposal. A simple and effective method was developed for decontamination of HLW canisters. This method of chemical decontamination is applicable to a wide variety of contaminated equipment found in the nuclear industry. The process employs a reduction-oxidation system [Ce(III)/Ce(IV)] in nitric acid solution to chemically mill the surface of stainless steel, similar to the electropolishing process, but without the need for an applied electrical current. Contaminated canisters are simply immersed in the solution at controlled temperature and Ce(IV) concentration levels

  18. Summary of Preliminary Criticality Analysis for Peach Bottom Fuel in the DOE Standardized Spent Nuclear Fuel Canister

    International Nuclear Information System (INIS)

    Henrikson, D.J.

    1999-01-01

    The Department of Energy's (DOE's) National Spent Nuclear Fuel Program is developing a standardized set of canisters for DOE spent nuclear fuel (SNF). These canisters will be used for DOE SNF handling, interim storage, transportation, and disposal in the national repository. Several fuels are being examined in conjunction with the DOE SNF canisters. This report summarizes the preliminary criticality safety analysis that addresses general fissile loading limits for Peach Bottom graphite fuel in the DOE SNF canister. The canister is considered both alone and inside the 5-HLW/DOE Long Spent Fuel Co-disposal Waste Package, and in intact and degraded conditions. Results are appropriate for a single DOE SNF canister. Specific facilities, equipment, canister internal structures, and scenarios for handling, storage, and transportation have not yet been defined and are not evaluated in this analysis. The analysis assumes that the DOE SNF canister is designed so that it maintains reasonable geometric integrity. Parameters important to the results are the canister outer diameter, inner diameter, and wall thickness. These parameters are assumed to have nominal dimensions of 45.7-cm (18.0-in.), 43.815-cm (17.25-in), and 0.953-cm (0.375-in.), respectively. Based on the analysis results, the recommended fissile loading for the DOE SNF canister is 13 Peach Bottom fuel elements if no internal steel is present, and 15 Peach Bottom fuel elements if credit is taken for internal steel

  19. Thermal and mechanical analyses of the spent nuclear fuel disposal canister and its barriers according to the design variable change

    International Nuclear Information System (INIS)

    Kwon, Young Joo

    2006-03-01

    This work constitutes a summary of research and development made for design and dimensioning of the spent nuclear fuel disposal canister. Since the spent nuclear fuel disposal emits high temperature heats and much radiation, its careful treatment is required. For that, a long term (usually 10,000 years) safe repository for the spent nuclear fuel disposal should be secured. Usually this repository is expected to locate at a depth of 500m underground. Many various analyses should be performed to secure the structural safety of the canister. For past years, these analyses have been performed to develop the canister model (so-called DKC-1 model). The diameter of the designed KDC-1 canister model is D=102m. However, there still remain some structural evaluations to make sure the structural safety of the designed KDC-1 canister mode. The one is the structural safety evaluation of the canister for the falling accident in the repository while handling the canister. There may happen two typical falling accidents in the repository. The one is the falling accident of the canister in the borehole while depositing the canister into the borehole. In these falling accidents the collision impact force between the canister and the surface of the ground or the bottom of the borehole may cause the structural damage onto the canister. However, the canister should be designed to withstand this impact force. Hence, the structural analysis of the canister for this impact force is required to guarantee the structural safety of the canister for this falling accident. Therefore in this report, the structural analyses of the KDC-1 canister model of the diameter of 102cm for two types of falling accidents are carried out for the impact forces while the canister collides onto the surface of the ground or the bottom of the borehole. The nonlinear structural analyses are performed for the canister to get the accurate analysis results assuming the materials composing canister parts as elasto

  20. Numerical study of canister filters with alternatives filter cap configurations

    Science.gov (United States)

    Mohammed, A. N.; Daud, A. R.; Abdullah, K.; Seri, S. M.; Razali, M. A.; Hushim, M. F.; Khalid, A.

    2017-09-01

    Air filtration system and filter play an important role in getting a good quality air into turbo machinery such as gas turbine. The filtration system and filter has improved the quality of air and protect the gas turbine part from contaminants which could bring damage. During separation of contaminants from the air, pressure drop cannot be avoided but it can be minimized thus helps to reduce the intake losses of the engine [1]. This study is focused on the configuration of the filter in order to obtain the minimal pressure drop along the filter. The configuration used is the basic filter geometry provided by Salutary Avenue Manufacturing Sdn Bhd. and two modified canister filter cap which is designed based on the basic filter model. The geometries of the filter are generated by using SOLIDWORKS software and Computational Fluid Dynamics (CFD) software is used to analyse and simulates the flow through the filter. In this study, the parameters of the inlet velocity are 0.032 m/s, 0.063 m/s, 0.094 m/s and 0.126 m/s. The total pressure drop produce by basic, modified filter 1 and 2 is 292.3 Pa, 251.11 Pa and 274.7 Pa. The pressure drop reduction for the modified filter 1 is 41.19 Pa and 14.1% lower compared to basic filter and the pressure drop reduction for modified filter 2 is 17.6 Pa and 6.02% lower compared to the basic filter. The pressure drops for the basic filter are slightly different with the Salutary Avenue filter due to limited data and experiment details. CFD software are very reliable in running a simulation rather than produces the prototypes and conduct the experiment thus reducing overall time and cost in this study.

  1. Manufacturing of the canister shells T54 and T55

    International Nuclear Information System (INIS)

    Raiko, H.

    2008-10-01

    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.)

  2. Corrosion studies on HGW-canister materials for marine disposal

    International Nuclear Information System (INIS)

    Taylor, K.J.; Bland, I.D.; Smith, S.; Marsh, G.P.

    1986-03-01

    Results are presented from theoretical and experimental work undertaken to investigate and assess the general corrosion behaviour of carbon steel canister/overpacks for heat generating nuclear waste under marine disposal conditions. The mean general corrosion rates of carbon steels, determined experimentally by polarisation resistance measurements on specimens in on-going immersion tests, are between 65-124 μm yr -1 at 90 0 C and 5-25 μm yr -1 at 25 0 C and are tending to increase with time. Anomalously high corrosion rates are being indicated by similar tests at 50 0 C. It is not clear what reliance should be placed on the polarisation resistance results, however, and therefore no conclusion will be drawn until the tests are dismantled and inspected in the 1985/86 programme. Tests with γ-radiation on forged carbon steel specimens immersed in deaerated seawater at 90 0 C show that this causes an acceleration of corrosion rate at the three dose rates down to at least 300 R h -1 . Deep ocean sediment from GME also accelerates the corrosion rate of carbon steel in deaerated seawater both with and without γ-radiation. The effect diminishes with continued exposure and is thought to be due to the presence of either an additional so far unidentified oxidising agent or some component which reduces the corrosion protection afforded by the build up of a corrosion product layer. Acquisition of improved electrochemical kinetic data for the mathematical model is now complete, and the model has been run for temperatures of 25 and 90 0 C, where it predicts steady corrosion rates of 19.3 and 180 μm/yr. The model has shown that the rate of attack is not influenced greatly by the depth of sediment, and that the component of corrosion caused by radiation is of the order of 7 mm over 1000 years. (author)

  3. Tests for manufacturing technology of disposal canisters for nuclear spent fuel

    International Nuclear Information System (INIS)

    Raiko, H.; Salonen, T.; Meuronen, I.; Lehto, K.

    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

  4. Modelling and analysis of canister and buffer for earthquake induced rock shear and glacial load

    International Nuclear Information System (INIS)

    Hernelind, Jan

    2010-08-01

    Existing fractures crossing a deposition hole may be activated and sheared by an earthquake. The effect of such a rock shear has been investigated by finite element calculations. The buffer material in a deposition hole acts as a cushion between the canister and the rock, which reduces the effect of a rock shear substantially. Lower density of the buffer yields softer material and reduced effect on the canister. However, at the high density that is suggested for a repository the stiffness of the buffer is rather high. The stiffness is also a function of the rate of shear, which means that there may be a substantial damage on the canister at very high shear rates. However, the earthquake induced rock shear velocity is lower than 1 m/s which is not considered to be very high. The rock shear has been modelled with finite element calculations with the code Abaqus. A three dimensional finite element mesh of the buffer and the canister has been created and simulation of a rock shear has been performed. The rock shear has been assumed to take place either perpendicular to the canister at the quarter point or at an inclined angle of 22.5 deg in tension. Furthermore horizontal shear has been studied using a vertical shear plane either at the centre or at 1/4-point for the canister. The shear calculations have been driven to a total shear of 10 cm. The canister also has to be designed to withstand the loads caused by a thick ice sheet. Besides rock shear the model has been used to analyse the effect of such glacial load (either combined with rock shear or without rock shear). This report also summarizes the effect when considering creep in the copper shell

  5. Transporting existing VSC-24 canisters using a risk-based licensing approach

    International Nuclear Information System (INIS)

    Srinivasan, R.; Sisley, S.E.; Hopf, J.E.

    2004-01-01

    The eventual disposition of the spent fuel assemblies loaded in canisters and casks currently designed and licensed only for on-site storage is an industry-wide issue. The canister-specific BUC evaluation approach developed by BFS can be used to license many of these storage canisters and casks for transportation. This will allow these storage canisters and casks to be transported intact to a long-term storage facility or repository, thereby minimizing fuel handling operations, impact on plant operations, and occupational exposure, as well as total infrastructure costs. Application of the proposed canister-specific BUC analysis approach to a preliminary evaluation of the 58 loaded MSBs demonstrates the benefits of this approach. The results of this preliminary evaluation show that a more rigorous analysis based on the known characteristics of the loaded spent fuel, rather than the design-basis fuel parameters, produces significantly lower maximum keff values and can be used to qualify many of the existing loaded storage canisters for transportation. Transportation certification for storage canisters having more reactive spent fuel payloads may require reliance on BUC approaches that are more aggressive than current NRC guidelines allow. Credit may be required for fission- product isotopes that do not have sufficient chemical assay data for benchmarking. In addition, reduced criticality safety margins may be required. For these more-aggressive BUC approaches, a risk assessment should be provided to support the NRC-approval basis. The risk assessment should evaluate the possibility and consequences of an accidental criticality event based upon inaccuracies in the characterization of the spent-fuel payloads

  6. Transporting existing VSC-24 canisters using a risk-based licensing approach

    Energy Technology Data Exchange (ETDEWEB)

    Srinivasan, R.; Sisley, S.E.; Hopf, J.E. [BNFL Fuel Solutions, Campbell, CA (United States)

    2004-07-01

    The eventual disposition of the spent fuel assemblies loaded in canisters and casks currently designed and licensed only for on-site storage is an industry-wide issue. The canister-specific BUC evaluation approach developed by BFS can be used to license many of these storage canisters and casks for transportation. This will allow these storage canisters and casks to be transported intact to a long-term storage facility or repository, thereby minimizing fuel handling operations, impact on plant operations, and occupational exposure, as well as total infrastructure costs. Application of the proposed canister-specific BUC analysis approach to a preliminary evaluation of the 58 loaded MSBs demonstrates the benefits of this approach. The results of this preliminary evaluation show that a more rigorous analysis based on the known characteristics of the loaded spent fuel, rather than the design-basis fuel parameters, produces significantly lower maximum keff values and can be used to qualify many of the existing loaded storage canisters for transportation. Transportation certification for storage canisters having more reactive spent fuel payloads may require reliance on BUC approaches that are more aggressive than current NRC guidelines allow. Credit may be required for fission- product isotopes that do not have sufficient chemical assay data for benchmarking. In addition, reduced criticality safety margins may be required. For these more-aggressive BUC approaches, a risk assessment should be provided to support the NRC-approval basis. The risk assessment should evaluate the possibility and consequences of an accidental criticality event based upon inaccuracies in the characterization of the spent-fuel payloads.

  7. Modelling and analysis of canister and buffer for earthquake induced rock shear and glacial load

    Energy Technology Data Exchange (ETDEWEB)

    Hernelind, Jan (5T Engineering AB (Sweden))

    2010-08-15

    Existing fractures crossing a deposition hole may be activated and sheared by an earthquake. The effect of such a rock shear has been investigated by finite element calculations. The buffer material in a deposition hole acts as a cushion between the canister and the rock, which reduces the effect of a rock shear substantially. Lower density of the buffer yields softer material and reduced effect on the canister. However, at the high density that is suggested for a repository the stiffness of the buffer is rather high. The stiffness is also a function of the rate of shear, which means that there may be a substantial damage on the canister at very high shear rates. However, the earthquake induced rock shear velocity is lower than 1 m/s which is not considered to be very high. The rock shear has been modelled with finite element calculations with the code Abaqus. A three dimensional finite element mesh of the buffer and the canister has been created and simulation of a rock shear has been performed. The rock shear has been assumed to take place either perpendicular to the canister at the quarter point or at an inclined angle of 22.5 deg in tension. Furthermore horizontal shear has been studied using a vertical shear plane either at the centre or at 1/4-point for the canister. The shear calculations have been driven to a total shear of 10 cm. The canister also has to be designed to withstand the loads caused by a thick ice sheet. Besides rock shear the model has been used to analyse the effect of such glacial load (either combined with rock shear or without rock shear). This report also summarizes the effect when considering creep in the copper shell

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

    International Nuclear Information System (INIS)

    Bermingham, A.J.; Booker, R.J.; Booth, H.R.; Ruehle, W.G.; Shevekov, S.; Silvester, A.G.; Tagart, S.W.; Thomas, J.A.; West, R.G.

    1978-01-01

    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. Dew point, internal gas pressure, and chemical composition of the gas within the free volume of DWPF canistered waste forms

    International Nuclear Information System (INIS)

    Harbour, J.R.; Herman, D.T.; Crump, S.; Miller, T.J.; McIntosh, J.

    1996-01-01

    The Defense Waste Processing Facility (DWPF) produced 55 canistered waste forms containing simulated waste glass during the four Waste Qualification campaigns of the DWPF Startup Test Program. Testing of the gas within the free volume of these canisters for dew point, internal gas pressure, and chemical composition was performed as part of a continuing effort to demonstrate compliance with the Waste Acceptance Product Specifications. Results are presented for six glass-filled canisters. The dew points within the canisters met the acceptance criterion of < 20 degrees C for all six canisters. Factors influencing the magnitude of the dew point are presented. The chemical composition of the free volume gas was indistinguishable from air for all six canisters. Hence, no foreign materials were present in the gas phase of these canisters. The internal gas pressures within the sealed canisters were < 1 atm at 25 degrees C for all six canisters which readily met the acceptance criterion of an internal gas pressure of less than 1.5 atm at 25 degrees C. These results provided the evidence required to demonstrate compliance with the Waste Acceptance Product Specifications

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

    Science.gov (United States)

    King, Fraser; Lilja, Christina; Vähänen, Marjut

    2013-07-01

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

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

    International Nuclear Information System (INIS)

    Turnbull, A.

    2009-07-01

    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

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

    International Nuclear Information System (INIS)

    Lanru Jing

    2004-04-01

    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

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

    International Nuclear Information System (INIS)

    Vafai, K.; Ettefagh, J.

    1986-07-01

    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

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

    International Nuclear Information System (INIS)

    Herlan Martono; Aisyah; Wati

    2007-01-01

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

  15. Corrosion experiments on stainless steels used in dry storage canisters of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ryskamp, J.M.; Adams, J.P.; Faw, E.M.; Anderson, P.A.

    1996-09-01

    Nonradioactive (cold) experiments have been set up in the Idaho Chemical Processing Plant (ICPP)-1634, and radioactive (hot) experiments have been set up in the Irradiated Fuel Storage Facility (IFSF) at ICPP. The objective of these experiments is to provide information on the interactions (corrosion) between the spent nuclear fuel currently stored at the ICPP and the dry storage canisters and containment materials in which this spent fuel will be stored for the next several decades. This information will be used to help select canister materials that will retain structural integrity over this period within economic, criticality, and other constraints. The two purposes for Dual Purpose Canisters (DPCs) are for interim storage of spent nuclear fuel and for shipment to a final geological repository. Information on how corrosion products, sediments, and degraded spent nuclear fuel may corrode DPCs will be required before the DPCs will be allowed to be shipped out of the State of Idaho. The information will also be required by the Nuclear Regulatory Commission (NRC) to support the licensing of DPCs. Stainless steels 304L and 316L are the most likely materials for dry interim storage canisters. Welded stainless steel coupons are used to represent the canisters in both hot and cold experiments.

  16. Drop of canistered spent fuel segments into a deep borehole and subsequent aerosol release

    International Nuclear Information System (INIS)

    Bantle, S.; Herbe, H.; Miu, J.

    1991-09-01

    The source term of the released aerosols is estimated. First, the number of failing canisters is calculated for the case of an axial symmetric canister (POLLUX) pile, and then, for the case of a 'zig-zag' pile, as found in reality. The weight-specific energy acting on the fuel - a measure for the degree of fuel fractioning - is determined from the acceleration acting on the pin segments. In the borehole prevails a steady-state flow pattern which is stimulated by the heat of the disposed waste canister, and is also influenced by the ventilation of the drift above the borehole. Based on this stationary flow pattern flow velocities are calculated by means of fluid mechanical methods. Further investigations deal with the unsteady case which occurs during and immediately after the canister drop as well as with the wake behind the canister. The most relevant result is that under the considered boundary conditions no release form the borehole into the repository is to be expected. (orig./HP) [de

  17. Corrosion experiments on stainless steels used in dry storage canisters of spent nuclear fuel

    International Nuclear Information System (INIS)

    Ryskamp, J.M.; Adams, J.P.; Faw, E.M.; Anderson, P.A.

    1996-09-01

    Nonradioactive (cold) experiments have been set up in the Idaho Chemical Processing Plant (ICPP)-1634, and radioactive (hot) experiments have been set up in the Irradiated Fuel Storage Facility (IFSF) at ICPP. The objective of these experiments is to provide information on the interactions (corrosion) between the spent nuclear fuel currently stored at the ICPP and the dry storage canisters and containment materials in which this spent fuel will be stored for the next several decades. This information will be used to help select canister materials that will retain structural integrity over this period within economic, criticality, and other constraints. The two purposes for Dual Purpose Canisters (DPCs) are for interim storage of spent nuclear fuel and for shipment to a final geological repository. Information on how corrosion products, sediments, and degraded spent nuclear fuel may corrode DPCs will be required before the DPCs will be allowed to be shipped out of the State of Idaho. The information will also be required by the Nuclear Regulatory Commission (NRC) to support the licensing of DPCs. Stainless steels 304L and 316L are the most likely materials for dry interim storage canisters. Welded stainless steel coupons are used to represent the canisters in both hot and cold experiments

  18. Characterization of projected DWPF glasses heat treated to simulate canister centerline cooling

    International Nuclear Information System (INIS)

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

    1992-05-01

    Liquid high-level nuclear waste will be immobilized at the Savannah River Site (SRS) by vitrification in borosilicate glass. The glass will be produced and poured into stainless steel canisters in the Defense Waste Processing Facility (DWPF). Eventually these canistered waste forms will be sent to a geologic repository for final disposal. In order to assure acceptability by the repository, the Department of Energy has defined requirements which DWPF canistered waste forms must meet. These requirements are the Waste Acceptance Preliminary Specifications (WAPS). The WAPS require DWPF to identify the crystalline phases expected to be present in the final glass product. Knowledge of the thermal history of the borosilicate glass during filling and cooldown of the canister is necessary to determine the amount and type of crystalline phases present in the final glass product. Glass samples of seven projected DWPF compositions were cooled following the same temperature profile as that of glass at the centerline of the full-scale DWPF canister. The glasses were characterized by x-ray diffraction and scanning electron microscopy to identify the crystalline phases present The volume percents of each crystalline phase present were determined by quantitative x-ray diffraction. The Product Consistency Test (PCI) was used to determine the durability of the heat-treated glasses

  19. 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

  20. Corrosion of the copper canister in the repository environment

    International Nuclear Information System (INIS)

    Hermansson, H.P.; Eriksson, Sture

    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 there is

  1. Fire simulation of the canister transfer and installation vehicle

    International Nuclear Information System (INIS)

    Peltokorpi, L.

    2012-12-01

    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 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 due

  2. Cost Comparison for the Transfer of Select Calcined Waste Canisters to the Monitored Geologic Repository at Yucca Mountain, NV

    International Nuclear Information System (INIS)

    Michael B. Heiser; Clark B. Millet

    2005-01-01

    This report performs a life-cycle cost comparison of three proposed canister designs for the shipment and disposition of Idaho National Laboratory high-level calcined waste currently in storage at the Idaho Nuclear Technology and Engineering Center to the proposed national monitored geologic repository at Yucca Mountain, Nevada. Concept A (2 x 10-ft) and Concept B (2 x 15-ft) canisters are comparable in design, but they differ in size and waste loading options and vary proportionally in weight. The Concept C (5.5 x 17.5-ft) canister (also called the ''super canister''), while similar in design to the other canisters, is considerably larger and heavier than Concept A and B canisters and has a greater wall thickness. This report includes estimating the unique life-cycle costs for the three canister designs. Unique life-cycle costs include elements such as canister purchase and filling at the Idaho Nuclear Technology and Engineering Center, cask preparation and roundtrip consignment costs, final disposition in the monitored geologic repository (including canister off-loading and placement in the final waste disposal package for disposition), and cask purchase. Packaging of the calcine ''as-is'' would save $2.9 to $3.9 billion over direct vitrification disposal in the proposed national monitored geologic repository at Yucca Mountain, Nevada. Using the larger Concept C canisters would use 0.75 mi less of tunnel space, cost $1.3 billion less than 10-ft canisters of Concept A, and would be complete in 6.2 years

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

    International Nuclear Information System (INIS)

    Choi, Youngchul; Yoon, Chanhoon; Lee, Jeaowan; Kim, Jinsup; Choi, Heuijoo

    2014-01-01

    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%

  4. Unreviewed safety question evaluation of 100 K West fuel canister gas and liquid sampling

    International Nuclear Information System (INIS)

    Alwardt, L.D.

    1995-01-01

    The purpose of this report is to provide the basis for answers to an Unreviewed Safety Question (USQ) safety evaluation for the gas and liquid sampling activities associated with the fuel characterization program at the 100 K West (KW) fuel storage basin. The scope of this safety evaluation is limited to the movement of canisters between the main storage basin, weasel pit, and south loadout pit transfer channel (also known as the decapping station); gas and liquid sampling of fuel canisters in the weasel pit; mobile laboratory preliminary sample analysis in or near the 105 KW basin building; and the placement of sample containers in an approved shipping container. It was concluded that the activities and potential accident consequences associated with the gas and liquid sampling of 100 KW fuel canisters are bounded by the current safety basis documents and do not constitute an Unreviewed Safety Question

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

    International Nuclear Information System (INIS)

    Ikonen, K.

    2005-06-01

    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

  6. Comparison of Tagging Technologies for Safeguards of Copper Canisters for Nuclear Spent Fuel.

    Science.gov (United States)

    Clementi, Chiara; Littmann, François; Capineri, Lorenzo

    2018-03-21

    Several countries are planning to store nuclear spent fuel in long term geological repositories, preserved by copper canisters with an iron insert. This new approach involves many challenging problems and one is to satisfy safeguards requirements: the Continuity of Knowledge (CoK) of the fuel must be kept from the encapsulation plant up to the final repository. To date, no measurement system has been suggested for a unique identification and authentication. Following the list of the most important safeguards, safety and security requirements for copper canisters identification and authentication, a review of conventional tagging technologies and measurement systems for nuclear items is reported in this paper. The aim of this study is to verify to what extent each technology could be potentially used for keeping the CoK of copper canisters. Several tagging methods are briefly described and compared, discussing advantages and disadvantages.

  7. Corrosion test plan to guide canister material selection and design for a tuff repository

    International Nuclear Information System (INIS)

    McCright, R.D.; van Konynenburg, R.A.; Ballou, L.B.

    1983-11-01

    Corrosion rates and the mode of corrosion attack form a most important basis for selection of canister materials and design of a nuclear waste package. Type 304L stainless steel was selected as the reference material for canister fabrication because of its generally excellent corrosion resistance in water, steam and air. However, 304L may be susceptible to localized and stress-assisted forms of corrosion under certain conditions. Alternative alloys are also investigated; these alloys were chosen because of their improved resistance to these forms of corrosion. The fabrication and welding processes, as well as the glass pouring operation for defense and commercial high-level wastes, may influence the susceptibility of the canister to localized and stress forms of corrosion. 12 references, 2 figures, 4 tables

  8. Iron-nickel alloys as canister material for radioactive waste disposal in underground repositories

    International Nuclear Information System (INIS)

    Apps, J.A.

    1982-01-01

    Canisters containing high-level radioactive waste must retain their integrity in an underground waste repository for at least one thousand years after burial (Nuclear Regulatory Commission, 1981). Since no direct means of verifying canister integrity is plausible over such a long period, indirect methods must be chosen. A persuasive approach is to examine the natural environment and find a suitable material which is thermodynamically compatible with the host rock under the environmental conditions with the host rock under the environmental conditions expected in a waste repository. Several candidates have been proposed, among them being iron-nickel alloys that are known to occur naturally in altered ultramafic rocks. The following review of stability relations among iron-nickel alloys below 350 0 C is the initial phase of a more detailed evaluation of these alloys as suitable canister materials

  9. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-30

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Jinsong [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.

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

    International Nuclear Information System (INIS)

    Jinsong Liu

    2006-04-01

    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 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 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 5 years

  13. Canisters for spent-fuel disposal: Design measures against localized corrosion

    International Nuclear Information System (INIS)

    Werme, L.O.; Oversby, V.M.

    2000-01-01

    Common to all high-level-waste disposal concepts is the encapsulation of the waste into metal canisters. The purpose of this waste canister is to isolate the radioactive waste from contact with its surroundings for a desired time period. The design service life ranges from hundreds to thousands of years depending on the disposal concept. After the isolation has been breached, other barriers in the disposal system will delay and attenuate the radioactive releases to acceptable levels. In a deep geologic repository, the waste package will be exposed to chemical attack and, depending on the type of repository, to mechanical stresses. Each of these factors will by itself or in combination inevitably lead to loss of confinement some time in the future. In the design of the Swedish waste canister, the corrosion resistance is provided by an outer shell of pure copper while an insert supplies the mechanical strength cast nodular iron. The close fit between the insert and the copper results in very small tensile stresses in the copper over very limited areas once the repository has been saturated. Measurements of stress corrosion crack growth show that annealed copper cannot maintain sufficiently high stress intensity factors for cracks to grow. For annealed copper, the stress intensity factor was limited to 25 MPa·m 1/2 because of extensive plastic deformation. For cold-worked copper, no crack growth could be observed for stress intensity factors 1/2 . Through the choices of canister material, canister, and repository design, and considering the expected chemical conditions, the risks for localized corrosion can be lowered to an acceptable level, if not eliminated altogether, and the releases from prematurely failed canisters can be kept well within acceptable dose levels

  14. Uncertainty analysis of multiple canister repository model by large-scale calculation

    International Nuclear Information System (INIS)

    Tsujimoto, K.; Okuda, H.; Ahn, J.

    2007-01-01

    A prototype uncertainty analysis has been made by using the multiple-canister radionuclide transport code, VR, for performance assessment for the high-level radioactive waste repository. Fractures in the host rock determine main conduit of groundwater, and thus significantly affect the magnitude of radionuclide release rates from the repository. In this study, the probability distribution function (PDF) for the number of connected canisters in the same fracture cluster that bears water flow has been determined in a Monte-Carlo fashion by running the FFDF code with assumed PDFs for fracture geometry. The uncertainty for the release rate of 237 Np from a hypothetical repository containing 100 canisters has been quantitatively evaluated by using the VR code with PDFs for the number of connected canisters and the near field rock porosity. The calculation results show that the mass transport is greatly affected by (1) the magnitude of the radionuclide source determined by the number of connected canisters by the fracture cluster, and (2) the canister concentration effect in the same fracture network. The results also show the two conflicting tendencies that the more fractures in the repository model space, the greater average value but the smaller uncertainty of the peak fractional release rate is. To perform a vast amount of calculation, we have utilized the Earth Simulator and SR8000. The multi-level hybrid programming method is applied in the optimization to exploit high performance of the Earth Simulator. The Latin Hypercube Sampling has been utilized to reduce the number of samplings in Monte-Carlo calculation. (authors)

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

    International Nuclear Information System (INIS)

    Bowyer, W.H.

    1997-02-01

    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

  16. Simulation of heat transfer around a canister placed horizontally in a drift

    International Nuclear Information System (INIS)

    Moujaes, S.; Bhargava, A.

    1994-01-01

    The Yucca Mountain Site Characterization Project is investigating the feasibility of locating a high level radioactive nuclear waste repository at Yucca Mountain, Nevada. The bore hole and the in-drift waste emplacement schemes are under evaluation as potential repository drift geometries. This paper presents a two-dimensional finite element thermal analysis of the nuclear waste canister placed horizontally in a drift. Simulation has been carried out for 1000 years and the peak temperatures at the walls of the drift and at the center of the canister have been determined. The effect of the three modes of heat transfer, conduction, natural convection and radiation, is also discussed

  17. 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

  18. Annular air space effects on nuclear waste canister temperatures in a deep geologic waste repository

    International Nuclear Information System (INIS)

    Lowry, W.E.; Cheung, H.; Davis, B.W.

    1980-01-01

    Air spaces in a deep geologic repository for nuclear high level waste will have an important effect on the long-term performance of the waste package. The important temperature effects of an annular air gap surrounding a high level waste canister are determined through 3-D numerical modeling. Air gap properties and parameters specifically analyzed and presented are the air gap size, surfaces emissivity, presence of a sleeve, and initial thermal power generation rate; particular emphasis was placed on determining the effect of these variables have on the canister surface temperature. Finally a discussion based on modeling results is presented which specifically relates the results to NRC regulatory considerations

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

    International Nuclear Information System (INIS)

    Chareyron, B.

    2010-01-01

    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)

  20. Description of a ceramic waste form and canister for Savannah River Plant high-level waste

    International Nuclear Information System (INIS)

    Butler, J.L.; Allender, J.S.; Gould, T.H. Jr.

    1982-04-01

    A canistered ceramic waste form for possible immobilization of Savannah River Plant (SRP) high-level radioactive wastes is described. Characteristics reported for the form include waste loading, chemical composition, heat content, isotope inventory, mechanical and thermal properties, and leach rates. A conceptual design of a potential production process for making this canistered form are also described. The ceramic form was selected in November 1981 as the primary alternative to the reference waste form, borosilicate glass, for making a final waste form decision for SRP waste by FY-1983. 11 tables

  1. 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

  2. Mechanical failure of SKB spent fuel disposal canisters. Mathematical modelling and scoping calculations

    International Nuclear Information System (INIS)

    Takase, Hiroyasu; Benbow, S.; Grindrod, P.

    1998-10-01

    According to the current design of SKB, a copper overpack with a cast steel inner component will be used as the disposal canister for spent nuclear fuel. A recent study considered the case of a breach in the copper overpack, through which groundwater could enter the canister. It has pointed out that hydrogen gas generated by an anaerobic corrosion could cushion the system and reduce or eventually stop further infiltration of water into the breached canister, and thence the spent fuel. One potential pitfall in this previous study lies in the fact that it did not consider any processes which might violate the following assumptions which are essential for the gas 'cushioning': 1. Hydrogen gas accumulated in the annular gap in the canister forms a free gas phase which is stable indefinitely into future; 2. Elevated gas pressure in the canister prevents further supply of groundwater except for diffusion of vapour. In the current study we developed a set of mathematical models for the above problem and applied it to carry out an independent assessment of the long-term behaviour of the canister. A key aim in this study was to clarify whether there are any alternative processes which may affect the result obtained by the previous study by violating one of the assumptions listed above. For this purpose, a scenario development exercise was conducted. The result supported the concept described in the previous study. One exception is that possible intrusion of bentonite gel followed by its desaturation could leave paths both for the gas and water simultaneously without forming a gas cushion. This is summarised in the first part of the report. In the second part, development of mathematical models and their applications are described. The key results are: 1. The model describing behaviour of gas and pore water in the canister and the buffer material reproduced the main results of the previous study; 2. The model considering intrusion of the bentonite gel pointed out possibility

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

    International Nuclear Information System (INIS)

    Bowyer, W.H.; Sjoblom, R.; Trolle, M.

    1996-01-01

    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

  4. Stress redistribution and void growth in butt-welded canisters for spent nuclear fuel

    International Nuclear Information System (INIS)

    Josefson, B.L.; Karlsson, L.; Haeggblad, H.Aa.

    1993-02-01

    The stress-redistribution in Cu-Fe canisters for spent nuclear fuel during waiting for deposition and after final deposition is calculated numerically. The constitutive equation modelling creep deformation during this time period employs values on materials parameters determined within the SKB-project on 'mechanical integrity of canisters for spent nuclear fuel'. The welding residual stresses are redistributed without lowering maximum values during the waiting period, a very low amount of void growth is predicted for this type of copper during the deposition period. This leads to an estimated very large rupture time

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

    International Nuclear Information System (INIS)

    Andrews, R.E.

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    Sande, W.E.; Bjorklund, W.J.; Brooks, N.A.

    1977-04-01

    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

  8. Spent Nuclear Fuel project stage and store K basin SNF in canister storage building functions and requirements. Revision 1

    International Nuclear Information System (INIS)

    Womack, J.C.

    1995-01-01

    This document establishes the functions and requirements baseline for the implementation of the Canister Storage Building Subproject. The mission allocated to the Canister Storage Building Subproject is to provide safe, environmentally sound staging and storage of K Basin SNF until a decision on the final disposition is reached and implemented

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

    International Nuclear Information System (INIS)

    Yi, Yun-Bo; Amme, Robert C.; Shayer, Zeev

    2008-01-01

    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

  10. A technical basis to relax the dew point specification for the environment in the vapor space in DWPF canisters

    International Nuclear Information System (INIS)

    Louthan, M.R. Jr.

    1995-05-01

    This memorandum establishes the technical basis to conclude that relaxing, from 0 C to 20 C, the dew point specification for the atmosphere in the vapor space (free volume) of a DWPF canister will not provide an environment that will cause significant amounts of corrosion induced degradation of the canister wall. The conclusion is based on engineering analysis, experience and review of the corrosion literature. The basic assumptions underlying the conclusion are: (1) the canister was fabricated from Type 304L stainless steel; (2) the corrosion behavior of the canister material, including base metal, fusion zones and heat effected zones, is typified by literature data for, and industrial experience with, 300 series austenitic stainless steels; and (3) the glass-metal crevices created during the pouring operation will not alter the basic corrosion resistance of the steel although such crevices might serve as sites for the initiation of minor amounts of corrosion on the canister wall

  11. 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.

  12. Status of Closure Welding Technology of Canister for Transportation and Storage of High Level Radioactive Material and Waste

    International Nuclear Information System (INIS)

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

    2010-10-01

    Closure seal welding is one of the key technologies in fabricating and handling the canister which is used for transportation and storage of high radioactive material and waste. Simple industrial fabrication processes are used before filling the radioactive waste into the canister. But, automatic and remote processes should be used after filling the radioactive material because the thickness of canister is not sufficient to shield the high radiation from filled material or waste. In order to simplify the welding process the closure structure of canister and the sealing method are investigated and developed properly. Two types of radioactive materials such as vitrified waste and compacted solid waste are produced in nuclear industry. Because the filling method of two types of waste is different, the shapes of closure and opening of canister and welding method is also different. The canister shape and sealing method should be standardized to standardize the handling facilities and inspection process such as leak test after closure welding. In order to improve the productivity of disposal and compatibility of the canister, the structure and shape of canister should be standardized considering the type of waste. Two kind of welding process such as arc welding and resistance welding are reported and used in the field. In the arc welding process GTAW and PAW are considered proper processes for closure welding. The closure seal welding process can be selected by considering material of canister, thickness of body, productivity, and applicable codes and rules. Because the storage time of nuclear waste in canister is very long, at least 20 years, the long-time corrosion at the weld should be estimated including mechanical integrity. Recently, the mitigation of residual stress around weld region, which causes stress corrosion cracking, is also interesting research issue

  13. Experimental assessment of the thermal performance of storage canister/holding fixture configurations for the Los Alamos Nuclear Materials Storage Facility

    International Nuclear Information System (INIS)

    Bernardin, J.D.; Naffziger, D.C.; Gregory, W.S.

    1997-11-01

    This report presents experimental results on the thermal performance of various nested canister configurations and canister holding fixtures to be used in the Los Alamos Nuclear Materials Storage Facility. The experiment consisted of placing a heated aluminum billet (to represent heat-generating nuclear material) inside curved- and flat-bottom canisters with and without holding plate fixtures and/or extended fin surfaces. Surface temperatures were measured at several locations on the aluminum billet, inner and outer canisters, and the holding plate fixture to assess the effectiveness of the various configurations in removing and distributing the heat from the aluminum billet. Results indicated that the curved-bottom canisters, with or without holding fixtures, were extremely ineffective in extracting heat from the aluminum billet. The larger thermal contact area provided by the flat-bottom canisters compared with the curved-bottom design, greatly enhanced the heat removal process and lowered the temperature of the aluminum billet considerably. The addition of the fixture plates to the flat-bottom canister geometry greatly enhances the heat removal rates and lowers the canister operating temperatures considerably. The addition of the fixture plates to the flat-bottom canister geometry greatly enhances the heat removal rates and lowers the canister operating temperatures considerably. Finally, the addition of extended fin surfaces to the outer flat-bottom canister positioned on a fixture plate, reduced the canister temperatures still further

  14. Quality Assurance Program Plan for Project W-379: Spent Nuclear Fuels Canister Storage Building Projec

    International Nuclear Information System (INIS)

    Duncan, D.W.

    1995-01-01

    This document describes the Quality Assurance Program Plan (QAPP) for the Spent Nuclear Fuels (SNF) Canister Storage Building (CSB) Project. The purpose of this QAPP is to control project activities ensuring achievement of the project mission in a safe, consistent and reliable manner

  15. 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.

  16. Acceptance Test Report for the high pressure water jet system canister cleaning fixture

    Energy Technology Data Exchange (ETDEWEB)

    Burdin, J.R.

    1995-10-25

    This Acceptance Test confirmed the test results and recommendations, documented in WHC-SD-SNF-DTR-001, Rev. 0 Development Test Report for the High Pressure Water Jet System Nozzles, for decontaminating empty fuel canisters in KE-Basin. Optimum water pressure, water flow rate, nozzle size and overall configuration were tested

  17. Spent Nuclear Fuel (SNF) Project Canister Storage Building (CSB) Process Flow Diagram Mass Balance Calculations

    International Nuclear Information System (INIS)

    KLEM, M.J.

    2000-01-01

    The purpose of these calculations is to develop the material balances for documentation of the Canister Storage Building (CSB) Process Flow Diagram (PFD) and future reference. The attached mass balances were prepared to support revision two of the PFD for the CSB. The calculations refer to diagram H-2-825869

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

    International Nuclear Information System (INIS)

    KLEM, M.J.

    2000-01-01

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

    International Nuclear Information System (INIS)

    Werme, Lars; Lilja, Christina

    2010-12-01

    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/

  20. The Characteristics of Welding Joint on Stainless Steel as a Candidate of High Level Waste Canister

    International Nuclear Information System (INIS)

    Aisyah; Herlan-Martono

    2000-01-01

    High level waste is the waste generated from reprocessing of the spent fuels. This type of waste is vitrified with borosilicate glass to become waste-glass. This waste glass is contained in a canister made of austenitic stainless steel. The canister material is subjected to be welded during fabrication and utilization. The character of the welding joint that is the function of the electrical current used in the welding process have been studied. The strength of the joint is tested mechanically i.e.: the tensile strength and hardness test. The result shows that the higher the current used in welding process, the better the strength of the joint and as well the tensile strength. The optimum current is 110 A. From the hardness test, it was figured that the length of the HAZ area is 14 mm. The material in HAZ area is the hardest compared to the others, it is due to the appearance of the chrome-carbide. The welding of the canister with such a condition, during fabrication as well as during the utilization of the canister for the container of the high level waste with the PWHT process gives better result. (author)

  1. Potential Multi-Canister Overpack (MCO) Cask Drop in the K West Basin South Loadout Pit

    International Nuclear Information System (INIS)

    POWERS, T.B.

    1999-01-01

    This calculation note documents the probabilistic calculation of a potential drop of a multi-canister overpack (MCO) cask or MCO cask and immersion pail at the K West Basin south loadout pit. The calculations are in support of the cask loading system (CLS) subproject alignment of CLS equipment in the K West Basin south loadout pit

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-12

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

  3. Canister Design for Deep Borehole Disposal of Nuclear Waste (CD-ROM)

    National Research Council Canada - National Science Library

    Hoag, Christopher I

    2006-01-01

    ...: 1 CD-ROM; 4 3/4 in.; 28.7 MB. ABSTRACT: 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...

  4. 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.

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

    International Nuclear Information System (INIS)

    Hamilton, L.

    2001-01-01

    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

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

    International Nuclear Information System (INIS)

    Hamilton, L.

    2001-01-01

    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

  7. Acceptance Test Report for the high pressure water jet system canister cleaning fixture

    International Nuclear Information System (INIS)

    Burdin, J.R.

    1995-01-01

    This Acceptance Test confirmed the test results and recommendations, documented in WHC-SD-SNF-DTR-001, Rev. 0 Development Test Report for the High Pressure Water Jet System Nozzles, for decontaminating empty fuel canisters in KE-Basin. Optimum water pressure, water flow rate, nozzle size and overall configuration were tested

  8. SPENT NUCLEAR FUEL NUMBER DENSITIES FOR MULTI-PURPOSE CANISTER CRITICALITY CALCULATIONS

    International Nuclear Information System (INIS)

    D. A. Thomas

    1996-01-01

    The purpose of this analysis is to calculate the number densities for spent nuclear fuel (SNF) to be used in criticality evaluations of the Multi-Purpose Canister (MPC) waste packages. The objective of this analysis is to provide material number density information which will be referenced by future MPC criticality design analyses, such as for those supporting the Conceptual Design Report

  9. 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/

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

    International Nuclear Information System (INIS)

    Andersson, Claes-Goeran

    2002-04-01

    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

  11. Simulation of residual stresses and deformations in electron beam-welded copper canisters

    International Nuclear Information System (INIS)

    Aronen, A.; Leikko, J.; Taskinen, P.; Karvinen, R.

    2013-07-01

    This report presents the modelling of residual stresses and deformations of an EB-welded copper canister. Two different mock-up lengths are modelled with the Abaqus FEA program, and the similarity of those results is studied. Canister mock-ups of 450 mm and 915 mm were chosen for the test cases. The heat treatment results presented in Taskinen 2009 are used as input data for the mechanical model. For the mechanical analysis some simplifications were made to the model. The contact surface between pipe and lid is assumed to be tied and support from the bottom surface is provided with four support points. Results show that, due to the similarity of 450 mm and 915 mm canisters, the short mock-up can be used to predict the stresses and deformation on a full-length canister (5000 mm). The similarity of the temperature fields has already been shown in the previous reports (Taskinen 2009). The main result in the deformation is the shape of the canister in the residual state. The top of the canister tries to shrink, resulting in the lid buckling inwards. The deformation of the lid of the canister is about 2.2 mm at the centre of the lid. The main results in the stresses are the stress level on the surface, the deviation of stresses over the circle and the stresses near the welding. On the surface there are areas where the circumferential stress is at tension. However, radial and axial stresses are usually in compression on the surface. The deviation of the stress level over the circle is quite small, except in the overlap area and near it. The residual stresses from 0 deg C to 45 deg C change remarkably, but over the rest of the area the stresses are more constant. Near the welding the stresses on the top surface are in compression, but in the centre of the welding the stresses are in tension. In the modelling, the possibility of calculating a mechanical model with the contact surface between pipe and lid, so that they could be separated during the welding, was also tested

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

    International Nuclear Information System (INIS)

    King, F.

    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 O/Cu(OH) 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 smaller discontinuities will die once they reach this maximum size. Death of propagating pits will be compounded by the decrease in oxygen flux to the canister as the repository environment becomes anoxic. Surface discontinuities could impact the SCC behaviour either through their effect on the local environment or via stress concentration or intensification. There is no evidence that surface discontinuities will affect the initiation of SCC by ennoblement of the corrosion potential or the formation of locally aggressive conditions. Stress concentration at pits could lead to crack initiation under some circumstances, but the stress intensity factor for the resultant cracks, or for pre-existing crack-like discontinuities, will be smaller than the

  13. Probabilistic analysis and material characterisation of canister insert for spent nuclear fuel. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Claes-Goeran [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden); Andersson, Mats; Erixon, Bo [AaF Industriteknik, Stockholm (Sweden); Bjoerkegren, Lars-Erik [Swedish Foundry Association, Stockholm (Sweden); Dillstroem, Peter [DNV Technology, Stockholm (Sweden); Minnebo, Philip

    2005-11-15

    The KBS-3 canister for geological disposal of spent nuclear fuel in Sweden consists of a ductile cast iron insert and a copper shielding. The canister should inhibit release of radionuclides for at least 100,000 years. The copper protects the canister from corrosion whereas the ductile cast iron insert provides the mechanical strength. In the repository the hydrostatic pressure from the groundwater and the swelling pressure from the surrounding bentonite, which in total results in a maximum pressure of 14 MPa, will load the canisters in compression. During the extreme time scales, ice ages are expected with a maximum ice thickness of 3,000 m resulting in an additional pressure of 30 MPa. The maximum design pressure for the KBS-3 canisters has therefore been set to be 44 MPa. A relatively large number of canisters have been manufactured as part of SKB's development programme. To verify the strength of the canisters at this stage of development SKB initiated a project in cooperation with the European commissions Joint Research Centre (JRC), Institute of Energy in Petten in the Netherlands, together with a number of other partners. Three inserts manufactured by different Swedish foundries 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 together with the results from stress and strain finite element analysis were subsequently used in probabilistic analysis to determine the probability for plastic collapse caused by high pressure or fracture by crack growth in regions with tensile stresses. The main conclusions from the probabilistic analysis are: 1. At the design pressure of 44 MPa, the probability of failure is insignificant ({approx}2x10{sup -9}). This is the case even though several conservative assumptions have been made. 2. The stresses in the insert caused by the outer pressure are mainly compressive. The regions with tensile

  14. Simulation of residual stresses and deformations in electron beam-welded copper canisters

    Energy Technology Data Exchange (ETDEWEB)

    Aronen, A.; Leikko, J.; Taskinen, P.; Karvinen, R. [Tampere Univ. of Technology (Finland)

    2013-07-15

    This report presents the modelling of residual stresses and deformations of an EB-welded copper canister. Two different mock-up lengths are modelled with the Abaqus FEA program, and the similarity of those results is studied. Canister mock-ups of 450 mm and 915 mm were chosen for the test cases. The heat treatment results presented in Taskinen 2009 are used as input data for the mechanical model. For the mechanical analysis some simplifications were made to the model. The contact surface between pipe and lid is assumed to be tied and support from the bottom surface is provided with four support points. Results show that, due to the similarity of 450 mm and 915 mm canisters, the short mock-up can be used to predict the stresses and deformation on a full-length canister (5000 mm). The similarity of the temperature fields has already been shown in the previous reports (Taskinen 2009). The main result in the deformation is the shape of the canister in the residual state. The top of the canister tries to shrink, resulting in the lid buckling inwards. The deformation of the lid of the canister is about 2.2 mm at the centre of the lid. The main results in the stresses are the stress level on the surface, the deviation of stresses over the circle and the stresses near the welding. On the surface there are areas where the circumferential stress is at tension. However, radial and axial stresses are usually in compression on the surface. The deviation of the stress level over the circle is quite small, except in the overlap area and near it. The residual stresses from 0 deg C to 45 deg C change remarkably, but over the rest of the area the stresses are more constant. Near the welding the stresses on the top surface are in compression, but in the centre of the welding the stresses are in tension. In the modelling, the possibility of calculating a mechanical model with the contact surface between pipe and lid, so that they could be separated during the welding, was also tested

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

    International Nuclear Information System (INIS)

    Smart, Nick; Reddy, Bharti; Rance, Andy

    2012-06-01

    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

  16. 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

  17. 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

  18. Technical note. A review of the mechanical integrity of the canister

    International Nuclear Information System (INIS)

    Segle, Peter

    2012-01-01

    Background: The Swedish Radiation Safety Authority (SSM) reviews the Swedish Nuclear Fuel Company's (SKB) applications under the Act on Nuclear Activities (SFS 1984:3) for the construction and operation of a repository for spent nuclear fuel and for an encapsulation facility. As part of the review, SSM commissions consultants to carry out work in order to obtain information on specific issues. The results from the consultants' tasks are reported in SSM's Technical Note series. Objectives of the project: This project is part of SSM:s review of SKB:s license application for final disposal of spent nuclear fuel. The assignment concerns a review of the mechanical integrity of the canister. Summary by the author: An introductory review of SR-Site has been conducted with respect to the mechanical integrity of the canister. The review is focused on the copper canister and the nodular cast iron insert. Review results show that a number of loads and loading scenarios for the copper canister has not been analysed by SKB. The importance of sufficient creep ductility of the copper material and sufficient ductility and fracture toughness of the nodular cast iron material is pointed out in the review. A sensitivity study is suggested where the impact of these properties on the mechanical integrity of the canister is investigated. It is also suggested that potential damage mechanisms influencing these properties are further investigated. SKB's modelling of creep elongation at rupture under repository conditions is questioned. Needs for complementary information from SKB for the main review of SR-Site is listed. A list of review topics for SSM is also suggested

  19. Model shear tests of canisters with smectite clay envelopes in deposition holes

    International Nuclear Information System (INIS)

    Boergesson, L.

    1986-01-01

    The consequences of rock displacement across a deposition hole has been investigated by some model tests. The model was scaled 1:10 to a real deposition hole. It was filled with a canister made of solid copper surrounded by highly compacted water saturated MX-80 bentonite. Before shear the swelling pressure was measured by six transducers in order to follow the water uptake process. During shear, pressure, strain, force and deformation were measured in altogether 18 points. The shearing was made at different rates in the various tests. An extensive sampling after shear was made through which the density, water content, degree of saturation, homogenization and the effect of shear on the bentonite and canister could be studied. One important conlusion from these tests was that the rate dependence is about 10% increased shear resistance per decade increased rate of shear. This resulted also in a very clear increase in strain in the canister with increased rate. The results also showed that the saturated bentonite has excellent stress distributing properties and that there is no risk of destroying the canister if the rock displacement is smaller than the thickness of the bentonite cover. The high density of the clay makes the bentonite produce such a high swelling pressure that the material will be very stiff. In the case of a larger shear deformation corresponding to ≅ 50% of the bentonite thickness the result will be a rather large deformation of the canister. A lower density would be preferable if it can be accepted with respect to other required isolating properties. The results also showed that three-dimensional FEM calculation using non-linear material properties is necessary to simulate the shear process. The rate dependence may be taken into account by adapting the properties to the actual rate of shear but might in a later stage be included in the model by giving the material viscous properties. (orig./HP)

  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. Hanford Waste Vitrification Plant: Preliminary description of waste form and canister

    International Nuclear Information System (INIS)

    Mitchell, D.E.

    1986-01-01

    In July 1985, the US Department of Energy's Office of Civilian Radioactive Waste Management established the Waste Acceptance Process as the means by which defense high-level waste producers, such as the Hanford Waste Vitrification Plant, will develop waste acceptance requirements with the candidate geologic repositories. A complete description of the Waste Acceptance Process is contained in the Preliminary Hanford Waste Vitrification Plant Waste Form Qualification Plan. The Waste Acceptance Process defines three documents that high-level waste producers must prepare as a part of the process of assuming that a high-level waste product will be acceptable for disposal in a geologic repository. These documents are the Description of Waste Form and Canister, Waste Compliance Plan, and Waste Qualification Report. This document is the Hanford Waste Vitrification Plant Preliminary Description of Waste Form and Canister for disposal of Neutralized Current Acid Waste. The Waste Acceptance Specifications for the Hanford Waste Vitrification Plant have not yet been developed, therefore, this document has been structured to corresponds to the Waste Acceptance Preliminary Specifications for the Defense Waste Processing Facility High-Level Waste Form. Not all of the information required by these specifications is appropriate for inclusion in this Preliminary Description of Waste Form and Canister. Rather, this description is limited to information that describes the physical and chemical characteristics of the expected high-level waste form. The content of the document covers three major areas: waste form characteristics, canister characteristics, and canistered waste form characteristics. This information will be used by the candidate geologic repository projects as the basis for preliminary repository design activities and waste form testing. Periodic revisions are expected as the Waste Acceptance Process progresses

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

    International Nuclear Information System (INIS)

    Ryu, Yong Ho

    1992-02-01

    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

  3. Deep Borehole Disposal Concept: Development of Universal Canister Concept of Operations

    Energy Technology Data Exchange (ETDEWEB)

    Rigali, Mark J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Applied Systems Analysis and Research; Price, Laura L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Applied Systems Analysis and Research

    2016-08-01

    This report documents key elements of the conceptual design for deep borehole disposal of radioactive waste to support the development of a universal canister concept of operations. A universal canister is a canister that is designed to be able to store, transport, and dispose of radioactive waste without the canister having to be reopened to treat or repackage the waste. This report focuses on the conceptual design for disposal of radioactive waste contained in a universal canister in a deep borehole. The general deep borehole disposal concept consists of drilling a borehole into crystalline basement rock to a depth of about 5 km, emplacing WPs in the lower 2 km of the borehole, and sealing and plugging the upper 3 km. Research and development programs for deep borehole disposal have been ongoing for several years in the United States and the United Kingdom; these studies have shown that deep borehole disposal of radioactive waste could be safe, cost effective, and technically feasible. The design concepts described in this report are workable solutions based on expert judgment, and are intended to guide follow-on design activities. Both preclosure and postclosure safety were considered in the development of the reference design concept. The requirements and assumptions that form the basis for the deep borehole disposal concept include WP performance requirements, radiological protection requirements, surface handling and transport requirements, and emplacement requirements. The key features of the reference disposal concept include borehole drilling and construction concepts, WP designs, and waste handling and emplacement concepts. These features are supported by engineering analyses.

  4. 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

  5. Coalbed gas desorption in canisters: Consumption of trapped atmospheric oxygen and implications for measured gas quality

    International Nuclear Information System (INIS)

    Jin, Hui; Schimmelmann, Arndt; Mastalerz, Maria; Pope, James; Moore, Tim A.

    2010-01-01

    Desorption canisters are routinely employed to quantify coalbed gas contents in coals. If purging with inert gas or water flooding is not used, entrapment of air with ∝ 78.08 vol.% nitrogen (N 2 ) in canisters during the loading of coal results in contamination by air and subsequent overestimates of N 2 in desorbed coalbed gas. Pure coalbed gas does not contain any elemental oxygen (O 2 ), whereas air contamination originally includes ∝ 20.95 vol.% O 2 and has a N 2 /O 2 volume ratio of ∝ 3.73. A correction for atmospheric N 2 is often attempted by quantifying O 2 in headspace gas and then proportionally subtracting atmospheric N 2 . However, this study shows that O 2 is not a conservative proxy for air contamination in desorption canisters. Time-series of gas chromatographic (GC) compositional data from several desorption experiments using high volatile bituminous coals from the Illinois Basin and a New Zealand subbituminous coal document that atmospheric O 2 was rapidly consumed, especially during the first 24 h. After about 2 weeks of desorption, the concentration of O 2 declined to near or below GC detection limits. Irreversible loss of O 2 in desorption canisters is caused by biological, chemical, and physical mechanisms. The use of O 2 as a proxy for air contamination is justified only immediately after loading of desorption canisters, but such rapid measurements preclude meaningful assessment of coalbed gas concentrations. With increasing time and progressive loss of O 2 , the use of O 2 content as a proxy for atmospheric N 2 results in overestimates of N 2 in desorbed coalbed gas. The indicated errors for nitrogen often range in hundreds of %. Such large analytical errors have a profound influence on market choices for CBM gas. An erroneously calculated N 2 content in CBM would not meet specifications for most pipeline-quality gas. (author)

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

    International Nuclear Information System (INIS)

    Geuther, J.; Conrad, E.A.; Rhoadarmer, D.

    2009-01-01

    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

  7. Analytical Evaluation of Preliminary Drop Tests Performed to Develop a Robust Design for the Standardized DOE Spent Nuclear Fuel Canister

    International Nuclear Information System (INIS)

    Ware, A.G.; Morton, D.K.; Smith, N.L.; Snow, S.D.; Rahl, T.E.

    1999-01-01

    The Department of Energy (DOE) has developed a design concept for a set of standard canisters for the handling, interim storage, transportation, and disposal in the national repository, of DOE spent nuclear fuel (SNF). The standardized DOE SNF canister has to be capable of handling virtually all of the DOE SNF in a variety of potential storage and transportation systems. It must also be acceptable to the repository, based on current and anticipated future requirements. This expected usage mandates a robust design. The canister design has four unique geometries, with lengths of approximately 10 feet or 15 feet, and an outside nominal diameter of 18 inches or 24 inches. The canister has been developed to withstand a drop from 30 feet onto a rigid (flat) surface, sustaining only minor damage - but no rupture - to the pressure (containment) boundary. The majority of the end drop-induced damage is confined to the skirt and lifting/stiffening ring components, which can be removed if de sired after an accidental drop. A canister, with its skirt and stiffening ring removed after an accidental drop, can continue to be used in service with appropriate operational steps being taken. Features of the design concept have been proven through drop testing and finite element analyses of smaller test specimens. Finite element analyses also validated the canister design for drops onto a rigid (flat) surface for a variety of canister orientations at impact, from vertical to 45 degrees off vertical. Actual 30-foot drop testing has also been performed to verify the final design, though limited to just two full-scale test canister drops. In each case, the analytical models accurately predicted the canister response

  8. Predicted peak temperature-rises around a high-level radioactive waste canister emplaced in the deep ocean bed

    International Nuclear Information System (INIS)

    Kipp, K.L.

    1978-06-01

    A simple mathematical model of heat conduction was used to evaluate the peak temperature-rise along the wall of a canister of high-level radioactive waste buried in deep ocean sediment. Three different amounts of vitrified waste, corresponding to standard Harvest, large Harvest, and AVM canisters, and three different waste loadings were studied. Peak temperature-rise was computed for the nine cases as a function of canister geometry and storage time between reprocessing and burial. Lower waste loadings or longer storage times than initially envisaged are necessary to prevent the peak temperature-rise from exceeding 200 0 C. The use of longer, thinner cylinders only modestly reduces the storage time for a given peak temperature. Effects of stacking of waste canisters and of close-packing were also studied. (author)

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

    International Nuclear Information System (INIS)

    Bowyer, W.H.

    1999-05-01

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

  10. Structural performance of a multipurpose canister shell for HLNW under normal handling conditions

    International Nuclear Information System (INIS)

    Ladkany, S.G.; Rajagopalan, R.

    1994-01-01

    A Multipurpose Canister (MPC) is analyzed for critical stresses that occur during normal handling conditions and accidental scenarios. Linear and Non-linear Finite Element Analysis is performed and the stresses at various critical locations in the MPC and its weldments are studied extensively. Progressive failure analysis of the MPC's groove and fillet welds, is presented. The structural response of the MPC to dynamic lifting loads, to loads resulting from an accidental slippage of a crane cable carrying the MPC, and from the impact between two canisters, is evaluated. Nonlinear structural analysis is used in the evaluation of the local buckling and the ultimate failure phenomena in the shell when the steel is in the strain hardening state during impact. Results make a case for increasing the thickness of the shell and all the welds

  11. The suitability of Titanium as a corrosion resistant canister for nuclear waste

    International Nuclear Information System (INIS)

    Henriksson, S.; Pettersson, K.

    1977-08-01

    A literature study and inventory of experience has been carried out, aimed at assessing the possibilities of unalloyed and Pd-alloyed titanium withstanding corrosion for 1000 - 10000 years in contact with Baltic Sea water at 100 percentC and pH 4 - 10. The following assesment can be made: -1. Pitting, crevice corrosion, stress corrosion cracking and corrosion fatigue constitute no problem if the canister is made of unalloyed titanium corresponding to ASTM Grade 1. 2. Linear extrapolation of reported corrosion rates for oxidation and general corrosion gives a life of between 1000 and 10000 years for a 5 mm thick canister. 3. Hydrogen embrittlement resulting from hydrogen pick-up from the deposition environment should not occur. Delayed failure caused by a redistribution of the hydrogen initially present in the titanium can be avoided if its concentration is maximized to 20 ppm. (author)

  12. FY17 Status Report: Research on Stress Corrosion Cracking of SNF Interim Storage Canisters.

    Energy Technology Data Exchange (ETDEWEB)

    Schindelholz, Eric John [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Alexander, Christopher L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-09-01

    This progress report describes work done in FY17 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. Work in FY17 refined our understanding of the chemical and physical environment on canister surfaces, and evaluated the relationship between chemical and physical environment and the form and extent of corrosion that occurs. The SNL corrosion work focused predominantly on pitting corrosion, a necessary precursor for SCC, and process of pit-to-crack transition; it has been carried out in collaboration with university partners. SNL is collaborating with several university partners to investigate SCC crack growth experimentally, providing guidance for design and interpretation of experiments.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-05-15

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

  14. PFPF canister counter for foreign plutonium (PCAS-3) hardware operations and procedures manual

    International Nuclear Information System (INIS)

    Menlove, H.O.; Baca, J.; Kroncke, K.E.; Miller, M.C.; Takahashi, S.; Seki, S.; Inose, S.; Yamamoto, T.

    1993-01-01

    A neutron coincidence counter has been designed for the measurement of plutonium powder contained in tall storage canisters. The counter was designed for installation in the Plutonium Fuel Production Facility fabrication plant. Each canister contains from one to five cans of PuO 2 . The neutron counter measures the spontaneous-fission rate from the plutonium and, when this is combined with the plutonium isotopic ratios, the plutonium mass is determined. The system can accommodate plutonium loadings up to 12 kg, with 10 kg being a typical loading. Software has been developed to permit the continuous operation of the system in an unattended mode. Authentication techniques have been developed for the system. This manual describes the system and its operation and gives performance and calibration parameters for typical applications

  15. Application of plutonium inventory measurement system (PIMS) and temporary canister verification system (TCVS) at RRP

    International Nuclear Information System (INIS)

    Noguchi, Yoshihiko; Nakamura, Hironobu; Adachi, Hideto; Iwamoto, Tomonori

    2004-01-01

    In U-Pu co-denitration area at Rokkasho Reprocessing Plant (RRP), Plutonium Inventory Measurement System (PIMS) and Temporary Canister Verification System (TCVS) are installed to provide efficient and effective safeguards. PIMS measures Pu quantity inside pipes and vessels installed in glove boxes by total neutron counting method. PIMS consists of total 142 neutron detector attached on the wall and top of glove boxes and neutron count rates of each detectors are related to each other to calculate Pu quantity of each process areas. In this moment, inactive calibration using Cf-source was completed. On the other hand, TCVS measures Pu quantity of canisters inside temporary storage by coincidence counting method and it will be installed before the active test. These systems have monitoring function as additional measures. This paper describes specification, performance and measurement principles of PIMS and TCVS. (author)

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

    International Nuclear Information System (INIS)

    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

  17. Mitigation of sliding motion of a cask-canister by fluid-structure interaction in an annular region - 59208

    International Nuclear Information System (INIS)

    Ito, Tomohiro; Fujiwara, Yoshihiro; Shintani, Atsuhiko; Nakagaw, Chihiro; Furuta, Kazuhisa

    2012-01-01

    The cask-canister system is a coaxial circular cylindrical structure in which several spent fuels are installed. This system is a free-standing structure thus, it is very important to reduce sliding motion for very large seismic excitations. In this study, we propose a mitigation method for sliding motion. Water is installed in an annular region between a cask and a canister. The equations of motion are derived taking fluid-structure interaction into consideration for nonlinear sliding motion analyses. Based on these equations, mitigation effects of sliding motions are studied analytically. Furthermore, a fundamental test model of a cask-canister system is fabricated and shaking table tests are conducted. From the analytical and test results, sliding motion mitigation effects are investigated. In this paper, the sliding motion of the cask-canister system subjected to a horizontal base excitation is studied and the effectiveness of water filled in the annular region between the cask and the canister is evaluated. This water brings inertia force coupling effect which is proportional to acceleration of the cask and the canister. Therefore, due to this fluid coupling, the cask and canister system couples through 3 types of forces, i.e., spring force, damping force and inertia force of the liquid. Equations of motion for the sliding motion are derived based on the fluid-structure coupling effects formulated by Fritz. Based on these equations of motion, nonlinear sliding motion of the cask-canister system is analyzed and the sliding suppression effects are investigated numerically. Furthermore, a fundamental test model of a cask-canister system is fabricated and the shaking table tests are conducted. From these analytical and test results, the sliding motion suppression effects due to fluid-structure coupling effects are investigated. As a result, it is confirmed that the inertia coupling effects due to water filled in the annular region are relatively large, and the

  18. 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

  19. Feasibility of using a high-level waste canister as an engineered barrier in disposal

    International Nuclear Information System (INIS)

    Slate, S.C.; Pitman, S.G.; Nesbitt, J.F.; Partain, W.L.

    1982-08-01

    The objective of this report is to evaluate the feasibility of designing a process canister that could also serve as a barrier canister. To do this a general set of performance criteria is assumed and several metal alloys having a high probability of demonstrating high corrosion resistance under repository conditions are evaluated in a qualitative design assessment. This assessment encompasses canister manufacture, the glass-filling process, interim storage, transportation, and to a limited extent, disposal in a repository. A series of scoping tests were carried out on two titanium alloys and Inconel 625 to determine if the high temperature inherent in the glass-fill processing would seriously affect either the strength or corrosion resistance of these metals. This is a process-related concern unique to the barrier canister concept. The material properties were affected by the heat treatments which simulated both the joule-heated glass melter process (titanium alloys and Inconel 625) and the in-can melter (ICM) process (Inconel 625). However, changes in the material properties were generally within 20% of the original specimens. Accelerated corrosion testing of the heat treated coupons in a highly oxygenated brine showed basic corrosion resistance of titanium grade 12 and Inconel 625 to compare favorably with that of the untreated coupons. The titanium grade 2 coupons experienced severe corrosion pitting. These corrosion tests were of a scoping nature and suitable primarily for the detection of gross sensitivity to the heat treatment inherent in the glass-fill process. They are only suggstive of repository performance since the tests do not adequately model the wide range of repository conditions that could conceivably occur

  20. A charcoal canister survey of radon emanation at the rehabilitated uranium mine site at Nabarlek

    International Nuclear Information System (INIS)

    Storm, J.R.; Patterson, J.R.

    1999-01-01

    This paper describes a recent survey of radon emanation measurements from the rehabilitated Nabarlek mine site. It was mined out in 1979, decommissioned in 1995 and provided a good test bed for assessment of rehabilitation in terms of radon flux attenuation. Measurements have been made with charcoal canisters. Studies to measure the radon-220 flux by observing Tl-208 progeny of thoron the effectiveness of trial covers and meteorological considerations will be reported

  1. TOUGH - a numerical model for nonisothermal unsaturated flow to study waste canister heating effects

    International Nuclear Information System (INIS)

    Pruess, K.; Wang, J.S.Y.

    1984-01-01

    The physical processes modeled and the mathematical and numerical methods employed in a simulator for non-isothermal flow of water, vapor, and air in permeable media are briefly summarized. The simulator has been applied to study thermohydrological conditions in the near vicinity of high-level nuclear waste packages emplaced in unsaturated rocks. The studies reported here specifically address the question whether or not the waste canister environment will dry up in the thermal phase. 13 references, 8 figures, 2 tables

  2. Monitored Retrievable Storage/Multi-Purpose Canister analysis: Simulation and economics of automation

    International Nuclear Information System (INIS)

    Bennett, P.C.; Stringer, J.B.

    1994-01-01

    Robotic automation is examined as a possible alternative to manual spent nuclear fuel, transport cask and Multi-Purpose canister (MPC) handling at a Monitored Retrievable Storage (MRS) facility. Automation of key operational aspects for the MRS/MPC system are analyzed to determine equipment requirements, through-put times and equipment costs is described. The economic and radiation dose impacts resulting from this automation are compared to manual handling methods

  3. A charcoal canister survey of radon emanation at the rehabilitated uranium mine site at Nabarlek

    Energy Technology Data Exchange (ETDEWEB)

    Storm, J R; Patterson, J R [University of Adelaide, Adelaide, SA (Australia). Department of Physics and Mathematical Physics

    1999-07-01

    This paper describes a recent survey of radon emanation measurements from the rehabilitated Nabarlek mine site. It was mined out in 1979, decommissioned in 1995 and provided a good test bed for assessment of rehabilitation in terms of radon flux attenuation. Measurements have been made with charcoal canisters. Studies to measure the radon-220 flux by observing Tl-208 progeny of thoron the effectiveness of trial covers and meteorological considerations will be reported.

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

    International Nuclear Information System (INIS)

    Drotning, W.D.

    1992-08-01

    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

  5. Numerical analysis of a natural convection cooling system for radioactive canisters storage

    Energy Technology Data Exchange (ETDEWEB)

    Tsal, R.J.; Anwar, S.; Mercada, M.G. [Fluor Daniel Inc., Irvine, CA (United States)

    1995-02-01

    This paper describes the use of numerical analysis for studying natural convection cooling systems for long term storage of heat producing radioactive materials, including special nuclear materials and nuclear waste. The paper explains the major design philosophy, and shares the experiences of numerical modeling. The strategy of storing radioactive material is to immobilize nuclear high-level waste by a vitrification process, convertion it into borosilicate glass, and cast the glass into stainless steel canisters. These canisters are seal welded, decontaminated, inspected, and temporarily stored in an underground vault until they can be sent to a geologic repository for permanent storage. These canisters generate heat by nuclear decay of radioactive isotopes. The function of the storage facility ventilation system is to ensure that the glass centerline temperature does not exceed the glass transition temperature during storage and the vault concrete temperatures remain within the specified limits. A natural convection cooling system was proposed to meet these functions. The effectiveness of a natural convection cooling system is dependent on two major factors that affect air movement through the vault for cooling the canisters: (1) thermal buoyancy forces inside the vault which create a stack effect, and (2) external wind forces, that may assist or oppose airflow through the vault. Several numerical computer models were developed to analyze the thermal and hydraulic regimes in the storage vault. The Site Model is used to simulate the airflow around the building and to analyze different air inlet/outlet devices. The Airflow Model simulates the natural convection, thermal regime, and hydraulic resistance in the vault. The Vault Model, internal vault temperature stratification; and, finally, the Hot Area Model is used for modeling concrete temperatures within the vault.

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

    Science.gov (United States)

    Kaufhold, Stephan; Hassel, Achim Walter; Sanders, Daniel; Dohrmann, Reiner

    2015-03-21

    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-silicate. Up to now it is not clear why and how the patina formed. It, however, may be relevant as a corrosion inhibitor. Copyright © 2014 Elsevier B.V. All rights reserved.

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

    International Nuclear Information System (INIS)

    Lidfors, E.D.; Tabe, T.; Johnson, H.M.

    1979-01-01

    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)

  8. 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.

  9. Comments on 'SKB RD and D-Programme 98'. Focused on canister integrity and corrosion

    International Nuclear Information System (INIS)

    Bowyer, W.H.; Hermansson, H.P.

    1999-04-01

    According to the Act on Nuclear Activities the nuclear utilities are requested to submit a comprehensive programme for research and development every third year, aiming at the safe storage of radioactive waste produced by the nuclear power plants. The latest was published by SKB in September 1998 and is called the 'RDandD Programme 98'. The work presented in the present report was commissioned by SKI and is a result of reading the 'RDandD Programme 98' and related reports with focus on canister production, integrity and corrosion. We find that those parts of the programme often are difficult to follow owing to the lack of detail in the Programme and in one of the supporting reports. In our opinion this will make the work difficult to monitor by SKI and SKB. We also feel that the interpretation of information already available often is overoptimistic. As a consequence the difficulties ahead are understated and the programme is allowed to converge too quickly. We agree that the materials choices for both the inner and outer canisters are appropriate providing they both can be produced commercially and in a satisfactory metallurgical condition, that they can be quality assured and that no further unforeseen difficulties arise. We also agree that alternative technologies merit consideration for production of the outer canister and that alternative joining processes should be studied. We are actually concerned that greater prominence is not given to the alternatives in the programme. We believe that it should be possible to develop a satisfactory canister for disposal of high level nuclear waste according to the general method proposed by SKB and with the proposed capacity within the time-scale of the overall programme. We do not believe, however, that all the difficulties have been recognised. As a consequence of this the results to date are interpreted optimistically. We believe that progress should be subjected to more professional review within SKB and that a higher

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

    International Nuclear Information System (INIS)

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

    1994-04-01

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

  11. Selection and evaluation of inner material candidates for Spanish high level radioactive waste canisters

    International Nuclear Information System (INIS)

    Puig, Francesc; Dies, Javier; Sevilla, Manuel; Pablo, Joan de; Pueyo, Juan Jose; Miralles, Lourdes; Martinez-Esparza, Aurora

    2007-01-01

    This paper summarizes the work carried out to analyse different alternatives related to the inner material selection of the Spanish high level waste canister for long term storage. The preliminary repository design considers granitic or clay formations, compacted bentonite sealing, corrosion allowing steel canisters and glass bead filling between the fuel assemblies and canister walls. This filling material will have the primary role of avoiding the possibility of a criticality event, which becomes an issue of major importance once the container is finally breached by corrosion and flooded by groundwater. In the first place, a complete set of requirements have been devised as evaluation criteria for candidate materials examination and selection; resulting in a compilation of demands significantly deeper and more exhaustive than any other similar work found in literature, including over 20 requirements and some other general aspects that could involve improvements in repository performance. Secondly, eight materials or material families (cast iron or steel, borosilicate glass, spinel, depleted uranium, dehydrated zeolites, hematite, phosphates and olivine) have been chosen and examined in detail, extracting some relevant conclusions. Either cast iron, borosilicate glass, spinel or depleted uranium are considered to look quite promising for the mentioned purpose. (authors)

  12. Feasibility of long-life and corrosion-resistant canister with titanium cladding

    International Nuclear Information System (INIS)

    Furuya, Masahiro; Tokiwai, Moriyasu; Saegusa, Toshiari

    2008-01-01

    In order to store nuclear spent fuels for a long term, we propose the concept of stainless steel canister with titanium cladding. The stainless canister is first brazed to titanium plates, and then the brazed joints are covered with other titanium plates. A MIG brazing for titanium and stainless steel was demonstrated with a brazing metal of Cu-1Mn-3Si alloy (MG960). JIS G 0601 shear strength, tensile shear stress and peel strength tests are conducted for the optimized MIG brazing conditions. These results showed the MIG brazing specimens possess adequate structural strength. After the salt spray test on the basis of JIS Z 2371, there were no pitting and general corrosions on a TIG welding specimen between titanium plates. The corrosion resistance is therefore, sufficiently high. Manufacturing cost estimation suggests that the titanium cladding concept is feasible thereby using 1-mm-thick titanium plates to reduce the material cost. In addition to this concept, we propose another concept of the canister by using titanium-stainless steel cladding plates to reduce a number of brazing joints. (author)

  13. Fracture toughness properties of candidate canister materials for spent fuel storage by concrete cask

    International Nuclear Information System (INIS)

    Arai, Taku; Mayuzumi, Masami; Libin, Niu; Takaku, Hiroshi

    2005-01-01

    It is very significant to clarify the fracture toughness properties of candidate canister materials to ensure the structural integrity against the accidents during handling in the storage facility. Fracture toughness tests on the CT specimens cut from base metal, heat affected zone (HAZ) and weld metal in the 2 types of weld joints made by candidate canister materials (SUS329J4L duplex stainless steel and YUS270 super stainless steel) were conducted under various test temperature between 233K and 473K. Stable ductile crack extensions were observed in all of the specimens. The fracture toughness J Q of the base metal and the HAZ of SUS329L4L showed the smallest value at 233K, and increased with temperature, then reached to the largest value at 298K. At the higher temperature, the value of J Q decreased slightly with temperature. While, the value of J Q in the weld metal increased with temperature. The value of J Q of YUS270 increased with temperature. The values of J Q for weld metal in both of the materials were not greater than those in base metal and HAZ at each test temperature. The values of J Q in weld metal of both materials at 213K and 473K were greater than applied J derived from postulated semi-elliptical surface flaw and maximum allowable stress in JSME design coed. This result suggested that these materials have enough toughness for use as the canister material. (author)

  14. Initial results from the canistered waste forms produced during the first campaign of the DWPF Startup Test Program

    International Nuclear Information System (INIS)

    Harbour, J.R.

    1995-01-01

    As part of the Defense Waste Processing Facility (DWPF) Startup Test Program, approximately 90 canisters will be filled with glass containing simulated radioactive waste during five separate campaigns. The first campaign is a facility acceptance test to demonstrate the operability of the facility and to collect initial data on the glass and the canistered waste forms. During the next four campaigns (the waste qualification campaigns) data will be obtained which will be used to demonstrate that the DWPF product meets DOE's Waste Acceptance Product Specifications (WAPS). Currently 12 of the 16 canisters have been filled with glass during the first campaign (FA-13). This paper describes the tests that have been carried out on these 12 glass-filled canisters and presents the data with reference to the acceptance criteria of the WAPS. These tests include measurement of canister dimensions prior to and after glass filling. dew point, composition, and pressure of the gas within the free volume of the canister, fill height, free volume, weight, leak rates of welds and temporary seals, and weld parameters

  15. 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.

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

    International Nuclear Information System (INIS)

    Rainer Senger

    2001-01-01

    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 and 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 and O 2000a), and is compliant with that plan. The application of the Quality Assurance (QA) program to the development of that plan (CRWMS M and O 2000a) and of this Analysis is

  17. Coalbed gas desorption in canisters: Consumption of trapped atmospheric oxygen and implications for measured gas quality

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Hui; Schimmelmann, Arndt [Indiana University, Dept. of Geological Sciences, Bloomington, IN 47405-1405 (United States); Mastalerz, Maria [Indiana University, Indiana Geological Survey, Bloomington, IN 47405-2208 (United States); Pope, James [CRL Energy Ltd., 123 Blenheim Road, Christchurch (New Zealand); University of Canterbury, Dept. of Geological Sciences, Christchurch (New Zealand); Moore, Tim A. [University of Canterbury, Dept. of Geological Sciences, Christchurch (New Zealand); P.T. Arrow Energy Indonesia, Wisma Anugraha, Jl. Taman Kemang No. 32B, Jakarta Selatan (Indonesia)

    2010-01-07

    Desorption canisters are routinely employed to quantify coalbed gas contents in coals. If purging with inert gas or water flooding is not used, entrapment of air with {proportional_to} 78.08 vol.% nitrogen (N{sub 2}) in canisters during the loading of coal results in contamination by air and subsequent overestimates of N{sub 2} in desorbed coalbed gas. Pure coalbed gas does not contain any elemental oxygen (O{sub 2}), whereas air contamination originally includes {proportional_to} 20.95 vol.% O{sub 2} and has a N{sub 2}/O{sub 2} volume ratio of {proportional_to} 3.73. A correction for atmospheric N{sub 2} is often attempted by quantifying O{sub 2} in headspace gas and then proportionally subtracting atmospheric N{sub 2}. However, this study shows that O{sub 2} is not a conservative proxy for air contamination in desorption canisters. Time-series of gas chromatographic (GC) compositional data from several desorption experiments using high volatile bituminous coals from the Illinois Basin and a New Zealand subbituminous coal document that atmospheric O{sub 2} was rapidly consumed, especially during the first 24 h. After about 2 weeks of desorption, the concentration of O{sub 2} declined to near or below GC detection limits. Irreversible loss of O{sub 2} in desorption canisters is caused by biological, chemical, and physical mechanisms. The use of O{sub 2} as a proxy for air contamination is justified only immediately after loading of desorption canisters, but such rapid measurements preclude meaningful assessment of coalbed gas concentrations. With increasing time and progressive loss of O{sub 2}, the use of O{sub 2} content as a proxy for atmospheric N{sub 2} results in overestimates of N{sub 2} in desorbed coalbed gas. The indicated errors for nitrogen often range in hundreds of %. Such large analytical errors have a profound influence on market choices for CBM gas. An erroneously calculated N{sub 2} content in CBM would not meet specifications for most pipeline

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

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Claes-Goeran

    2002-04-01

    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

  19. 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

  20. Preliminary design of the high-level waste canister storage system: Topical report for the period of January 1, 1987--September 30, 1987

    International Nuclear Information System (INIS)

    Peters, F.E.; Leap, D.R.

    1987-11-01

    The final stage of the West Valley solidification program will be to place the high-level waste canisters in interim storage until a federal repository is ready to receive them. The waste canisters will be stored in the largest former fuel reprocessing cell at West Valley modified for this purpose. This report provides a description of the preliminary design of the Waste Canister Storage Facility. 9 refs., 14 figs., 1 tab

  1. Research of radioactive waste storage cask/canister materials, spent nuclear fuels and various radioactive waste forms and development of their assessment methods. Final report for Stage 3

    International Nuclear Information System (INIS)

    Dobrev, D.; Balek, V.; Červinka, R.; Večerník, P.; Člupek, M.; Kouřil, M.; Novák, P.; Stoulil, J.; Silber, R.

    2013-08-01

    The main topics treated are: Research and development of methodologies for canister/cask material degradation assessment; Laboratory research of selected materials of canister/cask with radioactive waste; and Research and assessment of canister/cask materials in natural granite rocks. Two additional documents are appended: Corrosion rate determination for samples in compacted bentonite in anaerobic conditions (methodology), and Roll test for corrosion test in an occluded solution at the interface between a radioactive waste disposal canister and the bentonite cover. (P.A.)

  2. Thermal performance of a concrete cask: Methodology to model helium leakage from the steel canister

    International Nuclear Information System (INIS)

    Penalva, J.; Feria, F.; Herranz, L.E.

    2017-01-01

    Highlights: • A thermal analysis of the canister during a loss of leaktightness has been performed. • Methodologies that predict fuel temperatures and heat up rates have been developed. • Casks with heat loads below 20 kW would never exceed the thermal threshold. - Abstract: Concrete cask storage systems used in dry storage allocate spent fuel within containers that are usually filled with helium at a certain pressure. Potential leaks from the container would result in a cooling degradation of fuel that might jeopardize fuel integrity if temperature exceeded a threshold value. According to ISG-11, temperatures below 673 K ensure fuel integrity preservation. Therefore, the container thermal response to a loss of leaktightness is of utmost importance in terms of safety. In this work, a thermo-fluid dynamic analysis of the canister during a loss of leaktightness has been performed. To do so, steady-state and transient Computational Fluid Dynamics (CFD) simulations have been carried out. Likewise, it has been developed two methodologies capable of estimating peak fuel temperatures and heat up rates resulting from a postulated depressurization in a dry storage cask. One methodology is based on control theory and transfers functions, and the other methodology is based on a linear relationship between the inner pressure and the maximum temperature. Both methodologies have been verified through comparisons with CFD calculations. The period of time to achieve the temperature threshold (673 K) is a function of pressure loss rate and decay heat of the fuel stored in the container; in case of a fuel canister with 30 kW the period of time to reach the thermal limit takes between half day (fast pressure loss) and one week (slow pressure loss). In case of a 15% reduction of the decay heat, the period of time to achieve the thermal limit increase up to a few weeks. The results highlight that casks with heat loads below 20 kW would never exceed the thermal threshold (673 K).

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

    Energy Technology Data Exchange (ETDEWEB)

    PICKETT, W.W.

    2000-09-22

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

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

    International Nuclear Information System (INIS)

    PICKETT, W.W.

    2000-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    BAZINET, G.D.

    2001-05-15

    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. Revision 1 documented 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 requirements are noted in section 3

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

    International Nuclear Information System (INIS)

    BAZINET, G.D.

    2001-01-01

    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. Revision 1 documented 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 requirements are noted in section 3.1.5 and will be

  7. 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

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

    International Nuclear Information System (INIS)

    SWENSON, C.E.

    2000-01-01

    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 document SNF-6248

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

    International Nuclear Information System (INIS)

    Mecham, W.J.; Seefeldt, W.B.; Steindler, M.J.

    1976-08-01

    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

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

    International Nuclear Information System (INIS)

    Harrison, Tim

    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

  12. Analysis of heat and mass transport processes near an emplaced nuclear waste canister

    International Nuclear Information System (INIS)

    Keller, C.

    1990-01-01

    A review has been performed of the models and experimental plans for evaluation of the spent fuel canister environment in a nuclear repository, e.g., the planned Yucca Mountain facilities. Special emphasis was placed on the relevance of the models and experiments to the 100 to 10,000 year prediction. The question was addressed whether one could justify testing in materials other than Yucca Mountain rock and obtain results in a relatively short time which would be relevant to the long time in Yucca Mountain. The paper discusses steam evolution in calculations and experiments, fracture models, possible measurements of relative permeability, and long time scale effects. 5 figs. (MB)

  13. As-Built Verification Plan Spent Nuclear Fuel Canister Storage Building MCO Handling Machine

    International Nuclear Information System (INIS)

    SWENSON, C.E.

    2000-01-01

    This as-built verification plan outlines the methodology and responsibilities that will be implemented during the as-built field verification activity for the Canister Storage Building (CSB) MCO HANDLING MACHINE (MHM). This as-built verification plan covers THE ELECTRICAL PORTION of the CONSTRUCTION PERFORMED BY POWER CITY UNDER CONTRACT TO MOWAT. The as-built verifications will be performed in accordance Administrative Procedure AP 6-012-00, Spent Nuclear Fuel Project As-Built Verification Plan Development Process, revision I. The results of the verification walkdown will be documented in a verification walkdown completion package, approved by the Design Authority (DA), and maintained in the CSB project files

  14. Performance of CASTORR HAW Cask Cold Trials for Loading, Transport and Storage of HAW canisters

    International Nuclear Information System (INIS)

    Wilmsmeier, Marco; Vossnacke, Andre

    2008-01-01

    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 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 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 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 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 as well. GNS long-term CASTOR R

  15. TOUGH: a numerical model for nonisothermal unsaturated flow to study waste canister heating effects

    International Nuclear Information System (INIS)

    Pruess, K.; Wang, J.S.Y.

    1983-12-01

    The physical processes modeled and the mathematical and numerical methods employed in a simulator for non-isothermal flow of water, vapor, and air in permeable media are briefly summarized. The simulator has been applied to study thermo-hydrological conditions in the near vicinity of high-level nuclear waste packages emplaced in unsaturated rocks. The studies reported here specifically address the question whether or not the waste canister environment will dry up in the thermal phase. 13 references, 8 figures, 2 tables

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

    International Nuclear Information System (INIS)

    Garvin, L.J.

    1997-01-01

    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

  17. Using process instrumentation to obviate destructive examination of canisters of HLW glass

    International Nuclear Information System (INIS)

    Kuhn, W.L.; Slate, S.C.

    1983-01-01

    An important concern of a manufacturer of packages of solidified high-level waste (HLW) is quality assurance of the waste form. The vitrification of HLW as a borosilicate glass is considered, and, based on a reference vitrification process, it is proposed that information from process instrumentation may be used to assure quality without the need for additional information obtained by destructive examining (core drilling) canisters of glass. This follows mainly because models of product performance and process behavior must be previously established in order to confidently select the desired glass formulation, and to have confidence that the process is well enough developed to be installed and operated in a nuclear facility

  18. 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

  19. Corrosion resistant metallic canisters: an important element of a multibarrier disposal program

    International Nuclear Information System (INIS)

    Magnani, N.J.; Braithwaite, J.W.

    1979-01-01

    A program with the goal of qualifying a material for a 300 year lifetime as a nuclear waste canister is underway at Sandia Laboratories. The corrosion and stress corrosion cracking behavior of the leading candidate, TiCode-12 (Ti-0.8% Ni-0.3% Mo), is contrasted to that of a commonly used engineering alloy, 304 stainless steel. Experimental evidence is presented which shows the inadequacy of 304 stainless steel in simulated repository environments and shows that TiCode-12 may survive the desired 300 years. Further work required to qualify TiCode-12 is outlined

  20. Application of transient ignition model to multi-canister (MCO) accident analysis

    International Nuclear Information System (INIS)

    Kummerer, M.

    1996-01-01

    The potential for ignition of spent nuclear fuel in a Multi-Canister Overpack (MCO) is examined. A transient model is applied to calculate the highest ambient gas temperature outside an MCO wall tube or shipping cask for which a stable temperature condition exists. This integral analysis couples reaction kinetics with a description of the MCO configuration, heat and mass transfer, and fission product phenomena. It thereby allows ignition theory to be applied to various complex scenarios, including MCO water loss accidents and dry MCO air ingression

  1. Monitored retrievable storage and multi-purpose canister robotic applications: Feasibility, dose savings and cost analysis

    International Nuclear Information System (INIS)

    Bennett, P.C.

    1995-01-01

    Robotic automation is examined as a possible alternative to manual spent nuclear fuel, transport cask and Multi-Purpose Canister (MPC) handling at a Monitored Retrievable Storage (MRS) facility, and as an alternative to current MPC closure and welding methods at commercial nuclear reactor sites. Automation of key operational aspects is analyzed to determine equipment requirements, through-put times and equipment costs. The economic analysis approach is described, and economic and radiation dose impacts resulting from this automation are compared to manual handling methods. (author). 5 refs, 5 figs, 3 tabs

  2. Growth Protocols for Etiolated Soybeans Germinated within BRIC-60 Canisters Under Spaceflight Conditions

    Science.gov (United States)

    Levine, H. G.; Sharek, J. A.; Johnson, K. M.; Stryjewski, E. C.; Prima, V. I.; Martynenko, O. I.; Piastuch, W. C.

    As part of the GENEX (Gene Expression) spaceflight experiment, protocols were developed to optimize the inflight germination and subsequent growth of 192 soybean (Glycine max cv McCall) seeds during STS-87. We describe a method which provided uniform growth and development of etiolated seedlings while eliminating root and shoot restrictions for short-term (4-7 day) experiments. Final seedling growth morphologies and the gaseous CO2 and ethylene levels present both on the last day in space and at the time of recovery within the spaceflight and ground control BRIC-60 canisters are presented

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

    International Nuclear Information System (INIS)

    Zirker, L.R.; Rankin, R.A.; Ferrell, L.J.

    2002-01-01

    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

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

    International Nuclear Information System (INIS)

    Bajpai, R.K.; Verma, A.K.; Maheshwar, Sachin

    2016-01-01

    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

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

    International Nuclear Information System (INIS)

    Auerkari, P.; Leinonen, H.; Sandlin, S.

    1991-07-01

    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

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

    International Nuclear Information System (INIS)

    Garvin, L.J.

    1997-01-01

    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

  7. Enhanced Thermal Management System for Spent Nuclear Fuel Dry Storage Canister with Hybrid Heat Pipes

    International Nuclear Information System (INIS)

    Jeong, Yeong Shin; Bang, In Cheol

    2016-01-01

    Dry storage uses the gas or air as coolant within sealed canister with neutron shielding materials. Dry storage system for spent fuel is regarded as relatively safe and emits little radioactive waste for the storage, but it showed that the storage capacity and overall safety of dry cask needs to be enhanced for the dry storage cask for LWR in Korea. For safety enhancement of dry cask, previous studies of our group firstly suggested the passive cooling system with heat pipes for LWR spent fuel dry storage metal cask. As an extension, enhanced thermal management systems for the spent fuel dry storage cask for LWR was suggested with hybrid heat pipe concept, and their performances were analyzed in thermal-hydraulic viewpoint in this paper. In this paper, hybrid heat pipe concept for dry storage cask is suggested for thermal management to enhance safety margin. Although current design of dry cask satisfies the design criteria, it cannot be assured to have long term storage period and designed lifetime. Introducing hybrid heat pipe concept to dry storage cask designed without disrupting structural integrity, it can enhance the overall safety characteristics with adequate thermal management to reduce overall temperature as well as criticality control. To evaluate thermal performance of hybrid heat pipe according to its design, CFD simulation was conducted and previous and revised design of hybrid heat pipe was compared in terms of temperature inside canister

  8. Spent nuclear fuel project multi-canister overpack, additional NRC requirements

    International Nuclear Information System (INIS)

    Garvin, L.J.

    1998-01-01

    The US Department of Energy (DOE), established in the K Basin Spent Nuclear Fuel Project Regulatory Policy, dated August 4, 1995 (hereafter referred to as the Policy), the requirement for new Spent Nuclear Fuel (SNF) Project facilities to achieve nuclear safety equivalency to comparable US Nuclear Regulatory Commission (NRC)-licensed facilities. For activities other than during transport, when the Multi-Canister Overpack (MCO) is used and resides in the Canister Storage Building (CSB), Cold Vacuum Drying (CVD) facility or Hot Conditioning System, additional NRC requirements will also apply to the MCO based on the safety functions it performs and its interfaces with the SNF Project facilities. An evaluation was performed in consideration of the MCO safety functions to identify any additional NRC requirements needed, in combination with the existing and applicable DOE requirements, to establish nuclear safety equivalency for the MCO. The background, basic safety issues and general comparison of NRC and DOE requirements for the SNF Project are presented in WHC-SD-SNF-DB-002

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

    International Nuclear Information System (INIS)

    1990-03-01

    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

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

    International Nuclear Information System (INIS)

    McCright, R.D.; Weiss, H.; Juhas, M.C.; Logan, R.W.

    1983-11-01

    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

  11. A New Approach to Environmentally Safe Unique Identification of Long-Term Stored Copper Canisters

    International Nuclear Information System (INIS)

    Chernikova, D.; Axell, K.; Nordlund, A.

    2015-01-01

    A new approach to environmentally safe unique identification of long-term stored copper canisters is suggested in this paper. The approach is based on the use of a tungstenbased insert placed inside a copper cask between a top iron lid and a copper lid. The insert/label is marked with unique code in a form of binary number, which is implemented as a combination of holes in the tungsten plate. In order to provide a necessary redundancy of the identifier, the tungsten label marked with few identical binary codes. The position of code (i.e., holes in tungsten) corresponds to a predefined placement of the spent fuel assembles in the iron container. This is in order to avoid any non-uniformity of the gamma background at the canister surface caused by a presence of iron-filled spaces between spent nuclear fuel assembles. Due to the use of the tungsten material gamma rays emitted by the spent fuel assembles are collimated in a specific way because of strong attenuation properties of tungsten. As a result, the variation in the gamma-counting rate in a detector array placed on the top of copper lid provides the distribution of the holes in the tungsten insert or in other words the unique identifier. Thus, this way of identification of copper cask do not impair the integrity of the cask and it offers a way that the information about spent nuclear fuel is legible for a time scale up to a few thousands years. (author)

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

    International Nuclear Information System (INIS)

    Pike, S.; Allen, C.; Punshon, C.; Threadgill, P.; Gallegillo, M.; Holmes, B.; Nicholas, J.

    2010-03-01

    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

  13. 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.

  14. Multi Canister Overpack (MCO) Closure Welding Process Parameter Development and Qualification

    International Nuclear Information System (INIS)

    CANNELL, G.R.

    2003-01-01

    One of the Department of Energy's (DOE) top priorities at the Hanford Site (southeastern Washington state), is the processing of more than 2,000 tons of spent nuclear fuel (SNF) into large stainless steel containers called Multi-Canister Overpacks (MCO). Packaging into MCO's will assist in the safe and economic disposition of SNF and greatly reduce risk to the environment. Packaged fuel will be removed from close proximity to the Columbia River to a more suitable area of the site where it will be stored on an interim basis. Eventually, the fuel will be transferred to the federal geologic repository for long-term storage. One of the key elements in the SNF process is final closure of the MCO by welding. Fuel is loaded into the MCO (approximately 2 ft. in diameter and 13 ft. long) and a heavy shield plug inserted into the top, creating a mechanical seal. The plug contains several process ports for various operations, including vacuum drying and inert-gas backfilling of the packaged fuel. When fully processed, the Canister Cover Assembly (CCA) is placed over the shield plug and final closure made by welding. The following describes the effort to develop and qualify the root-pass technique associated with the MCO final closure weld

  15. 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.

  16. White Paper: Multi-purpose canister (MPC) for DOE-owned spent nuclear fuel (SNF)

    International Nuclear Information System (INIS)

    Knecht, D.A.

    1994-04-01

    The paper examines the issue, What are the advantages, disadvantages, and other considerations for using the MPC concept as part of the strategy for interim storage and disposal of DOE-owned SNF? The paper is based in part on the results of an evaluation made for the DOE National Spent Fuel Program by the Waste Form Barrier/Canister Team, which is composed of knowledgeable DOE and DOE-contractor personnel. The paper reviews the MPC and DOE SNF status, provides criteria and other considerations applicable to the issue, and presents an evaluation, conclusions, and recommendations. The primary conclusion is that while most of DOE SNF is not currently sufficiently characterized to be sealed into an MPC, the advantages of standardized packages in handling, reduced radiation exposure, and improved human factors should be considered in DOE SNF program planning. While the design of MPCs for DOE SNF are likely premature at this time, the use of canisters should be considered which are consistent with interim storage options and the MPC design envelope

  17. Investigation into the suitability of titanium as a corrosion resistant canister for nuclear waste

    International Nuclear Information System (INIS)

    Henriksson, S.; Pettersson, J.

    A literature study and inventory of experience has been carried out, aimed at assessing the possibilities of unalloyed and Pd-alloyed titanium withstanding corrosion for 1,000 to 10,000 years in contact with Baltic Sea water at 100 0 C and pH 4 to 10. Pitting, crevice corrosion, stress corrosion cracking and corrosion fatigue constitute no problem if the canister is made of unalloyed titanium corresponding to ASTM Grade 1. Titanium alloyed with palladium therefore need not be used. Linear extrapolation of reported corrosion rates for oxidation and general corrosion gives a life of between 1,000 and 10,000 years for a 5 mm thick canister. This estimate must be considered to be conservative since oxidation in fact follows a logarithmic law. Hydrogen embrittlement resulting from hydrogen pick-up from the deposition environment should not occur. Delayed failure caused by a redistribution of the hydrogen initially present in the titanium can be avoided if its concentration is maximized to 20 ppM. Pd-alloyed titanium is more sensitive than unalloyed titanium to hydrogen pick-up, especially in galvanic contact with less noble metals

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

    International Nuclear Information System (INIS)

    Auerkari, P.; Leinonen, H.; Sandlin, S.

    1991-09-01

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-10-01

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

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

    International Nuclear Information System (INIS)

    SMITH, K.E.

    1998-01-01

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

  1. Radioactive air emissions notice of construction for Canister Storage Building (revised sealing configuration for spent nuclear fuel) - Project W-379

    International Nuclear Information System (INIS)

    Kamberg, L.D.

    1998-01-01

    The purpose of this Notice of Construction (NOC) is to provide a rewritten NOC for obtaining regulatory approval for changes to the previous Canister Storage Building (CSB) NOCs (WDOH, 1996 and EPA, 1996) as were approved by the Washington State Department of Health (WDOH, 1996a) and US Environmental Protection Agency (EPA, 1996a). These changes are because of a revised sealing configuration of the multi-canister overpacks (MCOS) that are used to store the SNF. A flow schematic of the SNF Project is provided in Figure 1-1. A separate notification of startup will be provided apart from this NOC

  2. Galvanic corrosion of copper-cast iron couples in relation to the Swedish radioactive waste canister concept

    International Nuclear Information System (INIS)

    Smart, N.R.; Fennell, P.A.H.; Rance, A.P.; Werme, L.O.

    2004-01-01

    To ensure the safe encapsulation of spent nuclear fuel rods for geological disposal, SKB are considering using the Copper-Iron Canister, which consists of an outer copper canister and an inner cast iron container. The canister will be placed into boreholes in the bedrock of a geologic repository and surrounded by bentonite clay. In the unlikely event of the outer copper canister being breached, water could enter the annulus between the inner and outer canister and at points of contact between the two metals there would be a possibility of galvanic interactions. To study this effect, copper-cast iron galvanic couples were set up in a number of different environments representing possible conditions in the SKB repository. The tests investigated two artificial pore-waters and a bentonite slurry, under aerated and deaerated conditions, at 30 deg. C and 50 deg. C. The currents passing between the coupled electrodes and the potential of the couples were monitored for several months. In addition, some bimetallic crevice specimens based on the multi-crevice assembly (MCA) design were used to simulate the situation where the copper canister will be in direct contact with the cast iron inner vessel. The effect of growing an oxide film on the surface of the cast iron prior to coupling it with copper was also investigated. The electrochemical results are presented graphically in the form of electrode potentials and galvanic corrosion currents as a function of time. The galvanic currents in aerated conditions were much higher than in deaerated conditions. For example, at 30 deg. C, galvanic corrosion rates as low as 0.02 μm/year were observed for iron in groundwater after de-aeration, but of the order of 100 μm/year for the cast iron at 50 deg. C in the presence of oxygen. The galvanic currents were generally higher at 50 deg. C than at 30 deg. C. None of the MCA specimens exhibited any signs of crevice corrosion under deaerated conditions. It will be shown that in deaerated

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

    International Nuclear Information System (INIS)

    Bennett, D.G.; Hicks, T.W.

    2005-10-01

    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

  4. Application of a cold spray technique to the fabrication of a copper canister for the geological disposal of CANDU spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Heui-Joo, E-mail: hjchoi@kaeri.re.k [Korea Atomic Energy Research Institute, Radioactive Waste Management Technology Development, 150 Dukjin-dong, Yuseong, Daejon, 305-353 (Korea, Republic of); Lee, Minsoo; Lee, Jong Youl [Korea Atomic Energy Research Institute, Radioactive Waste Management Technology Development, 150 Dukjin-dong, Yuseong, Daejon, 305-353 (Korea, Republic of)

    2010-10-15

    A new method was proposed for the manufacture of a copper-cast iron canister for the spent fuel disposal based on the cold spray coating technique. The thickness of a copper shell could be fabricated to be as thin as 10 mm with the new method. Around 6 tons of copper could be saved with a 10 mm thick canister compared with a 50 mm thick canister. The electrochemical properties of the cold sprayed copper layer and forged copper were measured through a polarization test. The two copper layers showed very similar electrochemical properties. The lifetime of a 10 mm copper canister was estimated with a mathematical model based on the mass transport of sulfide ions through the buffer. The results showed that the canister lifetime was more than 140,000 years under the Korean granite groundwater condition. The thermal analysis with a current pre-conceptual design of a CANDU spent fuel canister showed that the maximum temperature between the canister and the saturated buffer was below the thermal criteria, 100 {sup o}C. Finally, the mechanical stability of the copper canister was confirmed with a computer program, ABAQUS, under the rock movement scenario.

  5. 2D and 3D thermal simulations for storage systems with internal natural convection for canistered spent fuel

    International Nuclear Information System (INIS)

    Yaksh, M.; Wang, C.

    2004-01-01

    In the US, the number of nuclear plants expected to implement on-site dry storage is increasing each year. As reactors burn advanced fuel assemblies to higher burnups, the dry storage systems will be required to accommodate higher heat loads. This is due to the increasing capacity of the systems and the need to store higher burnup fuel with reasonable cooling periods (i.e., five to six years). As the storage systems heat rejection design must be passive, natural convection is an efficient means for rejection of heat from the spent fuel to the surface of the canister boundary. The design presented in this paper is a canistered system that employs conduction, radiation and convection to reject heat from the canister, which is stored in a vertical concrete cask. The canister containing the spent fuel in this design is a right circular stainless steel vessel capable of storing 37 PWR fuel assemblies with a total canister heat load of 40 kW. Accompanying any design effort is the use of a numerical methodology that can accurately predict the peak-clad temperatures of the fuel and the structural components of the system. The main challenge to any analysis employing internal natural convection may be perceived as a practical limitation due to the size of the model. Since canisters are typically cylindrical, a two-dimensional model can be used to represent the canister. The fuel basket structure, which maintains the configuration of the spent fuel, is an array of square tubes, and is non-axisymmetric. Flow up through the fuel region in the basket encounters a complex cross section due to the fuel assembly rod array (up to 17 x 17). The flow region of the heated gas down the outside of the basket in the annulus between the canister shell and the basket assembly (downcomer) is also an irregular shaped area. To confirm that a two-dimensional (2D) modelling methodology is appropriate, a benchmark using results from a thermal test is required. The thermal test focuses on the

  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. Residual stress investigation of copper plate and canister EB-Welds Complementary Results

    International Nuclear Information System (INIS)

    Gripenberg, H.

    2009-03-01

    The residual stresses in copper as induced by EB-welding were studied by specimens where the weld had two configurations: either a linear or a circumferential weld. This report contains the residual stress measurements of two plates, containing linear welds, and the full-scale copper lid specimen to which a hollow cylinder section had been joined by a circumferential EB-weld. The residual stress state of the EB-welded copper specimens was investigated by X-ray diffraction (XRD), hole drilling (HD) ring core (RC) and contour method (CM). Three specimens, canister XK010 and plates X251 and X252, were subjected to a thorough study aiming at quantitative determination of the residual stress state in and around the EB-welds using XRD for surface and HD and RC for spatial stress analysis. The CM maps one stress component over a whole cross section. The surface residual stresses measured by XRD represent the machined condition of the copper material. The XRD study showed that the stress changes towards compression close to the weld in the hollow cylinder, which indicates shrinkage in the hoop direction. According to the same analogy, the shrinkage in the axial direction is much smaller. The HD measurements showed that the stress state in the base material is bi-axial and, in terms of von Mises stress, 50 MPa for the plates and 20 MPa for the cylinder part of the canister. The stress state in the EB-welds of all specimens differs clearly from the stress state in the base material being more tensile, with higher magnitudes of von Mises stress in the plate than in the canister welds. The HD and RC results were obtained using linear elastic theory. The RC measurements showed that the maximum principal stress in the BM is close to zero near the surface and it becomes slightly tensile, 10 MPa, deeper under the surface. Welding pushed the general stress state towards tension with the maximum principal stress reaching 50 MPa, deeper than 5 mm below the surface in the weld. The

  8. 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

  9. Friction Stir Welding of Copper Canisters Using Power and Temperature Control

    International Nuclear Information System (INIS)

    Cederqvist, Lars

    2011-01-01

    This thesis presents the development to reliably seal 50 mm thick copper canisters containing the Swedish nuclear waste using friction stir welding. To avoid defects and welding tool fractures, it is important to control the tool temperature within a process window of approximately 790 to 910 deg C. The welding procedure requires variable power input throughout the 45 minute long weld cycle to keep the tool temperature within its process window. This is due to variable thermal boundary conditions throughout the weld cycle. The tool rotation rate is the input parameter used to control the power input and tool temperature, since studies have shown that it is the most influential parameter, which makes sense since the product of tool rotation rate and spindle torque is power input. In addition to the derived control method, the reliability of the welding procedure was optimized by other improvements. The weld cycle starts in the lid above the joint line between the lid and the canister to be able to abort a weld during the initial phase without rejecting the canister. The tool shoulder geometry was modified to a convex scroll design that has shown a self-stabilizing effect on the power input. The use of argon shielding gas reduced power input fluctuations i.e. process disturbances, and the tool probe was strengthened against fracture by adding surface treatment and reducing stress concentrations through geometry adjustments. In the study, a clear relationship was shown between power input and tool temperature. This relationship can be used to more accurately control the process within the process window, not only for this application but for other applications where a slow responding tool temperature needs to be kept within a specified range. Similarly, the potential of the convex scroll shoulder geometry in force-controlled welding mode for use in applications with other metals and thicknesses is evident. The variable thermal boundary conditions throughout the weld

  10. Three Dimensional Modelling of Canister for Spent Nuclear Fuel - some migration studies

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, Antonio [AlbaNova Univ. Center, Stockholm (Sweden). Dept. of Physics

    2006-08-15

    Performance assessment transport models use extensively the concept of transport resistance in the calculation of breakthrough curves of radionuclide releases in the near field and geosphere. The aim of this work is to examine more closely the applicability of the transport resistance approach. Can the resistance approach be used in for the estimation of fluxes through a pinhole of a defected canister? Or for the estimation of fluxes as given by the resistance of a fracture that crosses a canister hole? And if so, what is the degree of conservatism (if any) introduced by the use of that concept? Two near-field 3D-models of the system consisting of canister, bentonite buffer and fracture have been developed. The goal is to examine the contribution to mass-transfer resistance of the interfaces between pinhole and bentonite buffer and between bentonite buffer and fracture respectively and to compare them with the resistance approach used by SKB in their compartment models of the near field. For this purpose we have developed two 3D models using the FEMLAB tool, to perform the set of calculations presented in this report. We estimate the above mentioned resistances separately for the interface between pinhole and bentonite buffer and for the interface between bentonite buffer and fracture respectively and we make a series of parameter variation studies. We conclude that the pinhole resistance used by SKB is a good approach to be used by compartment models even if some small discrepancy exists whenever the cross-section of the pinhole is larger than 10{sup -4} m{sup 2}. In respect to the fracture resistance parameterisation used in some SKB compartment models, the method is clearly conservative in many cases, with the exception for time points shorter than 200 years. This is due to the fact that the transient breakthrough curves cannot be described accurately by the parameterisation derived from the solution of the steady state equations used as the start point to

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

    International Nuclear Information System (INIS)

    Hermansson, H.P.; Koenig, M.

    2001-02-01

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

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

    International Nuclear Information System (INIS)

    BAZINET, G.D.

    2000-01-01

    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 requirements are noted

  13. 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

  14. Three Dimensional Modelling of a KBS-3 Canister for Spent Nuclear Fuel - some migration studies

    International Nuclear Information System (INIS)

    Pereira, Antonio

    2006-08-01

    Performance assessment transport models use extensively the concept of transport resistance in the calculation of breakthrough curves of radionuclide releases in the near field and geosphere. The aim of this work is to examine more closely the applicability of the transport resistance approach. Can the resistance approach be used in for the estimation of fluxes through a pinhole of a defected canister? Or for the estimation of fluxes as given by the resistance of a fracture that crosses a canister hole? And if so, what is the degree of conservatism (if any) introduced by the use of that concept? Two near-field 3D-models of the system consisting of canister, bentonite buffer and fracture have been developed. The goal is to examine the contribution to mass-transfer resistance of the interfaces between pinhole and bentonite buffer and between bentonite buffer and fracture respectively and to compare them with the resistance approach used by SKB in their compartment models of the near field. For this purpose we have developed two 3D models using the FEMLAB tool, to perform the set of calculations presented in this report. We estimate the above mentioned resistances separately for the interface between pinhole and bentonite buffer and for the interface between bentonite buffer and fracture respectively and we make a series of parameter variation studies. We conclude that the pinhole resistance used by SKB is a good approach to be used by compartment models even if some small discrepancy exists whenever the cross-section of the pinhole is larger than 10 -4 m 2 . In respect to the fracture resistance parameterisation used in some SKB compartment models, the method is clearly conservative in many cases, with the exception for time points shorter than 200 years. This is due to the fact that the transient breakthrough curves cannot be described accurately by the parameterisation derived from the solution of the steady state equations used as the start point to deduce the

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

    International Nuclear Information System (INIS)

    Bath, Adrian; Hermansson, Hans-Peter

    2009-08-01

    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 4 2- /HS - and Fe 3+ /Fe 2+ 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 sulphide

  16. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-02-01

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

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

    International Nuclear Information System (INIS)

    HEARD, F.J.

    2000-01-01

    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

  19. Preliminary assessment of the thermal effects of an annular air space surrounding an emplaced nuclear waste canister

    International Nuclear Information System (INIS)

    Davis, B.W.

    1979-01-01

    Modeling results have previously shown that the presence of a large air space (e.g., a repository room) within a nuclear waste repository is expected to cause a waste canister's temperature to remain cooler than it would otherwise be. Results presented herein show that an annular air space surrounding the waste canisters can have similar cooling effects under certain prescribable conditions; for a 16 ft x 1 ft diameter canister containing 650 PWR rods which initially generate a total of 4.61 kw, analysis will show that annular air spaces greater than 11 in will permit the canister surface to attain peak temperatures lower than that which would result from a zero-gap/perfect thermal contact. It was determined that the peak radial temperature gradient in the salt varies in proportion to the inverse of the drill hole radius. Thermal radiation is shown to be the dominant mode of heat transfer across an annular air space during the first two years after emplacement. Finally, a methodology is presented which will allow investigators to easily model radiation and convection heat transfer through air spaces by treating the space as a conduction element that possesses non-linear temperature dependent conductivity

  20. Canister materials proposed for final disposal of high level nuclear waste - a review with respect to corrosion resistance

    Energy Technology Data Exchange (ETDEWEB)

    Mattsson, E; Odoj, R; Merz, E [eds.

    1981-06-01

    Spent fuel from nuclear reactors has to be disposed of either after reprocessing or without such treatment. Due to toxic radiation the nuclear waste has to be isolated from the biosphere for 300-1000 years, or in extreme cases for more than 100,000 years. The nuclear waste will be enclosed in corrosion resistant canisters. These will be deposited in repositories in geological formations, such as granite, basalt, clay, bedded or domed salt, or the sediments beneath the deep ocean floor. There the canisters will be exposed to groundwater, brine or seawater at an elevated temperature. Species formed by radiolysis may affect the corrosivity of the agent. The corrosion resistance of candidate canister materials is evaluated by corrosion tests and by thermodynamic and mass transport calculations. Examination of ancient metal objects after long exposure in nature may give additional information. On the basis of the work carried out so far, the principal candidate canister materials are titanium materials, copper and high purity alumina.

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

    International Nuclear Information System (INIS)

    Cowan, R.G.

    1997-01-01

    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

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

    International Nuclear Information System (INIS)

    SWENSON, C.E.

    2000-01-01

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

  3. Corrosion of candidate materials for canister: applications in rock salt formations

    International Nuclear Information System (INIS)

    Azkarate, I.; Madina, V.; Barrio, A. del; Macarro, J.M.

    1994-01-01

    Previous corrosion studies carried out on various metallic materials in typical salt rock environments show that carbon steel and titanium alloys are the most promising candidates for canister applications in this geological formation. Although carbon steels have a low corrosion resistance, they are considered acceptable as corrosion-allowance materials for a thick walled container due to their practical immunity to the localized corrosion phenomena such as stress corrosion cracking, pitting or crevice corrosion. Aiming to improve the performances of these materials, studies on the effect of small additions of Ni and V on the general corrosion are in process. The improvement in the resistance to general corrosion should not be accompanied by a sensitivity to stress corrosion cracking. On the contrary, alfa titanium alloys are considered the most resistant materials to general corrosion in salt brines. However, pitting, are potential deficiencies of this corrosion-resistant materials for a thin walled container. (Author)

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

    International Nuclear Information System (INIS)

    Goldmann, L.H.

    1997-01-01

    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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

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

    International Nuclear Information System (INIS)

    GARRISON, R.C.

    2000-01-01

    This document implements the procedure for providing configuration control for the monitoring and control systems associated with the operation of the Canister Storage Building (CSB). It identifies and defines the configuration items in the monitoring and control systems, provides configuration control of these items throughout the system life cycle, provides configuration status accounting, physical protection and control, and verifies the completeness and correctness of the items. It is written to comply with HNF-SD-SNF-CM-001, Spent Nuclear Fuel Configuration Management Plan (Forehand 1998), HNF-PRO-309, Computer Software Quality Assurance Requirements, HNF-PRO-2778, IRM Application Software System Life Cycle Standards, and applicable sections of administrative procedure AP-CM-6-037-00, SNF Project Process Automation Software and Equipment Configuration Management

  7. The FUSE satellite is encased in a canister before being moved to the Launch Pad.

    Science.gov (United States)

    1999-01-01

    At Hangar AE, Cape Canaveral Air Station (CCAS), the last segment is lifted over the top of NASA's Far Ultraviolet Spectroscopic Explorer (FUSE) satellite already encased in a protective canister. The satellite will next be moved to Launch Pad 17A, CCAS, for its scheduled launch June 23 aboard a Boeing Delta II rocket. FUSE was developed by The Johns Hopkins University under contract to Goddard Space Flight Center, Greenbelt, Md., to investigate the origin and evolution of the lightest elements in the universe - hydrogen and deuterium. In addition, the FUSE satellite will examine the forces and process involved in the evolution of the galaxies, stars and planetary systems by investigating light in the far ultraviolet portion of the electromagnetic spectrum.

  8. Theoretical and experimental study of radon measurement with designing and calibration domestic canister with active charcoal

    International Nuclear Information System (INIS)

    Urosevic, V.; Nikezic, D.; Zekic, R.

    2005-01-01

    Radon concentration in air may change significantly large variation due to atmospheric variation. Measurement with active charcoal can be inaccurate because the variation in radon concentration. We made model to simulate radon measurements with active charcoal in order to optimize and improve integration characteristic. A numerical method and computer code based on the method of finite elements is developed for the case of variable radon concentration in air. This program simulates radon adsorption by the activated charcoal bed, enabling determination of sensitivity. The dependence of sensitivity on different parameters, such as temperature, thickness of the charcoal, etc. was studied using this program. Using results of theoretical investigation we designed and calibrated our canister with active charcoal for radon measurements. (author)

  9. Transportation system benefits of early deployment of a 75-ton multipurpose canister system

    International Nuclear Information System (INIS)

    Wankerl, M.W.; Schmid, S.P.

    1995-01-01

    In 1993 the US Civilian Radioactive Waste Management System (CRWMS) began developing two multipurpose canister (MPC) systems to provide a standardized method for interim storage and transportation of spent nuclear fuel (SNF) at commercial nuclear power plants. One is a 75-ton concept with an estimated payload of about 6 metric tons (t) of SNF, and the other is a 125-ton concept with an estimated payload of nearly 11 t of SNF. These payloads are two to three times the payloads of the largest currently certified US rail transport casks, the IF-300. Although is it recognized that a fully developed 125-ton MPC system is likely to provide a greater cost benefit, and radiation exposure benefit than the lower-capacity 75-ton MPC, the authors of this paper suggest that development and deployment of the 75-ton MPC prior to developing and deploying a 125-ton MPC is a desirable strategy. Reasons that support this are discussed in this paper

  10. 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

  11. Probabilistic sensitivity analysis for the 'initial defect in the canister' reference model

    International Nuclear Information System (INIS)

    Cormenzana, J. L.

    2013-08-01

    In Posiva Oy's Safety Case 'TURVA-2012' the repository system scenarios leading to radionuclide releases have been identified in Formulation of Radionuclide Release Scenarios. Three potential causes of canister failure and radionuclide release are considered: (i) the presence of an initial defect in the copper shell of one canister that penetrates the shell completely, (ii) corrosion of the copper overpack, that occurs more rapidly if buffer density is reduced, e.g. by erosion, (iii) shear movement on fractures intersecting the deposition hole. All three failure modes are analysed deterministically in Assessment of Radionuclide Release Scenarios, and for the 'initial defect in the canister' reference model a probabilistic sensitivity analysis (PSA) has been carried out. The main steps of the PSA have been: quantification of the uncertainties in the model input parameters through the creation of probability density distributions (PDFs), Monte Carlo simulations of the evolution of the system up to 106 years using parameters values sampled from the previous PDFs. Monte Carlo simulations with 10,000 individual calculations (realisations) have been used in the PSA, quantification of the uncertainty in the model outputs due to uncertainty in the input parameters (uncertainty analysis), and identification of the parameters whose uncertainty have the greatest effect on the uncertainty in the model outputs (sensitivity analysis) Since the biosphere is not included in the Monte Carlo simulations of the system, the model outputs studied are not doses, but total and radionuclide-specific normalised release rates from the near-field and to the biosphere. These outputs are calculated dividing the activity release rates by the constraints on the activity fluxes to the environment set out by the Finnish regulator. Two different cases are analysed in the PSA: (i) the 'hole forever' case, in which the small hole through the copper overpack remains unchanged during the assessment

  12. Experimental determination of the stress/strain situation in a sheared tunnel model with canister

    International Nuclear Information System (INIS)

    Pusch, R.

    1978-03-01

    A previous report concerned a technical matter which could be of great importance as regards the mechanical strength of canisters embedded in a bentonite/quartz buffer mass, i.e. the effect of a differential movement triggered by a critical deviatoric stress condition. Even if this is extremely unlikeley to occur it was considered to be of importance to verify the theoretical expressions for the maximum bending moment and maximum shear force. A special reason was to test the hypothesis that the contact pressure would soon reach a high value and then stay fairly constant when the displacement increased. The theoretical approach requires that the stress/strain properties of the fill are thoroghly investigated and described in therms of a mathematical model. Experience shows that this may be a tedions and difficult task. (L.E.)

  13. 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.

  14. DESIGN VERIFICATION REPORT SPENT NUCLEAR FUEL (SNF) PROJECT CANISTER STORAGE BUILDING (CSB)

    Energy Technology Data Exchange (ETDEWEB)

    BAZINET, G.D.

    2003-02-12

    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. Revision 1 documented verification actions that were pending at the time the initial report was prepared. Revision 3 of this document incorporates MCO Cover Cap Assembly welding verification activities. Verification activities for the installed and operational SSCs have been completed.

  15. DESIGN VERIFICATION REPORT SPENT NUCLEAR FUEL (SNF) PROJECT CANISTER STORAGE BUILDING (CSB)

    International Nuclear Information System (INIS)

    BAZINET, G.D.

    2003-01-01

    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. Revision 1 documented verification actions that were pending at the time the initial report was prepared. Revision 3 of this document incorporates MCO Cover Cap Assembly welding verification activities. Verification activities for the installed and operational SSCs have been completed

  16. Warehouse Plan for the Multi-Canister Overpacks (MC0) and Baskets

    International Nuclear Information System (INIS)

    MARTIN, M.K.

    2000-01-01

    The Multi-Canister Overpacks (MCO) will contain spent nuclear fuel (SNF) removed from the K East and West Basins. The SNF will be placed in fuel storage baskets that will be stacked inside the MCOs. Approximately 400 MCOs and 21 70 baskets will be fabricated for this purpose. These MCOs, loaded with SNF, will be placed in interim storage in the Canister Storage Building (CSB) located in the 200 Area of the Hanford Site. The MCOs consist of different components/sub-assemblies that will be manufactured by one or more vendors. All component/sub-assemblies will be shipped to the Hanford Site Central Stores Warehouse, 2355 Stevens Drive, Building 1163 in the 1100 Area, for inspection and storage until these components are required at the CSB and K Basins. The MCO fuel storage baskets will be manufactured in the MCO basket fabrication shop located in Building 328 of the Hanford Site 300 Area. The MCO baskets will be inspected at the fabrication shop before shipment to the Central Stores Warehouse for storage. The MCO components and baskets will be stored as received from the manufacturer with specified protective coatings, wrappings, and packaging intact to maintain mechanical integrity of the components and to prevent corrosion. The components and baskets will be shipped as needed from the warehouse to the CSB and K Basins. This warehouse plan includes the requirements for receipt of MCO components and baskets from the manufacturers and storage at the Hanford Site Central Stores Warehouse. Transportation of the MCO components and baskets from the warehouse, unwrapping, and assembly of the MCOs are the responsibility of SNF Operations and are not included in this plan

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

    International Nuclear Information System (INIS)

    Werme, Lars

    2006-10-01

    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. 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.

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

    International Nuclear Information System (INIS)

    Bowyer, William H.

    2003-04-01

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

  20. 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)

    Landolt, D.; Davenport, A.; Payer, J.; Shoesmith, D.

    2009-01-01

    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

  1. 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

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

    International Nuclear Information System (INIS)

    La Pointe, Paul R.; Cladouhos, Trenton T.; Outters, Nils; Follin, Sven

    2000-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-04-01

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

  4. Thermo-mechanical FE-analysis of butt-welding of a Cu-Fe canister for spent nuclear fuel

    International Nuclear Information System (INIS)

    Josefson, B.L.; Karlsson, L.; Lindgren, L.E.; Jonsson, M.

    1992-10-01

    In the Swedish nuclear waste program it has been proposed that spent nuclear fuel shall be placed in composite copper-steel canisters. These canisters will be placed in holes in tunnels located some 500 m underground in a rock storage. The canisters consists of two cylinders of 4850 mm length, one inner cylinder made of steel and one outer cylinder made of copper. The outer diameter of the canister is 880 mm and the wall thickness for each cylinder is 50 mm. At the storage, the steel cylinder, which contains the spent nuclear fuel, is placed inside the copper cylinder. Thereafter, a copper end is butt welded to the copper cylinder using electron beam welding. To obtain penetration through the thickness with this weld method a backing ring is placed at the inside of the copper cylinder. In the present paper, the temperature, strain and stress fields present during welding and after cooling after welding are calculated numerically using the FE-code NIKE-2D. The aim is to use the results of the present calculations to estimate the risk for creep fracture during the subsequent design life. A large strain formulation is employed for the calculation of transient and residual stresses in the canister based on the calculated history of the temperature field present in the canister during the welding process. The contact algorithm available in NIKE-2D is used to detect possible contact between the steel and copper cylinders during the welding. To simplify the numerical calculations and reduce the computational time, rotational symmetry is assumed. For large gap distances between the steel and copper cylinders the residual stress field is calculated to have a shape similar to that observed in butt welded pipes with maximum axial stress values at the yield stress level. For small gap distances the backing ring will come in contact with the steel cylinder which leads to large residual stresses in the backing ring. The maximum accumulated plastic strain in the weld metal and

  5. Dry storage technologies: keys to choosing among metal casks, concrete shielded steel canister modules and vaults

    International Nuclear Information System (INIS)

    Roland, V.; Solignac, Y.; Chiguer, M.; Guenon, Y.

    2003-01-01

    time. Then the key criterion is maximum modularity. Furthermore, the up front capital costs requirement for this type of solution is minimal, so depending on the chosen discount rate of the investor, they have an additional attraction. Those smaller modules allow to change course in back end policy more easily. Priority of modularity yi