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Sample records for instrumented fuel assembly

  1. Portable instrument for inspecting irradiated nuclear fuel assemblies

    DOEpatents

    Nicholson, Nicholas; Dowdy, Edward J.; Holt, David M.; Stump, Jr., Charles J.

    1985-01-01

    A portable instrument for measuring induced Cerenkov radiation associated with irradiated nuclear fuel assemblies in a water-filled storage pond is disclosed. The instrument includes a photomultiplier tube and an image intensifier which are operable in parallel and simultaneously by means of a field lens assembly and an associated beam splitter. The image intensifier permits an operator to aim and focus the apparatus on a submerged fuel assembly. Once the instrument is aimed and focused, an illumination reading can be obtained with the photomultiplier tube. The instrument includes a lens cap with a carbon-14/phosphor light source for calibrating the apparatus in the field.

  2. Precharacterization Report for Instrumented Fuel Assembly (IFA)-527

    SciTech Connect

    Cunningham, M. E.; Bradley, E. R.; Daniel, J. L.; Davis, N. C.; Lanning, D. D.; Williford, R. E.

    1981-07-01

    This report is a resource document covering the rationale, design, fabrication, and preirradiation characterization of instrumented fuel assembly (IFA)-527. This assembly is being irradiated in the Halden Boiling Water Reactor (HBWR) in Norway as part of the Experimental Support and Development of Single-Rod Fuel Codes Program conducted by Pacific Northwest laboratory (PNL) and sponsored by the Fuel Behavior Research Branch of the U.S. Nuclear Regulatory Commission (NRC). Data from this assembly will be used to better understand light water reactor (LWR) fuel behavior under normal operating conditions.

  3. Differential Die-Away Instrument: Report on Fuel Assembly Mock-up Measurements with Neutron Generator

    SciTech Connect

    Goodsell, Alison Victoria; Swinhoe, Martyn Thomas; Henzl, Vladimir; Rael, Carlos D.; Desimone, David J.

    2014-09-18

    Fresh fuel experiments for the differential die-away (DDA) project were performed using a DT neutron generator, a 15x15 PWR fuel assembly, and nine 3He detectors in a water tank inside of a shielded cell at Los Alamos National Laboratory (LANL). Eight different fuel enrichments were created using low enriched (LEU) and depleted uranium (DU) dioxide fuel rods. A list-mode data acquisition system recorded the time-dependent signal and analysis of the DDA signal die-away time was performed. The die-away time depended on the amount of fissile material in the fuel assembly and the position of the detector. These experiments were performed in support of the spent nuclear fuel Next Generation Safeguards Initiative DDA project. Lessons learned from the fresh fuel DDA instrument experiments and simulations will provide useful information to the spent fuel project.

  4. Determination of spent nuclear fuel assembly multiplication with the differential die-away self-interrogation instrument

    NASA Astrophysics Data System (ADS)

    Kaplan, Alexis C.; Henzl, Vladimir; Menlove, Howard O.; Swinhoe, Martyn T.; Belian, Anthony P.; Flaska, Marek; Pozzi, Sara A.

    2014-09-01

    We present a novel method for determining the multiplication of a spent nuclear fuel assembly with a Differential Die-Away Self-Interrogation (DDSI) instrument. The signal, which is primarily created by thermal neutrons, is measured with four 3He detector banks surrounding a spent fuel assembly. The Rossi-alpha distribution (RAD) at early times reflects coincident events from single fissions as well as fission chains. Because of this fact, the early time domain contains information about both the fissile material and spontaneous fission material in the assembly being measured. A single exponential function fit to the early time domain of the RAD has a die-away time proportional to the spent fuel assembly (SFA) multiplication. This correlation was tested by simulating assay of 44 different SFAs with the DDSI instrument. The SFA multiplication was determined with a variance of 0.7%.

  5. Portable instrument for inspecting irradiated nuclear-fuel assemblies in a water-filled storage pond by measurement of induced Cerenkov radiation

    DOEpatents

    Nicholson, N.; Dowdy, E.J.; Holt, D.M.; Stump, C.J. Jr.

    1982-05-13

    A portable instrument for measuring induced Cerenkov radiation associated with irradiated nuclear fuel assemblies in a water-filled storage pond is disclosed. The instrument includes a photomultiplier tube and an image intensifier which are operable in parallel and simultaneously by means of a field lens assembly and an associated beam splitter. The image intensifier permits an operator to aim and focus the apparatus on a submerged fuel assembly. Once the instrument is aimed and focused, an illumination reading can be obtained with the photomultiplier tube. The instrument includes a lens cap with a carbon-14/phosphor light source for calibrating the apparatus in the field.

  6. Determination of total plutonium content in spent nuclear fuel assemblies with the differential die-away self-interrogation instrument

    NASA Astrophysics Data System (ADS)

    Kaplan, Alexis C.; Henzl, Vladimir; Menlove, Howard O.; Swinhoe, Martyn T.; Belian, Anthony P.; Flaska, Marek; Pozzi, Sara A.

    2014-11-01

    As a part of the Next Generation Safeguards Initiative Spent Fuel project, we simulate the response of the Differential Die-away Self-Interrogation (DDSI) instrument to determine total elemental plutonium content in an assayed spent nuclear fuel assembly (SFA). We apply recently developed concepts that relate total plutonium mass with SFA multiplication and passive neutron count rate. In this work, the multiplication of the SFA is determined from the die-away time in the early time domain of the Rossi-Alpha distributions measured directly by the DDSI instrument. We utilize MCNP to test the method against 44 pressurized water reactor SFAs from a simulated spent fuel library with a wide dynamic range of characteristic parameters such as initial enrichment, burnup, and cooling time. Under ideal conditions, discounting possible errors of a real world measurement, a root mean square agreement between true and determined total Pu mass of 2.1% is achieved.

  7. Ultrasonic cleaning of fuel assemblies

    SciTech Connect

    Kondoh, Keisuke; Fujita, Chitoshi; Sakai, Hitoshi

    1994-12-31

    During fuel transportation, contamination of the transfer cask can lead to radiation dosage. That is radioactive crud becomes detached from the fuel surface and is deposited inside the cask. To avoid this at the Tsuruga Power Station Unit 1, crud was removed from fuel assemblies in advance of fuel transportation work. An ultrasonic cleaning process was adopted for this purpose; ultrasonic methods excel over other methods for this type of cleaning. Our process is also able to clean fuel assemblies without removing the channel box. Since this is the first time that the ultrasonic method was applied to fuel assemblies at the light water reactor in Japan on a large scale, the efficiency and the impact on plant instrumentation of the method were examined by performing preliminary test. Based on these tests, an optimum cleaning procedure was established.

  8. Monte Carlo simulations of differential die-away instrument for determination of fissile content in spent fuel assemblies

    NASA Astrophysics Data System (ADS)

    Lee, Tae-Hoon; Menlove, Howard O.; Swinhoe, Martyn T.; Tobin, Stephen J.

    2011-10-01

    The differential die-away (DDA) technique has been simulated by using the MCNPX code to quantify its capability of measuring the fissile content in spent fuel assemblies. For 64 different spent fuel cases of various initial enrichment, burnup and cooling time, the count rate and signal to background ratios of the DDA system were obtained, where neutron backgrounds are mainly coming from the 244Cm of the spent fuel. To quantify the total fissile mass of spent fuel, a concept of the effective 239Pu mass was introduced by weighing the relative contribution to the signal of 235U and 241Pu compared to 239Pu and the calibration curves of DDA count rate vs. 239Pu eff were obtained by using the MCNPX code. With a deuterium-tritium (DT) neutron generator of 10 9 n/s strength, signal to background ratios of sufficient magnitude are acquired for a DDA system with the spent fuel assembly in water.

  9. Carburetor fuel discharge assembly

    SciTech Connect

    Yost, R.M.

    1993-06-29

    An improved carburetor for use on an internal combustion engine is described, the carburetor having an airflow passage and fuel discharge means for admitting fuel into the airflow passage for mixing the fuel with air flowing in the airflow passage to form a fuel/air mixture to be supplied to the combustion chamber(s) of the engine, the fuel discharge means including a fuel discharge assembly which comprises a hollow discharge tube and fuel supplying means connected to the discharge tube for admitting fuel into the interior of the discharge tube, wherein the discharge tube has a longitudinal internal bore in fluid communication with the fuel supplying means, wherein the internal bore extends between an inlet that is closest to the fuel supplying means and an outlet that is furthest from the fuel supplying means with the outlet of the bore being located within the airflow passage of the carburetor to supply fuel into this passage after the fuel passes from the fuel supplying means through the internal bore of the discharge tube, wherein the improvement relates to the fuel discharge assembly and comprises: a hollow fuel flow guide tube telescopically received inside the internal bore of the discharge tube, wherein the fuel flow guide tube extends from approximately the location of the inlet of the bore up at least a portion of the length of the bore towards the outlet of the bore to conduct fuel from the fuel supplying means into the bore of the discharge tube.

  10. Fuel nozzle assembly

    DOEpatents

    Johnson, Thomas Edward; Ziminsky, Willy Steve; Lacey, Benjamin Paul; York, William David; Stevenson, Christian Xavier

    2011-08-30

    A fuel nozzle assembly is provided. The assembly includes an outer nozzle body having a first end and a second end and at least one inner nozzle tube having a first end and a second end. One of the nozzle body or nozzle tube includes a fuel plenum and a fuel passage extending therefrom, while the other of the nozzle body or nozzle tube includes a fuel injection hole slidably aligned with the fuel passage to form a fuel flow path therebetween at an interface between the body and the tube. The nozzle body and the nozzle tube are fixed against relative movement at the first ends of the nozzle body and nozzle tube, enabling the fuel flow path to close at the interface due to thermal growth after a flame enters the nozzle tube.

  11. FUEL ASSEMBLY SHAKER TEST SIMULATION

    SciTech Connect

    Klymyshyn, Nicholas A.; Sanborn, Scott E.; Adkins, Harold E.; Hanson, Brady D.

    2013-05-30

    This report describes the modeling of a PWR fuel assembly under dynamic shock loading in support of the Sandia National Laboratories (SNL) shaker test campaign. The focus of the test campaign is on evaluating the response of used fuel to shock and vibration loads that a can occur during highway transport. Modeling began in 2012 using an LS-DYNA fuel assembly model that was first created for modeling impact scenarios. SNL’s proposed test scenario was simulated through analysis and the calculated results helped guide the instrumentation and other aspects of the testing. During FY 2013, the fuel assembly model was refined to better represent the test surrogate. Analysis of the proposed loads suggested the frequency band needed to be lowered to attempt to excite the lower natural frequencies of the fuel assembly. Despite SNL’s expansion of lower frequency components in their five shock realizations, pretest predictions suggested a very mild dynamic response to the test loading. After testing was completed, one specific shock case was modeled, using recorded accelerometer data to excite the model. Direct comparison of predicted strain in the cladding was made to the recorded strain gauge data. The magnitude of both sets of strain (calculated and recorded) are very low, compared to the expected yield strength of the Zircaloy-4 material. The model was accurate enough to predict that no yielding of the cladding was expected, but its precision at predicting micro strains is questionable. The SNL test data offers some opportunity for validation of the finite element model, but the specific loading conditions of the testing only excite the fuel assembly to respond in a limited manner. For example, the test accelerations were not strong enough to substantially drive the fuel assembly out of contact with the basket. Under this test scenario, the fuel assembly model does a reasonable job of approximating actual fuel assembly response, a claim that can be verified through

  12. Direct Measurement of Initial Enrichment and Burn-up of Spent Fuel Assembly with a Differential Die-Away Technique Based Instrument

    SciTech Connect

    Henzl, Vladimir; Swinhoe, Martyn T.; Tobin, Stephen J.

    2012-07-16

    A key objective of the Next Generation Safeguards Initiative (NGSI) is to utilize non-destructive assay (NDA) techniques to determine the elemental plutonium (Pu) content in a commercial-grade nuclear spent fuel assembly (SFA). In the third year of the NGSI Spent Fuel NDA project, the research focus is on the integration of a few NDA techniques. One of the reoccurring challenges to the accurate determination of Pu content has been the explicit dependence of the measured signal on the presence of neutron absorbers which build up in the assembly in accordance with its operating and irradiation history. The history of any SFA is often summarized by the parameters of burn-up (BU), initial enrichment (IE) and cooling time (CT). While such parameters can typically be provided by the operator, the ability to directly measure and verify them would significantly enhance the autonomy of the IAEA inspectorate. Within this paper, we demonstrate that an instrument based on a Differential Die-Away technique is in principle capable of direct measurement of IE and, should the CT be known, also the BU.

  13. Determination of total Pu content in a Spent Fuel Assembly by Measuring Passive Neutron Count rate and Multiplication with the Differential Die-Away Instrument

    SciTech Connect

    Henzl, Vladimir; Croft, Stephen; Swinhoe, Martyn T.; Tobin, Stephen J.

    2012-07-18

    A key objective of the Next Generation Safeguards Initiative (NGSI) is to evaluate and develop non-destructive assay (NDA) techniques to determine the elemental plutonium content in a commercial-grade nuclear spent fuel assembly (SFA) [1]. Within this framework, we investigate by simulation a novel analytical approach based on combined information from passive measurement of the total neutron count rate of a SFA and its multiplication determined by the active interrogation using an instrument based on a Differential Die-Away technique (DDA). We use detailed MCNPX simulations across an extensive set of SFA characteristics to establish the approach and demonstrate its robustness. It is predicted that Pu content can be determined by the proposed method to a few %.

  14. Fuel assembly for nuclear reactors

    DOEpatents

    Creagan, Robert J.; Frisch, Erling

    1977-01-01

    A new and improved fuel assembly is formed to minimize the amount of parasitic structural material wherein a plurality of hollow tubular members are juxtaposed to the fuel elements of the assembly. The tubular members may serve as guide tubes for control elements and are secured to a number of longitudinally spaced grid members along the fuel assembly. The grid members include means thereon engaging each of the fuel elements to laterally position the fuel elements in a predetermined array. Openings in the bottom of each hollow member serve as a shock absorber to cushion shock transmitted to the structure when the control elements are rapidly inserted in their corresponding tubular members.

  15. Upgraded Fuel Assemblies for BWRs

    SciTech Connect

    Garner, N.L.; Rentmeister, T.; Lippert, H.J.; Mollard, P.

    2007-07-01

    Established with engineering and manufacturing operations in the US and Europe, AREVA NP has been and is supplying nuclear fuel assemblies and associated core components to light water reactors worldwide, representing today more than 170,000 fuel assemblies on the world market and more than 56,000 fuel assemblies for BWR plants. Since first delivered in 1992, ATRIUM{sup TM}(1)10 fuel assemblies have now been supplied to a total of 28 BWR plants in the US, Europe, and Asia resulting in an operating experience over 16 000 fuel assemblies. In the spring of 2001, a BWR record burnup of 71 MWd/kgU was reached by four lead fuel assemblies after eight operating cycles. More recently, ATRIUM 10XP and ATRIUM 10XM fuel assemblies featuring changes in their characteristics and exhibiting upgraded behavior have been delivered to several utilities worldwide. This success story has been made possible thanks to a continuous improvement process with the aim of further upgrading BWR fuel assembly performance and reliability. An overview is given on current AREVA advanced BWR fuel supply regarding: - advanced designs to tailor product selection to specific operating strategies; - performance capabilities of each advanced design option; - testing and operational experience for these advanced designs; - upgraded features available for inclusion with advanced designs. (authors)

  16. Direct Measurement of Initial Enrichment, Burn-up and Cooling Time of Spent Fuel Assembly with a Differential Die-Away Technique Based Instrument

    SciTech Connect

    Henzl, Vladimir; Swinhoe, Martyn T.; Tobin, Stephen J.

    2012-07-13

    An outline of this presentation of what a Differential Die-Away (DDA) instrument can do are: (1) Principle of operation of DDA instrument; (2) Determination of initial enrichment (IE) ({sigma} < 5%); (3) Determination of burn up (BU) ({sigma} {approx} 6%); (4) Determination of cooling time (CT) ({sigma} {approx} 20-50%); and (5) DDA instrument as a standalone device. DDA response (fresh fuel vs. spent fuel) is: (1) Fresh fuel => DDA response increases (die-away time is longer) with increasing fissile content; and (2) Spent fuel => DDA response decreases (die-away time is shorter) with higher burn-up (i.e. more neutron absorbers present).

  17. Fuel cell design and assembly

    NASA Technical Reports Server (NTRS)

    Myerhoff, Alfred (Inventor)

    1984-01-01

    The present invention is directed to a novel bipolar cooling plate, fuel cell design and method of assembly of fuel cells. The bipolar cooling plate used in the fuel cell design and method of assembly has discrete opposite edge and means carried by the plate defining a plurality of channels extending along the surface of the plate toward the opposite edges. At least one edge of the channels terminates short of the edge of the plate defining a recess for receiving a fastener.

  18. Microfabricated field calibration assembly for analytical instruments

    DOEpatents

    Robinson, Alex L.; Manginell, Ronald P.; Moorman, Matthew W.; Rodacy, Philip J.; Simonson, Robert J.

    2011-03-29

    A microfabricated field calibration assembly for use in calibrating analytical instruments and sensor systems. The assembly comprises a circuit board comprising one or more resistively heatable microbridge elements, an interface device that enables addressable heating of the microbridge elements, and, in some embodiments, a means for positioning the circuit board within an inlet structure of an analytical instrument or sensor system.

  19. Nuclear core and fuel assemblies

    DOEpatents

    Downs, Robert E.

    1981-01-01

    A fast flux nuclear core of a plurality of rodded, open-lattice assemblies having a rod pattern rotated relative to a rod support structure pattern. Elongated fuel rods are oriented on a triangular array and laterally supported by grid structures positioned along the length of the assembly. Initial inter-assembly contact is through strongbacks at the corners of the support pattern and peripheral fuel rods between adjacent assemblies are nested so as to maintain a triangular pitch across a clearance gap between the other portions of adjacent assemblies. The rod pattern is rotated relative to the strongback support pattern by an angle .alpha. equal to sin .sup.-1 (p/2c), where p is the intra-assembly rod pitch and c is the center-to-center spacing among adjacent assemblies.

  20. Pattern fuel assembly loading system

    SciTech Connect

    Ahmed, H.J.; Gerkey, K.S.; Miller, T.W.; Wylie, M.E.

    1986-12-02

    This patent describes an interactive system for facilitating preloading of fuel rods into magazines, which comprises: an operator work station adapted for positioning between a supply of fuel rods of predetermined types, and the magazine defining grid locations for a predetermined fuel assembly; display means associated with the work station; scanner means associated with the work station and adapted for reading predetermined information accompanying the fuel rods; a rectangular frame adapted for attachment to one end of the fuel assembly loading magazine; prompter/detector means associated with the frame for detecting insertion of a fuel rod into the magazine; and processing means responsive to the scanner means and the sensing means for prompting the operator via the display means to pre-load the fuel rods into desired grid locations in the magazine. An apparatus is described for facilitating pre-loading of fuel rods in predetermined grid locations of a fuel assembly loading magazine, comprising: a rectangular frame adapted for attachment to one end of the fuel assembly loading magazine; and means associated with the frame for detecting insertion of fuel rods into the magazine.

  1. FUEL ROD ASSEMBLY

    DOEpatents

    Hutter, E.

    1959-09-01

    A cluster of nuclear fuel rods aod a tubular casing through which a coolant flows in heat-change contact with the ruel rods are described. The casting is of trefoil section and carries the fuel rods, each of which has two fin engaging the serrated fins of the other two fuel rods, whereby the fuel rods are held in the casing and are interlocked against relative longitudinal movement.

  2. Determination of the plutonium content in a spent fuel assembly by passive and active interrogation using a differential die-away instrument

    NASA Astrophysics Data System (ADS)

    Henzl, V.; Croft, S.; Richard, J.; Swinhoe, M. T.; Tobin, S. J.

    2013-06-01

    In this paper, we present a novel approach to estimating the total plutonium content in a spent fuel assembly (SFA) that is based on combining information from a passive measurement of the total neutron count rate (PN) of the assayed SFA and a measure of its multiplication. While PN can be measured essentially with any non-destructive assay (NDA) technique capable of neutron detection, the measure of multiplication is, in our approach, determined by means of active interrogation using an instrument based on the Differential Die-Away technique (DDA). The DDA is a NDA technique developed within the U.S. Department of Energy's Next Generation Safeguards Initiative (NGSI) project focused on the utilization of NDA techniques to determine the elemental plutonium content in commercial nuclear SFA's [1]. This approach was adopted since DDA also allows determination of other SFA characteristics, such as burnup, initial enrichment, and cooling time, and also allows for detection of certain types of diversion of nuclear material. The quantification of total plutonium is obtained using an analytical correlation function in terms of the observed PN and active multiplication. Although somewhat similar approaches relating Pu content with PN have been adopted in the past, we demonstrate by extensive simulation of the fuel irradiation and NDA process that our analytical method is independent of explicit knowledge of the initial enrichment, burnup, and an absolute value of the SFA's reactivity (i.e. multiplication factor). We show that when tested with MCNPX™ simulations comprising the 64 SFA NGSI Spent Fuel Library-1 we were able to determine elemental plutonium content, using just a few calibration parameters, with an average variation in the prediction of around 1-2% across the wide dynamic range of irradiation history parameters used, namely initial enrichment (IE=2-5%), burnup (BU=15-60 GWd/tU) and cooling time (CT=1-80 y). In this paper we describe the basic approach and the

  3. Nuclear reactor composite fuel assembly

    DOEpatents

    Burgess, Donn M.; Marr, Duane R.; Cappiello, Michael W.; Omberg, Ronald P.

    1980-01-01

    A core and composite fuel assembly for a liquid-cooled breeder nuclear reactor including a plurality of elongated coextending driver and breeder fuel elements arranged to form a generally polygonal bundle within a thin-walled duct. The breeder elements are larger in cross section than the driver elements, and each breeder element is laterally bounded by a number of the driver elements. Each driver element further includes structure for spacing the driver elements from adjacent fuel elements and, where adjacent, the thin-walled duct. A core made up of the fuel elements can advantageously include fissile fuel of only one enrichment, while varying the effective enrichment of any given assembly or core region, merely by varying the relative number and size of the driver and breeder elements.

  4. Cooling assembly for fuel cells

    DOEpatents

    Kaufman, Arthur; Werth, John

    1990-01-01

    A cooling assembly for fuel cells having a simplified construction whereby coolant is efficiently circulated through a conduit arranged in serpentine fashion in a channel within a member of such assembly. The channel is adapted to cradle a flexible, chemically inert, conformable conduit capable of manipulation into a variety of cooling patterns without crimping or otherwise restricting of coolant flow. The conduit, when assembled with the member, conforms into intimate contact with the member for good thermal conductivity. The conduit is non-corrodible and can be constructed as a single, manifold-free, continuous coolant passage means having only one inlet and one outlet.

  5. NUCLEAR REACTOR FUEL ELEMENT ASSEMBLY

    DOEpatents

    Stengel, F.G.

    1963-12-24

    A method of fabricating nuclear reactor fuel element assemblies having a plurality of longitudinally extending flat fuel elements in spaced parallel relation to each other to form channels is presented. One side of a flat side plate is held contiguous to the ends of the elements and a welding means is passed along the other side of the platertransverse to the direction of the longitudinal extension of the elements. The setting and speed of travel of the welding means is set to cause penetration of the side plate with welds at bridge the gap in each channel between adjacent fuel elements with a weld-through bubble of predetermined size. The fabrication of a high strength, dependable fuel element is provided, and the reduction of distortion and high production costs are facilitated by this method. (AEC)

  6. Improved nuclear fuel assembly grid spacer

    DOEpatents

    Marshall, John; Kaplan, Samuel

    1977-01-01

    An improved fuel assembly grid spacer and method of retaining the basic fuel rod support elements in position within the fuel assembly containment channel. The improvement involves attachment of the grids to the hexagonal channel and of forming the basic fuel rod support element into a grid structure, which provides a design which is insensitive to potential channel distortion (ballooning) at high fluence levels. In addition the improved method eliminates problems associated with component fabrication and assembly.

  7. Internal reforming fuel cell assembly with simplified fuel feed

    DOEpatents

    Farooque, Mohammad; Novacco, Lawrence J.; Allen, Jeffrey P.

    2001-01-01

    A fuel cell assembly in which fuel cells adapted to internally reform fuel and fuel reformers for reforming fuel are arranged in a fuel cell stack. The fuel inlet ports of the fuel cells and the fuel inlet ports and reformed fuel outlet ports of the fuel reformers are arranged on one face of the fuel cell stack. A manifold sealing encloses this face of the stack and a reformer fuel delivery system is arranged entirely within the region between the manifold and the one face of the stack. The fuel reformer has a foil wrapping and a cover member forming with the foil wrapping an enclosed structure.

  8. Apparatus for shearing spent nuclear fuel assemblies

    DOEpatents

    Weil, Bradley S.; Metz, III, Curtis F.

    1980-01-01

    A method and apparatus are described for shearing spent nuclear fuel assemblies of the type comprising an array of fuel pins disposed within an outer metal shell or shroud. A spent fuel assembly is first compacted in a known manner and then incrementally sheared using fixed and movable shear blades having matched laterally projecting teeth which slidably intermesh to provide the desired shearing action. Incremental advancement of the fuel assembly after each shear cycle is limited to a distance corresponding to the lateral projection of the teeth to ensure fuel assembly breakup into small uniform segments which are amenable to remote chemical processing.

  9. Simulated nuclear reactor fuel assembly

    DOEpatents

    Berta, V.T.

    1993-04-06

    An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end.

  10. Simulated nuclear reactor fuel assembly

    DOEpatents

    Berta, Victor T.

    1993-01-01

    An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end.

  11. Fuel rod assembly to manifold attachment

    DOEpatents

    Donck, Harry A.; Veca, Anthony R.; Snyder, Jr., Harold J.

    1980-01-01

    A fuel element is formed with a plurality of fuel rod assemblies detachably connected to an overhead support with each of the fuel rod assemblies having a gas tight seal with the support to allow internal fission gaseous products to flow without leakage from the fuel rod assemblies into a vent manifold passageway system on the support. The upper ends of the fuel rod assemblies are located at vertically extending openings in the support and upper threaded members are threaded to the fuel rod assemblies to connect the latter to the support. The preferred threaded members are cap nuts having a dome wall encircling an upper threaded end on the fuel rod assembly and having an upper sealing surface for sealing contact with the support. Another and lower seal is achieved by abutting a sealing surface on each fuel rod assembly with the support. A deformable portion on the cap nut locks the latter against inadvertent turning off the fuel rod assembly. Orienting means on the fuel rod and support primarily locates the fuel rods azimuthally for reception of a deforming tool for the cap nut. A cross port in the fuel rod end plug discharges into a sealed annulus within the support, which serves as a circumferential chamber, connecting the manifold gas passageways in the support.

  12. Fuel assembly self-excited vibration and test methodology

    SciTech Connect

    Lu, R.Y.; Broach, K. D.; McEvoy, J. J.

    2004-07-01

    PWR fuel assemblies normally experience low amplitude, random vibration under normal reactor flow conditions. This normal fuel assembly vibration has almost no impact on grid-rod fretting wear. However, some fuel assembly designs experience a high resonant fuel assembly vibration under normal axial flow conditions. This anomalous fuel assembly vibration is defined as fuel assembly self-excitation vibration (FASE), because the assembly vibrates resonantly without any external periodic excitation force. Fuel assembly self-excitation vibration can cause severe grid-rod fretting if the assembly operates at the flow rate, which causes high fuel assembly vibration. This paper will describe the characteristics of fuel assembly self-excitation vibration and the test methodology to identify the fuel assembly vibration. Several fuel assembly designs are compared under standard test conditions. The causes for the fuel assembly self-excitation vibration are analyzed and discussed. The test acceptance criteria are defined for newly developed PWR fuel assemblies. (authors)

  13. Plumb nozzle for nuclear fuel assembly

    SciTech Connect

    Silverblatt, B.L.

    1986-02-25

    An elongated nuclear reactor fuel assembly is described having an asymmetric weight distribution across a cross section, including a nozzle affixed to one end of the fuel assembly having lifting surfaces formed thereon on which the fuel assembly can be supported when suspended from the surfaces. At least one of the lifting surfaces is located at a first elevation relative to the longitudinal axis of the fuel assembly. A second of the lifting surfaces is located at a second elevation different from the first elevation wherein the difference in the first and second elevations is sized to offset the asymmetric weight distribution when the fuel assembly is supported from the first and second surface so that when so supported the fuel assembly will hang plumb.

  14. Method for shearing spent nuclear fuel assemblies

    DOEpatents

    Weil, Bradley S.; Watson, Clyde D.

    1977-01-01

    A method is disclosed for shearing spent nuclear fuel assemblies of the type wherein a plurality of long metal tubes packed with ceramic fuel are supported in a spaced apart relationship within an outer metal shell or shroud which provides structural support to the assembly. Spent nuclear fuel assemblies are first compacted in a stepwise manner between specially designed gag-compactors and then sheared into short segments amenable to chemical processing by shear blades contoured to mate with the compacted surface of the fuel assembly.

  15. Fuel fire tests of selected assemblies

    NASA Astrophysics Data System (ADS)

    Kydd, G.; Spindola, K.; Askew, G. K.

    1982-04-01

    A varing assortment of clothing assemblies was tested in the Fuel Fire Test Facility at the Naval Air Development Center. Included was a Nomex-Kevlar Cloque Coverall which had relatively good protection from fuel flames.

  16. Locking support for nuclear fuel assemblies

    DOEpatents

    Ledin, Eric

    1980-01-01

    A locking device for supporting and locking a nuclear fuel assembly within a cylindrical bore formed by a support plate, the locking device including a support and locking sleeve having upwardly extending fingers forming wedge shaped contact portions arranged for interaction between an annular tapered surface on the fuel assembly and the support plate bore as well as downwardly extending fingers having wedge shaped contact portions arranged for interaction between an annularly tapered surface on the support plate bore and the fuel assembly whereby the sleeve tends to support and lock the fuel assembly in place within the bore by its own weight while facilitating removal and/or replacement of the fuel assembly.

  17. LMFBR fuel assembly design for HCDA fuel dispersal

    DOEpatents

    Lacko, Robert E.; Tilbrook, Roger W.

    1984-01-01

    A fuel assembly for a liquid metal fast breeder reactor having an upper axial blanket region disposed in a plurality of zones within the fuel assembly. The characterization of a zone is dependent on the height of the axial blanket region with respect to the active fuel region. The net effect of having a plurality of zones is to establish a dispersal flow path for the molten materials resulting during a core meltdown accident. Upward flowing molten material can escape from the core region and/or fuel assembly without solidifying on the surface of fuel rods due to the heat sink represented by blanket region pellets.

  18. Thermal Analysis of a TREAT Fuel Assembly

    SciTech Connect

    Papadias, Dionissios; Wright, Arthur E.

    2014-07-09

    The objective of this study was to explore options as to reduce peak cladding temperatures despite an increase in peak fuel temperatures. A 3D thermal-hydraulic model for a single TREAT fuel assembly was benchmarked to reproduce results obtained with previous thermal models developed for a TREAT HEU fuel assembly. In exercising this model, and variants thereof depending on the scope of analysis, various options were explored to reduce the peak cladding temperatures.

  19. Modular fuel-cell stack assembly

    DOEpatents

    Patel, Pinakin

    2010-07-13

    A fuel cell assembly having a plurality of fuel cells arranged in a stack. An end plate assembly abuts the fuel cell at an end of said stack. The end plate assembly has an inlet area adapted to receive an exhaust gas from the stack, an outlet area and a passage connecting the inlet area and outlet area and adapted to carry the exhaust gas received at the inlet area from the inlet area to the outlet area. A further end plate assembly abuts the fuel cell at a further opposing end of the stack. The further end plate assembly has a further inlet area adapted to receive a further exhaust gas from the stack, a further outlet area and a further passage connecting the further inlet area and further outlet area and adapted to carry the further exhaust gas received at the further inlet area from the further inlet area to the further outlet area.

  20. PWR and BWR spent fuel assembly gamma spectra measurements

    DOE PAGESBeta

    Vaccaro, S.; Tobin, Stephen J.; Favalli, Andrea; Grogan, Brandon R.; Jansson, Peter; Liljenfeldt, Henrik; Mozin, Vladimir; Hu, Jianwei; Schwalbach, P.; Sjoland, A.; et al

    2016-07-17

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detectmore » the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. As a result, to compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.« less

  1. A classification scheme for LWR fuel assemblies

    SciTech Connect

    Moore, R.S.; Williamson, D.A.; Notz, K.J.

    1988-11-01

    With over 100 light water nuclear reactors operating nationwide, representing designs by four primary vendors, and with reload fuel manufactured by these vendors and additional suppliers, a wide variety of fuel assembly types are in existence. At Oak Ridge National Laboratory, both the Systems Integration Program and the Characteristics Data Base project required a classification scheme for these fuels. This scheme can be applied to other areas and is expected to be of value to many Office of Civilian Radioactive Waste Management programs. To develop the classification scheme, extensive information on the fuel assemblies that have been and are being manufactured by the various nuclear fuel vendors was compiled, reviewed, and evaluated. It was determined that it is possible to characterize assemblies in a systematic manner, using a combination of physical factors. A two-stage scheme was developed consisting of 79 assembly types, which are grouped into 22 assembly classes. The assembly classes are determined by the general design of the reactor cores in which the assemblies are, or were, used. The general BWR and PWR classes are divided differently but both are based on reactor core configuration. 2 refs., 15 tabs.

  2. Polymer electrolyte membrane assembly for fuel cells

    NASA Technical Reports Server (NTRS)

    Yen, Shiao-Ping S. (Inventor); Kindler, Andrew (Inventor); Yavrouian, Andre (Inventor); Halpert, Gerald (Inventor)

    2002-01-01

    An electrolyte membrane for use in a fuel cell can contain sulfonated polyphenylether sulfones. The membrane can contain a first sulfonated polyphenylether sulfone and a second sulfonated polyphenylether sulfone, wherein the first sulfonated polyphenylether and the second sulfonated polyphenylether sulfone have equivalent weights greater than about 560, and the first sulfonated polyphenylether and the second sulfonated polyphenylether sulfone also have different equivalent weights. Also, a membrane for use in a fuel cell can contain a sulfonated polyphenylether sulfone and an unsulfonated polyphenylether sulfone. Methods for manufacturing a membrane electrode assemblies for use in fuel cells can include roughening a membrane surface. Electrodes and methods for fabricating such electrodes for use in a chemical fuel cell can include sintering an electrode. Such membranes and electrodes can be assembled into chemical fuel cells.

  3. Polymer electrolyte membrane assembly for fuel cells

    NASA Technical Reports Server (NTRS)

    Yen, Shiao-Ping S. (Inventor); Kindler, Andrew (Inventor); Yavrouian, Andre (Inventor); Halpert, Gerald (Inventor)

    2000-01-01

    An electrolyte membrane for use in a fuel cell can contain sulfonated polyphenylether sulfones. The membrane can contain a first sulfonated polyphenylether sulfone and a second sulfonated polyphenylether sulfone, wherein the first sulfonated polyphenylether and the second sulfonated polyphenylether sulfone have equivalent weights greater than about 560, and the first sulfonated polyphenylether and the second sulfonated polyphenylether sulfone also have different equivalent weights. Also, a membrane for use in a fuel cell can contain a sulfonated polyphenylether sulfone and an unsulfonated polyphenylether sulfone. Methods for manufacturing a membrane electrode assemblies for use in fuel cells can include roughening a membrane surface. Electrodes and methods for fabricating such electrodes for use in a chemical fuel cell can include sintering an electrode. Such membranes and electrodes can be assembled into chemical fuel cells.

  4. SOLID GAS SUSPENSION NUCLEAR FUEL ASSEMBLY

    DOEpatents

    Schluderberg, D.C.; Ryon, J.W.

    1962-05-01

    A fuel assembly is designed for use in a gas-suspension cooled nuclear fuel reactor. The coolant fluid is an inert gas such as nitrogen or helium with particles such as carbon suspended therein. The fuel assembly is contained within an elongated pressure vessel extending down into the reactor. The fuel portion is at the lower end of the vessel and is constructed of cylindrical segments through which the coolant passes. Turbulence promotors within the passageways maintain the particles in agitation to increase its ability to transfer heat away from the outer walls. Shielding sections and alternating passageways above the fueled portion limit the escape of radiation out of the top of the vessel. (AEC)

  5. Modular fuel-cell stack assembly

    DOEpatents

    Patel, Pinakin; Urko, Willam

    2008-01-29

    A modular multi-stack fuel-cell assembly in which the fuel-cell stacks are situated within a containment structure and in which a gas distributor is provided in the structure and distributes received fuel and oxidant gases to the stacks and receives exhausted fuel and oxidant gas from the stacks so as to realize a desired gas flow distribution and gas pressure differential through the stacks. The gas distributor is centrally and symmetrically arranged relative to the stacks so that it itself promotes realization of the desired gas flow distribution and pressure differential.

  6. Fuel cell with electrolyte matrix assembly

    DOEpatents

    Kaufman, Arthur; Pudick, Sheldon; Wang, Chiu L.

    1988-01-01

    This invention is directed to a fuel cell employing a substantially immobilized electrolyte imbedded therein and having a laminated matrix assembly disposed between the electrodes of the cell for holding and distributing the electrolyte. The matrix assembly comprises a non-conducting fibrous material such as silicon carbide whiskers having a relatively large void-fraction and a layer of material having a relatively small void-fraction.

  7. Instrumentation Report No. 3: performance and reliability of instrumentation deployed for the Spent Fuel Test - Climax

    SciTech Connect

    Patrick, W.C.; Rector, N.L.; Scarafiotti, J.J.

    1984-12-01

    A demonstration of the short-term storage and subsequent retrieval of spent nuclear fuel assemblies was successfully completed at the US Department of Energy`s Nevada Test Site. Nearly 1000 instruments were deployed to monitor the temperature of rock, air, and metallic components of the test; displacements and stress changes within the rock mass; radiation dosage to personnel and to the rock; thermal energy input; characteristics of the ventilation airstream; and the operational status of the test. Careful selection, installation, calibration, and maintenance of these instruments ensured the acquisition of about 15.3 x 10{sup 6} high-quality data points. With few exceptions, the performance and reliability of the instrumentation and associated data acquisition system (DAS) were within specified acceptable limits. Details of the performance and reliability of the instrumentation are discussed in this report. 42 figs., 32 tabs.

  8. Fuel cell assembly with electrolyte transport

    DOEpatents

    Chi, Chang V.

    1983-01-01

    A fuel cell assembly wherein electrolyte for filling the fuel cell matrix is carried via a transport system comprising a first passage means for conveying electrolyte through a first plate and communicating with a groove in a second plate at a first point, the first and second plates together sandwiching the matrix, and second passage means acting to carry electrolyte exclusively through the second plate and communicating with the groove at a second point exclusive of the first point.

  9. Membrane electrode assembly for a fuel cell

    NASA Technical Reports Server (NTRS)

    Prakash, Surya (Inventor); Narayanan, Sekharipuram R. (Inventor); Atti, Anthony (Inventor); Olah, George (Inventor); Smart, Marshall C. (Inventor)

    2006-01-01

    A catalyst ink for a fuel cell including a catalytic material and poly(vinylidene fluoride). The ink may be applied to a substrate to form an electrode, or bonded with other electrode layers to form a membrane electrode assembly (MEA).

  10. Measurement Protocols for Optimized Fuel Assembly Tags

    SciTech Connect

    Gerlach, David C.; Mitchell, Mark R.; Reid, Bruce D.; Gesh, Christopher J.; Hurley, David E.

    2008-11-01

    This report describes the measurement protocols for optimized tags that can be applied to standard fuel assemblies used in light water reactors. This report describes work performed by the authors at Pacific Northwest National Laboratory for NA-22 as part of research to identify specific signatures that can be developed to support counter-proliferation technologies.

  11. In-Situ Safeguards Verification of Low Burn-up Pressurized Water Reactor Spent Fuel Assemblies

    SciTech Connect

    Ham, Y S; Sitaraman, S; Park, I; Kim, J; Ahn, G

    2008-04-16

    A novel in-situ gross defect verification method for light water reactor spent fuel assemblies was developed and investigated by a Monte Carlo study. This particular method is particularly effective for old pressurized water reactor spent fuel assemblies that have natural uranium in their upper fuel zones. Currently there is no method or instrument that does verification of this type of spent fuel assemblies without moving the spent fuel assemblies from their storage positions. The proposed method uses a tiny neutron detector and a detector guiding system to collect neutron signals inside PWR spent fuel assemblies through guide tubes present in PWR assemblies. The data obtained in such a manner are used for gross defect verification of spent fuel assemblies. The method uses 'calibration curves' which show the expected neutron counts inside one of the guide tubes of spent fuel assemblies as a function of fuel burn-up. By examining the measured data in the 'calibration curves', the consistency of the operator's declaration is verified.

  12. Nuclear fuel assembly with coolant conducting tube

    SciTech Connect

    Dunlap, T. G.; Cearley, J. E.; Jameson, W. G. Jr.; Mefford, C. R.; Nelson, H. L.

    1983-12-13

    In a nuclear fuel assembly having a coolant conducting or water tube which also retains the spacers in axial position, the fuel rods experience greater axial growth with exposure than the water tube creating a risk that the water tube might become disengaged from the supporting tie plates. An arrangement for preventing such disengagement is described including lengthened end plug shanks for the water tube, a protective boss surrounding the lower end plug shank to protect it from flow induced vibration, a conical seat for the lower end plug and an arrangement for limiting upward movement of the water tube.

  13. FUEL ASSEMBLY SHAKER AND TRUCK TEST SIMULATION

    SciTech Connect

    Klymyshyn, Nicholas A.; Jensen, Philip J.; Sanborn, Scott E.; Hanson, Brady D.

    2014-09-25

    This study continues the modeling support of the SNL shaker table task from 2013 and includes analysis of the SNL 2014 truck test campaign. Detailed finite element models of the fuel assembly surrogate used by SNL during testing form the basis of the modeling effort. Additional analysis was performed to characterize and filter the accelerometer data collected during the SNL testing. The detailed fuel assembly finite element model was modified to improve the performance and accuracy of the original surrogate fuel assembly model in an attempt to achieve a closer agreement with the low strains measured during testing. The revised model was used to recalculate the shaker table load response from the 2013 test campaign. As it happened, the results remained comparable to the values calculated with the original fuel assembly model. From this it is concluded that the original model was suitable for the task and the improvements to the model were not able to bring the calculated strain values down to the extremely low level recorded during testing. The model needs more precision to calculate strains that are so close to zero. The truck test load case had an even lower magnitude than the shaker table case. Strain gage data from the test was compared directly to locations on the model. Truck test strains were lower than the shaker table case, but the model achieved a better relative agreement of 100-200 microstrains (or 0.0001-0.0002 mm/mm). The truck test data included a number of accelerometers at various locations on the truck bed, surrogate basket, and surrogate fuel assembly. This set of accelerometers allowed an evaluation of the dynamics of the conveyance system used in testing. It was discovered that the dynamic load transference through the conveyance has a strong frequency-range dependency. This suggests that different conveyance configurations could behave differently and transmit different magnitudes of loads to the fuel even when travelling down the same road at

  14. Advanced membrane electrode assemblies for fuel cells

    DOEpatents

    Kim, Yu Seung; Pivovar, Bryan S.

    2012-07-24

    A method of preparing advanced membrane electrode assemblies (MEA) for use in fuel cells. A base polymer is selected for a base membrane. An electrode composition is selected to optimize properties exhibited by the membrane electrode assembly based on the selection of the base polymer. A property-tuning coating layer composition is selected based on compatibility with the base polymer and the electrode composition. A solvent is selected based on the interaction of the solvent with the base polymer and the property-tuning coating layer composition. The MEA is assembled by preparing the base membrane and then applying the property-tuning coating layer to form a composite membrane. Finally, a catalyst is applied to the composite membrane.

  15. Advanced membrane electrode assemblies for fuel cells

    SciTech Connect

    Kim, Yu Seung; Pivovar, Bryan S

    2014-02-25

    A method of preparing advanced membrane electrode assemblies (MEA) for use in fuel cells. A base polymer is selected for a base membrane. An electrode composition is selected to optimize properties exhibited by the membrane electrode assembly based on the selection of the base polymer. A property-tuning coating layer composition is selected based on compatibility with the base polymer and the electrode composition. A solvent is selected based on the interaction of the solvent with the base polymer and the property-tuning coating layer composition. The MEA is assembled by preparing the base membrane and then applying the property-tuning coating layer to form a composite membrane. Finally, a catalyst is applied to the composite membrane.

  16. Fuel injection assembly for gas turbine engine combustor

    NASA Technical Reports Server (NTRS)

    Candy, Anthony J. (Inventor); Glynn, Christopher C. (Inventor); Barrett, John E. (Inventor)

    2002-01-01

    A fuel injection assembly for a gas turbine engine combustor, including at least one fuel stem, a plurality of concentrically disposed tubes positioned within each fuel stem, wherein a cooling supply flow passage, a cooling return flow passage, and a tip fuel flow passage are defined thereby, and at least one fuel tip assembly connected to each fuel stem so as to be in flow communication with the flow passages, wherein an active cooling circuit for each fuel stem and fuel tip assembly is maintained by providing all active fuel through the cooling supply flow passage and the cooling return flow passage during each stage of combustor operation. The fuel flowing through the active cooling circuit is then collected so that a predetermined portion thereof is provided to the tip fuel flow passage for injection by the fuel tip assembly.

  17. A precision press-fit instrument for assembling small parts

    NASA Astrophysics Data System (ADS)

    Lou, Zhifeng; Wang, Xiaodong; You, Bo; Xu, Yang

    2015-02-01

    In the paper, a precision press-fit instrument for assembling small interference fitting parts is introduced, which includes pressing module and parts alignment module. The pressing module was used to clamp and position parts, and parts alignment module was used for the two parts' alignment. Through analyzing press-fit control method, component alignment and adjustment strategy, and machine vision device calibration method, the instrument meets the pressing requirements of precision small components. Finite element method is used to predict the reasonable range of press-fit force, and pressing result of the instrument is tested by experiments.

  18. Fuel injection assembly for use in turbine engines and method of assembling same

    SciTech Connect

    Berry, Jonathan Dwight; Johnson, Thomas Edward; York, William David; Uhm, Jong Ho

    2015-12-15

    A fuel injection assembly for use in a turbine engine is provided. The fuel injection assembly includes an end cover, an endcap assembly, a fluid supply chamber, and a plurality of tube assemblies positioned at the endcap assembly. Each of the tube assemblies includes housing having a fuel plenum and a cooling fluid plenum. The cooling fluid plenum is positioned downstream from the fuel plenum and separated from the fuel plenum by an intermediate wall. The plurality of tube assemblies also include a plurality of tubes that extends through the housing. Each of the plurality of tubes is coupled in flow communication with the fluid supply chamber and a combustion chamber positioned downstream from the tube assembly. The plurality of tube assemblies further includes an aft plate at a downstream end of the cooling fluid plenum. The plate includes at least one aperture.

  19. Interface ring for gas turbine fuel nozzle assemblies

    DOEpatents

    Fox, Timothy A.; Schilp, Reinhard

    2016-03-22

    A gas turbine combustor assembly including a combustor liner and a plurality of fuel nozzle assemblies arranged in an annular array extending within the combustor liner. The fuel nozzle assemblies each include fuel nozzle body integral with a swirler assembly, and the swirler assemblies each include a bellmouth structure to turn air radially inwardly for passage into the swirler assemblies. A radially outer removed portion of each of the bellmouth structures defines a periphery diameter spaced from an inner surface of the combustor liner, and an interface ring is provided extending between the combustor liner and the removed portions of the bellmouth structures at the periphery diameter.

  20. Impact of Nuclear Data Uncertainties on Calculated Spent Fuel Nuclide Inventories and Advanced NDA Instrument Response

    SciTech Connect

    Hu, Jianwei; Gauld, Ian C.

    2014-12-01

    The U.S. Department of Energy’s Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) project is nearing the final phase of developing several advanced nondestructive assay (NDA) instruments designed to measure spent nuclear fuel assemblies for the purpose of improving nuclear safeguards. Current efforts are focusing on calibrating several of these instruments with spent fuel assemblies at two international spent fuel facilities. Modelling and simulation is expected to play an important role in predicting nuclide compositions, neutron and gamma source terms, and instrument responses in order to inform the instrument calibration procedures. As part of NGSI-SF project, this work was carried out to assess the impacts of uncertainties in the nuclear data used in the calculations of spent fuel content, radiation emissions and instrument responses. Nuclear data is an essential part of nuclear fuel burnup and decay codes and nuclear transport codes. Such codes are routinely used for analysis of spent fuel and NDA safeguards instruments. Hence, the uncertainties existing in the nuclear data used in these codes affect the accuracies of such analysis. In addition, nuclear data uncertainties represent the limiting (smallest) uncertainties that can be expected from nuclear code predictions, and therefore define the highest attainable accuracy of the NDA instrument. This work studies the impacts of nuclear data uncertainties on calculated spent fuel nuclide inventories and the associated NDA instrument response. Recently developed methods within the SCALE code system are applied in this study. The Californium Interrogation with Prompt Neutron instrument was selected to illustrate the impact of these uncertainties on NDA instrument response.

  1. Impact of Nuclear Data Uncertainties on Calculated Spent Fuel Nuclide Inventories and Advanced NDA Instrument Response

    DOE PAGESBeta

    Hu, Jianwei; Gauld, Ian C.

    2014-12-01

    The U.S. Department of Energy’s Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) project is nearing the final phase of developing several advanced nondestructive assay (NDA) instruments designed to measure spent nuclear fuel assemblies for the purpose of improving nuclear safeguards. Current efforts are focusing on calibrating several of these instruments with spent fuel assemblies at two international spent fuel facilities. Modelling and simulation is expected to play an important role in predicting nuclide compositions, neutron and gamma source terms, and instrument responses in order to inform the instrument calibration procedures. As part of NGSI-SF project, this work was carried outmore » to assess the impacts of uncertainties in the nuclear data used in the calculations of spent fuel content, radiation emissions and instrument responses. Nuclear data is an essential part of nuclear fuel burnup and decay codes and nuclear transport codes. Such codes are routinely used for analysis of spent fuel and NDA safeguards instruments. Hence, the uncertainties existing in the nuclear data used in these codes affect the accuracies of such analysis. In addition, nuclear data uncertainties represent the limiting (smallest) uncertainties that can be expected from nuclear code predictions, and therefore define the highest attainable accuracy of the NDA instrument. This work studies the impacts of nuclear data uncertainties on calculated spent fuel nuclide inventories and the associated NDA instrument response. Recently developed methods within the SCALE code system are applied in this study. The Californium Interrogation with Prompt Neutron instrument was selected to illustrate the impact of these uncertainties on NDA instrument response.« less

  2. Cross flow characteristics in a three fuel assemblies

    SciTech Connect

    Bae, J. H.; Euh, D. J.; Park, C. K.; Youn, Y. J.; Kwon, T. S.

    2012-07-01

    To evaluate the reactor thermal margin of APR+, reactor core flow distribution including both axial and lateral directional hydraulic resistances of fuel assemblies should be known. 3-Ch cross flow test facility has been constructed with three full-size fuel assemblies to investigate the cross flow characteristics. Performance tests have been performed. The axial and lateral directional hydraulic resistances of fuel assemblies have been measured. The test results have been compared to the CFD calculation. (authors)

  3. Fuel fire tests of selected assemblies. Interim report

    SciTech Connect

    Kydd, G.; Spindola, K.; Askew, G.K.

    1982-04-13

    A varing assortment of clothing assemblies was tested in the Fuel Fire Test Facility at the Naval Air Development Center. Included was a Nomex-Kevlar Cloque Coverall which had relatively good protection from fuel flames.

  4. Design of a full scale model fuel assembly for full power production reactor flow excursion experiments

    SciTech Connect

    Nash, C.A. ); Blake, J.E.; Rush, G.C. )

    1990-01-01

    A novel full scale production reactor fuel assembly model was designed and built to study thermal-hydraulic effects of postulated Savannah River Site (SRS) nuclear reactor accidents. The electrically heated model was constructed to simulate the unique annular concentric tube geometry of fuel assemblies in SRS nuclear production reactors. Several major design challenges were overcome in order to produce the prototypic geometry and thermal-hydraulic conditions. The two concentric heater tubes (total power over 6 MW and maximum heat flux of 3.5 MW/m{sup 2}) (1.1E+6 BTU/(ft{sup 2}hr)) were designed to closely simulate the thermal characteristics of SRS uranium-aluminum nuclear fuel. The paper discusses the design of the model fuel assembly, which met requirements of maintaining prototypic geometric and hydraulic characteristics, and approximate thermal similarity. The model had a cosine axial power profile and the electrical resistance was compatible with the existing power supply. The model fuel assembly was equipped with a set of instruments useful for code analysis, and durable enough to survive a number of LOCA transients. These instruments were sufficiently responsive to record the response of the fuel assembly to the imposed transient.

  5. Design of a full scale model fuel assembly for full power production reactor flow excursion experiments

    SciTech Connect

    Nash, C.A.; Blake, J.E.; Rush, G.C.

    1990-12-31

    A novel full scale production reactor fuel assembly model was designed and built to study thermal-hydraulic effects of postulated Savannah River Site (SRS) nuclear reactor accidents. The electrically heated model was constructed to simulate the unique annular concentric tube geometry of fuel assemblies in SRS nuclear production reactors. Several major design challenges were overcome in order to produce the prototypic geometry and thermal-hydraulic conditions. The two concentric heater tubes (total power over 6 MW and maximum heat flux of 3.5 MW/m{sup 2}) (1.1E+6 BTU/(ft{sup 2}hr)) were designed to closely simulate the thermal characteristics of SRS uranium-aluminum nuclear fuel. The paper discusses the design of the model fuel assembly, which met requirements of maintaining prototypic geometric and hydraulic characteristics, and approximate thermal similarity. The model had a cosine axial power profile and the electrical resistance was compatible with the existing power supply. The model fuel assembly was equipped with a set of instruments useful for code analysis, and durable enough to survive a number of LOCA transients. These instruments were sufficiently responsive to record the response of the fuel assembly to the imposed transient.

  6. Fuel injection assembly for use in turbine engines and method of assembling same

    DOEpatents

    Uhm, Jong Ho; Johnson, Thomas Edward

    2015-03-24

    A fuel injection assembly for use in a turbine engine is provided. The fuel injection assembly includes a plurality of tube assemblies, wherein each of the tube assemblies includes an upstream portion and a downstream portion. Each tube assembly includes a plurality of tubes that extend from the upstream portion to the downstream portion or from the upstream portion through the downstream portion. At least one injection system is coupled to at least one tube assembly of the plurality of tube assemblies. The injection system includes a fluid supply member that extends from a fluid source to the downstream portion of the tube assembly. The fluid supply member includes a first end portion located in the downstream portion of the tube assembly, wherein the first end portion has at least one first opening for channeling fluid through the tube assembly to facilitate reducing a temperature therein.

  7. Single PWR spent fuel assembly heat transfer data for computer code evaluations

    SciTech Connect

    Bates, J.M.

    1986-01-01

    The descriptions and results of two separate heat transfer tests designed to investigate the dry storage of commercial PWR spent fuel assemblies are presented. Presented first are descriptions and selected results from the Fuel Temperature Test performed at the Engine Maintenance and Disassembly facility on the Nevada Test Site. An actual spent fuel assembly from the Turkey Point Unit Number 3 Reactor with a decay heat level of 1.17 KW, was installed vertically in a test stand mounted canister/liner assembly. The boundary temperatures were controlled and the canister backfill gases were alternated between air, helium and vacuum to investigate the primary heat transfer mechanisms of convection, conduction and radiation. The assembly temperature profiles were experimentally measured using installed thermocouple instrumentation. Also presented are the results from the Single Assembly Heat Transfer Test designed and fabricated by Allied General Nuclear Services, under contract to the Department of Energy, and ultimately conducted by the Pacific Northwest Laboratory. For this test, an electrically heated 15 x 15 rod assembly was used to model a single PWR spent fuel assembly. The electrically heated model fuel assembly permitted various ''decay heat'', levels to be tested; 1.0 KW and 0.5 KW were used for these tests. The model fuel assembly was positioned within a prototypic fuel tube and in turn placed within a double-walled sealed cask. The complete test assembly could be positioned at any desired orientation (horizontal, vertical, and 25/sup 0/ from horizontal for the present work) and backfilled as desired (air, helium, or vacuum). Tests were run for all combinations of ''decay heat,'' backfill, and orientation. Boundary conditions were imposed by temperature controlled guard heaters installed on the cask exterior surface.

  8. Separator assembly for use in spent nuclear fuel shipping cask

    DOEpatents

    Bucholz, James A.

    1983-01-01

    A separator assembly for use in a spent nuclear fuel shipping cask has a honeycomb-type wall structure defining parallel cavities for holding nuclear fuel assemblies. Tubes formed of an effective neutron-absorbing material are embedded in the wall structure around each of the cavities and provide neutron flux traps when filled with water.

  9. Transient assembly of active materials fueled by a chemical reaction

    NASA Astrophysics Data System (ADS)

    Boekhoven, Job; Hendriksen, Wouter E.; Koper, Ger J. M.; Eelkema, Rienk; van Esch, Jan H.

    2015-09-01

    Fuel-driven self-assembly of actin filaments and microtubules is a key component of cellular organization. Continuous energy supply maintains these transient biomolecular assemblies far from thermodynamic equilibrium, unlike typical synthetic systems that spontaneously assemble at thermodynamic equilibrium. Here, we report the transient self-assembly of synthetic molecules into active materials, driven by the consumption of a chemical fuel. In these materials, reaction rates and fuel levels, instead of equilibrium composition, determine properties such as lifetime, stiffness, and self-regeneration capability. Fibers exhibit strongly nonlinear behavior including stochastic collapse and simultaneous growth and shrinkage, reminiscent of microtubule dynamics.

  10. Some methods for achieving more efficient performance of fuel assemblies

    NASA Astrophysics Data System (ADS)

    Boltenko, E. A.

    2014-07-01

    More efficient operation of reactor plant fuel assemblies can be achieved through the use of new technical solutions aimed at obtaining more uniform distribution of coolant over the fuel assembly section, more intense heat removal on convex heat-transfer surfaces, and higher values of departure from nucleate boiling ratio (DNBR). Technical solutions using which it is possible to obtain more intense heat removal on convex heat-transfer surfaces and higher DNBR values in reactor plant fuel assemblies are considered. An alternative heat removal arrangement is described using which it is possible to obtain a significantly higher power density in a reactor plant and essentially lower maximal fuel rod temperature.

  11. Spent fuel dry storage technology development: fuel temperature measurements under imposed dry storage conditions (I kW PWR spent fuel assembly)

    SciTech Connect

    Unterzuber, R.; Wright, J.B.

    1980-09-01

    A spent fuel assembly temperature test under imposed dry storage conditions was conducted at the Engine Maintenance Assembly and Disassembly (E-MAD) facility on the Nevada Test Site in support of spent fuel dry storage technology development. This document presents the test data and results obtained from an approximately 1.0 kW decay heat level PWR spent fuel assembly. A spent fuel test apparatus was designed to utilize a representative stainless steel spent fuel canister, a canister lid containing internal temperature instrumentation to measure fuel cladding temperatures, and a carbon steel liner that encloses the canister and lid. Electrical heaters along the liner length, on the lid, and below the canister are used to impose dry storage canister temperature profiles. Temperature instrumentation is provided on the liner and canister. The liner and canister are supported by a test stand in one of the large hot cells (West Process Cell) inside E-MAD. Fuel temperature measurements have been performed using imposed canister temperature profiles from the electrically heated and spent fuel drywell tests being conducted at E-MAD as well as for four constant canister temperature profiles, each with a vacuum, helium and air backfill. Computer models have been utilized in conjunction with the test to predict the thermal response of the fuel cladding. Computer predictions are presented, and they show good agreement with the test data.

  12. Solution High-Energy Burst Assembly (SHEBA) results from subprompt critical experiments with uranyl fluoride fuel

    SciTech Connect

    Cappiello, C.C.; Butterfield, K.B.; Sanchez, R.G.; Bounds, J.A.; Kimpland, R.H.; Damjanovich, R.P.; Jaegers, P.J.

    1997-08-01

    Experiments were performed to measure a variety of parameters for SHEBA: behavior of the facility during transient and steady-state operation; characteristics of the SHEBA fuel; delayed-critical solution height vs solution temperature; initial reactor period and reactivity vs solution height; calibration of power level vs reactor power instrumentation readings; flux profile in SHEBA; radiation levels and neutron spectra outside the assembly for code verification and criticality alarm and dosimetry purposes; and effect on reactivity of voids in the fuel.

  13. Description of fuel element brush assembly`s fabrication for 105-K west

    SciTech Connect

    Maassen, D. P.

    1997-10-15

    This report is a description of the process to redesign and fabricate, as well as, describe the features of the Fuel Element Brush Assembly used in the 105-K West Basin. This narrative description will identify problems that occurred during the redesigning and fabrication of the 105-K West Basin Fuel Element Brush Assembly and specifically address their solutions.

  14. Spent nuclear fuel assembly inspection using neutron computed tomography

    NASA Astrophysics Data System (ADS)

    Pope, Chad Lee

    The research presented here focuses on spent nuclear fuel assembly inspection using neutron computed tomography. Experimental measurements involving neutron beam transmission through a spent nuclear fuel assembly serve as benchmark measurements for an MCNP simulation model. Comparison of measured results to simulation results shows good agreement. Generation of tomography images from MCNP tally results was accomplished using adapted versions of built in MATLAB algorithms. Multiple fuel assembly models were examined to provide a broad set of conclusions. Tomography images revealing assembly geometric information including the fuel element lattice structure and missing elements can be obtained using high energy neutrons. A projection difference technique was developed which reveals the substitution of unirradiated fuel elements for irradiated fuel elements, using high energy neutrons. More subtle material differences such as altering the burnup of individual elements can be identified with lower energy neutrons provided the scattered neutron contribution to the image is limited. The research results show that neutron computed tomography can be used to inspect spent nuclear fuel assemblies for the purpose of identifying anomalies such as missing elements or substituted elements. The ability to identify anomalies in spent fuel assemblies can be used to deter diversion of material by increasing the risk of early detection as well as improve reprocessing facility operations by confirming the spent fuel configuration is as expected or allowing segregation if anomalies are detected.

  15. Criticality safety evaluation report for FFTF 42% fuel assemblies

    SciTech Connect

    Richard, R.F.

    1997-10-28

    An FFTF tritium/isotope production mission will require a new fuel supply. The reference design core will use a mixed oxide fuel nominally enriched to 40 wt% Pu. This enrichment is significantly higher than that of the standard Driver Fuel Assemblies used in past operations. Consequently, criticality safety for handling and storage of this fuel must be addressed. The purpose of this document is to begin the process by determining the minimum critical number for these new fuel assemblies in water, sodium and air. This analysis is preliminary and further work can be done to refine the results reported here. Analysis was initially done using 45 wt 5 PuO. Additionally, a preliminary assessment is done concerning storage of these fuel assemblies in Interim Decay Storage (IDS), Fuel Storage Facility (FSF), and Core Component Containers/Interim Storage Casks (CCC/ISC).

  16. Method and apparatus for close packing of nuclear fuel assemblies

    DOEpatents

    Newman, D.F.

    1993-03-30

    The apparatus of the present invention is a plate of neutron absorbing material. The plate may have a releasable locking feature permitting the plate to be secured within a nuclear fuel assembly between nuclear fuel rods during storage or transportation then removed for further use or destruction. The method of the present invention has the step of placing a plate of neutron absorbing material between nuclear fuel rods within a nuclear fuel assembly, preferably between the two outermost columns of nuclear fuel rods. Additionally, the plate may be releasably locked in place.

  17. Method and apparatus for close packing of nuclear fuel assemblies

    DOEpatents

    Newman, Darrell F.

    1993-01-01

    The apparatus of the present invention is a plate of neutron absorbing material. The plate may have a releasable locking feature permitting the plate to be secured within a nuclear fuel assembly between nuclear fuel rods during storage or transportation then removed for further use or destruction. The method of the present invention has the step of placing a plate of neutron absorbing material between nuclear fuel rods within a nuclear fuel assembly, preferably between the two outermost columns of nuclear fuel rods. Additionally, the plate may be releasably locked in place.

  18. SOURCE OF BURNUP VALUES FOR COMMERCIAL SPENT NUCLEAR FUEL ASSEMBLIES

    SciTech Connect

    BSC

    2004-12-01

    Waste packages are loaded with commercial spent nuclear fuel (SNF) that satisfies the minimum burnup requirements of a criticality loading curve. The burnup value assigned by the originating nuclear utility to each SNF assembly (assigned burnup) is used to load waste packages in compliance with a criticality loading curve. The burnup provided by a nuclear utility has uncertainties, so conservative calculation methods are used to characterize those uncertainties for incorporation into the criticality loading curves. Procedural safety controls ensure that the correct assembly is loaded into each waste package to prevent a misload that could create a condition affecting the safety margins. Probabilistic analyses show that procedural safety controls can minimize the chance of a misload but can not completely eliminate the possibility. Physical measurements of burnup with instrumentation in the surface facility are not necessary due to the conservative calculation methods used to produce the criticality loading curves. The reactor records assigned burnup of a commercial SNF assembly contains about two percent uncertainty, which is increased to five-percent to ensure conservatism. This five-percent uncertainty is accommodated by adjusting the criticality loading curve. Also, the record keeping methods of nuclear utilities are not uniform and the level of detail required by the NRC has varied over the last several decades. Thus, some SNF assemblies may have assigned burnups that are averages for a batch of assemblies with similar characteristics. Utilities typically have access to more detailed core-follow records that allow the batch average burnup to be changed to an assembly specific burnup. Alternatively, an additional safety margin is incorporated into the criticality loading curve to accommodate SNF assemblies with batch average burnups or greater uncertainties due to the methodology used by the nuclear utility. The utility records provide the assembly identifier

  19. Combustor with two stage primary fuel assembly

    DOEpatents

    Sharifi, Mehran; Zolyomi, Wendel; Whidden, Graydon Lane

    2000-01-01

    A combustor for a gas turbine having first and second passages for pre-mixing primary fuel and air supplied to a primary combustion zone. The flow of fuel to the first and second pre-mixing passages is separately regulated using a single annular fuel distribution ring having first and second row of fuel discharge ports. The interior portion of the fuel distribution ring is divided by a baffle into first and second fuel distribution manifolds and is located upstream of the inlets to the two pre-mixing passages. The annular fuel distribution ring is supplied with fuel by an annular fuel supply manifold, the interior portion of which is divided by a baffle into first and second fuel supply manifolds. A first flow of fuel is regulated by a first control valve and directed to the first fuel supply manifold, from which the fuel is distributed to first fuel supply tubes that direct it to the first fuel distribution manifold. From the first fuel distribution manifold, the first flow of fuel is distributed to the first row of fuel discharge ports, which direct it into the first pre-mixing passage. A second flow of fuel is regulated by a second control valve and directed to the second fuel supply manifold, from which the fuel is distributed to second fuel supply tubes that direct it to the second fuel distribution manifold. From the second fuel distribution manifold, the second flow of fuel is distributed to the second row of fuel discharge ports, which direct it into the second pre-mixing passage.

  20. Temperature measuring analysis of the nuclear reactor fuel assembly

    SciTech Connect

    Urban, F. E-mail: zdenko.zavodny@stuba.sk; Kučák, L. E-mail: zdenko.zavodny@stuba.sk; Bereznai, J. E-mail: zdenko.zavodny@stuba.sk; Závodný, Z. E-mail: zdenko.zavodny@stuba.sk; Muškát, P. E-mail: zdenko.zavodny@stuba.sk

    2014-08-06

    Study was based on rapid changes of measured temperature values from the thermocouple in the VVER 440 nuclear reactor fuel assembly. Task was to determine origin of fluctuations of the temperature values by experiments on physical model of the fuel assembly. During an experiment, heated water was circulating in the system and cold water inlet through central tube to record sensitivity of the temperature sensor. Two positions of the sensor was used. First, just above the central tube in the physical model fuel assembly axis and second at the position of the thermocouple in the VVER 440 nuclear reactor fuel assembly. Dependency of the temperature values on time are presented in the diagram form in the paper.

  1. Current conducting end plate of fuel cell assembly

    DOEpatents

    Walsh, Michael M.

    1999-01-01

    A fuel cell assembly has a current conducting end plate with a conductive body formed integrally with isolating material. The conductive body has a first surface, a second surface opposite the first surface, and an electrical connector. The first surface has an exposed portion for conducting current between a working section of the fuel cell assembly and the electrical connector. The isolating material is positioned on at least a portion of the second surface. The conductive body can have support passage(s) extending therethrough for receiving structural member(s) of the fuel cell assembly. Isolating material can electrically isolate the conductive body from the structural member(s). The conductive body can have service passage(s) extending therethrough for servicing one or more fluids for the fuel cell assembly. Isolating material can chemically isolate the one or more fluids from the conductive body. The isolating material can also electrically isolate the conductive body from the one or more fluids.

  2. Fuel burner and combustor assembly for a gas turbine engine

    DOEpatents

    Leto, Anthony

    1983-01-01

    A fuel burner and combustor assembly for a gas turbine engine has a housing within the casing of the gas turbine engine which housing defines a combustion chamber and at least one fuel burner secured to one end of the housing and extending into the combustion chamber. The other end of the fuel burner is arranged to slidably engage a fuel inlet connector extending radially inwardly from the engine casing so that fuel is supplied, from a source thereof, to the fuel burner. The fuel inlet connector and fuel burner coact to anchor the housing against axial movement relative to the engine casing while allowing relative radial movement between the engine casing and the fuel burner and, at the same time, providing fuel flow to the fuel burner. For dual fuel capability, a fuel injector is provided in said fuel burner with a flexible fuel supply pipe so that the fuel injector and fuel burner form a unitary structure which moves with the fuel burner.

  3. Adjustment-Free Mirage Instrument with Components Assembled in a Compact Package

    NASA Astrophysics Data System (ADS)

    Yarai, Atsushi; Yamaura, Yusuke; Nakanishi, Takuji

    2002-05-01

    This paper describes how to achieve an adjustment-free mirage instrument with components assembled in a compact package. The deflection angle amplifier that we previously developed for mirage detection is assembled in this instrument. The features of this instrument are: 1) no optical table, 2) adjustment-free design, and 3) an extremely compact structure for mobile use. The volume of our compact mirage instrument is at least 1/25 times that of the conventional mirage instrument. The capability of the compact mirage instrument is also verified. The results show good potential for using the technique for apparatus commercialization.

  4. The optimization of an AP1000 fuel assembly for the transmutation of plutonium and minor actinides

    NASA Astrophysics Data System (ADS)

    Washington, Jeremy A.

    The average nuclear power plant produces twenty metric tons of used nuclear fuel per year, containing approximately 95 wt% uranium, 1 wt% plutonium, and 4 wt% fission products and transuranic elements. Fast reactors are a preferred option for the transmutation of plutonium and minor actinides; however, an optimistic deployment time of at least 20 years indicates a need for a near-term solution. The goal of this thesis is to examine the potential of light water reactors for plutonium and minor actinides transmutation as a near-term solution. This thesis screens the available nuclear isotope database to identify potential absorbers as coatings on a transmutation fuel in a light water reactor. A spectral shift absorber coating tunes the neutron energy spectrum experienced by the underlying target fuel. Eleven different spectral shift absorbers (B4C, CdO, Dy2O3, Er 2O3, Eu2O3, Gd2O3, HfO2, In2O3, Lu2O3, Sm2O3, and TaC) have been selected for further evaluation. A model developed using the NEWT module of SCALE 6.1 code provided performance data for the burnup of the target fuel rods. Irradiation of the target fuels occurs in a Westinghouse 17x17 XL Robust Fuel Assembly over a 1400 Effective Full Power Days (EFPD) interval. The fuels evaluated in this thesis include PuO2, Pu3Si2, PuN, MOX, PuZrH, PuZrHTh, PuZrO 2, and PuUZrH. MOX (5 wt% PuO2), Pu0.31ZrH 1.6Th1.08, and PuZrO2MgO (8 wt%) are selected for detailed analysis in a multi-pin transmutation assembly. A coupled model optimized the resulting transmutation fuel elements. The optimization considered three stages of fuel assemblies containing target fuel pins. The first stage optimized four target fuel pins adjacent to the central instrumentation channel. The second stage evaluated a variety of assemblies with multiple target fuel pins and the third stage re-optimized target fuel pins in the second-stage assembly. A PuZrO2MgO (8 wt%) target fuel with a coating of Lu 2O3 resulted in the greatest reduction in curium-244

  5. Storage assembly for spent nuclear fuel

    SciTech Connect

    Lapides, M.E.

    1982-04-27

    A technique for storing spent fuel rods from a nuclear reactor is disclosed herein. This technique utilizes a housing including a closed inner chamber for containing the fuel rods and a thermally conductive member located partially within the housing chamber and partially outside the housing for transferring heat generated by the fuel rods from the chamber to the ambient surroundings. Particulate material is located within the chamber and surrounds the fuel rods contained therein. This material is selected to serve as a heat transfer media between the contained cells and the heat transferring member and, at the same time, stand ready to fuse into a solid mass around the contained cells if the heat transferring member malfunctions or otherwise fails to transfer the generated heat out of the housing chamber in a predetermined way.

  6. Hydrogen storage and integrated fuel cell assembly

    DOEpatents

    Gross, Karl J.

    2010-08-24

    Hydrogen is stored in materials that absorb and desorb hydrogen with temperature dependent rates. A housing is provided that allows for the storage of one or more types of hydrogen-storage materials in close thermal proximity to a fuel cell stack. This arrangement, which includes alternating fuel cell stack and hydrogen-storage units, allows for close thermal matching of the hydrogen storage material and the fuel cell stack. Also, the present invention allows for tailoring of the hydrogen delivery by mixing different materials in one unit. Thermal insulation alternatively allows for a highly efficient unit. Individual power modules including one fuel cell stack surrounded by a pair of hydrogen-storage units allows for distribution of power throughout a vehicle or other electric power consuming devices.

  7. Detachable connection for a nuclear reactor fuel assembly

    DOEpatents

    Christiansen, David W.; Karnesky, Richard A.

    1986-01-01

    A locking connection for releasably attaching a handling socket to the duct tube of a fuel assembly for a nuclear reactor. The connection comprises a load pad housing mechanically attached to the duct tube and a handling socket threadably secured within the housing. A retaining ring is interposed between the housing and the handling socket and is formed with a projection and depression engageable within a cavity and groove of the housing and handling socket, respectively, to form a detachable interlocked connection assembly.

  8. Detachable connection for a nuclear reactor fuel assembly

    DOEpatents

    Christiansen, D.W.; Karnesky, R.A.

    1983-08-29

    A locking connection for releasably attaching a handling socket to the duct tube of a fuel assembly for a nuclear reactor. The connection comprises a load pad housing mechanically attached to the duct tube and a handling socket threadably secured within the housing. A retaining ring is interposed between the housing and the handling socket and is formed with a projection and depression engagable within a cavity and groove of the housing and handling socket, respectively, to form a detachable interlocked connection assembly.

  9. DISSOLUTION OF IRRADIATED MURR FUEL ASSEMBLIES

    SciTech Connect

    Kyser, E.

    2010-06-17

    A literature survey on the dissolution of spent nuclear fuel from the University of Missouri Research Reactor (MURR) has been performed. This survey encompassed both internal and external literature sources for the dissolution of aluminum-clad uranium alloy fuels. The most limiting aspect of dissolution in the current facility configuration involves issues related to the control of the flammability of the off-gas from this process. The primary conclusion of this work is that based on past dissolution of this fuel in H-Canyon, four bundles of this fuel (initial charge) may be safely dissolved in a nitric acid flowsheet catalyzed with 0.002 M mercuric nitrate using a 40 scfm purge to control off-gas flammability. The initial charge may be followed by a second charge of up to five bundles to the same dissolver batch depending on volume and concentration constraints. The safety of this flowsheet relies on composite lower flammability limits (LFL) estimated from prior literature, pilot-scale work on the dissolution of site fuels, and the proposed processing flowsheet. Equipment modifications or improved LFL data offer the potential for improved processing rates. The fuel charging sequence, as well as the acid and catalyst concentrations, will control the dissolution rate during the initial portion of the cycle. These parameters directly impact the hydrogen and off-gas generation and, along with the purge flowrate determine the number of bundles that may be charged. The calculation approach within provides Engineering a means to determine optimal charging patterns. Downstream processing of this material should be similar to that of recent processing of site fuels requiring only minor adjustments of the existing flowsheet parameters.

  10. Design of a Prototype Differential Die-Away Instrument Proposed for Swedish Spent Nuclear Fuel Characterization

    NASA Astrophysics Data System (ADS)

    Martinik, Tomas; Henzl, Vladimir; Grape, Sophie; Jansson, Peter; Swinhoe, Martyn T.; Goodsell, Alison V.; Tobin, Stephen J.

    2016-06-01

    As part of the United States (US) Department of Energy's Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) project, the traditional Differential Die-Away (DDA) method that was originally developed for waste drum assay has been investigated and modified to provide a novel application to characterize or verify spent nuclear fuel (SNF). Following the promising, yet largely theoretical and simulation based, research of physics aspects of the DDA technique applied to SNF assay during the early stages of the NGSI-SF project, the most recent effort has been focused on the practical aspects of developing the first fully functional and deployable DDA prototype instrument for spent fuel. As a result of the collaboration among US research institutions and Sweden, the opportunity to test the newly proposed instrument's performance with commercial grade SNF at the Swedish Interim Storage Facility (Clab) emerged. Therefore the design of this instrument prototype has to accommodate the requirements of the Swedish regulator as well as specific engineering constrains given by the unique industrial environment. Within this paper, we identify key components of the DDA based instrument and we present methodology for evaluation and the results of a selection of the most relevant design parameters in order to optimize the performance for a given application, i.e. test-deployment, including assay of 50 preselected spent nuclear fuel assemblies of both pressurized (PWR) as well as boiling (BWR) water reactor type.

  11. Nuclear imaging of the fuel assembly in ignition experiments

    SciTech Connect

    Grim, G. P.; Guler, N.; Merrill, F. E.; Morgan, G. L.; Danly, C. R.; Volegov, P. L.; Wilde, C. H.; Wilson, D. C.; Batha, S.; Herrmann, H. W.; Kline, J. L.; Kyrala, G. A.; Clark, D. S.; Hinkel, D. E.; Jones, O. S.; Raman, K. S.; Izumi, N.; Fittinghoff, D. N.; Drury, O. B.; Alger, E. T.; and others

    2013-05-15

    First results from the analysis of neutron image data collected on implosions of cryogenically layered deuterium-tritium capsules during the 2011-2012 National Ignition Campaign are reported. The data span a variety of experimental designs aimed at increasing the stagnation pressure of the central hotspot and areal density of the surrounding fuel assembly. Images of neutrons produced by deuterium–tritium fusion reactions in the hotspot are presented, as well as images of neutrons that scatter in the surrounding dense fuel assembly. The image data are compared with 1D and 2D model predictions, and consistency checked using other diagnostic data. The results indicate that the size of the fusing hotspot is consistent with the model predictions, as well as other imaging data, while the overall size of the fuel assembly, inferred from the scattered neutron images, is systematically smaller than models' prediction. Preliminary studies indicate these differences are consistent with a significant fraction (20%–25%) of the initial deuterium-tritium fuel mass outside the compact fuel assembly, due either to low mode mass asymmetry or high mode 3D mix effects at the ablator-ice interface.

  12. Nuclear imaging of the fuel assembly in ignition experimentsa)

    NASA Astrophysics Data System (ADS)

    Grim, G. P.; Guler, N.; Merrill, F. E.; Morgan, G. L.; Danly, C. R.; Volegov, P. L.; Wilde, C. H.; Wilson, D. C.; Clark, D. S.; Hinkel, D. E.; Jones, O. S.; Raman, K. S.; Izumi, N.; Fittinghoff, D. N.; Drury, O. B.; Alger, E. T.; Arnold, P. A.; Ashabranner, R. C.; Atherton, L. J.; Barrios, M. A.; Batha, S.; Bell, P. M.; Benedetti, L. R.; Berger, R. L.; Bernstein, L. A.; Berzins, L. V.; Betti, R.; Bhandarkar, S. D.; Bionta, R. M.; Bleuel, D. L.; Boehly, T. R.; Bond, E. J.; Bowers, M. W.; Bradley, D. K.; Brunton, G. K.; Buckles, R. A.; Burkhart, S. C.; Burr, R. F.; Caggiano, J. A.; Callahan, D. A.; Casey, D. T.; Castro, C.; Celliers, P. M.; Cerjan, C. J.; Chandler, G. A.; Choate, C.; Cohen, S. J.; Collins, G. W.; Cooper, G. W.; Cox, J. R.; Cradick, J. R.; Datte, P. S.; Dewald, E. L.; Di Nicola, P.; Di Nicola, J. M.; Divol, L.; Dixit, S. N.; Dylla-Spears, R.; Dzenitis, E. G.; Eckart, M. J.; Eder, D. C.; Edgell, D. H.; Edwards, M. J.; Eggert, J. H.; Ehrlich, R. B.; Erbert, G. V.; Fair, J.; Farley, D. R.; Felker, B.; Fortner, R. J.; Frenje, J. A.; Frieders, G.; Friedrich, S.; Gatu-Johnson, M.; Gibson, C. R.; Giraldez, E.; Glebov, V. Y.; Glenn, S. M.; Glenzer, S. H.; Gururangan, G.; Haan, S. W.; Hahn, K. D.; Hammel, B. A.; Hamza, A. V.; Hartouni, E. P.; Hatarik, R.; Hatchett, S. P.; Haynam, C.; Hermann, M. R.; Herrmann, H. W.; Hicks, D. G.; Holder, J. P.; Holunga, D. M.; Horner, J. B.; Hsing, W. W.; Huang, H.; Jackson, M. C.; Jancaitis, K. S.; Kalantar, D. H.; Kauffman, R. L.; Kauffman, M. I.; Khan, S. F.; Kilkenny, J. D.; Kimbrough, J. R.; Kirkwood, R.; Kline, J. L.; Knauer, J. P.; Knittel, K. M.; Koch, J. A.; Kohut, T. R.; Kozioziemski, B. J.; Krauter, K.; Krauter, G. W.; Kritcher, A. L.; Kroll, J.; Kyrala, G. A.; Fortune, K. N. La; LaCaille, G.; Lagin, L. J.; Land, T. A.; Landen, O. L.; Larson, D. W.; Latray, D. A.; Leeper, R. J.; Lewis, T. L.; LePape, S.; Lindl, J. D.; Lowe-Webb, R. R.; Ma, T.; MacGowan, B. J.; MacKinnon, A. J.; MacPhee, A. G.; Malone, R. M.; Malsbury, T. N.; Mapoles, E.; Marshall, C. D.; Mathisen, D. G.; McKenty, P.; McNaney, J. M.; Meezan, N. B.; Michel, P.; Milovich, J. L.; Moody, J. D.; Moore, A. S.; Moran, M. J.; Moreno, K.; Moses, E. I.; Munro, D. H.; Nathan, B. R.; Nelson, A. J.; Nikroo, A.; Olson, R. E.; Orth, C.; Pak, A. E.; Palma, E. S.; Parham, T. G.; Patel, P. K.; Patterson, R. W.; Petrasso, R. D.; Prasad, R.; Ralph, J. E.; Regan, S. P.; Rinderknecht, H.; Robey, H. F.; Ross, G. F.; Ruiz, C. L.; Séguin, F. H.; Salmonson, J. D.; Sangster, T. C.; Sater, J. D.; Saunders, R. L.; Schneider, M. B.; Schneider, D. H.; Shaw, M. J.; Simanovskaia, N.; Spears, B. K.; Springer, P. T.; Stoeckl, C.; Stoeffl, W.; Suter, L. J.; Thomas, C. A.; Tommasini, R.; Town, R. P.; Traille, A. J.; Wonterghem, B. Van; Wallace, R. J.; Weaver, S.; Weber, S. V.; Wegner, P. J.; Whitman, P. K.; Widmann, K.; Widmayer, C. C.; Wood, R. D.; Young, B. K.; Zacharias, R. A.; Zylstra, A.

    2013-05-01

    First results from the analysis of neutron image data collected on implosions of cryogenically layered deuterium-tritium capsules during the 2011-2012 National Ignition Campaign are reported. The data span a variety of experimental designs aimed at increasing the stagnation pressure of the central hotspot and areal density of the surrounding fuel assembly. Images of neutrons produced by deuterium-tritium fusion reactions in the hotspot are presented, as well as images of neutrons that scatter in the surrounding dense fuel assembly. The image data are compared with 1D and 2D model predictions, and consistency checked using other diagnostic data. The results indicate that the size of the fusing hotspot is consistent with the model predictions, as well as other imaging data, while the overall size of the fuel assembly, inferred from the scattered neutron images, is systematically smaller than models' prediction. Preliminary studies indicate these differences are consistent with a significant fraction (20%-25%) of the initial deuterium-tritium fuel mass outside the compact fuel assembly, due either to low mode mass asymmetry or high mode 3D mix effects at the ablator-ice interface.

  13. Passive Tomography for Spent Fuel Verification: Analysis Framework and Instrument Design Study

    SciTech Connect

    White, Timothy A.; Svard, Staffan J.; Smith, Leon E.; Mozin, Vladimir V.; Jansson, Peter; Davour, Anna; Grape, Sophie; Trellue, H.; Deshmukh, Nikhil S.; Wittman, Richard S.; Honkamaa, Tapani; Vaccaro, Stefano; Ely, James

    2015-05-18

    The potential for gamma emission tomography (GET) to detect partial defects within a spent nuclear fuel assembly is being assessed through a collaboration of Support Programs to the International Atomic Energy Agency (IAEA). In the first phase of this study, two safeguards verification objectives have been identified. The first is the independent determination of the number of active pins that are present in the assembly, in the absence of a priori information. The second objective is to provide quantitative measures of pin-by-pin properties, e.g. activity of key isotopes or pin attributes such as cooling time and relative burnup, for the detection of anomalies and/or verification of operator-declared data. The efficacy of GET to meet these two verification objectives will be evaluated across a range of fuel types, burnups, and cooling times, and with a target interrogation time of less than 60 minutes. The evaluation of GET viability for safeguards applications is founded on a modelling and analysis framework applied to existing and emerging GET instrument designs. Monte Carlo models of different fuel types are used to produce simulated tomographer responses to large populations of “virtual” fuel assemblies. Instrument response data are processed by a variety of tomographic-reconstruction and image-processing methods, and scoring metrics specific to each of the verification objectives are defined and used to evaluate the performance of the methods. This paper will provide a description of the analysis framework and evaluation metrics, example performance-prediction results, and describe the design of a “universal” GET instrument intended to support the full range of verification scenarios envisioned by the IAEA.

  14. Individual source positioning mechanism for a nuclear reactor fuel assembly

    SciTech Connect

    Wilson, J.F.; Gjertsen, R.K.; Cerni, S.

    1987-07-07

    A nuclear reactor is described including a fuel assembly, at lest one elongated neutron source rod and an upper core plate. The fuel assembly has top and bottom nozzles with a guide thimbles extending between and interconnecting the nozzles. The upper core plate is positioned adjacent to and above the top nozzle of the fuel assembly and having flow openings to allow passage of coolant from the fuel assembly. At least some of the openings is aligned over respective ones of the guide thimbles with seating means defined about the openings on a lower side of the core plate, a separate mechanism for positioning each individual neutron source rod in a respective guide thimble aligned with one of the openings defined through the upper core plate, comprising: (a) locating means registering against the core plate seating means; and (b) resilient holddown means extending partially into the guide thimble and coupling the source rod with the locating means in a manner which restrains the source rod in a lateral direction and positions the rod in a stationary axial relationship within the guide thimble.

  15. Determination of BWR Spent Nuclear Fuel Assembly Effective Thermal Conductivity

    SciTech Connect

    Matthew D. Hinds

    2001-10-17

    The purpose of this calculation is to provide an effective thermal conductivity for use in predicting peak cladding temperatures in boiling water reactor (BWR) fuel assemblies with 7x7,8x8, and 9x9 rod arrays. The first objective of this calculation is to describe the development and application of a finite element representation that predicts peak spent nuclear fuel temperatures for BWR assemblies. The second objective is to use the discrete representation to develop a basis for determining an effective thermal conductivity (described later) for a BWR assembly with srneared/homogeneous properties and to investigate the thermal behavior of a spent fuel assembly. The scope of this calculation is limited to a steady-state two-dimensional representation of the waste package interior region. This calculation is subject to procedure AP-3.124, Calculations (Ref. 27) and guided by the applicable technical work plan (Ref. 14). While these evaluations were originally developed for the thermal analysis of conceptual waste package designs emplaced in the potential repository at Yucca Mountain, the methodology applies to storage and transportation thermal analyses as well. Note that the waste package sketch in Attachment V depicts a preliminary design, and should not be interpreted otherwise.

  16. Feasibility study of application of ductless fuel assembly to FBR

    SciTech Connect

    Itoh, K.; Shibahara, I.

    1996-07-01

    Feasibility studies on an application of the ductless fuel concept to an FBR core have been carried out in order to evaluate the basic features of the ductless core, especially in the fields of the thermal-hydraulic aspects and the mechanical behaviors. Regarding thermal-hydraulic aspects, analyses were performed by using a whole core thermal-hydraulic analysis code by making some modification for this study on the PWR code THINC. A small scaled ductless core model was prepared and a hydraulic experiment was carried out to study basic hydraulic characteristics of a ductless core. Core mechanical behaviors were analyzed focusing on the core irradiation bowing aspects and the seismic behaviors. Following features are revealed on the core structural behaviors: (1) the bowing stiffness of the ductless assembly is around 1/5 to 1/10 of that of the duct type assembly; (2) the contact loads between assemblies by the bowing effects are small through core cycles; (3) the damping of the ductless assemblies are so large that the seismic responses are small and the loads between assemblies are small due to occurring many contact points. Through this study it is expected that the concept of the ductless fuel can be applicable to FBR cores from the design view points of thermal-hydraulic and core mechanical behaviors.

  17. Valve assembly and fuel metering apparatus

    SciTech Connect

    Chute, R.

    1988-11-29

    This patent describes an improvement in a liquid flow valving assembly comprising valve body means, valve seating surface means carried by the body means, a purality of passages formed through the body means, each of the passages comprising an upstream inlet end generally surrounded by the seating surface means and a downstream outlet end, a valve member, the valve member comprising a valving surface means for at times sealingly engaging the seating surface means, the valve member being movable in a first direction for causing the valving surface means to sealingly engage the seating surface means to thereby terminate flow of liquid through each of the plurality of passages, the valve member being movable in a second direction opposite to the first direction to thereby open each of the passages to the flow of the liquid therethrough, wherein the first and second directions of movement comprise a single axis of movement, stationary stem-like guide means for guiding the valve member along the single axis of movement during the time that the valve member is moving in the first direction as well as during the time that the valve member is moving in the second direction, wherein the passages are each located in the body means as to be radially outwardly of the stem-like guide means and radially outwardly of the single axis of movement, means for causing the movement of the valve member along the stem-like guide means in the first and second directions, wherein the means for causing the movement of the valve member comprises electrically energizable coil means effective to cyclically produce a flux field for the corresponding cyclic movement of the valve member along the stem-like guide means and in the second direction, and wherein the valve member extends into the region of the coil means and the flux field as to be acted upon thereby.

  18. Fail-safe storage rack for irradiated fuel rod assemblies

    DOEpatents

    Lewis, D.R.

    1993-03-23

    A fail-safe storage rack is provided for interim storage of spent but radioactive nuclear fuel rod assemblies. The rack consists of a checkerboard array of substantially square, elongate receiving tubes fully enclosed by a double walled container, the outer wall of which is imperforate for liquid containment and the inner wall of which is provided with perforations for admitting moderator liquid flow to the elongate receiving tubes, the liquid serving to take up waste heat from the stored nuclear assemblies and dissipate same to the ambient liquid reservoir. A perforated cover sealing the rack facilitates cooling liquid entry and dissipation.

  19. Fail-safe storage rack for irradiated fuel rod assemblies

    DOEpatents

    Lewis, Donald R.

    1993-01-01

    A fail-safe storage rack is provided for interim storage of spent but radioactive nuclear fuel rod assemblies. The rack consists of a checkerboard array of substantially square, elongate receiving tubes fully enclosed by a double walled container, the outer wall of which is imperforate for liquid containment and the inner wall of which is provided with perforations for admitting moderator liquid flow to the elongate receiving tubes, the liquid serving to take up waste heat from the stored nuclear assemblies and dissipate same to the ambient liquid reservoir. A perforated cover sealing the rack facilitates cooling liquid entry and dissipation.

  20. Fail-safe storage rack for fuel rod assemblies

    SciTech Connect

    Lewis, D.R.

    1991-12-31

    This report discusses a fail-safe storage rack which is provided for interim storage of spent but radioactive nuclear fuel rod assemblies. The rack consists of a checkerboard array of substantially square, elongate receiving tubes fully enclosed by a double walled container, the outer wall of which is imperforate for liquid containment and the inner wall of which is provided with perforations for admitting moderator liquid flow to the elongate receiving tubes, the liquid serving to take up waste heat from the stored nuclear assemblies and dissipate same to the ambient liquid reservoir. A perforated cover sealing the rack facilitates cooling liquid entry and dissipation.

  1. Building and testing of MIDAS instrument sub-assemblies

    NASA Astrophysics Data System (ADS)

    Lewis, S. D.

    2001-09-01

    The MIDAS instrument is an atomic force microscope developed by ESTEC to fly on Rosetta. The purpose of the instrument is to sample and characterise cometary dust, which impinges upon a facetted wheel contained within the instrument enclosure. Due to its relative complexity, the long cruise phase of the Rosetta mission and the relatively novel use of piezomotors for all drive requirements the instrument has a number of interesting mechanisms engineering challenges. This paper describes the lubricant selection, EM and FM subassembly build and test campaigns carried out by AEA Technology Space in close support of the instrumentlevel activities which ran in parallel at ESTEC. The paper also identifies some lessons learned, which can be generally applied in other mechanism programmes.

  2. Micro-Pocket Fission Detectors (MPFD) For Fuel Assembly Analysis

    SciTech Connect

    Troy Unruh; Michael Reichenberger; Phillip Ugorowski

    2013-09-01

    Neutron sensors capable of real-time measurement of thermal flux, fast flux, and temperature in a single miniaturized probe are needed in irradiation tests required to demonstrate the performance of candidate new fuels, and cladding materials. In-core ceramic-based miniature neutron detectors or “Micro-Pocket Fission Detectors” (MPFDs) have been studied at Kansas State University (KSU). The first MPFD prototypes were tested in various neutron fields at the KSU TRIGA research reactor with successful results. Currently, a United States Department of Energy-sponsored joint KSU/Idaho National Laboratory (INL) effort is underway to develop a high-temperature, high-pressure version of the MPFD using radiation-resistant, high temperature materials, which would be capable of withstanding irradiation test conditions in high performance material and test reactors (MTRs). Ultimately, this more compact, more accurate, and longer lifetime flux sensor for critical mock-ups, existing and advanced reactor designs, high performance MTRs, and transient test reactors has the potential to lead to higher accuracy and resolution data from irradiation testing, more detailed core flux measurements and enhanced fuel assembly processing. Prior evaluations by KSU indicate that these sensors could also be used to monitor burn-up of nuclear fuel. If integrated into nuclear fuel assemblies, MPFDs offer several advantages to current spent fuel management systems.

  3. Fuel assembly design for APR1400 with low CBC

    SciTech Connect

    Hah, Chang Joo

    2015-04-29

    APR 1400 is a PWR (Pressurized Water Reactor) with rated power of 3983 MWth and 241 assemblies. Recently, demand for extremely longer cycle up to 24 months is increasing with challenge of higher critical boron concentration (CBC). In this paper, assembly design method of selecting Gd-rods is introduced to reduce CBC. The purpose of the method is to lower the critical boron concentration of the preliminary core loading pattern (PLP), and consequently to achieve more negative or less positive moderator temperature coefficient (MTC). In this method, both the ratio of the number of low-Gd rod to the number of high-Gd rod (r) and assembly average Gd wt% (w) are the decision variables. The target function is the amount of soluble boron concentration reduction, which can be converted to Δk{sub TARGET}. A set of new designed fuel assembly satisfies an objective function, min [f=∑{sub i}(Δk{sub FA}−Δk{sub i})], and enables a final loading pattern to reach a target CBC. The constraints required to determine a set of Δk are physically realizable pair, (r,w), and the sum of Δk of new designed assemblies as close to Δk{sub TARGET} as possible. New Gd-bearing assemblies selected based on valid pairs of (r,w) are replaced with existing assemblies in a PLP. This design methodology is applied to Shin-Kori Unit 3 Cycle 1 used as a reference model. CASMO-3/MASTER code is used for depletion calculation. CASMO-3/MASTER calculations with new designed assemblies produce lower CBC than the expected CBC, proving that the proposed method works successful.

  4. Fuel assembly design for APR1400 with low CBC

    NASA Astrophysics Data System (ADS)

    Hah, Chang Joo

    2015-04-01

    APR 1400 is a PWR (Pressurized Water Reactor) with rated power of 3983 MWth and 241 assemblies. Recently, demand for extremely longer cycle up to 24 months is increasing with challenge of higher critical boron concentration (CBC). In this paper, assembly design method of selecting Gd-rods is introduced to reduce CBC. The purpose of the method is to lower the critical boron concentration of the preliminary core loading pattern (PLP), and consequently to achieve more negative or less positive moderator temperature coefficient (MTC). In this method, both the ratio of the number of low-Gd rod to the number of high-Gd rod (r) and assembly average Gd wt% (w) are the decision variables. The target function is the amount of soluble boron concentration reduction, which can be converted to ΔkTARGET. A set of new designed fuel assembly satisfies an objective function, min [f =∑i (ΔkF A-Δki ) ] , and enables a final loading pattern to reach a target CBC. The constraints required to determine a set of Δk are physically realizable pair, (r,w), and the sum of Δk of new designed assemblies as close to ΔkTARGET as possible. New Gd-bearing assemblies selected based on valid pairs of (r,w) are replaced with existing assemblies in a PLP. This design methodology is applied to Shin-Kori Unit 3 Cycle 1 used as a reference model. CASMO-3/MASTER code is used for depletion calculation. CASMO-3/MASTER calculations with new designed assemblies produce lower CBC than the expected CBC, proving that the proposed method works successful.

  5. Monticello BWR spent fuel assembly decay heat predictions and measurements

    SciTech Connect

    McKinnon, M.A.; Doman, J.W.; Heeb, C.M.; Creer, J.M.

    1986-06-01

    This report compares pre-calorimetry predictions of rates of six 7 x 7 boiling water reactor (BWR) spent fuel assemblies with measured decay heat rates. The assemblies were from Northern States Power Company's Monticello Nuclear Generating Plant and had burnups of 9 to 21 GWd/MTU and cooling times of 9 to 10 years. Conclusions are: The agreement between ORIGEN2 predictions and decay heat measurements of Monticello spent fuel is dependent on the method used to calibrate the calorimeter and to make the decay heat measurements. The agreement between predictions and measurements of decay heat rates of Monticello fuel is the same as that for Cooper and Dresden fuel if the same measurement method is used. The predictions are within a standard deviation of +-15 W of the measurements. Using a different measurement method, ORIGEN2 underpredicts the measured decay heat output of Monticello fuel assemblies by a constant 20 +- 2 W. The 20-W offset appears to be an artifact of the calibration procedure. The constant term in the calibration curve (i.e., q/sub DH/ = mx + b) can account for measurement differences of 40 W based on the 1983, 1984, and 1985 calibration curves. The difference between ORIGEN2 predictions and calorimeter decay heat measurements does not appear to be dependent on the magnitude of decay heat output. Predicted axial decay heat profiles are in good agreement with measured axial gamma radiation profiles. Recommendations are: Predictions using other decay heat codes should be compared to experimental data contained in this report, to evaluate prediction capabilities. The source of the differences that exist among calorimeter calibration curves needs to be determined. Calorimeter operational methods need to be investigated further to determine cause and effect relationships between operational method and calorimeter precision and accuracy.

  6. Development of an ultrasonic cleaning method for fuel assemblies

    SciTech Connect

    Heki, H.; Komura, S.; Kato, H.; Sakai, H. ); Hattori, T. )

    1991-01-01

    Almost all radiation buildup in light water reactors is the result of the deposition of activated corrosion and wear products in out-of-core areas. After operation, a significant quantity of corrosion and wear products is deposited on the fuel rods as crud. At refueling shutdowns, these activation products are available for removal. If they can be quickly and easily removed, buildup of radioactivity on out-of-core surfaces and individual exposure dose can be greatly reduced. After studying various physical cleaning methods (e.g., water jet and ultrasonic), the ultrasonic cleaning method was selected as the most effective for fuel assembly cleaning. The ultrasonic cleaning method is especially able to efficiently clean the fuel without removing the channel box. The removed crud in the channel box would be swept out to the filtration unit. Parameter survey tests were carried out to evaluate the optimum conditions for ultrasonic cleaning using a mock-up of a short section of fuel assembly with the channel box. The ultrasonic device used was a 600-W ultrasonic transducer operating at 26-kHz ultrasonic frequency.

  7. Inactive end cell assembly for fuel cells for improved electrolyte management and electrical contact

    DOEpatents

    Yuh, Chao-Yi; Farooque, Mohammad; Johnsen, Richard

    2007-04-10

    An assembly for storing electrolyte in a carbonate fuel cell is provided. The combination of a soft, compliant and resilient cathode current collector and an inactive anode part including a foam anode in each assembly mitigates electrical contact loss during operation of the fuel cell stack. In addition, an electrode reservoir in the positive end assembly and an electrode sink in the negative end assembly are provided, by which ribbed and flat cathode members inhibit electrolyte migration in the fuel cell stack.

  8. Method and apparatus for assembling solid oxide fuel cells

    DOEpatents

    Szreders, Bernard E.; Campanella, Nicholas

    1989-01-01

    A plurality of jet air tubes are supported and maintained in a spaced matrix array by a positioning/insertion assembly for insertion in respective tubes of a solid oxide fuel cell (SOFC) in the assembly of an SOFC module. The positioning/insertion assembly includes a plurality of generally planar, elongated, linear vanes which are pivotally mounted at each end thereof to a support frame. The vanes, which each include a plurality of spaced slots along the facing edges thereof, may be pivotally displaced from a generally vertical orientation, wherein each jet air tube is positioned within and engaged by the aligned slots of a plurality of paired upper and lower vanes to facilitate their insertion in respective aligned SOFC tubes arranged in a matrix array, to an inclined orientation, wherein the jet air tubes may be removed from the positioning/insertion assembly after being inserted in the SOFC tubes. A rectangular compression assembly of adjustable size is adapted to receive and squeeze a matrix of SOFC tubes so as to compress the inter-tube nickel felt conductive pads which provide series/parallel electrical connection between adjacent SOFCs, with a series of increasingly larger retainer frames used to maintain larger matrices of SOFC tubes in position. Expansion of the SOFC module housing at the high operating temperatures of the SOFC is accommodated by conductive, flexible, resilient expansion, connector bars which provide support and electrical coupling at the top and bottom of the SOFC module housing.

  9. PWR-2 Blanket Fuel Assembly Removal Safety Basis Criteria Document

    SciTech Connect

    BUSHORE, R.P.

    2001-01-22

    This criteria document describes the proposed format, content, and schedule for the preparation of an amendment to the Interim Safety Basis for Solid Waste Facilities (T Plant) (ISB), (HNF-SD-WM-ISB-006), and to the T Plant Interim Operational Safety Requirements (IOSR) (''F-SD-WM-TSR-003). The amendments to these documents are intended to authorize removal of spent nuclear fuel (SNF) assemblies from the spent fuel pool in the Solid Waste Treatment Facility 221-T canyon for interim storage in the Canister Storage Building (CSB). The amendments will include a stand-alone safety assessment as well as revisions to these safety documents as needed to reflect the changes in work scope not currently authorized to accomplish the expected end-state of the Fuel Removal Project for the 221-T Facility.

  10. Simulation of differential die-away instrument's response to asymmetrically burned spent nuclear fuel

    NASA Astrophysics Data System (ADS)

    Martinik, Tomas; Henzl, Vladimir; Grape, Sophie; Svärd, Staffan Jacobsson; Jansson, Peter; Swinhoe, Martyn T.; Tobin, Stephen J.

    2015-07-01

    Previous simulation studies of Differential Die-Away (DDA) instrument's response to active interrogation of spent nuclear fuel from a pressurized water reactor (PWR) yielded promising results in terms of its capability to accurately measure or estimate basic spent fuel assembly (SFA) characteristics, such as multiplication, initial enrichment (IE) and burn-up (BU) as well as the total plutonium content. These studies were however performed only for a subset of idealized SFAs with a symmetric BU with respect to its longitudinal axis. Therefore, to complement the previous results, additional simulations have been performed of the DDA instrument's response to interrogation of asymmetrically burned spent nuclear fuel in order to determine whether detailed assay of SFAs from all 4 sides will be necessary in real life applications or whether a cost and time saving single sided assay could be used to achieve results of similar quality as previously reported in case of symmetrically burned SFAs. The results of this study suggest that DDA instrument response depends on the position of the individual neutron detectors and in fact can be split in two modes. The first mode, measured by the back detectors, is not significantly sensitive to the spatial distribution of fissile isotopes and neutron absorbers, but rather reflects the total amount of both contributors as in the cases of symmetrically burned SFAs. In contrary, the second mode, measured by the front detectors, yields certain sensitivity to the orientation of the asymmetrically burned SFA inside the assaying instrument. This study thus provides evidence that the DDA instrument can potentially be utilized as necessary in both ways, i.e. a quick determination of the average SFA characteristics in a single assay, as well as a more detailed characterization involving several DDA observables through assay of the SFA from all of its four sides that can possibly map the burn-up distribution and/or identify diversion or

  11. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal: Volume 1, Activation measurements and comparison with calculations for spent fuel assembly hardware

    SciTech Connect

    Luksic, A.

    1989-06-01

    Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly remains that is also radioactive and requires disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report presents a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volumes 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1. 5 refs., 4 figs., 21 tabs.

  12. Control assembly for controlling a fuel cell system during shutdown and restart

    DOEpatents

    Venkataraman, Ramki; Berntsen, George; Carlson, Glenn L.; Farooque, Mohammad; Beachy, Dan; Peterhans, Stefan; Bischoff, Manfred

    2010-06-15

    A fuel cell system and method in which the fuel cell system receives and an input oxidant gas and an input fuel gas, and in which a fuel processing assembly is provided and is adapted to at least humidify the input fuel gas which is to be supplied to the anode of the fuel cell of the system whose cathode receives the oxidant input gas via an anode oxidizing assembly which is adapted to couple the output of the anode of the fuel cell to the inlet of the cathode of the fuel cell during normal operation, shutdown and restart of the fuel cell system, and in which a control assembly is further provided and is adapted to respond to shutdown of the fuel cell system during which input fuel gas and input oxidant gas cease to be received by the fuel cell system, the control assembly being further adapted to, when the fuel cell system is shut down: control the fuel cell system so as to enable a purging gas to be able to flow through the fuel processing assembly to remove humidified fuel gas from the processing assembly and to enable a purging gas to be able to flow through the anode of the fuel cell.

  13. Natural convection heat transfer within horizontal spent nuclear fuel assemblies

    SciTech Connect

    Canaan, R.E.

    1995-12-01

    Natural convection heat transfer is experimentally investigated in an enclosed horizontal rod bundle, which characterizes a spent nuclear fuel assembly during dry storage and/or transport conditions. The basic test section consists of a square array of sixty-four stainless steel tubular heaters enclosed within a water-cooled rectangular copper heat exchanger. The heaters are supplied with a uniform power generation per unit length while the surrounding enclosure is maintained at a uniform temperature. The test section resides within a vacuum/pressure chamber in order to subject the assembly to a range of pressure statepoints and various backfill gases. The objective of this experimental study is to obtain convection correlations which can be used in order to easily incorporate convective effects into analytical models of horizontal spent fuel systems, and also to investigate the physical nature of natural convection in enclosed horizontal rod bundles in general. The resulting data consist of: (1) measured temperatures within the assembly as a function of power, pressure, and backfill gas; (2) the relative radiative contribution for the range of observed temperatures; (3) correlations of convective Nusselt number and Rayleigh number for the rod bundle as a whole; and (4) correlations of convective Nusselt number as a function of Rayleigh number for individual rods within the array.

  14. Nuclear mass inventory, photon dose rate and thermal decay heat of spent research reactor fuel assemblies

    SciTech Connect

    Pond, R.B.; Matos, J.E.

    1996-12-31

    This document has been prepared to assist research reactor operators possessing spent fuel containing enriched uranium of United States origin to prepare part of the documentation necessary to ship this fuel to the United States. Data are included on the nuclear mass inventory, photon dose rate, and thermal decay heat of spent research reactor fuel assemblies. Isotopic masses of U, Np, Pu and Am that are present in spent research reactor fuel are estimated for MTR, TRIGA and DIDO-type fuel assembly types. The isotopic masses of each fuel assembly type are given as functions of U-235 burnup in the spent fuel, and of initial U-235 enrichment and U-235 mass in the fuel assembly. Photon dose rates of spent MTR, TRIGA and DIDO-type fuel assemblies are estimated for fuel assemblies with up to 80% U-235 burnup and specific power densities between 0.089 and 2.857 MW/kg[sup 235]U, and for fission product decay times of up to 20 years. Thermal decay heat loads are estimated for spent fuel based upon the fuel assembly irradiation history (average assembly power vs. elapsed time) and the spent fuel cooling time.

  15. LWR fuel assembly designs for the transmutation of LWR Spent Fuel TRU with FCM and UO{sub 2}-ThO{sub 2} Fuels

    SciTech Connect

    Bae, G.; Hong, S. G.

    2013-07-01

    In this paper, transmutation of transuranic (TRU) nuclides from LWR spent fuels is studied by using LWR fuel assemblies which consist of UO{sub 2}-ThO{sub 2} fuel pins and FCM (Fully Ceramic Microencapsulated) fuel pins. TRU from LWR spent fuel is loaded in the kernels of the TRISO particle fuels of FCM fuel pins. In the FCM fuel pins, the TRISO particle fuels are distributed in SiC matrix having high thermal conductivity. The loading patterns of fuel pins and the fuel compositions are searched to have high transmutation rate and feasible neutronic parameters including pin power peaking, temperature reactivity coefficients, and cycle length. All studies are done only in fuel assembly calculation level. The results show that our fuel assembly designs have good transmutation performances without multi-recycling and without degradation of the safety-related neutronic parameters. (authors)

  16. Fabrication and Assembly of High-Precision Hinge and Latch Joints for Deployable Optical Instruments

    NASA Technical Reports Server (NTRS)

    Phelps, James E.

    1999-01-01

    Descriptions are presented of high-precision hinge and latch joints that have been co-developed, for application to deployable optical instruments, by NASA Langley Research Center and Nyma/ADF. Page-sized versions of engineering drawings are included in two appendices to describe all mechanical components of both joints. Procedures for assembling the mechanical components of both joints are also presented. The information herein is intended to facilitate the fabrication and assembly of the high-precision hinge and latch joints, and enable the incorporation of these joints into the design of deployable optical instrument systems.

  17. High Energy Absorption Top Nozzle For A Nuclaer Fuel Assembly

    DOEpatents

    Sparrow, James A.; Aleshin, Yuriy; Slyeptsov, Aleksey

    2004-05-18

    A high energy absorption top nozzle for a nuclear fuel assembly that employs an elongated upper tubular housing and an elongated lower tubular housing slidable within the upper tubular housing. The upper and lower housings are biased away from each other by a plurality of longitudinally extending springs that are restrained by a longitudinally moveable piston whose upward travel is limited within the upper housing. The energy imparted to the nozzle by a control rod scram is mostly absorbed by the springs and the hydraulic affect of the piston within the nozzle.

  18. Fuel cell cooler assembly and edge seal means therefor

    DOEpatents

    Breault, Richard D.; Roethlein, Richard J.; Congdon, Joseph V.

    1980-01-01

    A cooler assembly for a stack of fuel cells comprises a fibrous, porous coolant tube holder sandwiched between and bonded to at least one of a pair of gas impervious graphite plates. The tubes are disposed in channels which pass through the holder. The channels are as deep as the holder thickness, which is substantially the same as the outer diameter of the tubes. Gas seals along the edges of the holder parallel to the direction of the channels are gas impervious graphite strips.

  19. Gradient isolator for flow field of fuel cell assembly

    DOEpatents

    Ernst, W.D.

    1999-06-15

    Isolator(s) include isolating material and optionally gasketing material strategically positioned within a fuel cell assembly. The isolating material is disposed between a solid electrolyte and a metal flow field plate. Reactant fluid carried by flow field plate channel(s) forms a generally transverse electrochemical gradient. The isolator(s) serve to isolate electrochemically a portion of the flow field plate, for example, transversely outward from the channel(s), from the electrochemical gradient. Further, the isolator(s) serve to protect a portion of the solid electrolyte from metallic ions. 4 figs.

  20. Gradient isolator for flow field of fuel cell assembly

    DOEpatents

    Ernst, William D.

    1999-01-01

    Isolator(s) include isolating material and optionally gasketing material strategically positioned within a fuel cell assembly. The isolating material is disposed between a solid electrolyte and a metal flow field plate. Reactant fluid carried by flow field plate channel(s) forms a generally transverse electrochemical gradient. The isolator(s) serve to isolate electrochemically a portion of the flow field plate, for example, transversely outward from the channel(s), from the electrochemical gradient. Further, the isolator(s) serve to protect a portion of the solid electrolyte from metallic ions.

  1. Fluid flow plate for decreased density of fuel cell assembly

    DOEpatents

    Vitale, Nicholas G.

    1999-01-01

    A fluid flow plate includes first and second outward faces. Each of the outward faces has a flow channel thereon for carrying respective fluid. At least one of the fluids serves as reactant fluid for a fuel cell of a fuel cell assembly. One or more pockets are formed between the first and second outward faces for decreasing density of the fluid flow plate. A given flow channel can include one or more end sections and an intermediate section. An interposed member can be positioned between the outward faces at an interface between an intermediate section, of one of the outward faces, and an end section, of that outward face. The interposed member can serve to isolate the reactant fluid from the opposing outward face. The intermediate section(s) of flow channel(s) on an outward face are preferably formed as a folded expanse.

  2. Cerium migration during PEM fuel cell assembly and operation

    DOE PAGESBeta

    Baker, Andrew M.; Torraco, Dennis; Judge, Elizabeth J.; Spernjak, Dusan; Mukundan, Rangachary; Borup, Rod L.; Advani, Suresh G.; Prasad, Ajay K.

    2015-10-02

    Cerium migration between PEM fuel cell components is influenced by potential-driven mobility, ionic diffusion, and gradients in water content. These factors were investigated in ex situ experiments and in operating fuel cells. Potential-induced migration was measured ex situ in hydrated window cells. Cerium-containing MEAs were also fabricated and tested under ASTs. MEA disassembly and subsequent XRF analysis were used to observe rapid cerium migration during cell assembly and operation. During MEA hot pressing, humidification, and low RH operation at OCV, ionic diffusion causes uniform migration from the membrane into the catalyst layers. During high RH operation at OCV, in-plane ceriummore » gradients arise due to variations in water content. These gradients may diminish the scavenging efficacy of cerium by reducing its proximity to generated radicals.« less

  3. Cerium migration during PEM fuel cell assembly and operation

    SciTech Connect

    Baker, Andrew M.; Torraco, Dennis; Judge, Elizabeth J.; Spernjak, Dusan; Mukundan, Rangachary; Borup, Rod L.; Advani, Suresh G.; Prasad, Ajay K.

    2015-09-14

    Cerium migration between PEM fuel cell components is influenced by potential-driven mobility, ionic diffusion, and gradients in water content. These factors were investigated in ex situ experiments and in operating fuel cells. Potential-induced migration was measured ex situ in hydrated window cells. Cerium-containing MEAs were also fabricated and tested under ASTs. MEA disassembly and subsequent XRF analysis were used to observe rapid cerium migration during cell assembly and operation. During MEA hot pressing, humidification, and low RH operation at OCV, ionic diffusion causes uniform migration from the membrane into the catalyst layers. During high RH operation at OCV, in-plane cerium gradients arise due to variations in water content. These gradients may diminish the scavenging efficacy of cerium by reducing its proximity to generated radicals.

  4. Simplified process for leaching precious metals from fuel cell membrane electrode assemblies

    DOEpatents

    Shore, Lawrence; Matlin, Ramail

    2009-12-22

    The membrane electrode assemblies of fuel cells are recycled to recover the catalyst precious metals from the assemblies. The assemblies are cryogenically embrittled and pulverized to form a powder. The pulverized assemblies are then mixed with a surfactant to form a paste which is contacted with an acid solution to leach precious metals from the pulverized membranes.

  5. Membrane electrode assemblies for unitised regenerative polymer electrolyte fuel cells

    NASA Astrophysics Data System (ADS)

    Wittstadt, U.; Wagner, E.; Jungmann, T.

    Membrane electrode assemblies for regenerative polymer electrolyte fuel cells were made by hot pressing and sputtering. The different MEAs are examined in fuel cell and water electrolysis mode at different pressure and temperature conditions. Polarisation curves and ac impedance spectra are used to investigate the influence of the changes in coating technique. The hydrogen gas permeation through the membrane is determined by analysing the produced oxygen in electrolysis mode. The analysis shows, that better performances in both process directions can be achieved with an additional layer of sputtered platinum on the oxygen electrode. Thus, the electrochemical round-trip efficiency can be improved by more than 4%. Treating the oxygen electrode with PTFE solution shows better performance in fuel cell and less performance in electrolysis mode. The increase of the round-trip efficiency is negligible. A layer sputtered directly on the membrane shows good impermeability, and hence results in high voltages at low current densities. The mass transportation is apparently constricted. The gas diffusion layer on the oxygen electrode, in this case a titanium foam, leads to flooding of the cell in fuel cell mode. Stable operation is achieved after pretreatment of the GDL with a PTFE solution.

  6. METHOD AND MEANS FOR SUPPORTING REACTOR FUEL CONTAINERS IN AN ASSEMBLY

    DOEpatents

    Currier, E.L. Jr.; Nicklas, J.H.; Coombs, C.A.

    1962-12-11

    This patent relates to means for supporting fuelcontaining tubes in an assembly which include grid means at either end of the fuel element assembly antl improved grid means intermediate of the ends to provide support against lateral displacement. (AEC)

  7. Californium interrogation prompt neutron (CIPN) instrument for non-destructive assay of spent nuclear fuel-Design concept and experimental demonstration

    NASA Astrophysics Data System (ADS)

    Henzlova, D.; Menlove, H. O.; Rael, C. D.; Trellue, H. R.; Tobin, S. J.; Park, Se-Hwan; Oh, Jong-Myeong; Lee, Seung-Kyu; Ahn, Seong-Kyu; Kwon, In-Chan; Kim, Ho-Dong

    2016-01-01

    This paper presents results of the first experimental demonstration of the Californium Interrogation Prompt Neutron (CIPN) instrument developed within a multi-year effort launched by the Next Generation Safeguards Initiative Spent Fuel Project of the United States Department of Energy. The goals of this project focused on developing viable non-destructive assay techniques with capabilities to improve an independent verification of spent fuel assembly characteristics. For this purpose, the CIPN instrument combines active and passive neutron interrogation, along with passive gamma-ray measurements, to provide three independent observables. This paper describes the initial feasibility demonstration of the CIPN instrument, which involved measurements of four pressurized-water-reactor spent fuel assemblies with different levels of burnup and two initial enrichments. The measurements were performed at the Post-Irradiation Examination Facility at the Korea Atomic Energy Institute in the Republic of Korea. The key aim of the demonstration was to evaluate CIPN instrument performance under realistic deployment conditions, with the focus on a detailed assessment of systematic uncertainties that are best evaluated experimentally. The measurements revealed good positioning reproducibility, as well as a high degree of insensitivity of the CIPN instrument's response to irregularities in a radial burnup profile. Systematic uncertainty of individual CIPN instrument signals due to assembly rotation was found to be <4.5%, even for assemblies with fairly extreme gradients in the radial burnup profile. These features suggest that the CIPN instrument is capable of providing a good representation of assembly average characteristics, independent of assembly orientation in the instrument.

  8. An integrated approach for determining plutonium mass in spent fuel assemblies with nondestructive assay

    SciTech Connect

    Swinhoe, Martyn T; Tobin, Stephen J; Fensin, Mike L; Menlove, Howard O

    2009-01-01

    be part of a system that cost-effectively meets the burnup credit needs of a repository. Behind each of these reasons is a regulatory structure with MC&A requirements. In the case of the IAEA, the accountable quantity is elemental plutonium. The material in spent fuel (fissile isotopes, fission products, etc.) emits signatures that provide information about the content and history of the fuel. A variety of nondestructive assay (NDA) techniques are available to quantify these signatures. The effort presented in this paper is investigation of the capabilities of 12 NDA techniques. For these 12, none is conceptually capable of independently determining the Pu content in a spent fuel assembly while at the same time being able to detect the diversion of a significant quantity of rods. For this reason the authors are investigating the capability of 12 NDA techniques with the end goal of integrating a few techniques together into a system that is capable of measuring Pu mass in an assembly. The work described here is the beginning of what is anticipated to be a five year effort: (1) two years of modeling to select the best technologies, (2) one year fabricating instruments and (3) two years measuring spent fuel. This paper describes the first two years of this work. In order to cost effectively and robustly model the performance of the 12 NDA techniques, an 'assembly library' was created. The library contains the following: (a) A diverse range of PWR spent fuel assemblies (burnup, enrichment, cooling time) similar to that which exists in spent pools today and in the future. (b) Diversion scenarios that capture a range of possible rod removal options. (c) The spatial and isotopic detail needed to accurately quantify the capability of all the NDA techniques so as to enable integration. It is our intention to make this library available to other researchers in the field for inter-comparison purposes. The performance of each instrument will be quantified for the full assembly

  9. 96. SEED 1 FUEL ASSEMBLY FROM LOCATION L9 BEING REMOVED ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    96. SEED 1 FUEL ASSEMBLY FROM LOCATION L-9 BEING REMOVED FROM REACTOR VESSEL BY MEANS OF FUEL EXTRACTION CRANE, JANUARY 7, 1960 - Shippingport Atomic Power Station, On Ohio River, 25 miles Northwest of Pittsburgh, Shippingport, Beaver County, PA

  10. Between-cycle laser system for depressurization and resealing of modified design nuclear fuel assemblies

    DOEpatents

    Bradley, John G.

    1982-01-01

    A laser beam is used to puncture fuel cladding for release of contained pressurized fission gas from plenum sections or irradiated fuel pins. Exhausted fission gases are collected and trapped for safe disposal. The laser beam, adjusted to welding mode, is subsequently used to reseal the puncture holes. The fuel assembly is returned to additional irradiation or, if at end of reactivity lifetime, is routed to reprocess. The fuel assembly design provides graded cladding lengths, by rows or arrays, such that the cladding of each component fuel element of the assembly is accessible to laser beam reception.

  11. Differential Die-Away Instrument: Report on Initial Simulations of Spent Fuel Experiment

    SciTech Connect

    Goodsell, Alison V.; Henzl, Vladimir; Swinhoe, Martyn T.

    2014-04-01

    New Monte Carlo simulations of the differential die-away (DDA) instrument response to the assay of spent and fresh fuel helped to redefine the signal-to-Background ratio and the effects of source neutron tailoring on the system performance. Previously, burst neutrons from the neutron generator together with all neutrons from a fission chain started by a fast fission of 238U were considered to contribute to active background counts. However, through additional simulations, the magnitude of the 238U first fission contribution was found to not affect the DDA performance in reconstructing 239Pueff. As a result, the newly adopted DDA active background definition considers now any neutrons within a branch of the fission chain that does not include at least one fission event induced by a thermal neutron, before being detected, to be the active background. The active background, consisting thus of neutrons from a fission chain or its individual branches composed entirely of sequence of fast fissions on any fissile or fissionable nuclei, is not expected to change significantly with different fuel assemblies. Additionally, while source tailoring materials surrounding the neutron generator were found to influence and possibly improve the instrument performance, the effect was not substantial.

  12. Changes to Irradiation Conditions of VVER-1000 Surveillance Specimens Resulting from Fuel Assemblies with Greater Fuel Height

    NASA Astrophysics Data System (ADS)

    Panferov, Pavel; Kochkin, Viacheslav; Erak, Dmitry; Makhotin, Denis; Reshetnikov, Alexandr; Timofeev, Andrey

    2016-02-01

    The goal of the work was to obtain experimental data on the influence of newtype fuel assemblies with higher fuel rods on the irradiation conditions of surveillance specimens installed on the baffe of VVER-1000. For this purpose, two surveillance sets with container assemblies of the same design irradiated in reactors with different fuel assemblies in the core were investigated. Measurements of neutron dosimeters from these sets and retrospective measurements of 54Mn activity accumulated in each irradiated specimen allow a detailed distribution of the fast neutron flux in the containers to be obtained. Neutron calculations have been done using 3D discrete ordinate code KATRIN. On the basis of the obtained results, a change of the lead factor due to newtype fuel assemblies was evaluated for all types of VVER-1000 container assemblies.

  13. Cap assembly for a bundled tube fuel injector

    DOEpatents

    LeBegue, Jeffrey Scott; Melton, Patrick Benedict; Westmoreland, III, James Harold; Flanagan, James Scott

    2016-04-26

    A cap assembly for a bundled tube fuel injector includes an impingement plate and an aft plate that is disposed downstream from the impingement plate. The aft plate includes a forward side that is axially separated from an aft side. A tube passage extends through the impingement plate and the aft plate. A tube sleeve extends through the impingement plate within the tube passage towards the aft plate. The tube sleeve includes a flange at a forward end and an aft end that is axially separated from the forward end. A retention plate is positioned upstream from the impingement plate. A spring is disposed between the retention plate and the flange. The spring provides a force so as to maintain contact between at least a portion of the aft end of the tube sleeve and the forward side of the aft plate.

  14. Air Transport of Spent Nuclear Fuel (SNF) Assemblies

    SciTech Connect

    Haire, M.J.; Moses, S.D.; Shapovalov, V.I.; Morenko, A.

    2007-07-01

    Sometimes the only feasible means of shipping research reactor spent nuclear fuel (SNF) among countries is via air transport because of location or political conditions. The International Atomic Energy Agency (IAEA) has established a regulatory framework to certify air transport Type C casks. However, no such cask has been designed, built, tested, and certified. In lieu of an air transport cask, research reactor SNF has been transported using a Type B cask under an exemption with special arrangements for administrative and security controls. This work indicates that it may be feasible to transport commercial power reactor SNF assemblies via air, and that the cost is only about three times that of shipping it by railway. Optimization (i.e., reduction) of this cost factor has yet to be done. (authors)

  15. Measurement of gamma and neutron radiations inside spent fuel assemblies with passive detectors

    NASA Astrophysics Data System (ADS)

    Viererbl, L.; Lahodová, Z.; Voljanskij, A.; Klupák, V.; Koleška, M.; Cabalka, M.; Turek, K.

    2011-10-01

    During operation of a fission nuclear reactor, many radionuclides are generated in fuel by fission and activation of 235U, 238U and other nuclides present in the assembly. After removal of a fuel assembly from the core, these radionuclides are sources of different types of radiation. Gamma and neutron radiation emitted from an assembly can be non-destructively detected with different types of detectors. In this paper, a new method of measurement of radiation from a spent fuel assembly is presented. It is based on usage of passive detectors, such as alanine dosimeters for gamma radiation and track detectors for neutron radiation. Measurements are made on the IRT-2M spent fuel assemblies used in the LVR-15 research reactor. During irradiation of detectors, the fuel assembly is located in a water storage pool at a depth of 6 m. Detectors are inserted into central hole of the assembly, irradiated for a defined time interval, and after the detectors removed from the assembly, gamma dose or neutron fluence are evaluated. Measured profiles of gamma dose rate and neutron fluence rate inside of the spent fuel assembly are presented. This measurement can be used to evaluate relative fuel burn-up.

  16. Hydrogen fuel reforming in a fog cooled fuel cell power plant assembly

    SciTech Connect

    Levy, A.H.; Wertheim, R.J.

    1989-09-12

    This patent describes a high pressure phosphoric acid fuel cell stack assembly. The cell comprising: a stack of fuel cells for producing electricity. The stack including cathode means, anode means, and the stack being formed without a separate cooling system; means for delivering a pressurized air supply to the cathode means; means for delivering a hydrogen rich fuel gas to the anode means for electrochemically reacting with oxygen in the pressurized air to produce electricity and water; first exhaust means for removing a mixture of oxygen-depleted air and product water from the cathode means; means for delivering a water fog stream to the anode means for mixture with the hydrogen rich fuel gas. The water fog stream being evaporated in the anode means to cool the stack; means for exhausting a mixture of hydrogen-depleted gas and water vapor from the anode means; reformer means for reforming a raw hydrocarbon fuel to the hydrogen rich fuel gas; and means for delivering the mixture of hydrogen-depleted exhaust gas and water vapor to the reformer means to provide water for the reforming reaction.

  17. Modeling the effect of engine assembly mass on engine friction and vehicle fuel economy

    NASA Astrophysics Data System (ADS)

    An, Feng; Stodolsky, Frank

    An analytical model is developed to estimate the impact of reducing engine assembly mass (the term engine assembly refers to the moving components of the engine system, including crankshafts, valve train, pistons, and connecting rods) on engine friction and vehicle fuel economy. The relative changes in frictional mean effective pressure and fuel economy are proportional to the relative change in assembly mass. These changes increase rapidly as engine speed increases. Based on the model, a 25% reduction in engine assembly mass results in a 2% fuel economy improvement for a typical mid-size passenger car over the EPA Urban and Highway Driving Cycles.

  18. Buoyancy-driven flow excursions in fuel assemblies. Revision 1

    SciTech Connect

    Laurinat, J.E.; Paul, P.K.; Menna, J.D.

    1995-07-01

    A power limit criterion was developed for a postulated Loss of Pumping Accident (LOPA) in one of the recently shut down heavy water production reactors at the Savannah River Site. These reactors were cooled by recirculating heavy water moderator downward through channels in cylindrical fuel tubes. Powers were limited to safeguard against a flow excursion in one or more of these parallel channels. During full-power operation, limits safeguarded against a boiling flow excursion. At low flow rates, during the addition of emergency cooling water, buoyant forces reverse the flow in one of the coolant channels before boiling occurs. As power increases beyond the point of flow reversal, the maximum wall temperature approaches the fluid saturation temperature, and a thermal excursion occurs. The power limit criterion for low flow rates was the onset of flow reversal. To determine conditions for flow reversal, tests were performed in a mock-up of a fuel assembly that contained two electrically heated concentric tubes surrounded by three flow channels. These tests were modeled using a finite difference thermal-hydraulic code. According to code calculations, flow reversed in the outer flow channel before the maximum wall temperature reached the local fluid saturation temperature. Thermal excursions occurred when the maximum wall temperature approximately equaled the saturation temperature. For a postulated LOPA, the flow reversal criterion for emergency cooling water addition was more limiting than the boiling excursion criterion for full power operation. This criterion limited powers to 37% of the limiting power for previous long-term reactor operations.

  19. Buoyancy-driven flow excursions in fuel assemblies

    SciTech Connect

    Laurinat, J.E.; Paul, P.K.; Menna, J.D.

    1995-09-01

    A power limit criterion was developed for a postulated Loss of Pumping Accident (LOPA) in one of the recently shut down heavy water production reactors at the Savannah River Site. These reactors were cooled by recirculating heavy water moderator downward through channels in cylindrical fuel tubes. Powers were limited to safeguard against a flow excursion in one of more of these parallel channels. During-full-power operation, limits safeguarded against a boiling flow excursion. At low flow rates, during the addition of emergency cooling water, buoyant forces reverse the flow in one of the coolant channels before boiling occurs. As power increased beyond the point of flow reversal, the maximum wall temperature approaches the fluid saturation temperature, and a thermal excursion occurs. The power limit criterion for low flow rates was the onset of flow reversal. To determine conditions for flow reversal, tests were performed in a mock-up of a fuel assembly that contained two electrically heated concentric tubes surrounded by three flow channels. These tests were modeled using a finite difference thermal-hydraulic code. According to code calculations, flow reversed in the outer flow channel before the maximum wall temperature reached the local fluid saturation temperature. Thermal excursions occurred when the maximum wall temperature approximately equaled the saturation temperature. For a postulated LOPA, the flow reversal criterion for emergency cooling water addition was more limiting than the boiling excursion criterion for full power operation. This criterion limited powers to 37% of the limiting power for previous long-term reactor operations.

  20. Performance of boiling water reactor fuel lead test assemblies to 35 MWd/kg U

    SciTech Connect

    Rowland, T.C.; Ikemoto, R.N.; Gehl, S.

    1986-01-01

    This joint Electric Power Research Institute/General Electric (EPRI/GE) fuel performance program involved thorough preirradiation characterization of fuel used in lead test assemblies (LTAs), detailed surveillance of their operation, and interim site examinations of the assemblies during reactor outages. The program originally included four GE-5 LTAs operating in the Peach Bottom-2 (PB-2) reactor. The program was later modified to include the pressurized fuel rod test assembly in the Peach Bottom-3 (PB-3) reactor. The program modification also included extending the operation of the PB-2 and PB-3 LTA fuel beyond normal discharge exposures. Results are summarized in the paper.

  1. Sensitivity and System Response of Pin Power Peaking in VVER-1000 Fuel Assembly Using TSUNAMI-2D

    NASA Astrophysics Data System (ADS)

    Frybort, J.

    2014-04-01

    Pin power peaking in a VVER-1000 fuel assembly and its sensitivity and uncertainty was analyzed by TSUNAMI-2D code. Several types of fuel assemblies were considered. They differ in number and position of gadolinium fuel pins. The calculations were repeated for several fuel compositions obtained by fuel depletion calculation. The results are quantified sensitivity data, which can be used for enrichment profiling.

  2. A new optical axle measuring instrument for wheel alignment in assembly-line production

    SciTech Connect

    Bruhn, H.; Felske, A.

    1985-01-01

    The newly developed optical measuring system allows adjustment of front and rear wheel angularities - toe, camber, caster - in assembly-line production. There is no need to align the car, since the measuring base for the angle alignment is formed by the car itself. Defined spring compression values and direct caster angle determination lead to higher accuracy. Adjustment is carried out directly on the assembly line. Measuring pits are not required. The working time for each car and the working area required, which are important cost factors, are markedly lower than with conventional instruments. The axle measuring system was developed for the VW Vanagon, but can also be used for passenger car chassis. The measuring principle, the optical and mechanical design of the device, and a statistical analysis of over 100 cars aligned by means of this system are described in comparison with conventional measuring instruments.

  3. Analysis of a spacecraft instrument ball bearing assembly lubricated by a perfluoroalkylether

    NASA Technical Reports Server (NTRS)

    Morales, W.; Jones, W. R., Jr.; Buckley, D. H.

    1986-01-01

    An analysis of a spacecraft instrument ball bearing assembly, subjected to a scanning life test, was performed to determine the possible case of rotational problems involving these units aboard several satellites. The analysis indicated an ineffective transfer of a fluorinated liquid lubricant from a phenolic retainer to the bearing balls. Part of the analysis led to a novel HPLC separation method employing a fluorinated mobile phase in conjunction with silica based size exclusion columns.

  4. Saturn V Instrument Unit for the Apollo 4 Mission in the Vehicle Assembly Building

    NASA Technical Reports Server (NTRS)

    1967-01-01

    This photograph was taken during the final assembly operation of the Saturn V launch vehicle for the Apollo 4 (SA 501) mission. The instrument unit (IU) was mated atop the S-IC/S-II assembly in the Vehicle Assembly Building high bay at the Kennedy Space Center. The Apollo 4 mission was the first launch of the Saturn V launch vehicle. Objectives of the unmanned Apollo 4 test flight were to obtain flight information on launch vehicle and spacecraft structural integrity and compatibility, flight loads, stage separation, and subsystems operation including testing of restart of the S-IVB stage, and to evaluate the Apollo command module heat shield. The Apollo 4 was launched on November 9, 1967 from KSC.

  5. Saturn V Instrument Unit for the Apollo 4 Mission in the Vehicle Assembly Building

    NASA Technical Reports Server (NTRS)

    1967-01-01

    This photograph was taken during the final assembly operation of the Saturn V launch vehicle for the Apollo 4 (SA 501) mission. The instrument unit (IU) was hoisted to be mated to the S-IC/S-II assembly in the Vehicle Assembly Building high bay at the Kennedy Space Center. The Apollo 4 mission was the first launch of the Saturn V launch vehicle. Objectives of the unmanned Apollo 4 test flight were to obtain flight information on launch vehicle and spacecraft structural integrity and compatibility, flight loads, stage separation, and subsystems operation including testing of restart of the S-IVB stage, and to evaluate the Apollo command module heat shield. The Apollo 4 was launched on November 9, 1967 from KSC.

  6. Fuel assembly for the production of tritium in light water reactors

    DOEpatents

    Cawley, W.E.; Trapp, T.J.

    1983-06-10

    A nuclear fuel assembly is described for producing tritium in a light water moderated reactor. The assembly consists of two intermeshing arrays of subassemblies. The first subassemblies comprise concentric annular elements of an outer containment tube, an annular target element, an annular fuel element, and an inner neutron spectrums shifting rod. The second subassemblies comprise an outer containment tube and an inner rod of either fuel, target, or neutron spectrum shifting neutral.

  7. Fuel assembly for the production of tritium in light water reactors

    DOEpatents

    Cawley, William E.; Trapp, Turner J.

    1985-01-01

    A nuclear fuel assembly is described for producing tritium in a light water moderated reactor. The assembly consists of two intermeshing arrays of subassemblies. The first subassemblies comprise concentric annular elements of an outer containment tube, an annular target element, an annular fuel element, and an inner neutron spectrums shifting rod. The second subassemblies comprise an outer containment tube and an inner rod of either fuel, target, or neutron spectrum shifting neutral.

  8. A high converter concept for fuel management with blanket fuel assemblies in boiling water reactors

    SciTech Connect

    Martinez-Frances, N.; Timm, W.; Rossbach, D.

    2012-07-01

    Studies on the natural Uranium saving and waste reduction potential of a multiple-plant BWR system were performed. The BWR High Converter system should enable a multiple recycling of MOX fuel in current BWR plants by introducing blanket fuel assemblies and burning Uranium and MOX fuel separately. The feasibility of Uranium cores with blankets and full-MOX cores with Plutonium qualities as low as 40% were studied. The power concentration due to blanket insertion is manageable with modern fuel and acceptable values for the thermal limits and reactivity coefficients were obtained. While challenges remain, full-MOX cores also complied with the main design criteria. The combination of Uranium and Plutonium burners in appropriate proportions could enable obtaining as much as 40% more energy out of Uranium ore. Moreover, a proper adjustment of blanket average stay and Plutonium qualities could lead to a system with nearly no Plutonium left for final disposal. The achievement of such goals with current light water technology makes the BWR HC concept an attractive option to improve the fuel cycle until Gen-IV designs are mature. (authors)

  9. Refinements to temperature calculations of spent fuel assemblies when in a stagnant gas environment

    SciTech Connect

    Rhodes, C.A.; Haire, M.J.

    1984-01-01

    Undesirably high temperatures are possible in irradiated fuel assemblies because of the radioactive decay of fission products formed while in the reactor. The COXPRO computer code has been used for some time to calculate temperatures in spent fuel when the fuel is suspended in a stagnant gas environment. This code assumed radiation to be the only mode of heat dissipation within the fuel pin bundle. Refinements have been made to include conduction as well as radiation heat transfer within this code. Comparison of calculated and measured temperatures in four separate and independent tests indicate that maximum fuel assembly temperatures can be predicted to within about 6%. 2 references, 5 figures.

  10. Component-Level Electronic-Assembly Repair (CLEAR) Synthetic Instrument Capabilities Assessment and Test Report

    NASA Technical Reports Server (NTRS)

    Oeftering, Richard C.; Bradish, Martin A.

    2011-01-01

    The role of synthetic instruments (SIs) for Component-Level Electronic-Assembly Repair (CLEAR) is to provide an external lower-level diagnostic and functional test capability beyond the built-in-test capabilities of spacecraft electronics. Built-in diagnostics can report faults and symptoms, but isolating the root cause and performing corrective action requires specialized instruments. Often a fault can be revealed by emulating the operation of external hardware. This implies complex hardware that is too massive to be accommodated in spacecraft. The SI strategy is aimed at minimizing complexity and mass by employing highly reconfigurable instruments that perform diagnostics and emulate external functions. In effect, SI can synthesize an instrument on demand. The SI architecture section of this document summarizes the result of a recent program diagnostic and test needs assessment based on the International Space Station. The SI architecture addresses operational issues such as minimizing crew time and crew skill level, and the SI data transactions between the crew and supporting ground engineering searching for the root cause and formulating corrective actions. SI technology is described within a teleoperations framework. The remaining sections describe a lab demonstration intended to show that a single SI circuit could synthesize an instrument in hardware and subsequently clear the hardware and synthesize a completely different instrument on demand. An analysis of the capabilities and limitations of commercially available SI hardware and programming tools is included. Future work in SI technology is also described.

  11. Analysis of fuel relocation for the NRC/PNL Halden assemblies IFA-431, IFA-432, and IFA-513

    SciTech Connect

    Williford, R.E.; Mohr, C.L.; Lanning, D.D.; Cunningham, M.E.; Rausch, W.N.; Bradley, E.R.

    1980-04-01

    The effects of the thermally-induced cracking and subsequent relocation of UO/sub 2/ fuel pellets on the thermal and mechanical behavior of light-water reactor fuel rods during irradiation are quantified in this report. Data from the Nuclear Regulatory Commission/Pacific Northwest Laboratory Halden experiments on instrumented fuel assemblies (IFA) IFA-431, IFA-432, and IFA-513 are analyzed. Beginning-of-life in-reactor measurements of fuel center temperatures, linear heat ratings, and cladding axial elongations are used in a new model to solve for the effective thermal conductivity and elastic moduli of the cracked fuel column. The effective thermal conductivity and elastic moduli for the cracked fuel were found to be significantly reduced from the values for solid UO/sub 2/ pellets. The calculated fuel-cladding gap remained relatively constant (closed) with respect to power level, indicating that the fuel fragments do not retreat from the cladding when the power/temperature is reduced. Recommendations are made pertaining to the work required to further refine the model. 30 refs., 81 figs., 8 tabs.

  12. Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs

    DOE PAGESBeta

    Mertyurek, Ugur; Gauld, Ian C.

    2015-12-24

    In this research, ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup.more » The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. Finally, the nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.« less

  13. Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs

    SciTech Connect

    Mertyurek, Ugur; Gauld, Ian C.

    2015-12-24

    In this research, ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup. The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. Finally, the nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.

  14. Integrated Radiation Transport and Nuclear Fuel Performance for Assembly-Level Simulations

    SciTech Connect

    Clarno, Kevin T; Hamilton, Steven P; Philip, Bobby; Berrill, Mark A; Sampath, Rahul S; Allu, Srikanth; Pugmire, Dave; Dilts, Gary; Banfield, James E

    2012-02-01

    The Advanced Multi-Physics (AMP) Nuclear Fuel Performance code (AMPFuel) is focused on predicting the temperature and strain within a nuclear fuel assembly to evaluate the performance and safety of existing and advanced nuclear fuel bundles within existing and advanced nuclear reactors. AMPFuel was extended to include an integrated nuclear fuel assembly capability for (one-way) coupled radiation transport and nuclear fuel assembly thermo-mechanics. This capability is the initial step toward incorporating an improved predictive nuclear fuel assembly modeling capability to accurately account for source-terms and boundary conditions of traditional (single-pin) nuclear fuel performance simulation, such as the neutron flux distribution, coolant conditions, and assembly mechanical stresses. A novel scheme is introduced for transferring the power distribution from the Scale/Denovo (Denovo) radiation transport code (structured, Cartesian mesh with smeared materials within each cell) to AMPFuel (unstructured, hexagonal mesh with a single material within each cell), allowing the use of a relatively coarse spatial mesh (10 million elements) for the radiation transport and a fine spatial mesh (3.3 billion elements) for thermo-mechanics with very little loss of accuracy. In addition, a new nuclear fuel-specific preconditioner was developed to account for the high aspect ratio of each fuel pin (12 feet axially, but 1 4 inches in diameter) with many individual fuel regions (pellets). With this novel capability, AMPFuel was used to model an entire 17 17 pressurized water reactor fuel assembly with many of the features resolved in three dimensions (for thermo-mechanics and/or neutronics), including the fuel, gap, and cladding of each of the 264 fuel pins; the 25 guide tubes; the top and bottom structural regions; and the upper and lower (neutron) reflector regions. The final, full assembly calculation was executed on Jaguar using 40,000 cores in under 10 hours to model over 162

  15. Dendritic assembly of gold nanoparticles during fuel-forming electrocatalysis.

    PubMed

    Manthiram, Karthish; Surendranath, Yogesh; Alivisatos, A Paul

    2014-05-21

    We observe the dendritic assembly of alkanethiol-capped gold nanoparticles on a glassy carbon support during electrochemical reduction of protons and CO2. We find that the primary mechanism by which surfactant-ligated gold nanoparticles lose surface area is by taking a random walk along the support, colliding with their neighbors, and fusing to form dendrites, a type of fractal aggregate. A random walk model reproduces the fractal dimensionality of the dendrites observed experimentally. The rate at which the dendrites form is strongly dependent on the solubility of the surfactant in the electrochemical double layer under the conditions of electrolysis. Since alkanethiolate surfactants reductively desorb at potentials close to the onset of CO2 reduction, they do not poison the catalytic activity of the gold nanoparticles. Although catalyst mobility is typically thought to be limited for room-temperature electrochemistry, our results demonstrate that nanoparticle mobility is significant under conditions at which they electrochemically catalyze gas evolution, even in the presence of a high surface area carbon and binder. A careful understanding of the electrolyte- and polarization-dependent nanoparticle aggregation kinetics informs strategies for maintaining catalyst dispersion during fuel-forming electrocatalysis. PMID:24766431

  16. Numerical simulation of gas dynamics and heat exchange tasks in fuel assemblies of the nuclear reactors

    SciTech Connect

    Zhuchenko, S. V.

    2014-11-12

    This report presents a PC-based program for solution gas dynamics and heat exchange mathematical tasks in fuel assemblies of the fast-neutron nuclear reactors. A fuel assembly consisting of bulk heat-generating elements, which are integrated together by the system of supply and pressure manifolds, is examined. Spherical heat-generating microelements, which contain nuclear fuel, are pulled into the heat-generating elements. Gaseous coolant proceed from supply manifolds to heat-generating elements, where it withdraws the nuclear reaction heat and assembles in pressure manifolds.

  17. Space Qualification of the Optical Filter Assemblies for the ICESat-2/ATLAS Instrument

    NASA Technical Reports Server (NTRS)

    Troupaki, Elisavet; Denny, Zachary; Wu, Stewart; Bradshaw, Heather; Smith, Kevin; Hults, Judy; Ramos-Izquierdo, Luis; Cook, William

    2015-01-01

    The Advanced Topographic Laser Altimeter System (ATLAS) will be the only instrument on the Ice, Cloud, and Land Elevation Satellite -2 (ICESat-2). ICESat-2 is the 2nd-generation of the orbiting laser altimeter ICESat, which will continue polar ice topography measurements with improved precision laser-ranging techniques. In contrast to the original ICESat design, ICESat-2 will use a micro-pulse, multi-beam approach that provides dense cross-track sampling to help scientists determine a surface's slope with each pass of the satellite. The ATLAS laser will emit visible, green laser pulses at a wavelength of 532 nm and a rate of 10 kHz and will be split into 6 beams. A set of six identical, thermally-tuned etalon filter assemblies will be used to remove background solar radiation from the collected signal while transmitting the laser light to the detectors. A seventh etalon assembly will be used to monitor the laser center wavelength during the mission. In this paper, we present the design and optical performance measurements of the ATLAS optical filter assemblies (OFA) in air and in vacuum before integration on the ATLAS instrument.

  18. Assembly and test of MEDUSA, a multi-spectral instrument for stratospheric Earth observation

    NASA Astrophysics Data System (ADS)

    Delauré, B.; Van Achteren, T.; Everaerts, J.; Livens, S.; Beghuin, D.

    2008-10-01

    MEDUSA is a lightweight high resolution camera, designed to operate at stratospheric altitudes mounted on a solar-powered unmanned aerial vehicle (UAV). The MEDUSA instrument targets applications such as crisis monitoring and large scale mapping, requiring high resolution images with regional coverage, flexible flight patterns, high update rates and long mission lengths (weeks to months). The instrument is subject to severe constraints on mass (< 2,5 kg), volume, power consumption and survivability in the stratospheric environment. Operating temperatures within the payload vary over several tens of degrees Celsius over the day-night cycle. Nonetheless, the instrument will be able to provide panchromatic and color images of 30 cm ground resolution at an altitude of 18000 m and a wide swath of 3000m. This ESA-PRODEX (PROgramme de Développement d'Expériences scientifiques) funded project successfully passed the Critical Design Review in September 2007 and the assembly, integration and test (AIT) phase of the subsystems will be finalized by the end of 2008. Subsystem tests include an optical performance verification performed on optical compartment with a test-sensor assembly both at ambient and operational environmental conditions. The electronic subsystems and their interfaces are tested for functionality and performance in the operational temperature and pressure range. From early 2009 onwards, the MEDUSA system will be fully integrated including a custom designed wide swath MEDUSA CMOS frame sensor (10000x1200 pixels). The MEDUSA instrument will be ready for its first flight in spring of 2009. The detailed design of the optical instrument and its performances have been discussed in [1]. In this paper we will give an overview of the AIT status of the MEDUSA sensor and the optical system and an outlook on the system integration and test phase.

  19. Method for providing concentricity of pilot fuel assembly in a combustor

    NASA Technical Reports Server (NTRS)

    Halila, Ely E. (Inventor); Anderson, Michael (Inventor); Martus, James A. (Inventor)

    2003-01-01

    Concentric installation of a pilot fuel assembly in an opening in a gas turbine combustor casing is achieved by providing a boss having at least two flat surfaces which are perpendicular to each other on the combustor casing surrounding the opening and a mounting flange having at least two flat surfaces which are perpendicular to each other on the pilot fuel assembly. The pilot fuel assembly is concentrically installed to the combustor casing by inserting the assembly into the combustor casing opening, and moving the pilot fuel assembly as far as it will go in a first direction substantially parallel to one of the flat boss surfaces. The distance between the other flat boss surface and one of the flat flange surfaces is then taken. Next, the pilot fuel assembly is moved in the direction opposite the first direction, at which point, the distance between the same two flat surfaces is again measured. Lastly, the pilot fuel assembly is located at a position where the distance between the two measuring surfaces is equal to the average of the first and second measurements. If desired, these steps can be repeated back and forth along an axis perpendicular to the first and second directions.

  20. Fuel nozzle assembly for use as structural support for a duct structure in a combustor of a gas turbine engine

    DOEpatents

    Wiebe, David J; Fox, Timothy A

    2015-03-31

    A fuel nozzle assembly for use in a combustor apparatus of a gas turbine engine. An outer housing of the fuel nozzle assembly includes an inner volume and provides a direct structural connection between a duct structure and a fuel manifold. The duct structure defines a flow passage for combustion gases flowing within the combustor apparatus. The fuel manifold defines a fuel supply channel therein in fluid communication with a source of fuel. A fuel injector of the fuel nozzle assembly is provided in the inner volume of the outer housing and defines a fuel passage therein. The fuel passage is in fluid communication with the fuel supply channel of the fuel manifold for distributing the fuel from the fuel supply channel into the flow passage of the duct structure.

  1. Nuclear mass inventory, photon dose rate and thermal decay heat of spent research reactor fuel assemblies

    SciTech Connect

    Pond, R.B.; Matos, J.E.

    1996-05-01

    As part of the Department of Energy`s spent nuclear fuel acceptance criteria, the mass of uranium and transuranic elements in spent research reactor fuel must be specified. These data are, however, not always known or readily determined. It is the purpose of this report to provide estimates of these data for some of the more common research reactor fuel assembly types. The specific types considered here are MTR, TRIGA and DIDO fuel assemblies. The degree of physical protection given to spent fuel assemblies is largely dependent upon the photon dose rate of the spent fuel material. These data also, are not always known or readily determined. Because of a self-protecting dose rate level of radiation (dose rate greater than 100 ren-x/h at I m in air), it is important to know the dose rate of spent fuel assemblies at all time. Estimates of the photon dose rate for spent MTR, TRIGA and DIDO-type fuel assemblies are given in this report.

  2. The underwater coincidence counter for plutonium measurements in mixed-oxide fuel assemblies manual

    SciTech Connect

    G. W. Eccleston; H. O. Menlove; M. Abhold; M. Baker; J. Pecos

    1999-05-01

    This manual describes the Underwater Coincidence Counter (UWCC) that has been designed for the measurement of plutonium in mixed-oxide (MOX) fuel assemblies prior to irradiation. The UWCC uses high-efficiency {sup 3}He neutron detectors to measure the spontaneous-fission and induced-fission rates in the fuel assembly. Measurements can be made on MOX fuel assemblies in air or underwater. The neutron counting rate is analyzed for singles, doubles, and triples time correlations to determine the {sup 240}Pu effective mass per unit length of the fuel assembly. The system can verify the plutonium loading per unit length to a precision of less than 1% in a measurement time of 2 to 3 minutes. System design, components, performance tests, and operational characteristics are described in this manual.

  3. Nondestructive verification with minimal movement of irradiated light-water-reactor fuel assemblies

    SciTech Connect

    Phillips, J.R.; Bosler, G.E.; Halbig, J.K.; Klosterbuer, S.F.; Menlove, H.O.

    1982-10-01

    Nondestructive verification of irradiated light-water reactor fuel assemblies can be performed rapidly and precisely by measuring their gross gamma-ray and neutron signatures. A portable system measured fuel assemblies with exposures ranging from 18.4 to 40.6 GWd/tU and with cooling times ranging from 1575 to 2638 days. Differences in the measured results for side or corner measurements are discussed. 25 figures, 20 tables.

  4. Structural analysis of fuel assembly clads for the Upgraded Transient Reactor Test Facility (TREAT Upgrade)

    SciTech Connect

    Ewing, T.F.; Wu, T.S.

    1986-01-01

    The Upgraded Transient Reactor Test Facility (TREAT Upgrade) is designed to test full-length, pre-irradiated fuel pins of the type used in large LMFBRs under accident conditions, such as severe transient overpower and loss-of-coolant accidents. In TREAT Upgrade, the central core region is to contain new fuel assemblies of higher fissile loadings to maximize the energy deposition to the test fuel. These fuel assemblies must withstand normal peak clad temperatures of 850/sup 0/C for hundreds of test transients. Due to high temperatures and gradients predicted in the clad, creep and plastic strain effects are significant, and the clad structural behavior cannot be analyzed by conventional linear techniques. Instead, the detailed elastic-plastic-creep behavior must be followed along the time-dependent load history. This paper presents details of the structural evaluations of the conceptual TREAT Upgrade fuel assembly clads.

  5. Vibration Monitoring Using Fiber Optic Sensors in a Lead-Bismuth Eutectic Cooled Nuclear Fuel Assembly.

    PubMed

    De Pauw, Ben; Lamberti, Alfredo; Ertveldt, Julien; Rezayat, Ali; van Tichelen, Katrien; Vanlanduit, Steve; Berghmans, Francis

    2016-01-01

    Excessive fuel assembly vibrations in nuclear reactor cores should be avoided in order not to compromise the lifetime of the assembly and in order to prevent the occurrence of safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants, such as, for example, a molten lead-bismuth eutectic. The flow of molten heavy metal around and through the fuel assembly may cause the latter to vibrate and hence suffer degradation as a result of, for example, fretting wear or mechanical fatigue. In this paper, we demonstrate the use of optical fiber sensors to measure the fuel assembly vibration in a lead-bismuth eutectic cooled installation which can be used as input to assess vibration-related safety hazards. We show that the vibration characteristics of the fuel pins in the fuel assembly can be experimentally determined with minimal intrusiveness and with high precision owing to the small dimensions and properties of the sensors. In particular, we were able to record local strain level differences of about 0.2 μϵ allowing us to reliably estimate the vibration amplitudes and modal parameters of the fuel assembly based on optical fiber sensor readings during different stages of the operation of the facility, including the onset of the coolant circulation and steady-state operation. PMID:27110782

  6. Space Qualification of the Optical Filter Assemblies for the ICESat-2/ATLAS Instrument

    NASA Technical Reports Server (NTRS)

    Troupaki, E.; Denny, Z. H.; Wu, S.; Bradshaw, H. N.; Smith, K. A.; Hults, J. A.; Ramos-Izquierdo, L. A.; Cook, W. B.

    2015-01-01

    The Advanced Topographic Laser Altimeter System (ATLAS) will be the only instrument on the Ice, Cloud, and Land Elevation Satellite -2 (ICESat-2). ICESat-2 is the 2nd-generation of the orbiting laser altimeter ICESat, which will continue polar ice topography measurements with improved precision laser-ranging techniques. In contrast to the original ICESat design, ICESat-2 will use a micro-pulse, multi-beam approach that provides dense cross-track sampling to help scientists determine a surface's slope with each pass of the satellite. The ATLAS laser will emit visible, green laser pulses at a wavelength of 532 nm and a rate of 10 kHz and will be split into 6 beams. A set of six identical, thermally tuned optical filter assemblies (OFA) will be used to remove background solar radiation from the collected signal while transmitting the laser light to the detectors. A seventh assembly will be used to monitor the laser center wavelength during the mission. In this paper, we present the design and optical performance measurements of the ATLAS OFA in air and in vacuum prior to their integration on the ATLAS instrument.

  7. Space qualification of the optical filter assemblies for the ICESat-2/ATLAS instrument

    NASA Astrophysics Data System (ADS)

    Troupaki, E.; Denny, Z. H.; Wu, S.; Bradshaw, H. N.; Smith, K. A.; Hults, J. A.; Ramos-Izquierdo, L. A.; Cook, W. B.

    2015-02-01

    The Advanced Topographic Laser Altimeter System (ATLAS) will be the only instrument on the Ice, Cloud, and Land Elevation Satellite -2 (ICESat-2). ICESat-2 is the 2nd-generation of the orbiting laser altimeter ICESat, which will continue polar ice topography measurements with improved precision laser-ranging techniques. In contrast to the original ICESat design, ICESat-2 will use a micro-pulse, multi-beam approach that provides dense cross-track sampling to help scientists determine a surface's slope with each pass of the satellite. The ATLAS laser will emit visible, green laser pulses at a wavelength of 532 nm and a rate of 10 kHz and will be split into 6 beams. A set of six identical, thermally tuned optical filter assemblies (OFA) will be used to remove background solar radiation from the collected signal while transmitting the laser light to the detectors. A seventh assembly will be used to monitor the laser center wavelength during the mission. In this paper, we present the design and optical performance measurements of the ATLAS OFA in air and in vacuum prior to their integration on the ATLAS instrument.

  8. Integrated Radiation Transport and Nuclear Fuel Performance for Assembly-Level Simulations

    SciTech Connect

    Hamilton, Steven P; Clarno, Kevin T; Philip, Bobby; Berrill, Mark A; Sampath, Rahul S; Allu, Srikanth

    2012-01-01

    The Advanced Multi-Physics (AMP) Nuclear Fuel Performance code (AMPFuel) is focused on predicting the temperature and strain within a nuclear fuel assembly to evaluate the performance and safety of existing and advanced nuclear fuel bundles within existing and advanced nuclear reactors. AMPFuel was extended to include an integrated nuclear fuel assembly capability for (one-way) coupled radiation transport and nuclear fuel assembly thermo-mechanics. This capability is the initial step toward incorporating an improved predictive nuclear fuel assembly modeling capability to accurately account for source-terms, such as neutron flux distribution, coolant conditions and assembly mechanical stresses, of traditional (single-pin) nuclear fuel performance simulation. A novel scheme is introduced for transferring the power distribution from the Scale/Denovo (Denovo) radiation transport code (structured, Cartesian mesh with smeared materials within each cell) to AMPFuel (unstructured, hexagonal mesh with a single material within each cell), allowing the use of a relatively coarse spatial mesh (10 million elements) for the radiation transport and a fine spatial mesh (3.3 billion elements) for thermo-mechanics with very little loss of accuracy. With this novel capability, AMPFuel was used to model an entire 1717 pressurized water reactor fuel assembly with many of the features resolved in three dimensions (for thermo-mechanics and/or neutronics). A full assembly calculation was executed on Jaguar using 40,000 cores in under 10 hours to model over 160 billion degrees of freedom for 10 loading steps. The single radiation transport calculation required about 50% of the time required to solve the thermo-mechanics with a single loading step, which demonstrates that it is feasible to incorporate, in a single code, a high-fidelity radiation transport capability with a high-fidelity nuclear fuel thermo-mechanics capability and anticipate acceptable computational requirements. The

  9. Method of preparing gas tags for identification of single and multiple failures of nuclear reactor fuel assemblies

    DOEpatents

    McCormick, Norman J.

    1976-01-01

    For use in the identification of failed fuel assemblies in a nuclear reactor, the ratios of the tag gas isotopic concentrations are located on curved surfaces to enable the ratios corresponding to failure of a single fuel assembly to be distinguished from those formed from any combination of two or more failed assemblies.

  10. Determination of Sulfur in Fuel Oils: An Instrumental Analysis Experiment.

    ERIC Educational Resources Information Center

    Graham, Richard C.; And Others

    1982-01-01

    Chromatographic techniques are used in conjunction with a Parr oxygen combustion bomb to determine sulfur in fuel oils. Experimental procedures and results are discussed including an emphasis on safety considerations. (SK)

  11. 3D hydrodynamic lift force model for AREVA fuel assembly in EDF PWRs

    SciTech Connect

    Ekomie, S.; Bigot, J.; Dolleans, Ph.; Vallory, J.

    2007-07-01

    The accurate knowledge of the hydrodynamic lift force acting on a fuel assembly in PWR core is necessary to design the hold-down system of this assembly. This paper presents the model used by AREVA NP and EDF for computing this force. It results from a post-processing of sub-channel thermal-hydraulic codes respectively porous medium approach code THYC (EDF) and sub-channel type code FLICA III-F (AREVA NP). This model is based on the application of the Euler's theorem. Some hypotheses used to simplify the complexity of fuel assembly geometry are supported by CFD calculations. Then the model is compared to some experimental results obtained on a single fuel assembly inserted in the HERMES-T test facility located in CEA - Cadarache. Finally, the model is applied to calculate the lift force for the whole core. Various loading patterns including homogenous and mixed cores have been investigated and compared. (authors)

  12. Fuel Tank Assembly of the Saturn V S-IC Stage

    NASA Technical Reports Server (NTRS)

    1964-01-01

    The fuel tank assembly of the Saturn V S-IC (first) stage is readied to be mated to the liquid oxygen tank at the Marshall Space Flight Center. The fuel tank carried kerosene as its fuel. The S-IC stage utilized five F-1 engines that used kerosene and liquid oxygen as propellant. Each engine provided 1,500,000 pounds of thrust. This stage lifted the entire vehicle and Apollo spacecraft from the launch pad.

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

    SciTech Connect

    V. DeLa Brosse

    2003-03-27

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

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

    SciTech Connect

    T. Schmitt

    2005-08-17

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

  15. High mechanical performance of Areva upgraded fuel assemblies for PWR in USA

    SciTech Connect

    Gottuso, Dennis; Canat, Jean-Noel; Mollard, Pierre

    2007-07-01

    The merger of the product portfolios of the former Siemens and Framatome fuel businesses gave rise to a new family of PWR products which combine the best features of the different technologies to enhance the main performance of each of the existing products. In this way, the technology of each of the three main fuel assembly types usually delivered by AREVA NP, namely Mark-BW{sup TM}, HTP{sup TM} and AFA 3G{sup TM} has been enriched by one or several components from the others which contributes to improve their robustness and to enhance their performance. The combined experience of AREVA's products shows that the ROBUST FUELGUARD{sup TM}, the HMP{sup TM} end grid, the MONOBLOC{sup TM} guide tube, a welded structure, M5{sup R} material for every zirconium component and an upper QUICK-DISCONNECT{sup TM} are key features for boosting fuel assembly robustness. The ROBUST FUELGUARD benefits from a broad experience demonstrating its high efficiency in stopping debris. In addition, its mechanical strength has been enhanced and the proven blade design homogenizes the downstream flow distribution to strongly reduce excitation of fuel rods. The resistance to rod-to-grid fretting resistance of AREVA's new products is completed by the use of a lower HMP grid with 8 lines of contact to insure low wear. The Monobloc guide tube with a diameter maximized to strengthen the fuel assembly stiffness, excludes through its uniform outer geometry any local condition which could weaken guide tube straightness. The application of a welded cage to all fuel assemblies of the new family of products in combination with stiffer guide tubes and optimized hold-down assures each fuel assembly enhanced resistance to distortion. The combination of these features has been widely demonstrated as an effective method to reduce the risk of incomplete RCCA insertion and significantly reduce assembly distortion. Thanks to its enhanced performance, M5 alloy insures that all fuel assemblies in the family

  16. In Comparative Analysis for Fuel Burnup of Fuel Assembly Designs for the 300 kW Small Medical Reactor

    NASA Astrophysics Data System (ADS)

    Sambuu, Odmaa; Nanzad, Norov

    2009-03-01

    A 300 kW small medical reactor was designed to be used for boron neutron capture therapy (BNCT) at KAIST in 1996 [1]. In this paper, analysis for the core life cycle of the original design of the BNCT facility and modifications of the fuel assembly configuration and enrichment to get a proper life cycle were performed and a criticality, neutron flux distribution and fuel burnup calculations were carried out.

  17. In Comparative Analysis for Fuel Burnup of Fuel Assembly Designs for the 300 kW Small Medical Reactor

    SciTech Connect

    Sambuu, Odmaa; Nanzad, Norov

    2009-03-31

    A 300 kW small medical reactor was designed to be used for boron neutron capture therapy (BNCT) at KAIST in 1996. In this paper, analysis for the core life cycle of the original design of the BNCT facility and modifications of the fuel assembly configuration and enrichment to get a proper life cycle were performed and a criticality, neutron flux distribution and fuel burnup calculations were carried out.

  18. Analysis of subcritical experiments using fresh and spent research reactor fuel assemblies

    NASA Astrophysics Data System (ADS)

    Zino, John Frederick

    1999-11-01

    This research investigated the concepts associated with crediting the burnup of spent nuclear fuel assemblies for the purposes of criticality safety. To accomplish this, a collaborative experimental research program was undertaken between Westinghouse, the University of Missouri Research Reactor (MURR) facility and Oak Ridge National Laboratory (ORNL). The purpose of the program was to characterize the subcritical behavior of a small array of fresh and spent MURR fuel assemblies using the 252Cf Source-driven noise technique. An aluminum test rig was built which was capable of holding up to four, highly enriched (93.15 wt.% 235U) MURR fuel assemblies in a 2 x 2 array. The rig was outfitted with one source and four detector drywells which allowed researchers to perform active neutron noise measurements on the array of fuel assemblies. The 1 atmosphere gas 3He neutron detectors used to perform the measurements were quenched with CF4 gas to allow improved discrimination of the neutron signals in the very high gamma-ray fields associated with spent fuel (˜8000 R/hr). In addition, the detector drywells were outfitted with 1″ lead collars to provide additional gamma-ray shielding from the spent fuel. Reactivity changes were induced in the subcritical lattice by replacing individual fresh assemblies (in a 4-assembly array) with spent assemblies of known, maximum burnup (143 Mw-D). The absolute and relative measured reactivity changes were then compared to those predicted by three-dimensional Monte Carlo calculations. The purpose of these comparisons was to investigate the accuracy of modern transport theory depletion calculations to accurately simulate the reactivity effects of burnup in spent nuclear fuel. A total of seven subcritical measurements were performed at the MURR reactor facility on July 20th and 27th, 1998. These measurements generated several estimates of prompt neutron decay constants (alpha) and ratios of spectral densities through frequency correlations

  19. A U. S. Perspective on Fast Reactor Fuel Fabrication Technology and Experience Part I: Metal Fuels and Assembly Design

    SciTech Connect

    Douglas E. Burkes; Randall S. Fielding; Douglas L. Porter; Douglas C. Crawford; Mitchell K. Meyer

    2009-06-01

    This paper is Part I of a review focusing on the United States experience with metallic fast reactor fuel fabrication and assembly design for the Experimental Breeder Reactor-II and the Fast Flux Test Facility, and it also refers to the impact of development in other nations. Experience with metal fuel fabrication in the United States is extensive, including over 60 years of research conducted by the government, national laboratories, industry, and academia. This experience has culminated into a foundation of research and resulted in significant improvements to the technologies employed to fabricate metallic fast reactor fuel. This part of the review documents the current state of fuel fabrication technologies for metallic fuels, some of the challenges faced by previous researchers, and how these were overcome. Knowledge gained from reviewing previous investigations will aid both researchers and policy makers in forming future decisions relating to nuclear fuel fabrication technologies.

  20. A US perspective on fast reactor fuel fabrication technology and experience part I: metal fuels and assembly design

    NASA Astrophysics Data System (ADS)

    Burkes, Douglas E.; Fielding, Randall S.; Porter, Douglas L.; Crawford, Douglas C.; Meyer, Mitchell K.

    2009-06-01

    This paper is part I of a review focusing on the United States experience with metallic fast reactor fuel fabrication and assembly design for the Experimental Breeder Reactor-II (EBR-II) and the Fast Flux Test Facility (FFTF). Experience with metal fuel fabrication in the United States is extensive, including over 60 years of research conducted by the government, national laboratories, industry, and academia. This experience has culminated in a considerable amount of research that resulted in significant improvements to the technologies employed to fabricate metallic fast reactor fuel. This part of the review documents the current state of fuel fabrication technologies for metallic fuels, some of the challenges faced by previous researchers, and how these were overcome. Knowledge gained from reviewing previous investigations will aid both researchers and policy makers in forming future decisions relating to nuclear fuel fabrication technologies.

  1. Applicability of a set of tomographic reconstruction algorithms for quantitative SPECT on irradiated nuclear fuel assemblies

    NASA Astrophysics Data System (ADS)

    Jacobsson Svärd, Staffan; Holcombe, Scott; Grape, Sophie

    2015-05-01

    A fuel assembly operated in a nuclear power plant typically contains 100-300 fuel rods, depending on fuel type, which become strongly radioactive during irradiation in the reactor core. For operational and security reasons, it is of interest to experimentally deduce rod-wise information from the fuel, preferably by means of non-destructive measurements. The tomographic SPECT technique offers such possibilities through its two-step application; (1) recording the gamma-ray flux distribution around the fuel assembly, and (2) reconstructing the assembly's internal source distribution, based on the recorded radiation field. In this paper, algorithms for performing the latter step and extracting quantitative relative rod-by-rod data are accounted for. As compared to application of SPECT in nuclear medicine, nuclear fuel assemblies present a much more heterogeneous distribution of internal attenuation to gamma radiation than the human body, typically with rods containing pellets of heavy uranium dioxide surrounded by cladding of a zirconium alloy placed in water or air. This inhomogeneity severely complicates the tomographic quantification of the rod-wise relative source content, and the deduction of conclusive data requires detailed modelling of the attenuation to be introduced in the reconstructions. However, as shown in this paper, simplified models may still produce valuable information about the fuel. Here, a set of reconstruction algorithms for SPECT on nuclear fuel assemblies are described and discussed in terms of their quantitative performance for two applications; verification of fuel assemblies' completeness in nuclear safeguards, and rod-wise fuel characterization. It is argued that a request not to base the former assessment on any a priori information brings constraints to which reconstruction methods that may be used in that case, whereas the use of a priori information on geometry and material content enables highly accurate quantitative assessment, which

  2. Nonintrusive irradiated fuel inventory confirmation technique

    SciTech Connect

    Dowdy, E.J.; Nicholson, N.; Caldwell, J.T.

    1980-01-01

    Successful tests showing correlation between the intensity of the Cerenkov glow surrounding irradiated fuel assemblies in water-filled spent fuel storage ponds and the exposure and cooling times of assemblies have been concluded. Fieldable instruments used in subsequent tests confirmed that such measurements can be made easily and rapidly, without fuel assembly movement or the introduction of apparatus into the storage ponds.

  3. Partial Defect Verification of the Pressurized Water Reactor Spent Fuel Assemblies

    SciTech Connect

    Ham, Y S; Sitaraman, S

    2010-02-05

    The International Atomic Energy Agency (IAEA) has the responsibility to carry out independent inspections of all nuclear material and facilities subject to safeguards agreements in order to verify compliance with non-proliferation commitments. New technologies have been continuously explored by the IAEA and Member States to improve the verification measures to account for declared inventory of nuclear material and detect clandestine diversion and production of nuclear materials. Even with these efforts, a technical safeguards challenge has remained for decades for the case of developing a method in identifying possible diversion of nuclear fuel pins from the Light Water Reactor (LWR) spent fuel assemblies. We had embarked on this challenging task and successfully developed a novel methodology in detecting partial removal of fuel from pressurized water reactor spent fuel assemblies. The methodology uses multiple tiny neutron and gamma detectors in the form of a cluster and a high precision driving system to obtain underwater radiation measurements inside a Pressurized Water Reactor (PWR) spent fuel assembly without any movement of the fuel. The data obtained in such a manner can provide spatial distribution of neutron and gamma flux within a spent fuel assembly. The combined information of gamma and neutron signature is used to produce base signatures and they are principally dependent on the geometry of the detector locations, and exhibit little sensitivity to initial enrichment, burn-up or cooling time. A small variation in the fuel bundle such as a few missing pins changes the shape of the signature to enable detection. This resulted in a breakthrough method which can be used to detect pin diversion without relying on the nuclear power plant operator's declared operation data. Presented are the results of various Monte Carlo simulation studies and experiments from actual commercial PWR spent fuel assemblies.

  4. Development Tests of a Cryogenic Filter Wheel Assembly for the NIRCam Instrument

    NASA Technical Reports Server (NTRS)

    McCully, Sean; Clark, Charles; Schermerhorn, Michael; Trojanek, Filip; O'Hara, Mark; Williams, Jeff; Thatcher, John

    2006-01-01

    The James Webb Space Telescope is an infrared-optimized space telescope scheduled for launch in 201 3. Its 6.5-m diameter primary mirror will collect light from some of the first galaxies formed after the big bang. The Near Infrared camera (NIRCam) will detect the first light from these galaxies, provide the necessary tools for studying the formation of stars, aid in discovering planets around other stars, and adjust the wave front error on the primary mirror (Fig. 1). The instrument and its complement of mechanisms and optics will operate at a cryogenic temperature of 35 K. This paper describes tests and test results of the NIRCam Filter Wheel assembly prototype.

  5. Method of fabricating an integral gas seal for fuel cell gas distribution assemblies

    DOEpatents

    Dettling, Charles J.; Terry, Peter L.

    1988-03-22

    A porous gas distribution plate assembly for a fuel cell, such as a bipolar assembly, includes an inner impervious region wherein the bipolar assembly has good surface porosity but no through-plane porosity and wherein electrical conductivity through the impervious region is maintained. A hot-pressing process for forming the bipolar assembly includes placing a layer of thermoplastic sealant material between a pair of porous, electrically conductive plates, applying pressure to the assembly at elevated temperature, and allowing the assembly to cool before removing the pressure whereby the layer of sealant material is melted and diffused into the porous plates to form an impervious bond along a common interface between the plates holding the porous plates together. The distribution of sealant within the pores along the surface of the plates provides an effective barrier at their common interface against through-plane transmission of gas.

  6. Integral gas seal for fuel cell gas distribution assemblies and method of fabrication

    DOEpatents

    Dettling, Charles J.; Terry, Peter L.

    1985-03-19

    A porous gas distribution plate assembly for a fuel cell, such as a bipolar assembly, includes an inner impervious region wherein the bipolar assembly has good surface porosity but no through-plane porosity and wherein electrical conductivity through the impervious region is maintained. A hot-pressing process for forming the bipolar assembly includes placing a layer of thermoplastic sealant material between a pair of porous, electrically conductive plates, applying pressure to the assembly at elevated temperature, and allowing the assembly to cool before removing the pressure whereby the layer of sealant material is melted and diffused into the porous plates to form an impervious bond along a common interface between the plates holding the porous plates together. The distribution of sealant within the pores along the surface of the plates provides an effective barrier at their common interface against through-plane transmission of gas.

  7. LLNL MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    SciTech Connect

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R.

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. The DOE Office of Fissile Materials Disposition (DOE-MD) has developed a dual-path strategy for disposition of surplus weapons-grade plutonium. One of the paths is to disposition surplus plutonium through irradiation of MOX fuel in commercial nuclear reactors. MOX fuel consists of plutonium and uranium oxides (PuO{sub 2} and UO{sub 2}), typically containing 95% or more UO{sub 2}. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. LLNL has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. This includes receipt and storage of PuO{sub 2} powder, fabrication of MOX fuel pellets, assembly of fuel rods and bundles, and shipping of the packaged fuel to a commercial reactor site. Support activities will take place within a Category 1 area. Building 332 will be used to receive and store the bulk PuO{sub 2} powder, fabricate MOX fuel pellets, and assemble fuel rods. Building 334 will be used to assemble, store, and ship fuel bundles. Only minor modifications would be required of Building 332. Uncontaminated glove boxes would need to be removed, petition walls would need to be removed, and minor modifications to the ventilation system would be required.

  8. High conversion Th-U{sup 233} fuel assembly for current generation of PWRs

    SciTech Connect

    Baldova, D.; Fridman, E.

    2012-07-01

    This paper presents a preliminary design of a high conversion Th-U{sup 233} fuel assembly applicable for current generation of Pressurized Water Reactor (PWRs). The considered fuel assembly has a typical 17 x 17 PWR lattice. However in order to increase the conversion of Th{sup 232} to U{sup 233}, the assembly was subdivided into the two regions called seed and blanket. The central seed region has a higher than blanket U{sup 233} content and acts as a neutron source for the peripheral blanket region. The latest acts as a U{sup 233} breeder. While the seed fuel pins have a standard dimensions the blanket fuel radius was increased in order to reduce the moderation and to facilitate the resonance neutron absorption in blanket Th{sup 232}. The U{sup 233} content in the seed and blanket regions was optimized to achieve maximal initial to discharged fissile inventory ratio (FIR) taking into account the target fuel cycle length of 12 months with 3-batch reloading scheme. In this study the neutronic calculations were performed on the fuel assembly level using Helios deterministic lattice transport code. The fuel cycle length and the core k{sub eff} were estimated by applying the Non Linear Reactivity Model. The applicability of the HELIOS code for the analysis of the Th-based high conversion designs was confirmed with the help of continuous-energy Monte-Carlo code SERPENT. The results of optimization studies show that for the heterogeneous seed and blanket (SB) fuel assembly the FIR of about 0.95 can be achieved. (authors)

  9. Decay Heat Calculations for PWR and BWR Assemblies Fueled with Uranium and Plutonium Mixed Oxide Fuel using SCALE

    SciTech Connect

    Ade, Brian J; Gauld, Ian C

    2011-10-01

    In currently operating commercial nuclear power plants (NPP), there are two main types of nuclear fuel, low enriched uranium (LEU) fuel, and mixed-oxide uranium-plutonium (MOX) fuel. The LEU fuel is made of pure uranium dioxide (UO{sub 2} or UOX) and has been the fuel of choice in commercial light water reactors (LWRs) for a number of years. Naturally occurring uranium contains a mixture of different uranium isotopes, primarily, {sup 235}U and {sup 238}U. {sup 235}U is a fissile isotope, and will readily undergo a fission reaction upon interaction with a thermal neutron. {sup 235}U has an isotopic concentration of 0.71% in naturally occurring uranium. For most reactors to maintain a fission chain reaction, the natural isotopic concentration of {sup 235}U must be increased (enriched) to a level greater than 0.71%. Modern nuclear reactor fuel assemblies contain a number of fuel pins potentially having different {sup 235}U enrichments varying from {approx}2.0% to {approx}5% enriched in {sup 235}U. Currently in the United States (US), all commercial nuclear power plants use UO{sub 2} fuel. In the rest of the world, UO{sub 2} fuel is still commonly used, but MOX fuel is also used in a number of reactors. MOX fuel contains a mixture of both UO{sub 2} and PuO{sub 2}. Because the plutonium provides the fissile content of the fuel, the uranium used in MOX is either natural or depleted uranium. PuO{sub 2} is added to effectively replace the fissile content of {sup 235}U so that the level of fissile content is sufficiently high to maintain the chain reaction in an LWR. Both reactor-grade and weapons-grade plutonium contains a number of fissile and non-fissile plutonium isotopes, with the fraction of fissile and non-fissile plutonium isotopes being dependent on the source of the plutonium. While only RG plutonium is currently used in MOX, there is the possibility that WG plutonium from dismantled weapons will be used to make MOX for use in US reactors. Reactor-grade plutonium

  10. Silicon carbide composite for light water reactor fuel assembly applications

    NASA Astrophysics Data System (ADS)

    Yueh, Ken; Terrani, Kurt A.

    2014-05-01

    The feasibility of using SiCf-SiCm composites in light water reactor (LWR) fuel designs was evaluated. The evaluation was motivated by the desire to improve fuel performance under normal and accident conditions. The Fukushima accident once again highlighted the need for improved fuel materials that can maintain fuel integrity to higher temperatures for longer periods of time. The review identified many benefits as well as issues in using the material. Issues perceived as presenting the biggest challenges to the concept were identified to be flux gradient induced differential volumetric swelling, fragmentation and thermal shock resistance. The oxidation of silicon and its release into the coolant as silica has been identified as an issue because existing plant systems have limited ability for its removal. Detailed evaluation using available literature data and testing as part of this evaluation effort have eliminated most of the major concerns. The evaluation identified Boiling Water Reactor (BWR) channel, BWR fuel water tube, and Pressurized Water Reactor (PWR) guide tube as feasible applications for SiC composite. A program has been initiated to resolve some of the remaining issues and to generate physical property data to support the design of commercial fuel components.

  11. LANL MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    SciTech Connect

    Fisher, S.E.; Holdaway, R.; Ludwig, S.B.

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. LANL has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. This includes receipt and storage of PuO{sub 2} powder, fabrication of MOX fuel pellets, assembly of fuel rods and bundles, and shipping of the packaged fuel to a commercial reactor site. Support activities will take place within both Category 1 and 2 areas. Technical Area (TA) 55/Plutonium Facility 4 will be used to store the bulk PuO{sub 2} powder, fabricate MOX fuel pellets, assemble rods, and store fuel bundles. Bundles will be assembled at a separate facility, several of which have been identified as suitable for that activity. The Chemistry and Metallurgy Research Building (at TA-3) will be used for analytical chemistry support. Waste operations will be conducted in TA-50 and TA-54. Only very minor modifications will be needed to accommodate the LA program. These modifications consist mostly of minor equipment upgrades. A commercial reactor operator has not been identified for the LA irradiation. Postirradiation examination (PIE) of the irradiated fuel will take place at either Oak Ridge National Laboratory or ANL-W. The only modifications required at either PIE site would be to accommodate full-length irradiated fuel rods. Results from this program are critical to the overall plutonium distribution schedule.

  12. Neutronic optimization in high conversion Th-{sup 233}U fuel assembly with simulated annealing

    SciTech Connect

    Kotlyar, D.; Shwageraus, E.

    2012-07-01

    This paper reports on fuel design optimization of a PWR operating in a self sustainable Th-{sup 233}U fuel cycle. Monte Carlo simulated annealing method was used in order to identify the fuel assembly configuration with the most attractive breeding performance. In previous studies, it was shown that breeding may be achieved by employing heterogeneous Seed-Blanket fuel geometry. The arrangement of seed and blanket pins within the assemblies may be determined by varying the designed parameters based on basic reactor physics phenomena which affect breeding. However, the amount of free parameters may still prove to be prohibitively large in order to systematically explore the design space for optimal solution. Therefore, the Monte Carlo annealing algorithm for neutronic optimization is applied in order to identify the most favorable design. The objective of simulated annealing optimization is to find a set of design parameters, which maximizes some given performance function (such as relative period of net breeding) under specified constraints (such as fuel cycle length). The first objective of the study was to demonstrate that the simulated annealing optimization algorithm will lead to the same fuel pins arrangement as was obtained in the previous studies which used only basic physics phenomena as guidance for optimization. In the second part of this work, the simulated annealing method was used to optimize fuel pins arrangement in much larger fuel assembly, where the basic physics intuition does not yield clearly optimal configuration. The simulated annealing method was found to be very efficient in selecting the optimal design in both cases. In the future, this method will be used for optimization of fuel assembly design with larger number of free parameters in order to determine the most favorable trade-off between the breeding performance and core average power density. (authors)

  13. Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

    SciTech Connect

    Ragusa, Jean; Vierow, Karen

    2011-09-01

    The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzed advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.

  14. Process for recycling components of a PEM fuel cell membrane electrode assembly

    DOEpatents

    Shore, Lawrence

    2012-02-28

    The membrane electrode assembly (MEA) of a PEM fuel cell can be recycled by contacting the MEA with a lower alkyl alcohol solvent which separates the membrane from the anode and cathode layers of the assembly. The resulting solution containing both the polymer membrane and supported noble metal catalysts can be heated under mild conditions to disperse the polymer membrane as particles and the supported noble metal catalysts and polymer membrane particles separated by known filtration means.

  15. A qualitative analysis of the neutron population in fresh and spent fuel assemblies during simulated interrogation using the differential die-away technique

    DOE PAGESBeta

    Tobin, Stephen J.; Lundkvist, Niklas; Goodsell, Alison V.; Grape, Sophie; Hendricks, John S.; Henzl, Vladimir; Swinhoe, Martyn T.

    2015-01-01

    In this study, Monte Carlo simulations were performed for the differential die-away (DDA) technique to analyse the time-dependent behaviour of the neutron population in fresh and spent nuclear fuel assemblies as part of the Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) Project. Simulations were performed to investigate both a possibly portable as well as a permanent DDA instrument. Taking advantage of a custom made modification to the MCNPX code, the variation in the neutron population, simultaneously in time and space, was examined. The motivation for this research was to improve the design of the DDA instrument, as it is bemore » ing considered for possible deployment at the Central Storage of Spent Nuclear Fuel and Encapsulation Plant in Sweden (Clab), as well as to assist in the interpretation of the both simulated and measured signals.« less

  16. A qualitative analysis of the neutron population in fresh and spent fuel assemblies during simulated interrogation using the differential die-away technique

    SciTech Connect

    Tobin, Stephen J.; Lundkvist, Niklas; Goodsell, Alison V.; Grape, Sophie; Hendricks, John S.; Henzl, Vladimir; Swinhoe, Martyn T.

    2015-01-01

    In this study, Monte Carlo simulations were performed for the differential die-away (DDA) technique to analyse the time-dependent behaviour of the neutron population in fresh and spent nuclear fuel assemblies as part of the Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) Project. Simulations were performed to investigate both a possibly portable as well as a permanent DDA instrument. Taking advantage of a custom made modification to the MCNPX code, the variation in the neutron population, simultaneously in time and space, was examined. The motivation for this research was to improve the design of the DDA instrument, as it is be ing considered for possible deployment at the Central Storage of Spent Nuclear Fuel and Encapsulation Plant in Sweden (Clab), as well as to assist in the interpretation of the both simulated and measured signals.

  17. Verifying nuclear fuel assemblies in wet storages on a partial defect level: A software simulation tool for evaluating the capabilities of the Digital Cherenkov Viewing Device

    NASA Astrophysics Data System (ADS)

    Grape, Sophie; Jacobsson Svärd, Staffan; Lindberg, Bo

    2013-01-01

    The Digital Cherenkov Viewing Device (DCVD) is an instrument that records the Cherenkov light emitted from irradiated nuclear fuels in wet storages. The presence, intensity and pattern of the Cherenkov light can be used by the International Atomic Energy Agency (IAEA) inspectors to verify that the fuel properties comply with declarations. The DCVD is since several years approved by the IAEA for gross defect verification, i.e. to control whether an item in a storage pool is a nuclear fuel assembly or a non-fuel item [1]. Recently, it has also been endorsed as a tool for partial defect verification, i.e. to identify if a fraction of the fuel rods in an assembly have been removed or replaced. The latter recognition was based on investigations of experimental studies on authentic fuel assemblies and of simulation studies on hypothetic cases of partial defects [2]. This paper describes the simulation methodology and software which was used in the partial defect capability evaluations. The developed simulation procedure uses three stand-alone software packages: the ORIGEN-ARP code [3] used to obtain the gamma-ray spectrum from the fission products in the fuel, the Monte Carlo toolkit Geant4 [4] for simulating the gamma-ray transport in and around the fuel and the emission of Cherenkov light, and the ray-tracing programme Zemax [5] used to model the light transport through the assembly geometry to the DCVD and to mimic the behaviour of its lens system. Furthermore, the software allows for detailed information from the plant operator on power and/or burnup distributions to be taken into account to enhance the authenticity of the simulated images. To demonstrate the results of the combined software packages, simulated and measured DCVD images are presented. A short discussion on the usefulness of the simulation tool is also included.

  18. Dose rate estimates from irradiated light-water-reactor fuel assemblies in air

    SciTech Connect

    Lloyd, W.R.; Sheaffer, M.K.; Sutcliffe, W.G.

    1994-01-31

    It is generally considered that irradiated spent fuel is so radioactive (self-protecting) that it can only be moved and processed with specialized equipment and facilities. However, a small, possibly subnational, group acting in secret with no concern for the environment (other than the reduction of signatures) and willing to incur substantial but not lethal radiation doses, could obtain plutonium by stealing and processing irradiated spent fuel that has cooled for several years. In this paper, we estimate the dose rate at various distances and directions from typical pressurized-water reactor (PWR) and boiling-water reactor (BWR) spent-fuel assemblies as a function of cooling time. Our results show that the dose rate is reduced rapidly for the first ten years after exposure in the reactor, and that it is reduced by a factor of {approx}10 (from the one year dose rate) after 15 years. Even for fuel that has cooled for 15 years, a lethal dose (LD50) of 450 rem would be received at 1 m from the center of the fuel assembly after several minutes. However, moving from 1 to 5 m reduces the dose rate by over a factor of 10, and moving from 1 to 10 m reduces the dose rate by about a factor of 50. The dose rates 1 m from the top or bottom of the assembly are considerably less (about 10 and 22%, respectively) than 1 m from the center of the assembly, which is the direction of the maximum dose rate.

  19. Morphological features (defects) in fuel cell membrane electrode assemblies

    NASA Astrophysics Data System (ADS)

    Kundu, S.; Fowler, M. W.; Simon, L. C.; Grot, S.

    Reliability and durability issues in fuel cells are becoming more important as the technology and the industry matures. Although research in this area has increased, systematic failure analysis, such as a failure modes and effects analysis (FMEA), are very limited in the literature. This paper presents a categorization scheme of causes, modes, and effects related to fuel cell degradation and failure, with particular focus on the role of component quality, that can be used in FMEAs for polymer electrolyte membrane (PEM) fuel cells. The work also identifies component defects imparted on catalyst-coated membranes (CCM) by manufacturing and proposes mechanisms by which they can influence overall degradation and reliability. Six major defects have been identified on fresh CCM materials, i.e., cracks, orientation, delamination, electrolyte clusters, platinum clusters, and thickness variations.

  20. A new fast neutron collar for safeguards inspection measurements of fresh low enriched uranium fuel assemblies containing burnable poison rods

    NASA Astrophysics Data System (ADS)

    Evans, Louise G.; Swinhoe, Martyn T.; Menlove, Howard O.; Schwalbach, Peter; Baere, Paul De; Browne, Michael C.

    2013-11-01

    Safeguards inspection measurements must be performed in a timely manner in order to detect the diversion of significant quantities of nuclear material. A shorter measurement time can increase the number of items that a nuclear safeguards inspector can reliably measure during a period of access to a nuclear facility. In turn, this improves the reliability of the acquired statistical sample, which is used to inform decisions regarding compliance. Safeguards inspection measurements should also maintain independence from facility operator declarations. Existing neutron collars employ thermal neutron interrogation for safeguards inspection measurements of fresh fuel assemblies. A new fast neutron collar has been developed for safeguards inspection measurements of fresh low-enriched uranium (LEU) fuel assemblies containing gadolinia (Gd2O3) burnable poison rods. The Euratom Fast Collar (EFC) was designed with high neutron detection efficiency to make a fast (Cd) mode measurement viable whilst meeting the high counting precision and short assay time requirements of the Euratom safeguards inspectorate. A fast mode measurement reduces the instrument sensitivity to burnable poison rod content and therefore reduces the applied poison correction, consequently reducing the dependence on the operator declaration of the poison content within an assembly. The EFC non-destructive assay (NDA) of typical modern European pressurized water reactor (PWR) fresh fuel assembly designs have been simulated using Monte Carlo N-particle extended transport code (MCNPX) simulations. Simulations predict that the EFC can achieve 2% relative statistical uncertainty on the doubles neutron counting rate for a fast mode measurement in an assay time of 600 s (10 min) with the available 241AmLi (α,n) interrogation source strength of 5.7×104 s-1. Furthermore, the calibration range of the new collar has been extended to verify 235U content in variable PWR fuel designs in the presence of up to 32

  1. Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

    SciTech Connect

    Chodak, P. III

    1996-05-01

    This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO{sub 2} assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the {sup 239}Pu and {ge}90% {sub total}Pu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products.

  2. TRISO fuel compact thermal conductivity measurement instrument development

    NASA Astrophysics Data System (ADS)

    Jensen, Colby

    Thermal conductivity is an important thermophysical property needed for effectively predicting fuel performance. As part of the Next Generation Nuclear Plant (NGNP) program, the thermal conductivity of tri-isotropic (TRISO) fuel needs to be measured over a temperature range characteristic of its usage. The composite nature of TRISO fuel requires that measurement be performed over the entire length of the compact in a non-destructive manner. No existing measurement system is capable of performing such a measurement. A measurement system has been designed based on the steady-state, guarded-comparative-longitudinal heat flow technique. The system as currently designed is capable of measuring cylindrical samples with diameters ˜12.3-mm (˜0.5″) with lengths ˜25-mm (˜1″). The system is currently operable in a temperature range of 400 K to 1100 K for materials with thermal conductivities on the order of 10 W/m/K to 70 W/m/K. The system has been designed, built, and tested. An uncertainty analysis for the determinate errors of the system has been performed finding a result of 5.5%. Finite element modeling of the system measurement method has also been accomplished demonstrating optimal design, operating conditions, and associated bias error. Measurements have been performed on three calibration/validation materials: SS304, 99.95% pure iron, and inconel 625. In addition, NGNP graphite with ZrO2 particles and NGNP AGR-2 graphite matrix only, both in compact form, have been measured. Results from the SS304 sample show agreement of better than 3% for a 300--600°C temperature range. For iron between 100--600°C, the difference with published values is <8% for all temperatures. The maximum difference from published data for inconel 625 is 5.8%, near 600°C. Both NGNP samples were measured from 100--800°C. All results are presented and discussed. Finally, a discussion of ongoing work is included as well as a brief discussion of implementation under other operating

  3. MELCOR model for an experimental 17x17 spent fuel PWR assembly.

    SciTech Connect

    Cardoni, Jeffrey

    2010-11-01

    A MELCOR model has been developed to simulate a pressurized water reactor (PWR) 17 x 17 assembly in a spent fuel pool rack cell undergoing severe accident conditions. To the extent possible, the MELCOR model reflects the actual geometry, materials, and masses present in the experimental arrangement for the Sandia Fuel Project (SFP). The report presents an overview of the SFP experimental arrangement, the MELCOR model specifications, demonstration calculation results, and the input model listing.

  4. Fuel nozzle assembly for use in turbine engines and methods of assembling same

    DOEpatents

    Uhm, Jong Ho; Johnson, Thomas Edward

    2015-02-03

    A fuel nozzle for use with a turbine engine is described herein. The fuel nozzle includes a housing that is coupled to a combustor liner defining a combustion chamber. The housing includes an endwall that at least partially defines the combustion chamber. A plurality of mixing tubes extends through the housing for channeling fuel to the combustion chamber. Each mixing tube of the plurality of mixing tubes includes an inner surface that extends between an inlet portion and an outlet portion. The outlet portion is oriented adjacent the housing endwall. At least one of the plurality of mixing tubes includes a plurality of projections that extend outwardly from the outlet portion. Adjacent projections are spaced a circumferential distance apart such that a groove is defined between each pair of circumferentially-apart projections to facilitate enhanced mixing of fuel in the combustion chamber.

  5. Monte Carlo characterization of PWR spent fuel assemblies to determine the detectability of pin diversion

    NASA Astrophysics Data System (ADS)

    Burdo, James S.

    This research is based on the concept that the diversion of nuclear fuel pins from Light Water Reactor (LWR) spent fuel assemblies is feasible by a careful comparison of spontaneous fission neutron and gamma levels in the guide tube locations of the fuel assemblies. The goal is to be able to determine whether some of the assembly fuel pins are either missing or have been replaced with dummy or fresh fuel pins. It is known that for typical commercial power spent fuel assemblies, the dominant spontaneous neutron emissions come from Cm-242 and Cm-244. Because of the shorter half-life of Cm-242 (0.45 yr) relative to that of Cm-244 (18.1 yr), Cm-244 is practically the only neutron source contributing to the neutron source term after the spent fuel assemblies are more than two years old. Initially, this research focused upon developing MCNP5 models of PWR fuel assemblies, modeling their depletion using the MONTEBURNS code, and by carrying out a preliminary depletion of a ¼ model 17x17 assembly from the TAKAHAMA-3 PWR. Later, the depletion and more accurate isotopic distribution in the pins at discharge was modeled using the TRITON depletion module of the SCALE computer code. Benchmarking comparisons were performed with the MONTEBURNS and TRITON results. Subsequently, the neutron flux in each of the guide tubes of the TAKAHAMA-3 PWR assembly at two years after discharge as calculated by the MCNP5 computer code was determined for various scenarios. Cases were considered for all spent fuel pins present and for replacement of a single pin at a position near the center of the assembly (10,9) and at the corner (17,1). Some scenarios were duplicated with a gamma flux calculation for high energies associated with Cm-244. For each case, the difference between the flux (neutron or gamma) for all spent fuel pins and with a pin removed or replaced is calculated for each guide tube. Different detection criteria were established. The first was whether the relative error of the

  6. PWR internal flow modeling with fuel assemblies details

    SciTech Connect

    Popov, E.; Yan, J.; Karoutas, Z.; Gehin, J.; Brewster, R.; Baglietto, E.

    2012-07-01

    This study is an example of a massive parallel computing of the coolant flow in a nuclear reactor. It resolves the flow velocities in each assembly on pin level and predicts the flow distribution in complex geometries such as the lower and upper reactor plenums. The size of the developed model (1.035 billion cells) required the runs to be executed on the NCCS clusters (www.nccs.gov). STAR-CCM+ code (www.ed-adapco.com) was installed on two clusters: JAGUARXT5 and FROST, both of which were capable of executing this model. (authors)

  7. Fabrication of capsule assemblies, phase 3

    NASA Technical Reports Server (NTRS)

    Keeton, A. R.; Stemann, L. G.

    1973-01-01

    Thirteen capsule assemblies were fabricated for evaluation of fuel pin design concepts for a fast spectrum lithium cooled compact space power reactor. These instrumented assemblies were designed for real time test of prototype fuel pins. Uranium mononitride fuel pins were encased in AISI 304L stainless steel capsules. Fabrication procedures were fully qualified by process development and assembly qualification tests. Instrumentation reliability was achieved utilizing specially processed and closely controlled thermocouple hot zone fabrication and by thermal screening tests. Overall capsule reliability was achieved with an all electron beam welded assembly.

  8. Fuel Tank Assembly for the Saturn V S-IC Stage

    NASA Technical Reports Server (NTRS)

    1964-01-01

    The fuel tank assembly of the Saturn V S-IC (first) stage supported with the aid of a C frame on the transporter was readied to be transported to the Marshall Space Flight Center, building 4705. The fuel tank carried kerosene (RP-1) as its fuel. The S-IC stage utilized five F-1 engines that used kerosene and liquid oxygen as propellant and each engine provided 1,500,000 pounds of thrust. This stage lifted the entire vehicle and Apollo spacecraft from the launch pad.

  9. Code System to Calculate Cross Sections for PWR Fuel Assembly Calculations.

    Energy Science and Technology Software Center (ESTSC)

    1994-11-15

    Version 00 The MARIA System calculates cross sections for PWR fuel assembly calculations. It generates the cross sections library for the diffusion calculations with burnup and feedback effects (CARMEN System, NEA 0649 and RSIC CCC-487) and the k(infinite) and M**2 parameters for the nodal calculations (SIMULA, NEA 0768). MARIA includes three modules. PRELIM generates the input data for the fuel assembly calculation module, for all fuel assembly types in the core and at any conditionmore » of power rate and temperature. WIMS-TRACA is a modified version of the fuel assembly calculation program WIMS-D/4 (NEA 0329 and RSIC CCC-576), which generates the collapsed cross sections versus burn up needed by the CARMEN code (reference cell, boron, xenon, samarium, and light water). POSWIM calculates the transport corrections to the diffusion constant of the absorber materials generated by WIMS-TRACA, to be used directly in the diffusion code when rods or burnable absorber rods are present.« less

  10. PBF Reactor Building (PER620). Detail of fuel test assembly in ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    PBF Reactor Building (PER-620). Detail of fuel test assembly in preparation for test. When complete, it will fit into in-pile tube. The maximum outside diameter of which must be about 8.25 inches. Date: 1982. INEEL negative no. 82-4908 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

  11. Evaluation of sealed-tube neutron generators for the assay of fresh LWR fuel assemblies

    SciTech Connect

    Cutter, J.; Lee, D.; Lindquist, L.O.; Menlove, H.O.; Caldwell, J.T.; Atencio, J.D.; Kunz, W.E.

    1981-11-01

    The use of sealed-tube neutron generators for the active assay of fresh light-water reactor fuel assemblies has been investigated. The results of experimental tests of the Kaman 801 generator are presented. Neutron yields, source moderation, and transportability are discussed. A comparison is made with the AmLi neutron source for use in the Coincidence Collar.

  12. Use of Burnup Credit as a Safety Factor in Handling of NIST Fuel Assemblies in the L Basin of SRS

    SciTech Connect

    Eghbali, DA

    2004-01-07

    Burnup credit was recently used for the first time in criticality safety analysis to support the handling of the National Institute of Standards and Technology spent fuel assemblies in the L Basin of Savannah River Site. Previous criticality safety analyses were based on the fissile content of fresh, unirradiated fuel assemblies, resulting in handling of a group of 10 or less fuel assemblies at a time. Using burnup credit, it was demonstrated that an isolated configuration of up to 14 NITS fuel assemblies, the maximum number of fuel assemblies in a full basket, submerged in a concrete-lined, water-filled pool is subcritical, resulting in several administrative controls being modified or eliminated without compromising safety.

  13. Fuel cell system including a unit for electrical isolation of a fuel cell stack from a manifold assembly and method therefor

    DOEpatents

    Kelley; Dana A. , Farooque; Mohammad , Davis; Keith

    2007-10-02

    A fuel cell system with improved electrical isolation having a fuel cell stack with a positive potential end and a negative potential, a manifold for use in coupling gases to and from a face of the fuel cell stack, an electrical isolating assembly for electrically isolating the manifold from the stack, and a unit for adjusting an electrical potential of the manifold such as to impede the flow of electrolyte from the stack across the isolating assembly.

  14. Visual inspection system for radioactive fuel assemblies using fiberoptics

    SciTech Connect

    King, W.E.; Healy, L.G.

    1988-08-16

    A system is described for the visual inspection of a radioactive assembly of tubes in an underwater environment, comprising an elongated fiberoptic image guide for remotely transmitting an image of the tubes to be visually inspected, a light source for emitting light, an elongated fiberoptic light guide for transmitting the light emitted from the light source to the tubes. The distal end of the image guide is parallel and adjacent to one side of the light guide so that the distal ends of the image guide and light guide present an elongated cross section that facilitates the insertion of the distal ends in the spaces between the tubes, a first receiving means for remotely displaying the image conducted by the image guide, a second receiving means for remotely displaying the location of the distal ends of the adjacent image and light guides to facilitate the positioning thereof within the assembly, a positioning means for remotely positioning the distal ends of the image and light guide, and means for mechanically linking the second receiving means with the positioning means so that when the image and light guides are moved, the second receiving means is moved the same amount.

  15. New In-pile Instrumentation to Support Fuel Cycle Research and Development

    SciTech Connect

    J. Rempe; H. MacLean; R. Schley; D. Hurley; J. Daw; S. Taylor; J. Smith; J. Svoboda; D. Kotter; D. Knudson; M. Guers; S. C. Wilkins

    2011-01-01

    New and enhanced nuclear fuels are a key enabler for new and improved reactor technologies. For example, the goals of the next generation nuclear plant (NGNP) will not be met without irradiations successfully demonstrating the safety and reliability of new fuels. Likewise, fuel reliability has become paramount in ensuring the competitiveness of nuclear power plants. Recently, the Office of Nuclear Energy in the Department of Energy (DOE-NE) launched a new direction in fuel research and development that emphasizes an approach relying on first principle models to develop optimized fuel designs that offer significant improvements over current fuels. To facilitate this approach, high fidelity, real-time, data are essential for characterizing the performance of new fuels during irradiation testing. A three-year strategic research program is proposed for developing the required test vehicles with sensors of unprecedented accuracy and resolution for obtaining the data needed to characterize three-dimensional changes in fuel microstructure during irradiation testing. When implemented, this strategy will yield test capsule designs that are instrumented with new sensor technologies for the Advanced Test Reactor (ATR) and other irradiation locations for the Fuel Cycle Research and Development (FC R&D) program. Prior laboratory testing, and as needed, irradiation testing, of these sensors will have been completed to give sufficient confidence that the irradiation tests will yield the required data. Obtaining these sensors must draw upon the expertise of a wide-range of organizations not currently supporting nuclear fuels research. This document defines this strategic program and provides the necessary background information related to fuel irradiation testing, desired parameters for detection, and an overview of currently available in-pile instrumentation. In addition, candidate sensor technologies are identified in this document, and a list of proposed criteria for ranking

  16. An integrated approach for the verification of fresh mixed oxide fuel (MOX) assemblies at light water reactor MOX recycle reactors

    SciTech Connect

    Menlove, Howard O; Lee, Sang - Yoon

    2009-01-01

    This paper presents an integrated approach for the verification of mixed oxide (MOX) fuel assemblies prior to their being loaded into the reactor. There is a coupling of the verification approach that starts at the fuel fabrication plant and stops with the transfer of the assemblies into the thermal reactor. The key measurement points are at the output of the fuel fabrication plant, the receipt at the reactor site, and the storage in the water pool as fresh fuel. The IAEA currently has the capability to measure the MOX fuel assemblies at the output of the fuel fabrication plants using a passive neutron coincidence counting systems of the passive neutron collar (PNCL) type. Also. at the MOX reactor pool, the underwater coincidence counter (UWCC) has been developed to measure the MOX assemblies in the water. The UWCC measurement requires that the fuel assembly be lifted about two meters up in the storage rack to avoid interference from the fuel that is stored in the rack. This paper presents a new method to verify the MOX fuel assemblies that are in the storage rack without the necessity of moving the fuel. The detector system is called the Underwater MOX Verification System (UMVS). The integration and relationship of the three measurements systems is described.

  17. Evaluation of FSV-1 cask for the transport of LWR irradiated fuel assemblies

    SciTech Connect

    Not Available

    1980-05-01

    The Model FSV-1 spent fuel shipping cask was designed by General Atomic Company (GA) to service the Fort St. Vrain (FSV) nuclear generating station, a High Temperature Gas Reactor (HTGR) owned and operated by Public Service Company of Colorado (PSC). This report presents an evaluation of the suitability of the FSV-1 cask for the transport of irradiated Light Water Reactor (LWR) fuel assemblies from both Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). The FSV-1 cask evaluation parameters covered a wide spectrum of LWR fuel assemblies, based on burnup in Megawatt Days/Metric Ton of Heavy Metal (MWD/MTHM) and years of decay since irradiation. The criteria for suitability included allowable radiation dose rates, cask surface and interior temperatures and the Gross Vehicle Weight (GVW) of the complete shipping system.

  18. Welding fixture for nuclear fuel pin cladding assemblies

    DOEpatents

    Oakley, David J.; Feld, Sam H.

    1986-01-01

    A welding fixture for locating a driver sleeve about the open end of a nuclear fuel pin cladding. The welding fixture includes a holder provided with an open cavity having shoulders for properly positioning the driver sleeve, the end cap, and a soft, high temperature resistant plastic protective sleeve that surrounds a portion of the end cap stem. Ejected contaminant particles spewed forth by closure of the cladding by pulsed magnetic welding techniques are captured within a contamination trap formed in the holder for ultimate removal and disposal of contaminating particles along with the holder.

  19. Welding fixture for nuclear fuel pin cladding assemblies

    DOEpatents

    Oakley, D.J.; Feld, S.H.

    1984-02-22

    A welding fixture is described for locating a driver sleeve about the open end of a nuclear fuel pin cladding. The welding fixture includes a holder provided with an open cavity having shoulders for properly positioning the driver sleeve, the end cap, and a soft, high temperature resistant plastic protective sleeve that surrounds a portion of the end cap stem. Ejected contaminant particles spewed forth by closure of the cladding by pulsed magnetic welding techniques are captured within a contamination trap formed in the holder for ultimate removal and disposal of contaminating particles along with the holder.

  20. PEM fuel cell cost minimization using ``Design For Manufacture and Assembly`` techniques

    SciTech Connect

    Lomax, F.D. Jr.; James, B.D.; Mooradian, R.P.

    1997-12-31

    Polymer Electrolyte Membrane (PEM) fuel cells fueled with direct hydrogen have demonstrated substantial technical potential to replace Internal Combustion Engines (ICE`s) in light duty vehicles. Such a transition to a hydrogen economy offers the potential of substantial benefits from reduced criteria and greenhouse emissions as well as reduced foreign fuel dependence. Research conducted for the Ford Motor Co. under a US Department of Energy contract suggests that hydrogen fuel, when used in a fuel cell vehicle (FCV), can achieve a cost per vehicle mile less than or equal to the gasoline cost per mile when used in an ICE vehicle. However, fuel cost parity is not sufficient to ensure overall economic success: the PEM fuel cell power system itself must be of comparable cost to the ICE. To ascertain if low cost production of PEM fuel cells is feasible, a powerful set of mechanical engineering tools collectively referred to as Design for Manufacture and Assembly (DFMA) has been applied to several representative PEM fuel cell designs. The preliminary results of this work are encouraging, as presented.

  1. Experience gained from carrying out ultrasonic cleaning of fuel assemblies and control and protection system assemblies in the Novovoronezh NPP unit 3

    NASA Astrophysics Data System (ADS)

    Gorburov, V. I.; Shvarov, V. A.; Vitkovskii, S. L.

    2014-02-01

    A growth of deposits on fuel assembly elements was revealed during operation of the Novovoronezh NPP Unit 3 starting from 1997. This growth caused progressive reduction of coolant flow rate through the reactor core and increase of pressure difference across the assemblies, which eventually led to the need to reduce the power unit output and then to shut down the power unit. In view of these circumstances, it was decided to develop an installation for ultrasonic cleaning of fuel assemblies. The following conclusions were drawn with regard of this installation after completion of all stages of its development, commissioning, and improvement: no detrimental effect of ultrasound on the integrity of fuel assemblies was revealed, whereas the cleaning effect on the fuel assemblies subjected to ultrasonic treatment and improvement of their thermal-hydraulic characteristics are obvious. With these measures implemented, it became possible to clean all fuel assemblies in the core in 2011, to achieve better thermal-hydraulic characteristics, and to avoid reduction of power output and off-scheduled outages of Unit 3.

  2. Predicting fissile content of spent nuclear fuel assemblies with the passive neutron Albedo reactivity technique and Monte Carlo code emulation

    SciTech Connect

    Conlin, Jeremy Lloyd; Tobin, Stephen J

    2010-10-13

    There is a great need in the safeguards community to be able to nondestructively quantify the mass of plutonium of a spent nuclear fuel assembly. As part of the Next Generation of Safeguards Initiative, we are investigating several techniques, or detector systems, which, when integrated, will be capable of quantifying the plutonium mass of a spent fuel assembly without dismantling the assembly. This paper reports on the simulation of one of these techniques, the Passive Neutron Albedo Reactivity with Fission Chambers (PNAR-FC) system. The response of this system over a wide range of spent fuel assemblies with different burnup, initial enrichment, and cooling time characteristics is shown. A Monte Carlo method of using these modeled results to estimate the fissile content of a spent fuel assembly has been developed. A few numerical simulations of using this method are shown. Finally, additional developments still needed and being worked on are discussed.

  3. Th and U fuel photofission study by NTD for AD-MSR subcritical assembly

    NASA Astrophysics Data System (ADS)

    Sajo-Bohus, Laszlo; Greaves, Eduardo D.; Davila, Jesus; Barros, Haydn; Pino, Felix; Barrera, Maria T.; Farina, Fulvio

    2015-07-01

    During the last decade a considerable effort has been devoted for developing energy generating systems based on advanced nuclear technology within the design concepts of GEN-IV. Thorium base fuel systems such as accelerator driven nuclear reactors are one of the often mentioned attractive and affordable options. Several radiotherapy linear accelerators are on the market and due to their reliability, they could be employed as drivers for subcritical liquid fuel assemblies. Bremsstrahlung photons with energies above 5.5MeV, induce (γ,n) and (e,e'n) reactions in the W-target. Resulting gamma radiation and photo or fission neutrons may be absorbed in target materials such as thorium and uranium isotopes to induce sustained fission or nuclear transmutation in waste radioactive materials. Relevant photo driven and photo-fission reaction cross sections are important for actinides 232Th, 238U and 237Np in the radiotherapy machines energy range of 10-20 MV. In this study we employ passive nuclear track detectors (NTD) to determine fission rates and neutron production rates with the aim to establish the feasibility for gamma and photo-neutron driven subcritical assemblies. To cope with these objectives a 20 MV radiotherapy machine has been employed with a mixed fuel target. Results will support further development for a subcritical assembly employing a thorium containing liquid fuel. It is expected that acquired technological knowledge will contribute to the Venezuelan nuclear energy program.

  4. Th and U fuel photofission study by NTD for AD-MSR subcritical assembly

    SciTech Connect

    Sajo-Bohus, Laszlo; Greaves, Eduardo D.; Barros, Haydn; Pino, Felix; Barrera, Maria T.; Farina, Fulvio; Davila, Jesus

    2015-07-23

    During the last decade a considerable effort has been devoted for developing energy generating systems based on advanced nuclear technology within the design concepts of GEN-IV. Thorium base fuel systems such as accelerator driven nuclear reactors are one of the often mentioned attractive and affordable options. Several radiotherapy linear accelerators are on the market and due to their reliability, they could be employed as drivers for subcritical liquid fuel assemblies. Bremsstrahlung photons with energies above 5.5MeV, induce (γ,n) and (e,e’n) reactions in the W-target. Resulting gamma radiation and photo or fission neutrons may be absorbed in target materials such as thorium and uranium isotopes to induce sustained fission or nuclear transmutation in waste radioactive materials. Relevant photo driven and photo-fission reaction cross sections are important for actinides {sup 232}Th, {sup 238}U and {sup 237}Np in the radiotherapy machines energy range of 10-20 MV. In this study we employ passive nuclear track detectors (NTD) to determine fission rates and neutron production rates with the aim to establish the feasibility for gamma and photo-neutron driven subcritical assemblies. To cope with these objectives a 20 MV radiotherapy machine has been employed with a mixed fuel target. Results will support further development for a subcritical assembly employing a thorium containing liquid fuel. It is expected that acquired technological knowledge will contribute to the Venezuelan nuclear energy program.

  5. Method for recovering catalytic elements from fuel cell membrane electrode assemblies

    DOEpatents

    Shore, Lawrence; Matlin, Ramail; Heinz, Robert

    2012-06-26

    A method for recovering catalytic elements from a fuel cell membrane electrode assembly is provided. The method includes converting the membrane electrode assembly into a particulate material, wetting the particulate material, forming a slurry comprising the wetted particulate material and an acid leachate adapted to dissolve at least one of the catalytic elements into a soluble catalytic element salt, separating the slurry into a depleted particulate material and a supernatant containing the catalytic element salt, and washing the depleted particulate material to remove any catalytic element salt retained within pores in the depleted particulate material.

  6. Acceptance of non-fuel assembly hardware by the Federal Waste Management System

    SciTech Connect

    Not Available

    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: failed fuel; consolidated fuel and associated structural parts; non-fuel-assembly hardware; fuel in metal storage casks; fuel in multi-element sealed canisters; inspection and testing requirements for wastes; canister criteria; spent fuel selection for delivery; and defense and commercial high-level waste packages. 14 refs., 12 figs., 43 tabs.

  7. CFD prediction of flow and phase distribution in fuel assemblies with spacers

    SciTech Connect

    Anglart, H.; Nylund, O.; Kurul, N.

    1995-09-01

    This paper is concerned with the modeling and computation of multi-dimensional two-phase flows in BWR fuel assemblies. The modeling principles are presented based on using a two-fluid model in which lateral interfacial effects are accounted for. This model has been used to evaluate the velocity fields of both vapor and liquid phases, as well as phase distribution, between fuel elements in geometries similar to BWR fuel bundles. Furthermore, this model has been used to predict, in a detailed mechanistic manner, the effects of spacers on flow and phase distribution between, and pressure drop along, fuel elements. The related numerical simulations have been performed using a CFD computer code, CFDS-FLOW3D.

  8. Computed isotopic inventory and dose assessment for SRS fuel and target assemblies

    SciTech Connect

    Chandler, M.C.; Ketusky, E.T.; Thoman, D.C.

    1995-06-19

    Past studies have identified and evaluated important radionuclide contributors to dose from reprocessed spent fuel sent to waste for Mark 16B and 22 fuel assemblies and for Mark 31 A and 31B target assemblies. Fission-product distributions after a 5- and 15-year decay time were calculated for a ``representative`` set of irradiation conditions (i.e., reactor power, irradiation time, and exposure) for each type of assembly. The numerical calculations were performed using the SHIELD/GLASS system of codes. The sludge and supernate source terms for dose were studied separately with the significant radionuclide contributors for each identified and evaluated. Dose analysis considered both inhalation and ingestion pathways: The inhalation pathway was analyzed for both evaporative and volatile releases. Analysis of evaporative releases utilized release fractions for the individual radionuclides as defined in the ICRP-30 by DOE guidance. A release fraction of unity was assumed for each radionuclide under volatile-type releases, which would encompass internally initiated events (e.g., fires, explosions), process-initiated events, and externally initiated events. Radionuclides which contributed at least 1% to the overall dose were designated as significant contributors. The present analysis extends and complements the past analyses through considering a broader spectrum of fuel types and a wider range of irradiation conditions. The results provide for a more thorough understanding of the influences of fuel composition and irradiation parameters on fission product distributions (at 2 years or more). Additionally, the present work allows for a more comprehensive evaluation of radionuclide contributions to dose and an estimation of the variability in the radionuclide composition of the dose source term that results from the spent fuel sent to waste encompassing a broad spectrum of fuel compositions and irradiation conditions.

  9. Solution High-Energy Burst Assembly (SHEBA) results from subprompt critical experiments with uranyl fluoride fuel

    SciTech Connect

    Cappiello, C.C.; Butterfield, K.B.; Sanchez, R.G.

    1997-10-01

    The Solution High-Energy Burst Assembly (SHEBA) was originally constructed during 1980 and was designed to be a clean free-field geometry, right-circular, cylindrically symmetric critical assembly employing U(5%)O{sub 2}F{sub 2} solution as fuel. A second version of SHEBA, employing the same fuel but equipped with a fuel pump and shielding pit, was commissioned in 1993. This report includes data and operating experience for the 1993 SHEBA only. Solution-fueled benchmark work focused on the development of experimental measurements of the characterization of SHEBA; a summary of the results are given. A description of the system and the experimental results are given in some detail in the report. Experiments were designed to: (1) study the behavior of nuclear excursions in a low-enrichment solution, (2) evaluate accidental criticality alarm detectors for fuel-processing facilities, (3) provide radiation spectra and dose measurements to benchmark radiation transport calculations on a low-enrichment solution system similar to centrifuge enrichment plants, and (4) provide radiation fields to calibrate personnel dosimetry. 15 refs., 37 figs., 10 tabs.

  10. Container for reprocessing and permanent storage of spent nuclear fuel assemblies

    DOEpatents

    Forsberg, Charles W.

    1992-01-01

    A single canister process container for reprocessing and permanent storage of spent nuclear fuel assemblies comprising zirconium-based cladding and fuel, which process container comprises a collapsible container, having side walls that are made of a high temperature alloy and an array of collapsible support means wherein the container is capable of withstanding temperature necessary to oxidize the zirconium-based cladding and having sufficient ductility to maintain integrity when collapsed under pressure. The support means is also capable of maintaining their integrity at temperature necessary to oxide the zirconium-based cladding. The process container also has means to introduce and remove fluids to and from the container.

  11. CERCA LEU fuel assemblies testing in Maria Reactor - safety analysis summary and testing program scope.

    SciTech Connect

    Pytel, K.; Mieleszczenko, W.; Lechniak, J.; Moldysz, A.; Andrzejewski, K.; Kulikowska, T.; Marcinkowska, A.; Garner, P. L.; Hanan, N. A.; Nuclear Engineering Division; Institute of Atomic Energy

    2010-03-01

    The presented paper contains neutronic and thermal-hydraulic (for steady and unsteady states) calculation results prepared to support annex to Safety Analysis Report for MARIA reactor in order to obtain approval for program of testing low-enriched uranium (LEU) lead test fuel assemblies (LTFA) manufactured by CERCA. This includes presentation of the limits and operational constraints to be in effect during the fuel testing investigations. Also, the scope of testing program (which began in August 2009), including additional measurements and monitoring procedures, is described.

  12. COBRA-SFS predictions of single assembly spent fuel heat transfer data

    SciTech Connect

    Lombardo, N.J.; Michener, T.E.; Wheeler, C.L.; Rector, D.R.

    1986-04-01

    The study reported here is one of several efforts to evaluate and qualify the COBRA-SFS computer code for use in spent fuel storage system thermal analysis. The ability of COBRA-SFS to predict the thermal response of two single assembly spent fuel heat transfer tests was investigated through comparisons of predictions with experimental test data. From these comparisons, conclusions regarding the computational treatment of the physical phenomena occurring within a storage system can be made. This objective was successfully accomplished as reasonable agreement between predictions and data were obtained for the 21 individual test cases of the two experiments.

  13. Nuclear reactor fuel assembly duct-tube-to-inlet-nozzle attachment system

    DOEpatents

    Christiansen, David W.; Smith, Bob G.

    1982-01-01

    A reusable system for removably attaching the lower end 21 of a nuclear reactor fuel assembly duct tube to an upper end 11 of a nuclear reactor fuel assembly inlet nozzle. The duct tube's lower end 21 has sides terminating in locking tabs 22 which end in inwardly-extending flanges 23. The flanges 23 engage recesses 13 in the top section 12 of the inlet nozzle's upper end 11. A retaining collar 30 slides over the inlet nozzle's upper end 11 to restrain the flanges 23 in the recesses 13. A locking nut 40 has an inside threaded portion 41 which engages an outside threaded portion 15 of the inlet nozzle's upper end 11 to secure the retaining collar 30 against protrusions 24 on the duct tube's sides.

  14. Development of techniques for joining fuel rod simulators to test assemblies

    SciTech Connect

    Moorhead, A.J.; Reed, R.W.

    1980-01-01

    A unique tubular electrode carrier is described for gas tungsten-arc welding small-diameter nuclear fuel rod simulators to the tubesheet of a test assembly. Both the close-packed geometry of the array of simulators and the extension of coaxial electrical conductors from each simulator hindered access to the weld joint. Consequently, a conventional gas tungsten-arc torch could not be used. Two seven-rod assemblies that were mockups of the simulator-to-tubesheet joint area were welded and successfully tested. Modified versions of the electrode carrier for brazing electrical leads to the upper ends of the fuel pin simulators are also described. Satisfactory brazes have been made on both single-rod mockups and an array of 25 simulators by using the modified electrode carrier and a filler metal with a composition of 71.5 Ag-28 Cu-0.5 Ni.

  15. Two dimensional, two fluid model for sodium boiling in LMFBR fuel assemblies

    SciTech Connect

    Granziera, M.R.; Kazimi, M.S.

    1980-05-01

    A two dimensional numerical model for the simulation of sodium boiling transient was developed using the two fluid set of conservation equations. A semiimplicit numerical differencing scheme capable of handling the problems associated with the ill-posedness implied by the complex characteristic roots of the two fluid problems was used, which took advantage of the dumping effect of the exchange terms. Of particular interest in the development of the model was the identification of the numerical problems caused by the strong disparity between the axial and radial dimensions of fuel assemblies. A solution to this problem was found which uses the particular geometry of fuel assemblies to accelerate the convergence of the iterative technique used in the model. Three sodium boiling experiments were simulated with the model, with good agreement between the experimental results and the model predictions.

  16. Performance Spec. for Fuel Drying and Canister Inerting System for PWR Core 2 Blanket Fuel Assemblies Stored within Shipping Port Spent Fuel Canisters

    SciTech Connect

    JOHNSON, D.M.

    2000-03-14

    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.

  17. Highly-optimized membrane electrode assembly for direct methanol fuel cell prepared by sedimentation method

    NASA Astrophysics Data System (ADS)

    Liu, Jing Hua; Jeon, Min Ku; Choi, Won Choon; Woo, Seong Ihl

    An electrode for a direct methanol fuel cell (DMFC) is prepared by means of the sedimentation method. A suspension containing Pt black, PTFE and water was filtered through a polycarbonate film and a thin catalyst layer remains on this film. This catalyst layer is then transferred to a gas-diffusion layer by applying a pressure to the assembly and then peeling off the filter film. For the anode catalyst layer, the suspension contained Pt-Ru black and water. The preparation process is optimized and single-cell performance is examined under different operating conditions. Operated at 60 °C, the output power density of the membrane electrode assembly (MEA) fabricated by the sedimentation method is 70% higher than that for an assembly prepared by the conventional brushing technique.

  18. Fuel cell assembly unit for promoting fluid service and electrical conductivity

    DOEpatents

    Jones, Daniel O.

    1999-01-01

    Fluid service and/or electrical conductivity for a fuel cell assembly is promoted. Open-faced flow channel(s) are formed in a flow field plate face, and extend in the flow field plate face between entry and exit fluid manifolds. A resilient gas diffusion layer is located between the flow field plate face and a membrane electrode assembly, fluidly serviced with the open-faced flow channel(s). The resilient gas diffusion layer is restrained against entering the open-faced flow channel(s) under a compressive force applied to the fuel cell assembly. In particular, a first side of a support member abuts the flow field plate face, and a second side of the support member abuts the resilient gas diffusion layer. The support member is formed with a plurality of openings extending between the first and second sides of the support member. In addition, a clamping pressure is maintained for an interface between the resilient gas diffusion layer and a portion of the membrane electrode assembly. Preferably, the support member is spikeless and/or substantially flat. Further, the support member is formed with an electrical path for conducting current between the resilient gas diffusion layer and position(s) on the flow field plate face.

  19. Iterative ct reconstruction from few projections for the nondestructive post irradiation examination of nuclear fuel assemblies

    NASA Astrophysics Data System (ADS)

    Abir, Muhammad Imran Khan

    The core components (e.g. fuel assemblies, spacer grids, control rods) of the nuclear reactors encounter harsh environment due to high temperature, physical stress, and a tremendous level of radiation. The integrity of these elements is crucial for safe operation of the nuclear power plants. The Post Irradiation Examination (PIE) can reveal information about the integrity of the elements during normal operations and off?normal events. Computed tomography (CT) is a tool for evaluating the structural integrity of elements non-destructively. CT requires many projections to be acquired from different view angles after which a mathematical algorithm is adopted for reconstruction. Obtaining many projections is laborious and expensive in nuclear industries. Reconstructions from a small number of projections are explored to achieve faster and cost-efficient PIE. Classical reconstruction algorithms (e.g. filtered back projection) cannot offer stable reconstructions from few projections and create severe streaking artifacts. In this thesis, conventional algorithms are reviewed, and new algorithms are developed for reconstructions of the nuclear fuel assemblies using few projections. CT reconstruction from few projections falls into two categories: the sparse-view CT and the limited-angle CT or tomosynthesis. Iterative reconstruction algorithms are developed for both cases in the field of compressed sensing (CS). The performance of the algorithms is assessed using simulated projections and validated through real projections. The thesis also describes the systematic strategy towards establishing the conditions of reconstructions and finds the optimal imaging parameters for reconstructions of the fuel assemblies from few projections.

  20. A CFD M&S PROCESS FOR FAST REACTOR FUEL ASSEMBLIES

    SciTech Connect

    Kurt D. Hamman; Ray A. Berry

    2008-09-01

    A CFD modeling and simulation process for large-scale problems using an arbitrary fast reactor fuel assembly design was evaluated. Three dimensional flow distributions of sodium for several fast reactor fuel assembly pin spacing configurations were simulated on high performance computers using commercial CFD software. This research focused on 19-pin fuel assembly “benchmark” geometry, similar in design to the Advanced Burner Test Reactor, where each pin is separated by helical wire-wrap spacers. Several two-equation turbulence models including the k-e and SST (Menter) k-? were evaluated. Considerable effort was taken to resolve the momentum boundary layer, so as to eliminate the need for wall functions and reduce computational uncertainty. High performance computers were required to generate the hybrid meshes needed to predict secondary flows created by the wire-wrap spacers; computational meshes ranging from 65 to 85 million elements were common. A general validation methodology was followed, including mesh refinement and comparison of numerical results with empirical correlations. Predictions for velocity, temperature, and pressure distribution are shown. The uncertainty of numerical models, importance of high fidelity experimental data, and the challenges associated with simulating and validating large production-type problems are presented.

  1. Human factors and safety issues associated with actinide retrieval from spent light water reactor fuel assemblies

    SciTech Connect

    Spelt, P.F.

    1992-01-01

    A major problem in environmental restoration and waste management is the disposition of used fuel assemblies from the many light water reactors in the United States, which present a radiation hazard to those whose job is to dispose of them, with a similar threat to the general environment associated with long-term storage in fuel repositories around the country. Actinides resident in the fuel pins as a result of their use in reactor cores constitute a significant component of this hazard. Recently, the Department of Energy has initiated an Actinide Recycle Program to study the feasibility of using pyrochemical (molten salt) processes to recover actinides from the spent fuel assemblies of commercial reactors. This project concerns the application of robotics technology to the operation and maintenance functions of a plant whose objective is to recover actinides from spent fuel assemblies, and to dispose of the resulting hardware and chemical components from this process. Such a procedure involves a number of safety and human factors issues. The purpose of the project is to explore the use of robotics and artificial intelligence to facilitate accomplishment of the program goals while maintaining the safety of the humans doing the work and the integrity of the environment. This project will result in a graphic simulation on a Silicon Graphics workstation as a proof of principle demonstration of the feasibility of using robotics along with an intelligent operator interface. A major component of the operator-system interface is a hybrid artificial intelligence system developed at Oak Ridge National Laboratory, which combines artificial neural networks and an expert system into a hybrid, self-improving computer-based system interface. 10 refs.

  2. Human factors and safety issues associated with actinide retrieval from spent light water reactor fuel assemblies

    SciTech Connect

    Spelt, P.F.

    1992-08-01

    A major problem in environmental restoration and waste management is the disposition of used fuel assemblies from the many light water reactors in the United States, which present a radiation hazard to those whose job is to dispose of them, with a similar threat to the general environment associated with long-term storage in fuel repositories around the country. Actinides resident in the fuel pins as a result of their use in reactor cores constitute a significant component of this hazard. Recently, the Department of Energy has initiated an Actinide Recycle Program to study the feasibility of using pyrochemical (molten salt) processes to recover actinides from the spent fuel assemblies of commercial reactors. This project concerns the application of robotics technology to the operation and maintenance functions of a plant whose objective is to recover actinides from spent fuel assemblies, and to dispose of the resulting hardware and chemical components from this process. Such a procedure involves a number of safety and human factors issues. The purpose of the project is to explore the use of robotics and artificial intelligence to facilitate accomplishment of the program goals while maintaining the safety of the humans doing the work and the integrity of the environment. This project will result in a graphic simulation on a Silicon Graphics workstation as a proof of principle demonstration of the feasibility of using robotics along with an intelligent operator interface. A major component of the operator-system interface is a hybrid artificial intelligence system developed at Oak Ridge National Laboratory, which combines artificial neural networks and an expert system into a hybrid, self-improving computer-based system interface. 10 refs.

  3. The Euratom Fast Collar (EFC): A Safeguards Instrument Design to Address Future Fuel Measurement Challenges

    SciTech Connect

    Evans, Louise; Swinhoe, Martyn T.; Menlove, Howard O.; Browne, Michael C.

    2012-08-13

    Summary of this presentation: (1) EFC instrument design for {sup 235}U verification measurements issued to EURATOM to issue a call for commercial tender; (2) Achieved a fast (Cd mode) measurement with less than 2% relative uncertainty in the doubles neutron counting rate in 10 minutes using a standard source strength; (3) Assay time in fast mode consistent with the needs of an inspector; (4) Extended to realistic calibration range for modern fuel designs - Relatively insensitive to gadolinia content for fuel designs with up to 32 burnable poison rods and 15 wt % gadolinia concentration, which is a realistic maximum for modern PWR fuel; (5) Improved performance over the standard thermal neutron collar with greater than twice the efficiency of the original design; (6) Novel tube pattern to reduce the impact of accidental pile-up; and (7) Joint test of prototype unit - EURATOM-LANL.

  4. Controlling the hydrophilicity and contact resistance of fuel cell bipolar plate surfaces using layered nanoparticle assembly

    NASA Astrophysics Data System (ADS)

    Wang, Feng

    Hybrid nanostructured coatings exhibiting the combined properties of electrical conductivity and surface hydrophilicity were obtained by using Layer-by-Layer (LBL) assembly of cationic polymer, silica nanospheres, and carbon nanoplatelets. This work demonstrates that by controlling the nanoparticle zeta (zeta) potential through the suspension parameters (pH, organic solvent type and amount, and ionic content) as well as the assembly sequence, the nanostructure and composition of the coatings may be adjusted to optimize the desired properties. Two types of silica nanospheres were evaluated as the hydrophilic component: X-TecRTM 3408 from Nano-X Corporation, with a diameter of about 20 nm, and polishing silica from Electron Microscopy Supply, with diameter of about 65 nm. Graphite nanoplatelets with a thickness of 5~10nm (Aquadag RTM E from Acheson Industries) were used as electrically conductive filler. A cationic copolymer of acrylamide and a quaternary ammonium salt (SuperflocRTM C442 from Cytec Corporation) was used as the binder for the negatively charged nanoparticles. Coatings were applied to gold-coated stainless steel substrates presently used a bipolar plate material for proton exchange membrane (PEM) fuel cells. Coating thickness was found to vary nearly linearly with the number of polymer-nanoparticle layers deposited while a monotonic increase in coating contact resistance was observed for all heterogeneous and pure silica coatings. Thickness increased if the difference in the oppositely charged zeta potentials of the adsorbing components was enhanced through alcohol addition. Interestingly, an opposite effect was observed if the zeta potential difference was increased through pH variation. This previously undocumented difference in adsorption behavior is herein related to changes to the surface chemical heterogeneity of the nanoparticles. Coating contact resistance and surface wettability were found to have a more subtle dependence on the assembly

  5. Evaluation of the magnitude and effects of bundle duct interaction in fuel assemblies at developmental plant conditions

    SciTech Connect

    Serell, D.C.; Kaplan, S.

    1980-09-01

    Purpose of this evaluation is to estimate the magnitude and effects of irradiation and creep induced fuel bundle deformations in the developmental plant. This report focuses on the trends of the results and the ability of present models to evaluate the assembly temperatures in the presence of bundle deformation. Although this analysis focuses on the developmental plant, the conclusions are applicable to LMFBR fuel assemblies in general if they have wire spacers.

  6. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    NASA Astrophysics Data System (ADS)

    Ohshima, Hiroyuki; Uwaba, Tomoyuki; Hashimoto, Akihiko; Imai, Yasutomo; Ito, Masahiro

    2015-12-01

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  7. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    SciTech Connect

    Ohshima, Hiroyuki; Uwaba, Tomoyuki; Hashimoto, Akihiko; Imai, Yasutomo; Ito, Masahiro

    2015-12-31

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  8. Estimation of critical flow velocity for collapse of booster fuel assembly

    SciTech Connect

    Donna Guillen; Mark J. Russell

    2005-09-01

    A Gas Test Loop (GTL) system is currently being designed to provide a high intensity fast-flux irradiation environment for testing fuels and materials for advanced concept nuclear reactors. To assess the performance of candidate reactor fuels, these fuels must be irradiated under actual fast reactor flux conditions and operating environments, preferably in an existing irradiation facility. The GTL system is being designed for operation in the northwest test lobe of the Advanced Test Reactor (ATR) at the Idaho National Laboratory. The Technical and Functional Requirements (T&FRs) for the GTL stipulate a minimum neutron flux intensity (10{sup 15} n/cm{sup 2} {center_dot} s) and fast to thermal neutron ratio (>15) for the test environment. The incorporation of booster fuel within the test lobe is necessary to achieve these neutron flux requirements. The current design of the booster fuel assembly for the GTL calls for 3 concentric rings of 4 ft long uranium silicide fuel plates clad with 6061 aluminum.

  9. Vibration Monitoring Using Fiber Optic Sensors in a Lead-Bismuth Eutectic Cooled Nuclear Fuel Assembly

    PubMed Central

    De Pauw, Ben; Lamberti, Alfredo; Ertveldt, Julien; Rezayat, Ali; van Tichelen, Katrien; Vanlanduit, Steve; Berghmans, Francis

    2016-01-01

    Excessive fuel assembly vibrations in nuclear reactor cores should be avoided in order not to compromise the lifetime of the assembly and in order to prevent the occurrence of safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants, such as, for example, a molten lead-bismuth eutectic. The flow of molten heavy metal around and through the fuel assembly may cause the latter to vibrate and hence suffer degradation as a result of, for example, fretting wear or mechanical fatigue. In this paper, we demonstrate the use of optical fiber sensors to measure the fuel assembly vibration in a lead-bismuth eutectic cooled installation which can be used as input to assess vibration-related safety hazards. We show that the vibration characteristics of the fuel pins in the fuel assembly can be experimentally determined with minimal intrusiveness and with high precision owing to the small dimensions and properties of the sensors. In particular, we were able to record local strain level differences of about 0.2 μϵ allowing us to reliably estimate the vibration amplitudes and modal parameters of the fuel assembly based on optical fiber sensor readings during different stages of the operation of the facility, including the onset of the coolant circulation and steady-state operation. PMID:27110782

  10. Forced-to-natural convection transition tests in parallel simulated liquid metal reactor fuel assemblies

    SciTech Connect

    Levin, A.E. ); Montgomery, B.H. )

    1990-01-01

    The Thermal-Hydraulic Out of Reactor Safety (THORS) Program at Oak Ridge National Laboratory (ORNL) had as its objective the testing of simulated, electrically heated liquid metal reactor (LMR) fuel assemblies in an engineering-scale, sodium loop. Between 1971 and 1985, the THORS Program operated 11 simulated fuel bundles in conditions covering a wide range of normal and off-normal conditions. The last test series in the Program, THORS-SHRS Assembly 1, employed two parallel, 19-pin, full-length, simulated fuel assemblies of a design consistent with the large LMR (Large Scale Prototype Breeder -- LSPB) under development at that time. These bundles were installed in the THORS Facility, allowing single- and parallel-bundle testing in thermal-hydraulic conditions up to and including sodium boiling and dryout. As the name SHRS (Shutdown Heat Removal System) implies, a major objective of the program was testing under conditions expected during low-power reactor operation, including low-flow forced convection, natural convection, and forced-to-natural convection transition at various powers. The THORS-SHRS Assembly 1 experimental program was divided up into four phases. Phase 1 included preliminary and shakedown tests, including the collection of baseline steady-state thermal-hydraulic data. Phase 2 comprised natural convection testing. Forced convection testing was conducted in Phase 3. The final phase of testing included forced-to-natural convection transition tests. Phases 1, 2, and 3 have been discussed in previous papers. The fourth phase is described in this paper. 3 refs., 2 figs.