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Note: This page contains sample records for the topic "fuel assemblies cooling" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


1

Cooling assembly for fuel cells  

DOE Patents (OSTI)

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.

Kaufman, Arthur (West Orange, NJ); Werth, John (Princeton, NJ)

1990-01-01T23:59:59.000Z

2

Implementation of vented fuel assemblies in the supercritical CO?-cooled fast reactor  

E-Print Network (OSTI)

Analysis has been undertaken to investigate the utilization of fuel assembly venting in the reference design of the gas-cooled fast reactor under study as part of the larger research effort at MIT under Gen-IV NERI Project ...

McKee, Stephanie A

2008-01-01T23:59:59.000Z

3

Fuel cell sub-assembly  

DOE Patents (OSTI)

A fuel cell sub-assembly comprising a plurality of fuel cells, a first section of a cooling means disposed at an end of the assembly and means for connecting the fuel cells and first section together to form a unitary structure.

Chi, Chang V. (Brookfield, CT)

1983-01-01T23:59:59.000Z

4

Apparatus for in situ determination of burnup, cooling time and fissile content of an irradiated nuclear fuel assembly in a fuel storage pond  

DOE Patents (OSTI)

A detector head for in situ inspection of irradiated nuclear fuel assemblies submerged in a water-filled nuclear fuel storage pond. The detector head includes two parallel arms which extend from a housing and which are spaced apart so as to be positionable on opposite sides of a submerged fuel assembly. Each arm includes an ionization chamber and two fission chambers. One fission chamber in each arm is enclosed in a cadmium shield and the other fission chamber is unshielded. The ratio of the outputs of the shielded and unshielded fission chambers is used to determine the boron content of the pond water. Correcting for the boron content, the neutron flux and gamma ray intensity are then used to verify the declared exposure, cooling time and fissile material content of the irradiated fuel assembly.

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

1984-01-01T23:59:59.000Z

5

Apparatus for in situ determination of burnup, cooling time and fissile content of an irradiated nuclear fuel assembly in a fuel storage pond  

DOE Patents (OSTI)

A detector head for in situ inspection of irradiated nuclear fuel assemblies submerged in a water-filled nuclear fuel storage pond. The detector head includes two parallel arms which extend from a housing and which are spaced apart so as to be positionable on opposite sides of a submerged fuel assembly. Each arm includes an ionization chamber and two fission chambers. One fission chamber in each arm is enclosed in a cadmium shield and the other fission chamber is unshielded. The ratio of the outputs of the shielded and unshielded fission chambers is used to determine the boron content of the pond water. Correcting for the boron content, the neutron flux and gamma ray intensity are then used to verify the declared exposure, cooling time and fissile material content of the irradiated fuel assembly.

Phillips, John R. (Los Alamos, NM); Halbig, James K. (Los Alamos, NM); Menlove, Howard O. (Los Alamos, NM); Klosterbuer, Shirley F. (Los Alamos, NM)

1985-01-01T23:59:59.000Z

6

HEAT-TRANSFER EXPERIMENTS ON A PROPOSED FUEL ASSEMBLY FOR THE EXPERIMENTAL GAS COOLED REACTOR. SECTION II FO FUEL-ASSEMBLY HEAT-TRANSFER AND CHANNEL PRESSURE-DROP EXPERIMENT FOR THE EGCR RESEARCH AND DEVELOPMENT PROGRAM  

SciTech Connect

Heat-transfer data are presented for the Experimental Gas Cooled Reactor Title I seven-rod fuel-assembly design. The effect on heat transfer of (1) the radial location of the outer six rods of the seven-fuel-rod cluster and of (2) the addition of helical-finned spacers at the midpoint of each of the seven fuel rods is discussed. The heattransfer data were obtained to verify preliminary general assumptions pertaining to the heat-transfer characteristics of the seven- rod fuel-assembly design and to obtain local heat-transfer correlations. The heat-transfer tests were performed at near-atmospheric pressure using air as the coolant medium. Plots and equations of heattransfer correlations over a Reynolds Number range from 12,000 to 80,000 are included. The test set-up and test procedure are also described. (auth)

Beaudoin, C.L.; Higgins, R.M.

1960-04-12T23:59:59.000Z

7

Nuclear reactor composite fuel assembly  

DOE Patents (OSTI)

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.

Burgess, Donn M. (Richland, WA); Marr, Duane R. (West Richland, WA); Cappiello, Michael W. (Richland, WA); Omberg, Ronald P. (Richland, WA)

1980-01-01T23:59:59.000Z

8

Fuel assembly transfer basket for pool type nuclear reactor vessels  

DOE Patents (OSTI)

A fuel assembly transfer basket for a pool type, liquid metal cooled nuclear reactor having a side access loading and unloading port for receiving and relinquishing fuel assemblies during transfer.

Fanning, Alan W. (San Jose, CA); Ramsour, Nicholas L. (San Jose, CA)

1991-01-01T23:59:59.000Z

9

BWR Fuel Assembly BWR Fuel Assembly PWR Fuel Assembly  

National Nuclear Security Administration (NNSA)

Spacer Grid Structural Guide Tube End Fitting Fuel Rod Upper Tie Plate ULTRAFLOW Spacer Water Channel Part-length Fuel Rod Lower Tie Plate PWR pressurized water reactor BWR ...

10

Variable area fuel cell cooling  

DOE Patents (OSTI)

A fuel cell arrangement having cooling fluid flow passages which vary in surface area from the inlet to the outlet of the passages. A smaller surface area is provided at the passage inlet, which increases toward the passage outlet, so as to provide more uniform cooling of the entire fuel cell. The cooling passages can also be spaced from one another in an uneven fashion.

Kothmann, Richard E. (Churchill Borough, PA)

1982-01-01T23:59:59.000Z

11

Fuel assembly for nuclear reactors  

DOE Patents (OSTI)

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.

Creagan, Robert J. (Pitcairn, PA); Frisch, Erling (Pittsburgh, PA)

1977-01-01T23:59:59.000Z

12

Spiral cooled fuel nozzle  

Science Conference Proceedings (OSTI)

A fuel nozzle for delivery of fuel to a gas turbine engine. The fuel nozzle includes an outer nozzle wall and a center body located centrally within the nozzle wall. A gap is defined between an inner wall surface of the nozzle wall and an outer body surface of the center body for providing fuel flow in a longitudinal direction from an inlet end to an outlet end of the fuel nozzle. A turbulating feature is defined on at least one of the central body and the inner wall for causing at least a portion of the fuel flow in the gap to flow transverse to the longitudinal direction. The gap is effective to provide a substantially uniform temperature distribution along the nozzle wall in the circumferential direction.

Fox, Timothy; Schilp, Reinhard

2012-09-25T23:59:59.000Z

13

Nuclear core and fuel assemblies  

DOE Patents (OSTI)

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.

Downs, Robert E. (Monroeville, PA)

1981-01-01T23:59:59.000Z

14

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

SciTech Connect

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).

Henzl, Vladimir [Los Alamos National Laboratory; Swinhoe, Martyn T. [Los Alamos National Laboratory; Tobin, Stephen J. [Los Alamos National Laboratory

2012-07-13T23:59:59.000Z

15

Nuclear fuel assembly spacer  

Science Conference Proceedings (OSTI)

In a fuel assembly for a nuclear reactor including a plurality of elongated elements, a spacer is described for retaining the elements in lateral position. The spacer consists of: an array of laterally positioned, cojoined tubular ferrules, each of the ferrules providing a passage for one of the elements, laterally oriented leaf spring members, each of the spring members spanning two adjacent ones of the ferrules and extending therein to engage and laterally support the elements extending through the adjacent ferrules, facing sides of the adjacent ferrules being formed with cutouts to receive and support the spring member. The sides of the ferrules opposite the facing sides are formed with openings to receive and restrain the ends of the spring member, the spring member being formed with a generally V-shaped central portion with an apex extending toward the adjacent sides of the adjacent ferrules whereby in the absence of elements through the adjacent ferrules the central portion contacts the adjacent sides to provide a preload on the spring member and limit the amount of projection of the spring member into the ferrules whereby the insertion of the elements through the ferrules is facilitated. The central portion of the spring member is unrestrained in the presence of the elements through the ferrules, the spring member having left and right arms extending outward from the V-shaped central portion, each of the arms including a relatively long center portion for contacting a respective one of the elements. A shorter end portion is angled toward the ferrules and a tab of reduced height at the end of each arm engaging a respective one of the openings whereby the resulting shoulders at the ends of the spring member engage the inner surface of the ferrules adjacent the openings to laterally locate and retain the spring member.

Johanssen, E.B.; Matzner, B.

1986-02-18T23:59:59.000Z

16

FUEL ASSEMBLY SHAKER TEST SIMULATION  

SciTech Connect

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 direct comparison of model results to recorded test results. This does not offer validation for the fuel assembly model in all conceivable cases, such as high kinetic energy shock cases where the fuel assembly might lift off the basket floor to strike to basket ceiling. This type of nonlinear behavior was not witnessed in testing, so the model does not have test data to be validated against.a basis for validation in cases that substantially alter the fuel assembly response range. This leads to a gap in knowledge that is identified through this modeling study. The SNL shaker testing loaded a surrogate fuel assembly with a certain set of artificially-generated time histories. One thing all the shock cases had in common was an elimination of low frequency components, which reduces the rigid body dynamic response of the system. It is not known if the SNL test cases effectively bound all highway transportation scenarios, or if significantly greater rigid body motion than was tested is credible. This knowledge gap could be filled through modeling the vehicle dynamics of a used fuel conveyance, or by collecting acceleration time history data from an actual conveyance under highway conditions.

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

2013-05-30T23:59:59.000Z

17

NUCLEAR REACTOR FUEL ELEMENT ASSEMBLY  

DOE Patents (OSTI)

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)

Stengel, F.G.

1963-12-24T23:59:59.000Z

18

Improved nuclear fuel assembly grid spacer  

DOE Patents (OSTI)

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.

Marshall, John (San Jose, CA); Kaplan, Samuel (Los Gatos, CA)

1977-01-01T23:59:59.000Z

19

Apparatus for shearing spent nuclear fuel assemblies  

DOE Patents (OSTI)

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.

Weil, Bradley S. (Knoxville, TN); Metz, III, Curtis F. (Knoxville, TN)

1980-01-01T23:59:59.000Z

20

Simulated nuclear reactor fuel assembly  

DOE Patents (OSTI)

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.

Berta, Victor T. (Idaho Falls, ID)

1993-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "fuel assemblies cooling" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


21

Fuel rod assembly to manifold attachment  

DOE Patents (OSTI)

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.

Donck, Harry A. (San Diego, CA); Veca, Anthony R. (San Diego, CA); Snyder, Jr., Harold J. (San Diego, CA)

1980-01-01T23:59:59.000Z

22

Internal reforming fuel cell assembly with simplified fuel feed  

DOE Patents (OSTI)

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.

Farooque, Mohammad (Huntington, CT); Novacco, Lawrence J. (Brookfield, CT); Allen, Jeffrey P. (Naugatuck, CT)

2001-01-01T23:59:59.000Z

23

Method for shearing spent nuclear fuel assemblies  

DOE Patents (OSTI)

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.

Weil, Bradley S. (Oak Ridge, TN); Watson, Clyde D. (Knoxville, TN)

1977-01-01T23:59:59.000Z

24

Fuel cell crimp-resistant cooling device with internal coil  

DOE Patents (OSTI)

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. The conduit has an internal coil means which enables it to be bent in small radii without crimping.

Wittel, deceased, Charles F. (late of Linden, NJ)

1986-01-01T23:59:59.000Z

25

Locking support for nuclear fuel assemblies  

DOE Patents (OSTI)

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.

Ledin, Eric (San Diego, CA)

1980-01-01T23:59:59.000Z

26

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

SciTech Connect

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.

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

1996-12-31T23:59:59.000Z

27

Fail-safe storage rack for irradiated fuel rod assemblies  

DOE Patents (OSTI)

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.

Lewis, D.R.

1993-03-23T23:59:59.000Z

28

Fail-safe storage rack for fuel rod assemblies  

DOE Patents (OSTI)

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.

Lewis, D.R.

1991-12-31T23:59:59.000Z

29

A classification scheme for LWR fuel assemblies  

Science Conference Proceedings (OSTI)

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.

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

1988-11-01T23:59:59.000Z

30

Fuel Assembly Shaker Test for Determining Loads on a PWR Assembly...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Fuel Assembly Shaker Test for Determining Loads on a PWR Assembly under Surrogate Normal Conditions of Truck Transport R0.1 Fuel Assembly Shaker Test for Determining Loads on a PWR...

31

Lead Fuel Assembly Programs Analysis: Utility Perspectives  

Science Conference Proceedings (OSTI)

Licensees, in association with nuclear fuel vendors, conduct lead fuel assembly (LFA) programs to test new design features prior to batch implementation. A limited number of LFAs are irradiated to obtain data and to confirm successful operation in the host reactor environment. The new LFA design features range from minor changes of dimensions and/or materials to an entirely new design from an alternate fuel vendor. LFA program elements can consist of design activities, methods development, analysis, ...

2013-10-17T23:59:59.000Z

32

PNNL Helps the Navy Stay Cool and Conserve Fuel | Department...  

NLE Websites -- All DOE Office Websites (Extended Search)

Helps the Navy Stay Cool and Conserve Fuel PNNL Helps the Navy Stay Cool and Conserve Fuel July 30, 2013 - 3:33pm Addthis As a Laboratory Fellow at the Energy Department's Pacific...

33

Advanced water-cooled phosphoric acid fuel cell development  

DOE Green Energy (OSTI)

The Advanced Water Cooled Phosphoric Acid Fuel Cell Development program is being conducted by International Fuel Cells Corporation (IFC) to improve the performance and minimize the cost of water-cooled, electric utility phosphoric acid fuel cell stacks. The program adapts the existing on-site Configuration B cell design to electric utility operating conditions and introduces additional new design features. Task 1 consists of the conceptual design of a full-scale electric utility cell stack that meets program objectives. Tasks 2 and 3 develop the materials and processes required to fabricate the components that meet the program objective. The design of the small area and two 10-ft[sup 2] short stacks is conducted in Task 4. The conceptual design also is updated to incorporate the results of material and process developments, as well as results of stack tests conducted in Task 6. Fabrication and assembly of the short stacks are conducted in Task 5 and subsequent tests are conducted in Task 6. The Contractor expects to enter into a contract with the Electric Power Research Institute (EPRI) to assemble and endurance test the second 10-ft[sup 2] short stack. The management and reporting functions of Task 7 provide DOE/METC with program visibility through required documentation and program reviews. This report describes the cell design and development effort that is being conducted to demonstrate, by subscale stack test, the technical achievements made toward the above program objectives.

Not Available

1992-07-01T23:59:59.000Z

34

Fuel cell with electrolyte matrix assembly  

DOE Patents (OSTI)

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.

Kaufman, Arthur (West Orange, NJ); Pudick, Sheldon (Sayreville, NJ); Wang, Chiu L. (Edison, NJ)

1988-01-01T23:59:59.000Z

35

Fuel cell assembly with electrolyte transport  

DOE Patents (OSTI)

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.

Chi, Chang V. (Brookfield, CT)

1983-01-01T23:59:59.000Z

36

Bore tube assembly for steam cooling a turbine rotor  

SciTech Connect

An axial bore tube assembly for a turbine is provided to supply cooling steam to hot gas components of the turbine wheels and return the spent cooling steam. A pair of inner and outer tubes define a steam supply passage concentric about an inner return passage. The forward ends of the tubes communicate with an end cap assembly having sets of peripheral holes communicating with first and second sets of radial tubes whereby cooling steam from the concentric passage is supplied through the end cap holes to radial tubes for cooling the buckets and return steam from the buckets is provided through the second set of radial tubes through a second set of openings of the end cap into the coaxial return passage. A radial-to-axial flow transitioning device, including anti-swirling vanes is provided in the end cap. A strut ring adjacent the aft end of the bore tube assembly permits axial and radial thermal expansion of the inner tube relative to the outer tube.

DeStefano, Thomas Daniel (Ballston Lake, NY); Wilson, Ian David (Clifton Park, NY)

2002-01-01T23:59:59.000Z

37

Measurement Protocols for Optimized Fuel Assembly Tags  

Science Conference Proceedings (OSTI)

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.

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

2008-11-01T23:59:59.000Z

38

Nuclear Fuel Assembly and Related Methods for Spent Nuclear ...  

Nuclear Fuel Assembly and Related Methods for Spent Nuclear Fuel Reprocessing and Management Note: The technology described above is an early stage ...

39

An evolutionary fuel assembly design for high power density BWRs  

E-Print Network (OSTI)

An evolutionary BWR fuel assembly design was studied as a means to increase the power density of current and future BWR cores. The new assembly concept is based on replacing four traditional assemblies and large water gap ...

Karahan, Aydin

2007-01-01T23:59:59.000Z

40

Selected Isotopes for Optimized Fuel Assembly Tags  

SciTech Connect

In support of our ongoing signatures project we present information on 3 isotopes selected for possible application in optimized tags that could be applied to fuel assemblies to provide an objective measure of burnup. 1. Important factors for an optimized tag are compatibility with the reactor environment (corrosion resistance), low radioactive activation, at least 2 stable isotopes, moderate neutron absorption cross-section, which gives significant changes in isotope ratios over typical fuel assembly irradiation levels, and ease of measurement in the SIMS machine 2. From the candidate isotopes presented in the 3rd FY 08 Quarterly Report, the most promising appear to be Titanium, Hafnium, and Platinum. The other candidate isotopes (Iron, Tungsten, exhibited inadequate corrosion resistance and/or had neutron capture cross-sections either too high or too low for the burnup range of interest.

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

2008-10-01T23:59:59.000Z

Note: This page contains sample records for the topic "fuel assemblies cooling" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


41

Fuel leak detection apparatus for gas cooled nuclear reactors  

SciTech Connect

Apparatus is disclosed for detecting nuclear fuel leaks within nuclear power system reactors, such as high temperature gas cooled reactors. The apparatus includes a probe assembly that is inserted into the high temperature reactor coolant gaseous stream. The probe has an aperture adapted to communicate gaseous fluid between its inside and outside surfaces and also contains an inner tube for sampling gaseous fluid present near the aperture. A high pressure supply of noncontaminated gas is provided to selectively balance the pressure of the stream being sampled to prevent gas from entering the probe through the aperture. The apparatus includes valves that are operable to cause various directional flows and pressures, which valves are located outside of the reactor walls to permit maintenance work and the like to be performed without shutting down the reactor.

Burnette, Richard D. (San Diego, CA)

1977-01-01T23:59:59.000Z

42

Advanced Water-Cooled Phosphoric Acid Fuel Cell Development  

DOE Green Energy (OSTI)

This program is being conducted to improve the performance and minimize the cost of water cooled, electric utility phosphoric acid fuel cell stacks. The program adapts the existing on-site Configuration B cell design to electric utility operating conditions and introduces additional new design features. Task 1 consists of the conceptual design of a full-scale electric utility cell stack that meets program objectives. Tasks 2 and 3 develop the materials and processes requested to fabricate the components that meet the program objective. The design of the small area and two 10-ft[sup 2] short stacks is conducted in Task 4. The conceptual design also is updated to incorporate the results of material and process developments, as well as results of stack tests conducted in Task 6. Fabrication and assembly of the short stacks are conducted in Task 5 and subsequent tests are conducted in Task 6. The Contractor expects to enter into a contract with the Electric Power Research Institute (EPRI) to assemble and endurance test the second 10-ft[sup 2] short stack. The management and reporting functions of Task 7 provide DOE/METC with program visibility through required documentation and program reviews. This report describes the cell design and development effort that is being conducted to demonstrate, by subscale stack test, the technical achievements made toward the above program objectives.

Not Available

1992-05-01T23:59:59.000Z

43

Preparation of high temperature gas-cooled reactor fuel element  

DOE Patents (OSTI)

This invention relates to a method for the preparation of high temperature gas-cooled reactor (HTGR) fuel elements wherein uncarbonized fuel rods are inserted in appropriate channels of an HTGR fuel element block and the entire block is inserted in an autoclave for in situ carbonization under high pressure. The method is particularly applicable to remote handling techniques.

Bradley, Ronnie A. (Oak Ridge, TN); Sease, John D. (Knoxville, TN)

1976-01-01T23:59:59.000Z

44

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

DOE Patents (OSTI)

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.

Dettling, Charles J. (E. Hanover, NJ); Terry, Peter L. (Chatham Township, Morris County, NJ)

1985-03-19T23:59:59.000Z

45

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

DOE Patents (OSTI)

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.

Dettling, Charles J. (E. Hanover, NJ); Terry, Peter L. (Chathum, NJ)

1988-03-22T23:59:59.000Z

46

Fuel Assembly Shaker Test for Determining Loads on a PWR Assembly under  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Assembly Shaker Test for Determining Loads on a PWR Assembly Assembly Shaker Test for Determining Loads on a PWR Assembly under Surrogate Normal Conditions of Truck Transport R0.1 Fuel Assembly Shaker Test for Determining Loads on a PWR Assembly under Surrogate Normal Conditions of Truck Transport R0.1 The United States current approach of long-term storage at its nuclear power plants and independent spent fuel storage installation, and deferred transportation of used nuclear fuel (UNF), along with the trend of nuclear power plants using reactor fuel for a longer time, creates questions concerning the ability of this aged, high-burnup fuel to withstand stresses and strains seen during normal conditions of transport from its current location to a future consolidated storage facility or permanent repository. UNFD R&D conducted testing employing surrogate instrumented

47

Separator assembly for use in spent nuclear fuel shipping cask  

DOE Patents (OSTI)

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.

Bucholz, James A. (Oak Ridge, TN)

1983-01-01T23:59:59.000Z

48

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

SciTech Connect

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.

Conlin, Jeremy Lloyd [Los Alamos National Laboratory; Tobin, Stephen J [Los Alamos National Laboratory

2010-10-13T23:59:59.000Z

49

Method and apparatus for close packing of nuclear fuel assemblies  

DOE Patents (OSTI)

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.

Newman, D.F.

1993-03-30T23:59:59.000Z

50

Method and apparatus for close packing of nuclear fuel assemblies  

DOE Patents (OSTI)

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.

Newman, Darrell F. (Richland, WA)

1993-01-01T23:59:59.000Z

51

Advanced water-cooled phosphoric acid fuel cell development  

DOE Green Energy (OSTI)

This program was conducted to improve the performance and minimize the cost of existing water-cooled phosphoric acid fuel cell stacks for electric utility and on-site applications. The goals for the electric utility stack technology were a power density of at least 175 watts per square foot over a 40,000-hour useful life and a projected one-of-a-kind, full-scale manufactured cost of less than $400 per kilowatt. The program adapted the existing on-site Configuration-B cell design to electric utility operating conditions and introduced additional new design features. Task 1 consisted of the conceptual design of a full-scale electric utility cell stack that meets program objectives. The conceptual design was updated to incorporate the results of material and process developments in Tasks 2 and 3, as well as results of stack tests conducted in Task 6. Tasks 2 and 3 developed the materials and processes required to fabricate the components that meet the program objectives. The design of the small area and 10-ft{sup 2} stacks was conducted in Task 4. Fabrication and assembly of the short stacks were conducted in Task 5 and subsequent tests were conducted in Task 6. The management and reporting functions of Task 7 provided DOE/METC with program visibility through required documentation and program reviews. This report describes the cell design and development effort that was conducted to demonstrate, by subscale stack test, the technical achievements made toward the above program objectives.

Not Available

1992-09-01T23:59:59.000Z

52

Combustor with two stage primary fuel assembly  

DOE Patents (OSTI)

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.

Sharifi, Mehran (Winter Springs, FL); Zolyomi, Wendel (Oviedo, FL); Whidden, Graydon Lane (Orlando, FL)

2000-01-01T23:59:59.000Z

53

THERMAL ANALYSIS OF A PROPOSED TRANSPORT CASK FOR THREE ADVANCED BURNER REACTOR USED FUEL ASSEMBLIES  

SciTech Connect

Preliminary studies of used fuel generated in the US Department of Energy’s Advanced Fuel Cycle Initiative have indicated that current used fuel transport casks may be insufficient for the transportation of said fuel. This work considers transport of three 5-year-cooled oxide Advanced Burner Reactor used fuel assemblies with a burn-up of 160 MWD/kg. A transport cask designed to carry these assemblies is proposed. This design employs a 7-cm-thick lead gamma shield and a 20-cm-thick NS-4-FR composite neutron shield. The temperature profile within the cask, from its center to its exterior surface, is determined by two dimensional computational fluid dynamics simulations of conduction, convection, and radiation within the cask. Simulations are performed for a cask with a smooth external surface and various neutron shield thicknesses. Separate simulations are performed for a cask with a corrugated external surface and a neutron shield thickness that satisfies shielding constraints. Resulting temperature profiles indicate that a three-assembly cask with a smooth external surface will meet fuel cladding temperature requirements but will cause outer surface temperatures to exceed the regulatory limit. A cask with a corrugated external surface will not exceed the limits for both the fuel cladding and outer surface temperatures.

T. Bullard; M. Greiner; M. Dennis; S. Bays; R. Weiner

2010-09-01T23:59:59.000Z

54

Fuel burner and combustor assembly for a gas turbine engine  

DOE Patents (OSTI)

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.

Leto, Anthony (Franklin Lakes, NJ)

1983-01-01T23:59:59.000Z

55

Portable instrument for inspecting irradiated nuclear fuel assemblies  

DOE Patents (OSTI)

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.

Nicholson, Nicholas (Los Alamos, NM); Dowdy, Edward J. (Los Alamos, NM); Holt, David M. (Los Alamos, NM); Stump, Jr., Charles J. (Santa Fe, NM)

1985-01-01T23:59:59.000Z

56

Noncontacting Laser Scanner for Fuel Assembly Distortion Measurement  

Science Conference Proceedings (OSTI)

The distortion of fuel assemblies and bundles due to in-reactor exposure is a problem that has the potential to significantly affect the operations of both BWRs and PWRs. The distortion can take many shapes of axial, lateral and torsional displacements, with common distortions manifested as bows (such as C, S, and W shapes). Such distortion has implications from fuel ...

2012-09-30T23:59:59.000Z

57

Fuel Rod Cooling in Natural Uranium Reactors  

SciTech Connect

An analysis is presented of the transfer of heat from a cylindrical fuel rod surrounded by a fast flowing coolant in an annular duct, with maximum power output limited by fuel rod temperatures, coolant pressure drop and pumping power requirements. A method is also presented for comparing and evaluating various liquid and gaseous coolants within these limitations. The report also shows and discusses some calculated results obtained for the systems considred in the study of natural U reactors for the production of Pu and useful power (NAA-SR-137).

Trilling, C.A.

1952-01-28T23:59:59.000Z

58

Loss of spent fuel pool cooling PRA: Model and results  

Science Conference Proceedings (OSTI)

This letter report documents models for quantifying the likelihood of loss of spent fuel pool cooling; models for identifying post-boiling scenarios that lead to core damage; qualitative and quantitative results generated for a selected plant that account for plant design and operational practices; a comparison of these results and those generated from earlier studies; and a review of available data on spent fuel pool accidents. The results of this study show that for a representative two-unit boiling water reactor, the annual probability of spent fuel pool boiling is 5 {times} 10{sup {minus}5} and the annual probability of flooding associated with loss of spent fuel pool cooling scenarios is 1 {times} 10{sup {minus}3}. Qualitative arguments are provided to show that the likelihood of core damage due to spent fuel pool boiling accidents is low for most US commercial nuclear power plants. It is also shown that, depending on the design characteristics of a given plant, the likelihood of either: (a) core damage due to spent fuel pool-associated flooding, or (b) spent fuel damage due to pool dryout, may not be negligible.

Siu, N.; Khericha, S.; Conroy, S.; Beck, S.; Blackman, H.

1996-09-01T23:59:59.000Z

59

VENTED FUEL ELEMENT FOR GAS-COOLED NEUTRONIC REACTORS  

DOE Patents (OSTI)

A hollow, porous-walled fuel element filled with fissionable fuel and provided with an outlet port through its wall is described. In operation in a gas-cooled reactor, the element is connected, through its outlet port, to the vacuum side of a pump that causes a portion of the coolant gas flowing over the exterior surface of the element to be drawn through the porous walls thereof and out through the outlet port. This continuous purging gas flow sweeps away gaseous fission products as they are released by the fissioning fuel. (AEC) A fuel element for a nuclear reactor incorporating a body of metal of melting point lower than the temperature of operation of the reactor and a nuclear fuel in finely divided form dispersed in the body of metal as a settled slurry is presented. (AEC)

Furgerson, W.T.

1963-12-17T23:59:59.000Z

60

[Gas cooled fuel cell systems technology development program  

DOE Green Energy (OSTI)

Objective is the development of a gas-cooled phosphoric acid fuel cell for electric utility power plant application. Primary objectives are to: demonstrate performance endurance in 10-cell stacks at 70 psia, 190 C, and 267 mA/cm[sup 2]; improve cell degradation rate to less than 8 mV/1000 hours; develop cost effective criteria, processes, and design configurations for stack components; design multiple stack unit and a single 100 kW fuel cell stack; design a 375 kW fuel cell module and demonstrate average cell beginning-of-use performance; manufacture four 375-kW fuel cell modules and establish characteristics of 1.5 MW pilot power plant. The work is broken into program management, systems engineering, fuel cell development and test, facilities development.

Not Available

1988-03-01T23:59:59.000Z

Note: This page contains sample records for the topic "fuel assemblies cooling" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


61

Detachable connection for a nuclear reactor fuel assembly  

DOE Patents (OSTI)

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.

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

1983-08-29T23:59:59.000Z

62

Hydrogen storage and integrated fuel cell assembly  

DOE Patents (OSTI)

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.

Gross, Karl J. (Fremont, CA)

2010-08-24T23:59:59.000Z

63

DISSOLUTION OF IRRADIATED MURR FUEL ASSEMBLIES  

DOE Green Energy (OSTI)

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.

Kyser, E.

2010-06-17T23:59:59.000Z

64

IN-PILE GAS-COOLED FUEL ELEMENT TEST FACILITY  

SciTech Connect

Paper presented at American Nuclear Society Meeting, June I8-21, 1962, Boston, Mass. Design and operating problems of unclad and ceramic gas-cooled reactor fuels in high temperature circulating gas systems will be studied using a test facility now nearing completion at the Oak Ridge Research Reactor. A shielded air-tight cell houses a closed circuit gas system equipped for dealing with fission products circulating in the gas. Experiments can be conducted on fuel element performance and stability, fission product deposition, gas clean up, activity levels, component and system performance and shielding, and decontamination and maintenance of system hardware. (auth)

Zasler, J.; Huntley, W.R.; Gnadt, P.A.; Kress, T.S.

1962-07-10T23:59:59.000Z

65

Method and apparatus for assembling solid oxide fuel cells  

DOE Patents (OSTI)

This invention relates generally to solid oxide fuel power generators and is particularly directed to improvements in the assembly and coupling of solid oxide fuel cell modules. 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. 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. 17 figs.

Szreders, B.E.; Campanella, N.

1988-05-11T23:59:59.000Z

66

Tools for LWR spent fuel characterization: Assembly classes and fuel designs  

Science Conference Proceedings (OSTI)

The Characteristics Data Base (CDB) is sponsored by the DOE's Office of Civilian Radioactive Waste Management (OCRWM). The CDB provides a single, comprehensive source of data pertaining to radioactive wastes that will or may require geologic disposal, including detailed data describing the physical, quantitative, and radiological characteristics of light-water reactor (LWR) spent fuel. In developing the CDB, tools for the classification of fuel assembly types have been developed. The assembly class scheme is particularly useful for size- and handling-based describes these tools and presents results of their applications in the areas of fuel assembly type identification, characterization of projected discharges, cask accommodation analyses, and defective fuel analyses. Suggestions for additional applications are also made. 7 refs., 1 fig., 2 tabs.

Moore, R.S. (Oak Ridge National Lab., TN (USA) Automated Sciences Group, Inc., Oak Ridge, TN (USA)); Notz, K.J. (Oak Ridge National Lab., TN (USA))

1991-01-01T23:59:59.000Z

67

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

Science Conference Proceedings (OSTI)

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.

Troy Unruh; Michael Reichenberger; Phillip Ugorowski

2013-09-01T23:59:59.000Z

68

Computational fluid dynamics modeling of two-phase flow in a BWR fuel assembly. Final CRADA Report.  

Science Conference Proceedings (OSTI)

A direct numerical simulation capability for two-phase flows with heat transfer in complex geometries can considerably reduce the hardware development cycle, facilitate the optimization and reduce the costs of testing of various industrial facilities, such as nuclear power plants, steam generators, steam condensers, liquid cooling systems, heat exchangers, distillers, and boilers. Specifically, the phenomena occurring in a two-phase coolant flow in a BWR (Boiling Water Reactor) fuel assembly include coolant phase changes and multiple flow regimes which directly influence the coolant interaction with fuel assembly and, ultimately, the reactor performance. Traditionally, the best analysis tools for this purpose of two-phase flow phenomena inside the BWR fuel assembly have been the sub-channel codes. However, the resolution of these codes is too coarse for analyzing the detailed intra-assembly flow patterns, such as flow around a spacer element. Advanced CFD (Computational Fluid Dynamics) codes provide a potential for detailed 3D simulations of coolant flow inside a fuel assembly, including flow around a spacer element using more fundamental physical models of flow regimes and phase interactions than sub-channel codes. Such models can extend the code applicability to a wider range of situations, which is highly important for increasing the efficiency and to prevent accidents.

Tentner, A.; Nuclear Engineering Division

2009-10-13T23:59:59.000Z

69

Method and apparatus for assembling solid oxide fuel cells  

DOE Patents (OSTI)

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.

Szreders, Bernard E. (Oakdale, CT); Campanella, Nicholas (O' Fallon, MO)

1989-01-01T23:59:59.000Z

70

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  

Science Conference Proceedings (OSTI)

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.

Luksic, A.

1989-06-01T23:59:59.000Z

71

Burnup verification measurements on spent fuel assemblies at Arkansas Nuclear One  

SciTech Connect

Burnup verification measurements have been performed using the Fork system at Arkansas Nuclear One, Units 1 and 2, operated by Energy Operations, Inc. Passive neutron and gamma-ray measurements on individual spent fuel assemblies were correlated with the reactor records for burnup, cooling time, and initial enrichment. The correlation generates an internal calibration for the system in the form of a power law determined by a least squares fit to the neutron data. The values of the exponent in the power laws were 3.83 and 4.35 for Units 1 and 2, respectively. The average deviation of the reactor burnup records from the calibration determined from the measurements is a measure of the random error in the burnup records. The observed average deviations were 2.7% and 3.5% for assemblies at Units 1 and 2, respectively, indicating a high degree of consistency in the reactor records. Two non-standard assemblies containing neutron sources were studied at Unit 2. No anomalous measurements were observed among the standard assemblies at either Unit. The effectiveness of the Fork system for verification of reactor records is due to the sensitivity of the neutron yield to burnup, the self-calibration generated by a series of measurements, the redundancy provided by three independent detection systems, and the operational simplicity and flexibility of the design.

Ewing, R.I.

1995-09-01T23:59:59.000Z

72

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

DOE Patents (OSTI)

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.

Shore, Lawrence (Edison, NJ); Matlin, Ramail (Berkeley Heights, NJ)

2009-12-22T23:59:59.000Z

73

METHOD AND MEANS FOR SUPPORTING REACTOR FUEL CONTAINERS IN AN ASSEMBLY  

DOE Patents (OSTI)

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)

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

1962-12-11T23:59:59.000Z

74

Gradient isolator for flow field of fuel cell assembly  

DOE Patents (OSTI)

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.

Ernst, W.D.

1999-06-15T23:59:59.000Z

75

High Energy Absorption Top Nozzle For A Nuclaer Fuel Assembly  

DOE Patents (OSTI)

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.

Sparrow, James A. (Irmo, SC); Aleshin, Yuriy (Columbia, SC); Slyeptsov, Aleksey (Columbia, SC)

2004-05-18T23:59:59.000Z

76

SOURCE OF BURNUP VALUES FOR COMMERCIAL SPENT NUCLEAR FUEL ASSEMBLIES  

SciTech Connect

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, initial {sup 235}U enrichment, and time of discharge from the reactor as well as the assigned burnup, but the distribution. of burnup axially along the assembly length is not provided. The axial burnup profile is maintained within acceptable bounds by the operating conditions of the nuclear reactor and is calculated during preparations to reload a reactor, but the actual burnup profile is not measured. The axial burnup profile is important to the determination of the reactivity of a waste package, so a conservative evaluation of the calculated axial profiles for a large database of SNF has been performed. The product of the axial profile evaluation is a profile that is conservative. Thus, there is no need for physical measurement of the axial profile. The assembly identifier is legible on each SNF assembly and the utility records provide the associated characteristics of the assembly. The conservative methodologies used to determine the criticality loading curve for a waste package provide sufficient margin so that criticality safety is assured for preclosure operations even in the event of a misload. Consideration of misload effects for postclosure time periods is provided by the criticality Features, Events, and Processes (FEPs) analysis. The conservative approaches used to develop and apply the criticality loading curve are thus sufficiently robust that the utility assigned burnup is an adequate source of burnup values, and additional means of verification of assigned burnup through physical measurements are not needed.

BSC

2004-12-01T23:59:59.000Z

77

An Investigation of Different Methods of Fabricating Membrane Electrode Assemblies for Methanol Fuel Cells  

E-Print Network (OSTI)

Methanol fuel cells are electrochemical conversion devices that produce electricity from methanol fuel. The current process of fabricating membrane electrode assemblies (MEAs) is tedious and if it is not sufficiently ...

Hall, Kwame (Kwame J.)

2009-01-01T23:59:59.000Z

78

Advanced water-cooled phosphoric acid fuel cell development. Quarterly technical progress report No. 50, April--June 1992  

DOE Green Energy (OSTI)

The Advanced Water Cooled Phosphoric Acid Fuel Cell Development program is being conducted by International Fuel Cells Corporation (IFC) to improve the performance and minimize the cost of water-cooled, electric utility phosphoric acid fuel cell stacks. The program adapts the existing on-site Configuration B cell design to electric utility operating conditions and introduces additional new design features. Task 1 consists of the conceptual design of a full-scale electric utility cell stack that meets program objectives. Tasks 2 and 3 develop the materials and processes required to fabricate the components that meet the program objective. The design of the small area and two 10-ft{sup 2} short stacks is conducted in Task 4. The conceptual design also is updated to incorporate the results of material and process developments, as well as results of stack tests conducted in Task 6. Fabrication and assembly of the short stacks are conducted in Task 5 and subsequent tests are conducted in Task 6. The Contractor expects to enter into a contract with the Electric Power Research Institute (EPRI) to assemble and endurance test the second 10-ft{sup 2} short stack. The management and reporting functions of Task 7 provide DOE/METC with program visibility through required documentation and program reviews. This report describes the cell design and development effort that is being conducted to demonstrate, by subscale stack test, the technical achievements made toward the above program objectives.

Not Available

1992-07-01T23:59:59.000Z

79

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

DOE Patents (OSTI)

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.

Bradley, John G. (Richland, WA)

1982-01-01T23:59:59.000Z

80

Spent Fuel Pool Cooling and Cleanup During Decommissioning: Experience at Trojan Nuclear Power Plant  

Science Conference Proceedings (OSTI)

Operation of original in-plant spent fuel pool facilities at shutdown power plants is expensive compared to available alternatives and can interfere with the decommissioning process. This report describes the approach taken in the Trojan Decommissioning Project to establish independent cooling and cleanup services for the fuel pool until the spent fuel is placed in dry storage.

1999-03-15T23:59:59.000Z

Note: This page contains sample records for the topic "fuel assemblies cooling" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


81

Separator assembly for use in spent-nuclear-fuel shipping cask. [Patent application  

DOE Patents (OSTI)

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.

Bucholz, J.A.

1981-04-24T23:59:59.000Z

82

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

DOE Green Energy (OSTI)

In this paper, 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.

An, Feng [University of California, Riverside, CA (United States); Stodolsky, F. [Argonne National Lab., IL (United States)

1995-06-01T23:59:59.000Z

83

Methods for disassembling, replacing and assembling parts of a steam cooling system for a gas turbine  

SciTech Connect

The steam cooling circuit for a gas turbine includes a bore tube assembly supplying steam to circumferentially spaced radial tubes coupled to supply elbows for transitioning the radial steam flow in an axial direction along steam supply tubes adjacent the rim of the rotor. The supply tubes supply steam to circumferentially spaced manifold segments located on the aft side of the 1-2 spacer for supplying steam to the buckets of the first and second stages. Spent return steam from these buckets flows to a plurality of circumferentially spaced return manifold segments disposed on the forward face of the 1-2 spacer. Crossover tubes couple the steam supply from the steam supply manifold segments through the 1-2 spacer to the buckets of the first stage. Crossover tubes through the 1-2 spacer also return steam from the buckets of the second stage to the return manifold segments. Axially extending return tubes convey spent cooling steam from the return manifold segments to radial tubes via return elbows. The bore tube assembly, radial tubes, elbows, manifold segments and crossover tubes are removable from the turbine rotor and replaceable.

Wilson, Ian D. (Mauldin, SC); Wesorick, Ronald R. (Albany, NY)

2002-01-01T23:59:59.000Z

84

Development of a High Temperature Gas-Cooled Reactor TRISO-coated particle fuel chemistry model  

E-Print Network (OSTI)

The first portion of this work is a comprehensive analysis of the chemical environment in a High Temperature Gas-Cooled Reactor TRISO fuel particle. Fission product inventory versus burnup is calculated. Based on those ...

Diecker, Jane T

2005-01-01T23:59:59.000Z

85

Fuel assembly for the production of tritium in light water reactors  

DOE Patents (OSTI)

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.

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

1983-06-10T23:59:59.000Z

86

Fuel assembly for the production of tritium in light water reactors  

DOE Patents (OSTI)

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.

Cawley, William E. (Richland, WA); Trapp, Turner J. (Richland, WA)

1985-01-01T23:59:59.000Z

87

EVIDENCE FOR WIDESPREAD COOLING IN AN ACTIVE REGION OBSERVED WITH THE SDO ATMOSPHERIC IMAGING ASSEMBLY  

SciTech Connect

A well-known behavior of EUV light curves of discrete coronal loops is that the peak intensities of cooler channels or spectral lines are reached at progressively later times than hotter channels. This time lag is understood to be the result of hot coronal loop plasma cooling through these lower respective temperatures. However, loops typically comprise only a minority of the total emission in active regions (ARs). Is this cooling pattern a common property of AR coronal plasma, or does it only occur in unique circumstances, locations, and times? The new Solar Dynamics Observatory/Atmospheric Imaging Assembly (SDO/AIA) data provide a wonderful opportunity to answer this question systematically for an entire AR. We measure the time lag between pairs of SDO/AIA EUV channels using 24 hr of images of AR 11082 observed on 2010 June 19. We find that there is a time-lag signal consistent with cooling plasma, just as is usually found for loops, throughout the AR including the diffuse emission between loops for the entire 24 hr duration. The pattern persists consistently for all channel pairs and choice of window length within the 24 hr time period, giving us confidence that the plasma is cooling from temperatures of greater than 3 MK, and sometimes exceeding 7 MK, down to temperatures lower than {approx}0.8 MK. This suggests that the bulk of the emitting coronal plasma in this AR is not steady; rather, it is dynamic and constantly evolving. These measurements provide crucial constraints on any model which seeks to describe coronal heating.

Viall, Nicholeen M.; Klimchuk, James A. [NASA Goddard Space Flight Center, Solar Physics Laboratory, Greenbelt, MD 20706 (United States)

2012-07-01T23:59:59.000Z

88

Operating experience feedback report: Assessment of spent fuel cooling. Volume 12  

SciTech Connect

This report documents the results of an independent assessment by a team from the Office of Analysis and Evaluation of Operational Data of spent-fuel-pool (SFP) cooling in operating nuclear power plants. The team assessed the likelihood and consequences of an extended loss of SFP cooling and suggested corrective actions, based on their findings.

Ibarra, J.G.; Jones, W.R.; Lanik, G.F.; Ornstein, H.L.; Pullani, S.V.

1997-02-01T23:59:59.000Z

89

PNNL Helps the Navy Stay Cool and Conserve Fuel | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

PNNL Helps the Navy Stay Cool and Conserve Fuel PNNL Helps the Navy Stay Cool and Conserve Fuel PNNL Helps the Navy Stay Cool and Conserve Fuel July 30, 2013 - 3:33pm Addthis As a Laboratory Fellow at the Energy Department's Pacific Northwest National Laboratory, Pete McGrail and his team are working to develop a more efficient adsorption chiller that could help the Navy cut its fuel costs. | Photo courtesy of Pacific Northwest National Laboratory. As a Laboratory Fellow at the Energy Department's Pacific Northwest National Laboratory, Pete McGrail and his team are working to develop a more efficient adsorption chiller that could help the Navy cut its fuel costs. | Photo courtesy of Pacific Northwest National Laboratory. Andrew Gumbiner Contractor, Advanced Research Projects Agency-Energy. What are the key facts?

90

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

SciTech Connect

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)

Martinez-Frances, N.; Timm, W.; Rossbach, D. [AREVA, AREVA NP, Erlangen (Germany)

2012-07-01T23:59:59.000Z

91

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

SciTech Connect

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 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 results of the full assembly simulation clearly show the axial, radial, and azimuthal variation of the neutron flux, power, temperature, and deformation of the assembly, highlighting behavior that is neglected in traditional axisymmetric fuel performance codes that do not account for assembly features, such as guide tubes and control rods.

Clarno, Kevin T [ORNL; Hamilton, Steven P [ORNL; Philip, Bobby [ORNL; Berrill, Mark A [ORNL; Sampath, Rahul S [ORNL; Allu, Srikanth [ORNL; Pugmire, Dave [ORNL; Dilts, Gary [Los Alamos National Laboratory (LANL); Banfield, James E [ORNL

2012-02-01T23:59:59.000Z

92

Development of a Safeguards Verification Method and Instrument to Detect Pin Diversion from Pressurized Water Reactor (PWR) Spent Fuel Assemblies Phase I Study  

SciTech Connect

A novel methodology to detect diversion of spent fuel from Pressurized Water Reactors (PWR) has been developed in order to address a long unsolved safeguards verification problem for international safeguards community such as International Atomic Energy Agency (IAEA) or European Atomic Energy Community (EURATOM). The concept involves inserting tiny neutron and gamma detectors into the guide tubes of a spent fuel assembly and measuring the signals. The guide tubes form a quadrant symmetric pattern in the various PWR fuel product lines and the neutron and gamma signals from these various locations are processed to obtain a unique signature for an undisturbed fuel assembly. Signatures based on the neutron and gamma signals individually or in a combination can be developed. Removal of fuel pins from the assembly will cause the signatures to be visibly perturbed thus enabling the detection of diversion. All of the required signal processing to obtain signatures can be performed on standard laptop computers. Monte Carlo simulation studies and a set of controlled experiments with actual commercial PWR spent fuel assemblies were performed and validated this novel methodology. Based on the simulation studies and benchmarking measurements, the methodology developed promises to be a powerful and practical way to detect partial defects that constitute 10% or more of the total active fuel pins. This far exceeds the detection threshold of 50% missing pins from a spent fuel assembly, a threshold defined by the IAEA Safeguards Criteria. The methodology does not rely on any operator provided data like burnup or cooling time and does not require movement of the fuel assembly from the storage rack in the spent fuel pool. A concept was developed to build a practical field device, Partial Defect Detector (PDET), which will be completely portable and will use standard radiation measuring devices already in use at the IAEA. The use of the device will not require any information provided by the operator or any prior knowledge of the spent fuel assembly. The device can also be operated without any movement of the spent fuel from its storage position. Based on parametric studies conducted via computer simulation, the device should be able to detect diversion of as low as ten percent of the missing or replaced fuel from an assembly regardless of the location of the missing fuel within the assembly, of the cooling time, initial fuel enrichment or burnup levels. Conditions in the spent fuel pool such as clarity of the water or boron content are also not issues for this device. The shape of the base signature is principally dependent on the layout of the guide tubes in the various types of PWR fuel assemblies and perturbations in the form of replaced fuel pins will distort this signature. These features of PDET are all unique and overcome limitation and disadvantages presented by currently used devices such as the Fork detector or the Cerenkov Viewing Device. Thus, this device when developed and tested could fill an important need in the safeguards area for partial defect detection, a technology that the IAEA has been seeking for the past few decades.

Ham, Y S; Sitaraman, S

2008-12-24T23:59:59.000Z

93

Impact of Fission Products Impurity on the Plutonium Content of Metal- and Oxide- Fuels in Sodium Cooled Fast Reactors  

Science Conference Proceedings (OSTI)

This short report presents the neutronic analysis to evaluate the impact of fission product impurity on the Pu content of Sodium-cooled Fast Reactor (SFR) metal- and oxide- fuel fabrication. The similar work has been previously done for PWR MOX fuel [1]. The analysis will be performed based on the assumption that the separation of the fission products (FP) during the reprocessing of UOX spent nuclear fuel assemblies is not perfect and that, consequently, a certain amount of FP goes into the Pu stream used to fabricate SFR fuels. Only non-gaseous FPs have been considered (see the list of 176 isotopes considered in the calculations in Appendix 1 of Reference 1). Throughout of this report, we define the mixture of Pu and FPs as PuFP. The main objective of this analysis is to quantify the increase of the Pu content of SFR fuels necessary to maintain the same average burnup at discharge independently of the amount of FP in the Pu stream, i.e. independently of the PuFP composition. The FP losses are considered element-independent, i.e., for example, 1% of FP losses mean that 1% of all non-gaseous FP leak into the Pu stream.

Hikaru Hiruta; Gilles Youinou

2013-09-01T23:59:59.000Z

94

Steep-Slope Assembly Testing of Clay and Concrete Tile With and Without Cool Pigmented Colors  

Science Conference Proceedings (OSTI)

Cool color pigments and sub-tile venting of clay and concrete tile roofs significantly impact the heat flow crossing the roof deck of a steep-slope roof. Field measures for the tile roofs revealed a 70% drop in the peak heat flow crossing the deck as compared to a direct-nailed asphalt shingle roof. The Tile Roofing Institute (TRI) and its affiliate members are keenly interested in documenting the magnitude of the drop for obtaining solar reflectance credits with state and federal "cool roof" building efficiency standards. Tile roofs are direct-nailed or are attached to a deck with batten or batten and counter-batten construction. S-Misson clay and concrete tile roofs, a medium-profile concrete tile roof, and a flat slate tile roof were installed on fully nstrumented attic test assemblies. Temperature measures of the roof, deck, attic, and ceiling, heat flows, solar reflectance, thermal emittance, and the ambient weather were recorded for each of the tile roofs and also on an adjacent attic cavity covered with a conventional pigmented and directnailed asphalt shingle roof. ORNL measured the tile's underside temperature and the bulk air temperature and heat flows just underneath the tile for batten and counter-batten tile systems and compared the results to the conventional asphalt shingle.

Miller, William A [ORNL

2005-11-01T23:59:59.000Z

95

Advanced water-cooled phosphoric acid fuel cell development. Final report  

DOE Green Energy (OSTI)

This program was conducted to improve the performance and minimize the cost of existing water-cooled phosphoric acid fuel cell stacks for electric utility and on-site applications. The goals for the electric utility stack technology were a power density of at least 175 watts per square foot over a 40,000-hour useful life and a projected one-of-a-kind, full-scale manufactured cost of less than $400 per kilowatt. The program adapted the existing on-site Configuration-B cell design to electric utility operating conditions and introduced additional new design features. Task 1 consisted of the conceptual design of a full-scale electric utility cell stack that meets program objectives. The conceptual design was updated to incorporate the results of material and process developments in Tasks 2 and 3, as well as results of stack tests conducted in Task 6. Tasks 2 and 3 developed the materials and processes required to fabricate the components that meet the program objectives. The design of the small area and 10-ft{sup 2} stacks was conducted in Task 4. Fabrication and assembly of the short stacks were conducted in Task 5 and subsequent tests were conducted in Task 6. The management and reporting functions of Task 7 provided DOE/METC with program visibility through required documentation and program reviews. This report describes the cell design and development effort that was conducted to demonstrate, by subscale stack test, the technical achievements made toward the above program objectives.

Not Available

1992-09-01T23:59:59.000Z

96

Thermal-Hydraulic Analysis of Advanced Mixed-Oxide Fuel Assemblies with VIPRE-01  

E-Print Network (OSTI)

Two new fuel assembly designs for light water reactors using advanced mixed-oxide fuels have been proposed to reduce the radiotoxicity of used nuclear fuel discharged from nuclear power plants. The research efforts of this thesis are the first to consider the effects of burnup on advanced mixed-oxide fuel assembly performance and thermal safety margin over an assembly?s expected operational burnup lifetime. In order to accomplish this, a new burnup-dependent thermal-hydraulic analysis methodology has been developed. The new methodology models many of the effects of burnup on an assembly design by including burnup-dependent variations in fuel pin relative power from neutronic calculations, assembly power reductions due to fissile content depletion and core reshuffling, and fuel material thermal-physical properties. Additionally, a text-based coupling method is developed to facilitate the exchange of information between the neutronic code DRAGON and thermal-hydraulic code VIPRE-01. The new methodology effectively covers the entire assembly burnup lifetime and evaluates the thermal-hydraulic performance against ANS Condition I, II, and III events with respect to the minimum departure from nucleate boiling ratio, peak cladding temperatures, and fuel centerline temperatures. A comprehensive literature survey on the thermal conductivity of posed fuel materials with burnup-dependence has been carried out to model the advanced materials in the thermal-hydraulic code VIPRE-01. Where documented conductivity values are not available, a simplified method for estimating the thermal conductivity has been developed. The new thermal conductivity models are based on established FRAPCON-3 fuel property models used in the nuclear industry, with small adjustments having been made to account for actinide additions. Steady-state and transient thermal-hydraulic analyses are performed with VIPRE- 01 for a reference UO2 assembly design, and two advanced mixed-oxide fuel assembly designs using the new burnup-dependent thermal-hydraulic analysis methodology. All three designs maintain a sufficiently large thermal margin with respect to the minimum departure from nucleate boiling ratio, and maximum cladding and fuel temperatures during partial and complete loss-of-flow accident scenarios. The presence of a thin (Am,Zr)O2 outer layer on the fuel pellet in the two advanced mixed-oxide fuel assembly designs increases maximum fuel temperatures during transient conditions, but does not otherwise greatly compromise the thermal margin of the new designs.

Bingham, Adam R.

2009-05-01T23:59:59.000Z

97

Deposition of Alternative (Syngas) Fuels on Turbine Blades with Film Cooling  

NLE Websites -- All DOE Office Websites (Extended Search)

ACERC ACERC Dr. Jeffrey Bons and Dr. Thomas Fletcher BRIGHAM YOUNG UNIVERSITY SCIES Project 05-01-SR-120 with support from General Electric, Siemens-Westinghouse, Solar Turbines, Praxair UTSR Peer Workshop III, Clemson University, SC Oct. 18-20, 2005 Deposition of Alternative ( Deposition of Alternative ( Syngas Syngas ) Fuels on ) Fuels on Turbine Blades with Film Cooling Turbine Blades with Film Cooling Alternate fuels (e.g. coal, petcoke, and biomass) are being cons Alternate fuels (e.g. coal, petcoke, and biomass) are being cons idered to idered to produce produce syngas syngas fuels to replace natural gas in power turbines fuels to replace natural gas in power turbines Despite gas cleanup, small levels of airborne particulate (e.g. Despite gas cleanup, small levels of airborne particulate (e.g. 0.1 0.1 ppmw

98

New Fuel Cycle and Fuel Management Options in Heavy Liquid Metal-Cooled Reactors  

Science Conference Proceedings (OSTI)

Technical Paper / Advances in Nuclear Fuel Management - Fuel Management of Reactors Other Than Light Water Reactors

Ehud Greenspan; Pavel Hejzlar; Hiroshi Sekimoto; Georgy Toshinsky; David Wade

99

Durable Fuel Cell Membrane Electrode Assembly (MEA) - Energy ...  

Technology Marketing Summary The membrane electrode assembly (MEA) is an essential, yet highly expensive component of any polymer electrolyte membrane ...

100

Assemblies with both target and fuel pins in an isotope-production reactor  

DOE Patents (OSTI)

A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium target material is placed in pins adjacent to fuel pins in order to increase the tritium production rate.

Cawley, W.E.; Omberg, R.P.

1982-08-19T23:59:59.000Z

Note: This page contains sample records for the topic "fuel assemblies cooling" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


101

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

SciTech Connect

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.

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

1996-05-01T23:59:59.000Z

102

Numerical Modeling Study and Assessment of PWR Fuel Rod and Assembly Distortion  

Science Conference Proceedings (OSTI)

Fuel assembly and rod distortion experienced in pressurized water reactors (PWRs) result in numerous operational challenges to plant operators such as mechanical interference between distorted assembly and control rods, difficulties in unloading and reloading cores during outages, and possibly anomalous fuel performance due to atypical water gaps. Therefore, an improved understanding of the various parameters contributing to distortion is important in order to manage or otherwise eliminate these ...

2012-11-30T23:59:59.000Z

103

Passive cooling system for top entry liquid metal cooled nuclear reactors  

DOE Patents (OSTI)

A liquid metal cooled nuclear fission reactor plant having a top entry loop joined satellite assembly with a passive auxiliary safety cooling system for removing residual heat resulting from fuel decay during shutdown, or heat produced during a mishap. This satellite type reactor plant is enhanced by a backup or secondary passive safety cooling system which augments the primary passive auxiliary cooling system when in operation, and replaces the primary cooling system when rendered inoperative.

Boardman, Charles E. (Saratoga, CA); Hunsbedt, Anstein (Los Gatos, CA); Hui, Marvin M. (Cupertino, CA)

1992-01-01T23:59:59.000Z

104

A Technical Review of Non-Destructive Assay Research for the Characterization of Spent Nuclear Fuel Assemblies Being Conducted Under the US DOE NGSI - 11544  

E-Print Network (OSTI)

the Characterization of Spent Nuclear Fuel Assemblies BeingSociety’s Advances in Nuclear Fuel Management IV, HiltonPlutonium Mass in Spent Nuclear Fuel,” 2010 ANS Annual

Croft, S.

2012-01-01T23:59:59.000Z

105

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

DOE Patents (OSTI)

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.

McCormick, Norman J. (Seattle, WA)

1976-01-01T23:59:59.000Z

106

Spent Fuel Pool Accident Characteristics  

Science Conference Proceedings (OSTI)

Spent fuel pools (SFPs) at nuclear reactor sites contain used fuel assemblies, control rods, used radioactive sources, and used instrumentation. Cooling of the used fuel is required to remove the decay heat generated by radioactive decay.BackgroundThe SFPs include heat removal systems to provide methods to cool the used fuel and inventory makeup systems as backup methods to preserve water inventory if the SFP cooling system is ineffective. These two methods ...

2013-05-27T23:59:59.000Z

107

Development and Evaluation of a Safeguards System Concept for a Pebble-Fueled High Temperature Gas-cooled Reactor  

E-Print Network (OSTI)

Pebble-fueled high temperature gas-cooled reactor (HTGR) technology was first developed by the Federal Republic of Germany in the 1950s. More recently, the design has been embraced by the People's Republic of China and the Republic of South Africa. Unlike light water reactors that generate heat from fuel assemblies comprised of fuel rods, pebble-fueled HTGRs utilize thousands of 60-mm diameter fuel spheres (pebbles) comprised of thousands of TRISO particles. As this reactor type is deployed across the world, adequate methods for safeguarding the reactor must be developed. Current safeguards methods for the pebble-fueled HTGR focus on extensive, redundant containment and surveillance (C/S) measures or a combination of item-type and bulk-type material safeguards measures to deter and detect the diversion of fuel pebbles. The disadvantages to these approaches are the loss of continuity of knowledge (CoK) when C/S systems fail, or are compromised, and the introduction of material unaccounted for (MUF). Either vulnerability can be exploited by an adversary to divert fuel pebbles from the reactor system. It was determined that a solution to maintaining CoK is to develop a system to identify each fuel pebble that is inserted and removed from the reactor. Work was performed to develop and evaluate the use of inert microspheres placed in each fuel pebble, whose random placement could be used as a fingerprint to identify the fuel pebble. Ultrasound imaging of 1 mm zirconium oxide microspheres was identified as a possible imaging system and microsphere material for the new safeguards system concept. The system concept was evaluated, and it was found that a minimum of three microspheres are necessary to create enough random fingerprints for 10,000,000 pebbles. It was also found that, over the lifetime of the reactor, less than 0.01% of fuel pebbles can be expected to have randomly the same microsphere fingerprint. From an MCNP 5.1 model, it was determined that less than fifty microspheres in each pebble will have no impact on the reactivity or temperature coefficient of reactivity of the reactor system. Finally, using an ultrasound system it was found that ultrasound waves can penetrate thin layers of graphite to image the microsphere fingerprint.

Gitau, Ernest Travis Ngure

2011-08-01T23:59:59.000Z

108

Advanced water-cooled phosphoric acid fuel cell development  

DOE Green Energy (OSTI)

Fabrication of repeat parts for the small area short stack is nearing completion and assembly activities are being initiated. Electrolyte reservoir plates (ERPs) were completed and processed into integral separator plates, and acid fill of parts was initiated. Fabrication of electrodes was also completed, including catalyzation and applications of seals and matrices.

Not Available

1990-01-01T23:59:59.000Z

109

Deep Burn Develpment of Transuranic Fuel for High-Temperature Helium-Cooled Reactors - July 2010  

SciTech Connect

The DB Program Quarterly Progress Report for April - June 2010, ORNL/TM/2010/140, was distributed to program participants on August 4. This report discusses the following: (1) TRU (transuranic elements) HTR (high temperature helium-cooled reactor) Fuel Modeling - (a) Thermochemical Modeling, (b) 5.3 Radiation Damage and Properties; (2) TRU HTR Fuel Qualification - (a) TRU Kernel Development, (b) Coating Development, (c) ZrC Properties and Handbook; and (3) HTR Fuel Recycle - (a) Recycle Processes, (b) Graphite Recycle, (c) Pyrochemical Reprocessing - METROX (metal recovery from oxide fuel) Process Development.

Snead, Lance Lewis [ORNL; Besmann, Theodore M [ORNL; Collins, Emory D [ORNL; Bell, Gary L [ORNL

2010-08-01T23:59:59.000Z

110

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

SciTech Connect

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 results of the full assembly simulation clearly show the axial, radial, and azimuthal variation of the neutron flux, power, temperature, and deformation of the assembly, highlighting behavior that is neglected in traditional axisymmetric fuel performance codes that do not account for assembly features, such as guide tubes and control rods.

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

2012-01-01T23:59:59.000Z

111

Method of cooling gas only nozzle fuel tip  

DOE Patents (OSTI)

A diffusion flame nozzle gas tip is provided to convert a dual fuel nozzle to a gas only nozzle. The nozle tip diverts compressor discharge air from the passage feeding the diffusion nozzle air swirl vanes to a region vacated by removal of the dual fuel components, so that the diverted compressor discharge air can flow to and through effusion holes in the end cap plate of the nozzle tip. In a preferred embodiment, the nozzle gas tip defines a cavity for receiving the compressor discharge air from a peripheral passage of the nozzle for flow through the effusion openings defined in the end cap plate.

Bechtel, William Theodore (Scotia, NY); Fitts, David Orus (Ballston Spa, NY); DeLeonardo, Guy Wayne (Glenville, NY)

2002-01-01T23:59:59.000Z

112

A feasibility study of internal evaporative cooling for proton exchange membrane fuel cells  

E-Print Network (OSTI)

An investigation was conducted to determine the feasibility of using the technique of ultrasonic nebulization of water into the anode gas stream for evaporative cooling of a Proton Exchange Membrane (PEM) fuel cell. The basic concept of this form of internal evaporative cooling of the PEM fuel cell is to introduce finely atomized liquid water into the anode gas stream, so that the finely atomized liquid water adsorbs onto the anode and then moves to the cathode via electro-osmotic drag, where this water then evaporates into the relatively dry cathode gas stream, carrying with it the waste thermal energy generated within the fuel cell. The thermal and electrical performance of a 50 cm2 PEM fuel cell utilizing this technique was compared to the performance obtained with conventional water management. Both techniques were compared over a range of humidification chamber temperatures for both the anode and cathode gas streams so as to determine the robustness of the proposed method. The proposed method produced only meager levels of evaporative cooling (at best 2 watts, for which a minimum of 30 watts was required for adequate cooling), but the average cell voltage increased considerably (as much as a 10% gain), and the technique increased the fault tolerance of the fuel cell (the Nafion� membrane did not dry out even if cell temperature went well in excess of 70° C despite both anode and cathode humidification temperatures of 55° C). An interesting phenomena was also observed wherein the fuel cell voltage oscillated regularly with a period of tens of seconds, and that the amplitude of this oscillation corresponded inversely with the level of humidification received by the fuel cell.

Snyder, Loren E

2004-12-01T23:59:59.000Z

113

On the impact of variability and assembly on turbine blade cooling flow and oxidation life  

E-Print Network (OSTI)

The life of a turbine blade is dependent on the quantity and temperature of the cooling flow sup- plied to the blade. The focus of this thesis is the impact of variability on blade cooling flow and, subsequently, its impact ...

Sidwell, Carroll Vincent, 1972-

2004-01-01T23:59:59.000Z

114

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

SciTech Connect

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.

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

1998-08-01T23:59:59.000Z

115

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

SciTech Connect

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)

Baldova, D.; Fridman, E. [Reactor Safety Div., Helmholtz-Zentrum Dresden-Rossendorf, POB 510119, Dresden, 01314 (Germany)

2012-07-01T23:59:59.000Z

116

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

Science Conference Proceedings (OSTI)

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.

Shore, Lawrence (Edison, NJ)

2012-02-28T23:59:59.000Z

117

Thermal-Hydraulic Analysis of Seed-Blanket Unit Duplex Fuel Assemblies with VIPRE-01  

E-Print Network (OSTI)

One of the greatest challenges facing the nuclear power industry is the final disposition of nuclear waste. To meet the needs of the nuclear power industry, a new fuel assembly design, called DUPLEX, has been developed which provides higher fuel burnups, burns transuranic waste while reducing minor actinides, reduces the long term radiotoxicity of spent nuclear fuel, and was developed for use in current light water reactors. The DUPLEX design considered in this thesis is based on a seed and blanket unit (SBU) configuration, where the seed region contains standard UO2 fuel, and the blanket region contains an inert matrix (Pu,Np,Am)O2-MgO-ZrO2 fuel. The research efforts of this thesis are first to consider the higher burnup effects on DUPLEX assembly thermal-hydraulic performance and thermal safety margin over the assembly’s expected operational lifetime. In order to accomplish this, an existing burnup-dependent thermal-hydraulic methodology for conventional homogeneous fuel assemblies has been updated to meet the modeling needs specific to SBU-type assemblies. The developed framework dramatically expands the capabilities of the latest thermal-hydraulic evaluation framework such that the most promising and unique DUPLEX fuel design can be evaluated. As part of this updated methodology, the posed DUPLEX design is evaluated with respect to the minimum departure from nucleate boiling ratio, peak fuel temperatures for both regions, and the peak cladding temperatures, under ANS Condition I, II, and III transient events with the thermal-hydraulic code VIPRE-01. Due to difficulty in the fabrication and handling of minor actinide dioxides, documented thermal conductivity values for the considered IMF design are unavailable. In order to develop a representative thermal conductivity model for use in VIPRE-01, an extensive literature survey on the thermal conductivity of (Pu,Np,Am)O2-MgO-ZrO2 component materials and a comprehensive review of combinatory models was performed. Using the updated methodology, VIPRE-01 is used to perform steady-state and transient thermal hydraulic analyses for the DUPLEX fuel assembly. During loss-of-flow accident scenarios, the DUPLEX design is shown to meet imposed safety criteria. However, using the most conservative thermal conductivity modeling approach for (Pu,Np,Am)O2-MgO-ZrO2, the blanket region fuel temperatures remain only slightly below the design limit.

McDermott, Patrick 1987-

2012-12-01T23:59:59.000Z

118

Methods for manufacturing porous nuclear fuel elements for high-temperature gas-cooled nuclear reactors  

SciTech Connect

Methods for manufacturing porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's). Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, a thin coating of nuclear fuel may be deposited inside of a highly porous skeletal structure made, for example, of reticulated vitreous carbon foam.

Youchison, Dennis L. (Albuquerque, NM); Williams, Brian E. (Pocoima, CA); Benander, Robert E. (Pacoima, CA)

2010-02-23T23:59:59.000Z

119

Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors  

SciTech Connect

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

Youchison, Dennis L. (Albuquerque, NM); Williams, Brian E. (Pacoima, CA); Benander, Robert E. (Pacoima, CA)

2011-03-01T23:59:59.000Z

120

Feasibility study on the verification of fresh fuel assemblies in shipping containers  

SciTech Connect

The purpose of this study was to examine the feasibility of using various nondestructive measurement techniques to determine the presence of fuel assemblies inside shipping containers and to examine the feasibility of measuring the fissile content of the containers. Passive and active techniques based on both gamma and neutron assay were examined. In addition, some experiments and calculations were performed to evaluate neutron techniques. Passive counting of the 186 keV gamma from {sup 235}U is recommended for use as an attributes measurement technique. Experiments and studies indicated that a bismuth germanate (BGO) scintillator is the preferred detector. A properly designed system based on this detector will provide a compact detector that can selectively verify fuel assemblies within a shipping container while the container is in a stack of similarly loaded containers. Missing fuel assemblies will be readily detected, but gamma counting of assemblies cannot detect changes in the fissile content of the inner rods in an assembly. If a variables technique is required, it is recommended that more extensive calculations be performed and removal of the outer shipping container be considered. Marking (sealing) of the assemblies with a uniquely identifiable transponder was also considered. This would require the development of procedures that would assure proper application and removal of the seal. When change to a metal outer container occurs, the technique will no longer be useful unless a radiolucent window is included in the container. 20 refs., 7 figs., 2 tabs.

Swinth, K.L.; Tanner, J.E.

1990-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "fuel assemblies cooling" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


121

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

SciTech Connect

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.

Fisher, S.E.; Holdaway, R.; Ludwig, S.B. [and others

1998-08-01T23:59:59.000Z

122

Transmutation of Transuranic Elements in Advanced MOX and IMF Fuel Assemblies Utilizing Multi-recycling Strategies  

E-Print Network (OSTI)

The accumulation of spent nuclear fuel may be hindering the expansion of nuclear electricity production. However, the reprocessing and recycling of spent fuel may reduce its volume and environmental burden. Although fast spectrum reactors are the preferred modality for transuranic element transmutation, such fast spectrum systems are in very short supply. It is therefore legitimate to investigate the recycling potential of thermal spectrum systems, which constitute the overwhelming majority of nuclear power plants worldwide. To do so efficiently, several new fuel assembly designs are proposed in this Thesis: these include (1) Mixed Oxide Fuel (MOX), (2) MOX fuel with Americium coating, (3) Inert-Matrix Fuel (IMF) with UOX as inner zone, and (4) IMF with MOX as inner zone. All these designs are investigated in a multi-recycling strategy, whereby the spent fuel from a given generation is re-used for the next generation. The accumulation of spent nuclear fuel may be hindering the expansion of nuclear electricity production. However, the reprocessing and recycling of spent fuel may reduce its volume and environmental burden. Although fast spectrum reactors are the preferred modality for transuranic element transmutation, such fast spectrum systems are in very short supply. It is therefore legitimate to investigate the recycling potential of thermal spectrum systems, which constitute the overwhelming majority of nuclear power plants worldwide. To do so efficiently, several new fuel assembly designs are proposed in this Thesis: these include (1) Mixed Oxide Fuel (MOX), (2) MOX fuel with Americium coating, (3) Inert-Matrix Fuel (IMF) with UOX as inner zone, and (4) IMF with MOX as inner zone. All these designs are investigated in a multi-recycling strategy, whereby the spent fuel from a given generation is re-used for the next generation.

Zhang, Yunhuang

2009-12-01T23:59:59.000Z

123

Advanced Water-Cooled Phosphoric Acid Fuel Cell Development. Quarterly technical progress report No. 47, January--March, 1992  

DOE Green Energy (OSTI)

This program is being conducted to improve the performance and minimize the cost of water cooled, electric utility phosphoric acid fuel cell stacks. The program adapts the existing on-site Configuration B cell design to electric utility operating conditions and introduces additional new design features. Task 1 consists of the conceptual design of a full-scale electric utility cell stack that meets program objectives. Tasks 2 and 3 develop the materials and processes requested to fabricate the components that meet the program objective. The design of the small area and two 10-ft{sup 2} short stacks is conducted in Task 4. The conceptual design also is updated to incorporate the results of material and process developments, as well as results of stack tests conducted in Task 6. Fabrication and assembly of the short stacks are conducted in Task 5 and subsequent tests are conducted in Task 6. The Contractor expects to enter into a contract with the Electric Power Research Institute (EPRI) to assemble and endurance test the second 10-ft{sup 2} short stack. The management and reporting functions of Task 7 provide DOE/METC with program visibility through required documentation and program reviews. This report describes the cell design and development effort that is being conducted to demonstrate, by subscale stack test, the technical achievements made toward the above program objectives.

Not Available

1992-05-01T23:59:59.000Z

124

Technology gap analysis on sodium-cooled reactor fuel handling system supporting advanced burner reactor development.  

Science Conference Proceedings (OSTI)

The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand in an environmentally sustainable manner, to address nuclear waste management issues without making separated plutonium, and to address nonproliferation concerns. The advanced burner reactor (ABR) is a fast reactor concept which supports the GNEP fuel cycle system. Since the integral fast reactor (IFR) and advanced liquid-metal reactor (ALMR) projects were terminated in 1994, there has been no major development on sodium-cooled fast reactors in the United States. Therefore, in support of the GNEP fast reactor program, the history of sodium-cooled reactor development was reviewed to support the initiation of this technology within the United States and to gain an understanding of the technology gaps that may still remain for sodium fast reactor technology. The fuel-handling system is a key element of any fast reactor design. The major functions of this system are to receive, test, store, and then load fresh fuel into the core; unload from the core; then clean, test, store, and ship spent fuel. Major requirements are that the system must be reliable and relatively easy to maintain. In addition, the system should be designed so that it does not adversely impact plant economics from the viewpoints of capital investment or plant operations. In this gap analysis, information on fuel-handling operating experiences in the following reactor plants was carefully reviewed: EBR-I, SRE, HNPF, Fermi, SEFOR, FFTF, CRBR, EBR-II, DFR, PFR, Rapsodie, Phenix, Superphenix, KNK, SNR-300, Joyo, and Monju. The results of this evaluation indicate that a standardized fuel-handling system for a commercial fast reactor is yet to be established. However, in the past sodium-cooled reactor plants, most major fuel-handling components-such as the rotatable plug, in-vessel fuel-handling machine, ex-vessel fuel transportation cask, ex-vessel sodium-cooled storage, and cleaning stations-have accumulated satisfactory construction and operation experiences. In addition, two special issues for future development are described in this report: large capacity interim storage and transuranic-bearing fuel handling.

Chikazawa, Y.; Farmer, M.; Grandy, C.; Nuclear Engineering Division

2009-03-01T23:59:59.000Z

125

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

Science Conference Proceedings (OSTI)

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 in MOX fuel is generally obtained from reprocessed irradiated nuclear fuel, whereas weapons-grade plutonium is obtained from decommissioned nuclear weapons material and thus has a different plutonium (and other actinides) concentration. Using MOX fuel instead of UOX fuel has potential impacts on the neutronic performance of the nuclear fuel and the design of the nuclear fuel must take these differences into account. Each of the plutonium sources (RG and WG) has different implications on the neutronic behavior of the fuel because each contains a different blend of plutonium nuclides. The amount of heat and the number of neutrons produced from fission of plutonium nuclides is different from fission of {sup 235}U. These differences in UOX and MOX do not end at discharge of the fuel from the reactor core - the short- and long-term storage of MOX fuel may have different requirements than UOX fuel because of the different discharged fuel decay heat characteristics. The research documented in this report compares MOX and UOX fuel during storage and disposal of the fuel by comparing decay heat rates for typical pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies with and without weapons-grade (WG) and reactor-grade (RG) MOX fuel.

Ade, Brian J [ORNL; Gauld, Ian C [ORNL

2011-10-01T23:59:59.000Z

126

OXIDATION OF FUELS IN THE COOL FLAME REGIME FOR COMBUSTION AND REFORMING FOR FUEL CELLS.  

DOE Green Energy (OSTI)

THE REVIEW INTEGRATES RECENT INVESTIGATIONS ON AUTO OXIDATION OF FUEL OILS AND THEIR REFORMING INTO HYDROGEN RICH GAS THAT COULD SERVE AS A FEED FOR FUEL CELLS AND COMBUSTION SYSTEMS.

NAIDJA,A.; KRISHNA,C.R.; BUTCHER,T.; MAHAJAN,D.

2002-08-01T23:59:59.000Z

127

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

SciTech Connect

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.

Cardoni, Jeffrey

2010-11-01T23:59:59.000Z

128

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

SciTech Connect

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.

Chodak, P. III

1996-05-01T23:59:59.000Z

129

Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank  

DOE Patents (OSTI)

The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps.

Corletti, Michael M. (New Kensington, PA); Lau, Louis K. (Monroeville, PA); Schulz, Terry L. (Murrysville Boro, PA)

1993-01-01T23:59:59.000Z

130

Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank  

DOE Patents (OSTI)

The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps. 1 figures.

Corletti, M.M.; Lau, L.K.; Schulz, T.L.

1993-12-14T23:59:59.000Z

131

Multi-Recycling of Transuranic Elements in a Modified PWR Fuel Assembly  

E-Print Network (OSTI)

The nuclear waste currently generated in the United States is stored in spent fuel pools and dry casks throughout the country awaiting a permanent disposal solution. One efficient solution would be to remove the actinides from the waste and transmute these isotopes in a fast spectrum reactor. Currently this technology is unavailable on a commercial scale and a considerable amount of research and development is still required. An alternate solution is to reprocess and recycle the used fuel in thermal reactors, creating new fuel while reducing the amount of waste and its impact to the environment. This thesis examines the possibility of multi-recycling the transuranics (Pu, Np, Am, and Cm) in a standard pressurized water reactor (PWR). Two types of recycling strategies will be examined: one where Pu, Np, and Am are recycled (TRU-Cm) and a second where the previous isotopes as well as Cm are recycled (TRU+Cm). To offset the hardened neutron spectrum that results from the inclusion of the transuranics, a smaller fuel pin is employed to provide additional moderation. Computer simulations are used to model the in-reactor physics and long-term isotopic decay. Each fuel type is assessed based on the required U-235 enrichment, void coefficient, transuranic production/destruction, and radiotoxicity reduction as compared to a UOX and MOX assembly. It is found that the most beneficial recycling strategy is the one where all of the transuranics are recycled. The inclusion of Cm reduces the required U-235 enrichment, compared to the other multi-recycled fuel and, after a significant number of recycles, can result in the required enrichment to decrease. This fuel type also maintains a negative void coefficient for each recycle. The void coefficient of the fuel type without Cm becomes positive after the third cycle. The transmutation destruction of the two multi-recycled assemblies is less than that of a MOX assembly, but the transmutation efficiency of the multi-recycled assemblies exceeds the MOX assemblies. The radiotoxicity of both multi-recycled assemblies is significantly lower than the UOX and MOX with the TRU+Cm fuel being the lowest. When Curium is recycled only 28,000 years are required for the radiotoxicity of the waste to reach that of natural Uranium and when Cm is not recycled, the amount of time increases to 57,000 years.

Chambers, Alex

2011-08-01T23:59:59.000Z

132

Pwr fuel assembly optimization using adaptive simulated annealing coupled with translat  

E-Print Network (OSTI)

Optimization methods have been developed and refined throughout many scientific fields of study. This work utilizes one such developed technique of optimization called simulated annealing to produce optimal operation parameters for a 15x15 fuel assembly to be used in an operating nuclear power reactor. The two main cases of optimization are: one that finds the optimal 235U enrichment layout of the fuel pins in the assembly and another that finds both the optimal 235U enrichments where gadolinium burnable absorber pins are also inserted. Both of these optimizations can be performed by coupling Adaptive Simulated Annealing to TransLAT which successfully searches the optimization space for a fuel assembly layout that produces the minimized pin power peaking factor. Within given time constraints this package produces optimal layouts within a given set of assumptions and constraints. Each layout is forced to maintain the fuel assembly average 235U enrichment as a constraint. Reductions in peaking factors that are produced through this method are on the order of 2% to 3% when compared to the baseline results. As with any simulated annealing approach, families of optimal layouts are produced that can be used at the engineer’s discretion.

Rogers, Timothy James

2008-08-01T23:59:59.000Z

133

RELIABILITY of FUEL ASSEMBLY EFFLUENT TEMPERATURES UNDER L0CA/LOPA CONDITIONS  

SciTech Connect

The purpose of this study was to ascertain whether or not the K-Reactor safety computers could calculate primarily false positive, but also false negative, and ''on-scale'' misleading fuel assembly average effluent temperatures (AETs) due to relatively large temperature changes in or flooding of the -36 foot elevation isothermal box during a LOCA/LOPA.

Sachs, A.D.

1999-06-21T23:59:59.000Z

134

On0Line Fuel Failure Monitor for Fuel Testing and Monitoring of Gas Cooled Very High Temperature Reactor  

Science Conference Proceedings (OSTI)

IVery High Temperature Reactors (VHTR) utilize the TRISO microsphere as the fundamental fuel unit in the core. The TRISO microsphere (~ 1- mm diameter) is composed of a UO2 kernel surrounded by a porous pyrolytic graphite buffer, an inner pyrolytic graphite layer, a silicon carbide (SiC) coating, and an outer pyrolytic graphite layer. The U-235 enrichment of the fuel is expected to range from 4% – 10% (higher enrichments are also being considered). The layer/coating system that surrounds the UO2 kernel acts as the containment and main barrier against the environmental release of radioactivity. To understand better the behavior of this fuel under in-core conditions (e.g., high temperature, intense fast neutron flux, etc.), the US Department of Energy (DOE) is launching a fuel testing program that will take place at the Advanced Test Reactor (ATR) located at Idaho National Laboratory (INL). During this project North Carolina State University (NCSU) researchers will collaborate with INL staff for establishing an optimized system for fuel monitoring for the ATR tests. In addition, it is expected that the developed system and methods will be of general use for fuel failure monitoring in gas cooled VHTRs.

Ayman I. Hawari; Mohamed A. Bourham

2010-04-22T23:59:59.000Z

135

Fretting wear behaviors of a dual-cooled nuclear fuel rod under a simulated rod vibration  

Science Conference Proceedings (OSTI)

Recently, a dual-cooled fuel (i.e., annular fuel) that is compatible with current operating PWR plants has been proposed in order to realize both a considerable amount of power uprating and an increase of safety margins. As the design concept should be compatible with current operating PWR plants, however, it shows a narrow gap between the fuel rods when compared with current solid nuclear fuel arrays and needs to modify the spacer grid shapes and their positions. In this study, fretting wear tests have been performed to evaluate the wear resistance of a dual-cooled fuel by using a proposed spring and dimple of spacer grids that have a cantilever type and hemispherical shape, respectively. As a result, the wear volume of the spring specimen gradually increases as the contact condition is changed from a certain gap, just contact to positive force. However, in the dimple specimen, just contact condition shows a large wear volume. In addition, a circular rod motion at upper region of contact surface is gradually increased and its diametric size depends on the wear depth increase. Based on the test results, the fretting wear resistance of the proposed spring and dimple is analyzed by comparing the wear measurement results and rod motion in detail.

Lee, Young-Ho; Kim, Hyung-Kyu; Kang, Heung-Seok; Yoon, Kyung-Ho; Kim, Jae-Yong; Lee, Kang-Hee [Innovative Nuclear Fuel Division, Korea Atomic Energy Research Institute 1045 Daedeokdaero, Yuseong, Daejeon, 305-353 (Korea, Republic of)

2012-06-06T23:59:59.000Z

136

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

Science Conference Proceedings (OSTI)

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.

Menlove, Howard O [Los Alamos National Laboratory; Lee, Sang - Yoon [Los Alamos National Laboratory

2009-01-01T23:59:59.000Z

137

EBR-II argon cooling system restricted fuel handling I and C upgrade  

SciTech Connect

The instrumentation and control of the Argon Cooling System (ACS) restricted fuel handling control system at Experimental Breeder Reactor II (EBR-II) is being upgraded from a system comprised of many discrete components and controllers to a computerized system with a graphical user interface (GUI). This paper describes the aspects of the upgrade including reasons for the upgrade, the old control system, upgrade goals, design decisions, philosophies and rationale, and the new control system hardware and software.

Start, S.E.; Carlson, R.B.; Gehrman, R.L. [Argonne National Lab., Idaho Falls, ID (United States). Engineering Div.

1995-06-01T23:59:59.000Z

138

Processing of FRG high-temperature gas-cooled reactor fuel elements at General Atomic under the US/FRG cooperative agreement for spent fuel elements  

Science Conference Proceedings (OSTI)

The Federal Republic of Germany (FRG) and the United States (US) are cooperating on certain aspects of gas-cooled reactor technology under an umbrella agreement. Under the spent fuel treatment development section of the agreement, both FRG mixed uranium/ thorium and low-enriched uranium fuel spheres have been processed in the Department of Energy-sponsored cold pilot plant for high-temperature gas-cooled reactor (HTGR) fuel processing at General Atomic Company in San Diego, California. The FRG fuel spheres were crushed and burned to recover coated fuel particles suitable for further treatment for uranium recovery. Successful completion of the tests described in this paper demonstrated certain modifications to the US HTGR fuel burining process necessary for FRG fuel treatment. Results of the tests will be used in the design of a US/FRG joint prototype headend facility for HTGR fuel.

Holder, N.D.; Strand, J.B.; Schwarz, F.A.; Drake, R.N.

1981-11-01T23:59:59.000Z

139

Combustor assembly for use in a turbine engine and methods of assembling same  

SciTech Connect

A fuel nozzle assembly for use with a turbine engine is described herein. The fuel nozzle assembly includes a plurality of fuel nozzles positioned within an air plenum defined by a casing. Each of the plurality of fuel nozzles is coupled to a combustion liner defining a combustion chamber. Each of the plurality of fuel nozzles includes a housing that includes an inner surface that defines a cooling fluid plenum and a fuel plenum therein, and a plurality of mixing tubes extending through the housing. Each of the mixing tubes includes an inner surface defining a flow channel extending between the air plenum and the combustion chamber. At least one mixing tube of the plurality of mixing tubes including at least one cooling fluid aperture for channeling a flow of cooling fluid from the cooling fluid plenum to the flow channel.

Uhm, Jong Ho; Johnson, Thomas Edward

2013-05-14T23:59:59.000Z

140

Qualification of the Frequency-Scanning Eddy Current Technique for LWR Fuel Assembly Structural Components Under Simulated Poolside Conditions  

Science Conference Proceedings (OSTI)

There is a growing need for nondestructive characterization of hydrogen content in zirconium alloys used in light water reactor (LWR) fuel assembly components. Pending revisions to current regulations and industry emphasis on determining fuel reliability margins have driven the development of poolside-deployable nondestructive characterization techniques to inspect fuel assembly materials. Poolside inspection reduces risk, dose, time and costs related to the transportation of these materials for ...

2013-10-29T23:59:59.000Z

Note: This page contains sample records for the topic "fuel assemblies cooling" from the National Library of EnergyBeta (NLEBeta).
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141

Automated closure system for nuclear reactor fuel assemblies  

DOE Patents (OSTI)

A welder for automated closure of fuel pins by a pulsed magnetic process in which the open end of a length of cladding is positioned within a complementary tube surrounded by a pulsed magnetic welder. Seals are provided at each end of the tube, which can be evacuated or can receive tag gas for direct introduction to the cladding interior. Loading of magnetic rings and end caps is accomplished automatically in conjunction with the welding steps carried out within the tube.

Christiansen, David W. (Kennewick, WA); Brown, William F. (West Richland, WA)

1985-01-01T23:59:59.000Z

142

Welding fixture for nuclear fuel pin cladding assemblies  

DOE Patents (OSTI)

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.

Oakley, David J. (Richland, WA); Feld, Sam H. (West Richland, WA)

1986-01-01T23:59:59.000Z

143

Welding fixture for nuclear fuel pin cladding assemblies  

DOE Patents (OSTI)

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.

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

1984-02-22T23:59:59.000Z

144

Market share elasticities for fuel and technology choice in home heating and cooling  

Science Conference Proceedings (OSTI)

A new technique for estimating own- and cross-elasticities of market share for fuel and technology choices in home heating and cooling is presented. We simulate changes in economic conditions and estimate elasticities by calculating predicted changes in fuel and technology market shares. Elasticities are found with respect to household income, equipment capital cost, and equipment capital cost, and equipment operating cost (including fuel price). The method is applied to a revised and extended version of a study by the Electric Power Research Institute (EPRI). Data for that study are drawn primarily from the 1975--1979 Annual Housing Surveys. Results are generally similar to previous studies, although our estimates of elasticities are somewhat lower. We feel the superior formulation of consumer choice and the currency of data in EPRI's work produce reliable estimates of market share elasticities. 18 refs., 1 fig., 6 tabs.

Wood, D.J.; Ruderman, H.; McMahon, J.E.

1989-05-01T23:59:59.000Z

145

Deep Burn: Development of Transuranic Fuel for High-Temperature Helium-Cooled Reactors- Monthly Highlights October 2010  

Science Conference Proceedings (OSTI)

The DB Program monthly highlights report for September 2010, ORNL/TM-2010/252, was distributed to program participants by email on October 26. This report discusses: (1) Core and Fuel Analysis; (2) Spent Fuel Management; (3) Fuel Cycle Integration of the HTR (high temperature helium-cooled reactor); (4) TRU (transuranic elements) HTR Fuel Qualification; (5) HTR Spent Fuel Recycle - (a) TRU Kernel Development (ORNL), (b) Coating Development (ORNL), (c) Characterization Development and Support, (d) ZrC Properties and Handbook; and (6) HTR Fuel Recycle.

Snead, Lance Lewis [ORNL; Besmann, Theodore M [ORNL; Collins, Emory D [ORNL; Bell, Gary L [ORNL

2010-11-01T23:59:59.000Z

146

Method for recovering catalytic elements from fuel cell membrane electrode assemblies  

DOE Patents (OSTI)

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.

Shore, Lawrence (Edison, NJ); Matlin, Ramail (Berkeley Heights, NJ); Heinz, Robert (Ludwigshafen, DE)

2012-06-26T23:59:59.000Z

147

Emergency core cooling system  

DOE Patents (OSTI)

A liquid metal cooled fast breeder reactor provided with an emergency core cooling system includes a reactor vessel which contains a reactor core comprising an array of fuel assemblies and a plurality of blanket assemblies. The reactor core is immersed in a pool of liquid metal coolant. The reactor also includes a primary coolant system comprising a pump and conduits for circulating liquid metal coolant to the reactor core and through the fuel and blanket assemblies of the core. A converging-diverging venturi nozzle with an intermediate throat section is provided in between the assemblies and the pump. The intermediate throat section of the nozzle is provided with at least one opening which is in fluid communication with the pool of liquid sodium. In normal operation, coolant flows from the pump through the nozzle to the assemblies with very little fluid flowing through the opening in the throat. However, when the pump is not running, residual heat in the core causes fluid from the pool to flow through the opening in the throat of the nozzle and outwardly through the nozzle to the assemblies, thus providing a means of removing decay heat.

Schenewerk, William E. (Sherman Oaks, CA); Glasgow, Lyle E. (Westlake Village, CA)

1983-01-01T23:59:59.000Z

148

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

SciTech Connect

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.

Cappiello, C.C.; Butterfield, K.B.; Sanchez, R.G. [and others

1997-10-01T23:59:59.000Z

149

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

Science Conference Proceedings (OSTI)

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.

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

1980-05-01T23:59:59.000Z

150

Effect of Fuel Fraction on Small Modified CANDLE Burn-up Based Gas Cooled Fast Reactors  

Science Conference Proceedings (OSTI)

A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE Burn-up has been performed. The objective of this research is to get optimal design parameters of such type reactors. The parameters of nuclear design including the critical condition, conversion ratio, and burn-up level were compared. These parameters are calculated by variation in the fuel fraction 47.5% up to 70%. Two dimensional full core multi groups diffusion calculations was performed by CITATION code. Group constant preparations are performed by using SRAC code system with JENDL-3.2 nuclear data library. In this design the reactor cores with cylindrical cell two dimensional R-Z core models are subdivided into several parts with the same volume in the axial directions. The placement of fuel in core arranged so that the result of plutonium from natural uranium can be utilized optimally for 10 years reactor operation. Modified CANDLE burn-up was established successfully in a core radial width 1.4 m. Total thermal power output for reference core is 550 MW. Study on the effect of fuel to coolant ratio shows that effective multiplication factor (k{sub eff}) is in almost linear relations with the change of the fuel volume to coolant ratio.

Ariani, Menik [Departmen of Physics Bandung Institute of Technology, Jl. Ganesha 10, Bandung 40134 (Indonesia); Physics Department, Sriwijaya University, Kampus Indralaya, Ogan Ilir, Sumatera Selatan (Indonesia); Su'ud, Zaki; Waris, Abdul; Asiah, Nur [Departmen of Physics Bandung Institute of Technology, Jl. Ganesha 10, Bandung 40134 (Indonesia); Shafii, M. Ali [Departmen of Physics Bandung Institute of Technology, Jl. Ganesha 10, Bandung 40134 (Indonesia); Physics Department, Andalas University, Kampus Limau Manis, Padang, Sumatera Barat (Indonesia); Khairurrijal

2010-12-23T23:59:59.000Z

151

Estimation of critical flow velocity for collapse of booster fuel assembly  

Science Conference Proceedings (OSTI)

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.

Donna Guillen; Mark J. Russell

2005-09-01T23:59:59.000Z

152

Radial blanket assembly orificing arrangement  

DOE Patents (OSTI)

A nuclear reactor core for a liquid metal cooled fast breeder reactor is described in which means are provided for increasing the coolant flow through the reactor fuel assemblies as the reactor ages by varying the coolant flow rate with the changing coolant requirements during the core operating lifetime. (auth)

Patterson, J.F.

1975-07-01T23:59:59.000Z

153

Market Share Elasticities for Fuel and Technology Choice in Home Heating and Cooling  

E-Print Network (OSTI)

level, the choice alternatives are cooling and no cooling.to zero in central cooling alternative Income ($1000) in airalternatives are conventional air conditioning and heat pump, given the cooling

Wood, D.J.

2010-01-01T23:59:59.000Z

154

Non-fuel assembly components: 10 CFR 61.55 classification for waste disposal  

SciTech Connect

This document reports the results of laboratory radionuclide measurements on a representative group of non-fuel assembly (NFA) components for the purposes of waste classification. This document also provides a methodology to estimate the radionuclide inventory of NFA components, including those located outside the fueled region of a nuclear reactor. These radionuclide estimates can then be used to determine the waste classification of NFA components for which there are no physical measurements. Previously, few radionuclide inventory measurements had been performed on NFA components. For this project, recommended scaling factors were selected for the ORIGEN2 computer code that result in conservative estimates of radionuclide concentrations in NFA components. These scaling factors were based upon experimental data obtained from the following NFA components: (1) a pressurized water reactor (PWR) burnable poison rod assembly, (2) a PVM rod cluster control assembly, and (3) a boiling water reactor cruciform control rod blade. As a whole, these components were found to be within Class C limits. Laboratory radionuclide measurements for these components are provided in detail.

Migliore, R.J.; Reid, B.D.; Fadeff, S.K.; Pauley, K.A.; Jenquin, U.P.

1994-09-01T23:59:59.000Z

155

Fuel Cycle Optimization of a Helium-Cooled, Sub-Critical, Fast Transmutation of Waste Reactor with a Fusion Neutron Source.  

E-Print Network (OSTI)

??Possible fuel cycle scenarios for a helium-cooled, sub-critical, fast reactor with a fusion neutron source for the transmutation of spent nuclear fuel have been analyzed.… (more)

Maddox, James Warren

2006-01-01T23:59:59.000Z

156

Development of coupled SCALE4.2/GTRAN2 computational capability for advanced MOX fueled assembly designs  

Science Conference Proceedings (OSTI)

An advanced assembly code system that can efficiently and accurately analyze various designs (current and advanced) proposed for plutonium disposition is being developed by {open_quotes}marrying{close_quotes} two existing state-of-the-art methodologies-GTRAN2 and SCALE 4.2. The resulting code system, GT-SCALE, posses several unique characteristics: exact 2D representation of a complete fuel assembly, while preserving the heterogeniety of each of its pin cells; flexibility in the energy group structure, the present upper limit being 218 groups; a comprehensive cross-section library and material data base; and accurate burnup calculations. The resulting GT-SCALE is expected to be very useful for a wide variety of applications, including the analysis of very heterogeneous UO{sub 2} fueled LWR fuel assemblies; of hexagonal shaped fuel assemblies as of the Russian LWRs; of fuel assemblies for HTGRs; as well as for the analysis of criticality safety and for calculation of the source term of spent fuel.

Vujic, J.; Greenspan, E.; Slater, Postma, T.; Casher, G.; Soares, I. [Univ. of California, Berkeley, CA (United States). Dept. of Nuclear Engineering; Leal, L. [Oak Ridge National Lab., TN (United States)

1995-05-01T23:59:59.000Z

157

Hanford MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement  

SciTech Connect

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. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. Six initial site combinations were proposed: (1) Argonne National Laboratory-West (ANL-W) with support from Idaho National Engineering and Environmental Laboratory (INEEL), (2) Hanford, (3) Los Alamos National Laboratory (LANL) with support from Pantex, (4) Lawrence Livermore National Laboratory (LLNL), (5) Oak Ridge Reservation (ORR), and (6) Savannah River Site (SRS). After further analysis by the sites and DOE-MD, five site combinations were established as possible candidates for producing MOX LAs: (1) ANL-W with support from INEEL, (2) Hanford, (3) LANL, (4) LLNL, and (5) SRS. Hanford has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. An alternate approach would allow fabrication of fuel pellets and assembly of fuel rods in an S and S Category 1 facility. In all, a total of three LA MOX fuel fabrication options were identified by Hanford that could accommodate the program. In every case, only minor modification would be required to ready any of the facilities to accept the equipment necessary to accomplish the LA program.

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

1998-08-01T23:59:59.000Z

158

Supplemental information for a notice of construction for the Fueled Clad Fabrication System, the Radioisotope Power Systems Facility, and the Fuel Assembly Area  

Science Conference Proceedings (OSTI)

This ''Notice of Construction'' has been submitted by the US Department of Energy-Richland Operations Office (P.O. Box 550, Richland, Washington 99352), pursuant to WAC 402-80-070, for three new sources of radionuclide emissions at the Hanford Site in Washington State (Figure 1). The three new sources, the Fueled Clad Fabrication System (FCFS) the Radioisotope Power Systems Facility (RPSF) and the Fuel Assembly Area (FAA) will be located in one facility, the Fuels and materials Examination Facility (FMEF) of the 400 Area. The FMEF was originally designed to provide for post- irradiation examination and fabrication of breeder reactor fuels. These FMEF missions were cancelled before the introduction of any fuel materials or any irradiated material. The current plans are to use the facility to fabricate power supplies to be used in space applications and to produce Fast Flux Test Facility (FFTF) fuel and target assemblies. The FCFS and the RPSF will produce materials and assemblies for application in space. The FAA project will produce FFTF fuel and target assemblies. The FCFS and the RPSF will share the same building, stack, and, in certain cases, the same floor space. Given this relationship, to the extent possible, these systems will be dealt with separately. The FAA is a comparatively independent operation though it will share the FMEF complex.

Not Available

1989-08-01T23:59:59.000Z

159

Prevention of significant deterioration permit application for the Fueled Clad Fabrication System, the Radioisotope Power Systems Facility, and the Fuel Assembly Area  

SciTech Connect

This New Source Review'' has been submitted by the US Department of Energy-Richland Operations Office (PO Box 550, Richland, Washington 99352), pursuant to WAC 173-403-050 and in compliance with the Department of Ecology Guide to Processing A Prevention Of Significant Deterioration (PSD) Permit'' for three new sources of radionuclide emissions at the Hanford Site in Washington State. The three new sources, the Fueled Clad Fabrication System (FCFS), the Radioisotope Power Systems Facility (RPSF), and the Fuel Assembly Area (FAA), will be located in one facility, the Fuels and Materials Examination Facility (FMEF) of the 400 Area. The FMEF was originally designed to provide for post-irradiation examination and fabrication of breeder reactor fuels. These FMEF missions were cancelled before the introduction of any fuel materials or any irradiated material. The current plans are to use the facility to fabricate power supplies for use in space applications and to produce Fast Flux Test Facility (FFTF) fuel and target assemblies. The FCFS and the RPSF will produce materials and assemblies for application in space. The FAA project will produce FFTF fuel and target assemblies. The FCFS and the RPSF will share the same building, stack, and, in certain cases, the same floor space. Given this relationship, these systems will be dealt with separately to the extent possible. The FAA is a comparatively independent operation though it will share the FMEF complex.

Not Available

1989-08-01T23:59:59.000Z

160

Deep Burn: Development of Transuranic Fuel for High-Temperature Helium-Cooled Reactors- Monthly Highlights September 2010  

SciTech Connect

The DB Program monthly highlights report for August 2010, ORNL/TM-2010/184, was distributed to program participants by email on September 17. This report discusses: (1) Core and Fuel Analysis - (a) Core Design Optimization in the HTR (high temperature helium-cooled reactor) Prismatic Design (Logos), (b) Core Design Optimization in the HTR Pebble Bed Design (INL), (c) Microfuel analysis for the DB HTR (INL, GA, Logos); (2) Spent Fuel Management - (a) TRISO (tri-structural isotropic) repository behavior (UNLV), (b) Repository performance of TRISO fuel (UCB); (3) Fuel Cycle Integration of the HTR (high temperature helium-cooled reactor) - Synergy with other reactor fuel cycles (GA, Logos); (4) TRU (transuranic elements) HTR Fuel Qualification - (a) Thermochemical Modeling, (b) Actinide and Fission Product Transport, (c) Radiation Damage and Properties; (5) HTR Spent Fuel Recycle - (a) TRU Kernel Development (ORNL), (b) Coating Development (ORNL), (c) Characterization Development and Support, (d) ZrC Properties and Handbook; and (6) HTR Fuel Recycle - (a) Graphite Recycle (ORNL), (b) Aqueous Reprocessing, (c) Pyrochemical Reprocessing METROX (metal recovery from oxide fuel) Process Development (ANL).

Snead, Lance Lewis [ORNL; Besmann, Theodore M [ORNL; Collins, Emory D [ORNL; Bell, Gary L [ORNL

2010-10-01T23:59:59.000Z

Note: This page contains sample records for the topic "fuel assemblies cooling" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


161

Apparatus and method for preventing the rotation of rods used in nuclear fuel assemblies  

Science Conference Proceedings (OSTI)

Apparatus and method for preventing the rotation of one or more elongated rods used in nuclear fuel assemblies include an end plug secured to one longitudinal end of such an elongated rod and having an out-of-cavity, non-round structure affixed thereto and configured to mate with a complementary shaped structure in a lower tie plate of a nuclear fuel assembly in such a manner as to prevent the rotation of the rod about its longitudinal axis. In one embodiment, the end plug includes a pair of flats formed on a portion of the end plug and configured to abut against a pair of flats formed on the outer surface of a cylindrical boss or sleeve of the lower tie plate, thereby to prevent the rotation of the rod. In another embodiment, four grooves, disposed 90/sup 0/ apart about the periphery of an end plug of a rod form a spline. The grooves are configured to receive four, radially inwardly protruding, key members disposed 90/sup 0/ apart about the periphery of a sleeve secured to the lower tie plate, thereby to prevent the rotation of the rod. In a further embodiment, a sleeve is secured to an end plug of a rod and includes four elongated slots disposed 90/sup 0/ apart about the periphery of the sleeve and configured in width, depth and spacing to receive and mate with four web portions of the lower tie plate of the nuclear fuel assembly, thereby to secure the rod against rotation about its longitudinal axis.

Pilgrim Jr., L. G.; Jackson, L. F.

1985-01-08T23:59:59.000Z

162

Forced-convection boiling tests performed in parallel simulated LMR fuel assemblies  

SciTech Connect

Forced-convection tests have been carried out using parallel simulated Liquid Metal Reactor fuel assemblies in an engineering-scale sodium loop, the Thermal-Hydraulic Out-of-Reactor Safety facility. The tests, performed under single- and two-phase conditions, have shown that for low forced-convection flow there is significant flow augmentation by thermal convection, an important phenomenon under degraded shutdown heat removal conditions in an LMR. The power and flows required for boiling and dryout to occur are much higher than decay heat levels. The experimental evidence supports analytical results that heat removal from an LMR is possible with a degraded shutdown heat removal system.

Rose, S.D.; Carbajo, J.J.; Levin, A.E.; Lloyd, D.B.; Montgomery, B.H.; Wantland, J.L.

1985-04-21T23:59:59.000Z

163

COMMERCIAL SPENT NUCLEAR FUEL ASSEMBLY WASTE VOLUME AND WEIGHT ESTIMATION FOR 63,000 MTU  

Science Conference Proceedings (OSTI)

The purpose of this calculation is to create a high-level estimation of the weights and volume of the commercial spent nuclear fuel (SNF) assemblies, at the time of repository receipt, that will comprise 63,000 metric tons of uranium (MTU) waste. The results of this calculation are for informational purposes only and are not intended to be used as input to design documents. This calculation was prepared in accordance with procedure AP-3.12Q REV 00 ICN 0, Calculations.

B.A. Colton

1999-07-22T23:59:59.000Z

164

Hydraulically actuated fuel injector including a pilot operated spool valve assembly and hydraulic system using same  

DOE Patents (OSTI)

The present invention relates to hydraulic systems including hydraulically actuated fuel injectors that have a pilot operated spool valve assembly. One class of hydraulically actuated fuel injectors includes a solenoid driven pilot valve that controls the initiation of the injection event. However, during cold start conditions, hydraulic fluid, typically engine lubricating oil, is particularly viscous and is often difficult to displace through the relatively small drain path that is defined past the pilot valve member. Because the spool valve typically responds slower than expected during cold start due to the difficulty in displacing the relatively viscous oil, accurate start of injection timing can be difficult to achieve. There also exists a greater difficulty in reaching the higher end of the cold operating speed range. Therefore, the present invention utilizes a fluid evacuation valve to aid in displacement of the relatively viscous oil during cold start conditions.

Shafer, Scott F. (Morton, IL)

2002-01-01T23:59:59.000Z

165

Container for reprocessing and permanent storage of spent nuclear fuel assemblies  

DOE Patents (OSTI)

A single canister process container is described 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 its integrity at a temperature necessary to oxidize the zirconium-based cladding. The process container also has means to introduce and remove fluids to and from the container. 10 figs.

Forsberg, C.W.

1992-03-24T23:59:59.000Z

166

Depletion Analysis of Modular High Temperature Gas-cooled Reactor Loaded with LEU/Thorium Fuel  

SciTech Connect

Thorium based fuel has been considered as an option to uranium-based fuel, based on considerations of resource utilization (Thorium is more widely available when compared to Uranium). The fertile isotope of Thorium (Th-232) can be converted to fissile isotope U-233 by neutron capture during the operation of a suitable nuclear reactor such as High Temperature Gas-cooled Reactor (HTGR). However, the fertile Thorium needs a fissile supporter to start and maintain the conversion process such as U-235 or Pu-239. This report presents the results of a study that analyzed the thorium utilization in a prismatic HTGR, namely Modular High Temperature Gas-Cooled Reactor (MHTGR) that was designed by General Atomics (GA). The collected for the modeling of this design come from Chapter 4 of MHTGR Preliminary Safety Information Document that GA sent to Department of Energy (DOE) on 1995. Both full core and unit cell models were used to perform this analysis using SCALE 6.1 and Serpent 1.1.18. Because of the long mean free paths (and migration lengths) of neutrons in HTRs, using a unit cell to represent a whole core can be non-trivial. The sizes of these cells were set to match the spectral index between unit cell and full core domains. It was found that for the purposes of this study an adjusted unit cell model is adequate. Discharge isotopics and one-group cross-sections were delivered to the transmutation analysis team. This report provides documentation for these calculations

Sonat Sen; Gilles Youinou

2013-02-01T23:59:59.000Z

167

Irradiated fuel monitoring by Cerenkov glow intensity measurements  

SciTech Connect

Attribute measurement techniques for confirmation of declared irradiated fuel inventories at nuclear installations under safeguards surveillance are being investigated. High-gain measurements of the intensity of the Cerenkov glow from exposed assemblies in water-filled storage ponds are promising for this purpose. Such measurements have been made of Materials Testing Reactor plate-type fuel assemblies and Pressurized Water Reactor pin-type fuel assemblies. The measured intensities depend on cooling times as calculations predict.

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

1979-09-01T23:59:59.000Z

168

Market Share Elasticities for Fuel and Technology Choice in Home Heating and Cooling  

E-Print Network (OSTI)

Choice in Home Heating and Cooling D.J. Wood, H. Ruderman,IN HOME HEATING AND COOLING* David J. Wood, Henry RudermanIN HOME HEATING AND COOLING David J. Wood, Henry Ruderman,

Wood, D.J.

2010-01-01T23:59:59.000Z

169

Characterization of candidate DOE sites for fabricating MOX fuel for lead assemblies  

Science Conference Proceedings (OSTI)

The Office of Fissile Materials Disposition (MD) of the Department of Energy (DOE) is directing the program to disposition US surplus weapons-usable plutonium. For the reactor option for disposition of this surplus plutonium, MD is seeking to contract with a consortium, which would include a mixed-oxide (MOX) fuel fabricator and a commercial US reactor operator, to fabricate and burn MOX fuel in existing commercial nuclear reactors. This option would entail establishing a MOX fuel fabrication facility under the direction of the consortium on an existing DOE site. Because of the lead time required to establish a MOX fuel fabrication facility and the need to qualify the MOX fuel for use in a commercial reactor, MD is considering the early fabrication of lead assemblies (LAs) in existing DOE facilities under the technical direction of the consortium. The LA facility would be expected to produce a minimum of 1 metric ton heavy metal per year and must be operational by June 2003. DOE operations offices were asked to identify candidate sites and facilities to be evaluated for suitability to fabricate MOX fuel LAs. Savannah River Site, Argonne National Laboratory-West, Hanford, Lawrence Livermore National Laboratory, and Los Alamos National Laboratory were identified as final candidates to host the LA project. A Site Evaluation Team (SET) worked with each site to develop viable plans for the LA project. SET then characterized the suitability of each of the five plans for fabricating MOX LAs using 28 attributes and documented the characterization to aid DOE and the consortium in selecting the site for the LA project. SET concluded that each option has relative advantages and disadvantages in comparison with other options; however, each could meet the requirements of the LA project as outlined by MD and SET.

Holdaway, R.F.; Miller, J.W.; Sease, J.D.; Moses, R.J.; O`Connor, D.G. [Oak Ridge National Lab., TN (United States); Carrell, R.D. [Technical Resources International, Inc., Richland, WA (United States); Jaeger, C.D. [Sandia National Labs., Albuquerque, NM (United States); Thompson, M.L.; Strasser, A.A. [Delta-21 Resources, Inc., Oak Ridge, TN (United States)

1998-03-01T23:59:59.000Z

170

Modeling and Implementation of a 1 kW, Air Cooled HTPEM Fuel Cell in a Hybrid Electrical Vehicle  

E-Print Network (OSTI)

Modeling and Implementation of a 1 kW, Air Cooled HTPEM Fuel Cell in a Hybrid Electrical Vehicle engine vehicles (1). Hybrid systems of many kinds, combining a primary energy source having a high energy://www.ecsdl.org/terms_use.jsp #12;article, a model of a hybrid vehicle, including a HTPEM with lead acid batteries, is de- veloped

Nielsen, Mads Pagh

171

LLNL Site plan for a MOX fuel lead assembly mission in support of surplus plutonium disposition  

SciTech Connect

The principal facilities that LLNL would use to support a MOX Fuel Lead Assembly Mission are Building 332 and Building 334. Both of these buildings are within the security boundary known as the LLNL Superblock. Building 332 is the LLNL Plutonium Facility. As an operational plutonium facility, it has all the infrastructure and support services required for plutonium operations. The LLNL Plutonium Facility routinely handles kilogram quantities of plutonium and uranium. Currently, the building is limited to a plutonium inventory of 700 kilograms and a uranium inventory of 300 kilograms. Process rooms (excluding the vaults) are limited to an inventory of 20 kilograms per room. Ongoing operations include: receiving SSTS, material receipt, storage, metal machining and casting, welding, metal-to-oxide conversion, purification, molten salt operations, chlorination, oxide calcination, cold pressing and sintering, vitrification, encapsulation, chemical analysis, metallography and microprobe analysis, waste material processing, material accountability measurements, packaging, and material shipping. Building 334 is the Hardened Engineering Test Building. This building supports environmental and radiation measurements on encapsulated plutonium and uranium components. Other existing facilities that would be used to support a MOX Fuel Lead Assembly Mission include Building 335 for hardware receiving and storage and TRU and LLW waste storage and shipping facilities, and Building 331 or Building 241 for storage of depleted uranium.

Bronson, M.C.

1997-10-01T23:59:59.000Z

172

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

SciTech Connect

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.

Henzl, Vladimir [Los Alamos National Laboratory; Swinhoe, Martyn T. [Los Alamos National Laboratory; Tobin, Stephen J. [Los Alamos National Laboratory

2012-07-16T23:59:59.000Z

173

A Pressurized Water Reactor Plutonium Incinerator Based on Thorium Fuel and Seed-Blanket Assembly Geometry  

Science Conference Proceedings (OSTI)

A pressurized water reactor (PWR) fuel cycle is proposed, whose purpose is the elimination and degradation of weapons-grade plutonium. This Radkowsky thorium-fuel Pu incinerator (RTPI) cycle is based on a core and assemblies retrofittable to a Westinghouse-type PWR. The RTPI assembly, however, is a seed-blanket unit. The seed is supercritical, loaded with Pu-Zr alloy as fuel in a high moderator-to-fuel ratio configuration. The blanket is subcritical, loaded mainly with ThO{sub 2}, generating and burning {sup 233}U in situ. Blankets are loaded once every 6 yr. The seed fuel management scheme is based on three batches, with one-third of the seed modules replaced every year. The core generates 1100 MW(electric). Equilibrium conditions are achieved with the second seed loading. For equilibrium conditions, the annual average of disposed (loaded) Pu is 1210 kg, of which 702 kg are completely eliminated, and 508 kg are discharged, but with significantly degraded isotopics (i.e., with a high percentage of even mass isotopes). Spontaneous fissions per second in a gram of this degraded Pu are {approx}500, resulting in significantly increased proliferation resistance.Every 6 yr the blanket discharge contains 780 kg of {sup 233}U (including {sup 233}Pa) and 36 kg of {sup 235}U. However, the blankets are initially loaded with an amount of natural uranium selected such that these U fissile isotopes constitute only 12% of the total U discharge, a percentage equivalent to 20% {sup 235}U enrichment; hence, both the discharged uranium isotopics satisfy proliferation-resistant criteria.The RTPI control variables, namely, the moderator temperature coefficient, the reactivity per ppm boron, and the control rods worth, are about equal to those of a PWR. The RTPI spent-fuel stockpile ingestion toxicity over a period of ten million years is about the same as the counterpart toxicities of a regular, or a mixed-oxide (MOX), PWR. Compared with known PWR MOX variants, the RTPI is, per 1000 MW(electric) and per annum, a significantly more efficient incinerator of weapons-grade plutonium.

Galperin, A. [Ben-Gurion University of the Negev (Israel); Segev, M. [Ben-Gurion University of the Negev (Israel); Todosow, M. [Brookhaven National Laboratory (United States)

2000-11-15T23:59:59.000Z

174

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

Science Conference Proceedings (OSTI)

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

Lata

1996-09-26T23:59:59.000Z

175

[Gas cooled fuel cell systems technology development program]. Quarterly technical progress narrative No. 21, December 1, 1987--February 29, 1988  

DOE Green Energy (OSTI)

Objective is the development of a gas-cooled phosphoric acid fuel cell for electric utility power plant application. Primary objectives are to: demonstrate performance endurance in 10-cell stacks at 70 psia, 190 C, and 267 mA/cm{sup 2}; improve cell degradation rate to less than 8 mV/1000 hours; develop cost effective criteria, processes, and design configurations for stack components; design multiple stack unit and a single 100 kW fuel cell stack; design a 375 kW fuel cell module and demonstrate average cell beginning-of-use performance; manufacture four 375-kW fuel cell modules and establish characteristics of 1.5 MW pilot power plant. The work is broken into program management, systems engineering, fuel cell development and test, facilities development.

Not Available

1988-03-01T23:59:59.000Z

176

Fluoride-Salt-Cooled High-Temperature Reactor (FHR) with Silicon-Carbide-Matrix Coated-Particle Fuel  

Science Conference Proceedings (OSTI)

The FHR is a new reactor concept that uses coated-particle fuel and a low-pressure liquid-salt coolant. Its neutronics are similar to a high-temperature gas-cooled reactor (HTGR). The power density is 5 to 10 times higher because of the superior cooling properties of liquids versus gases. The leading candidate coolant salt is a mixture of {sup 7}LiF and BeF{sub 2} (FLiBe) possessing a boiling point above 1300 C and the figure of merit {rho}C{sub p} (volumetric heat capacity) for the salt slightly superior to water. Studies are underway to define a near-term base-line concept while understanding longer-term options. Near-term options use graphite-matrix coated-particle fuel where the graphite is both a structural component and the primary neutron moderator. It is the same basic fuel used in HTGRs. The fuel can take several geometric forms with a pebble bed being the leading contender. Recent work on silicon-carbide-matrix (SiCm) coated-particle fuel may create a second longer-term fuel option. SiCm coated-particle fuels are currently being investigated for use in light-water reactors. The replacement of the graphite matrix with a SiCm creates a new family of fuels. The first motivation behind the effort is to take advantage of the superior radiation resistance of SiC compared to graphite in order to provide a stable matrix for hosting coated fuel particles. The second motivation is a much more rugged fuel under accident, repository, and other conditions.

Forsberg, C. W. [Massachusetts Institute of Technology (MIT); Terrani, Kurt A [ORNL; Snead, Lance Lewis [ORNL; Katoh, Yutai [ORNL

2012-01-01T23:59:59.000Z

177

Characteristics of spent fuel, high-level waste, and other radioactive wastes which may require long-term isolation: Appendix 2A, Physical descriptions of LWR (Light-Water Reactor) fuel assemblies  

SciTech Connect

This appendix includes a four-page Physical Description report for each assembly type identified from the current data. Where available, a drawing of an assembly follows the appropriate Physical Description report. If no drawing is available for an assembly, a cross-reference to a similar assembly is provided if possible. For Advanced Nuclear Fuels, Babcock and Wilcox, Combustion Engineering, and Westinghouse assemblies, information was obtained via subcontracts with these fuel vendors. Data for some assembly types are not available. For such assemblies, the information shown in this report was obtained from the open literature and by inference from reload fuels made by other vendors. Efforts to obtain additional information are continuing. Individual Physical Description reports can be generated interactively through the menu-driven LWR Assemblies Data Base system. These reports can be viewed on the screen or directed to a printer. Special reports and compilations of specific data items can be produced on request.

Not Available

1987-12-01T23:59:59.000Z

178

Estimation of Critical Flow Velocity for Collapse of Gas Test Loop Booster Fuel Assembly  

Science Conference Proceedings (OSTI)

This paper presents calculations performed to determine the critical flow velocity for plate collapse due to static instability for the Gas Test Loop booster fuel assembly. Long, slender plates arranged in a parallel configuration can experience static divergence and collapse at sufficiently high coolant flow rates. Such collapse was exhibited by the Oak Ridge High Flux Reactor in the 1940s and the Engineering Test Reactor at the Idaho National Laboratory in the 1950s. Theoretical formulas outlined by Miller, based upon wide-beam theory and Bernoulli’s equation, were used for the analysis. Calculations based upon Miller’s theory show that the actual coolant flow velocity is only 6% of the predicted critical flow velocity. Since there is a considerable margin between the theoretically predicted plate collapse velocity and the design velocity, the phenomena of plate collapse due to static instability is unlikely.

Guillen; Mark J. Russell

2006-07-01T23:59:59.000Z

179

Parametric Study of CHF Data, Volume 2: A Generalized Subchannel CHF Correlation for PWR and BWR Fuel Assemblies  

Science Conference Proceedings (OSTI)

This volume describes the development of a generalized subchannel critical heat flux (CHF) correlation for PWR and BWR fuel assemblies. The effects of nonuniform axial heat flux, cold walls, and grid spacers are discussed, and the correlation's performance is compared with a wide range of data.

1983-01-01T23:59:59.000Z

180

Micro Cooling, Heating, and Power (Micro-CHP) and Bio-Fuel Center, Mississippi State University  

Science Conference Proceedings (OSTI)

Initially, most micro-CHP systems will likely be designed as constant-power output or base-load systems. This implies that at some point the power requirement will not be met, or that the requirement will be exceeded. Realistically, both cases will occur within a 24-hour period. For example, in the United States, the base electrical load for the average home is approximately 2 kW while the peak electrical demand is slightly over 4 kW. If a 3 kWe micro- CHP system were installed in this situation, part of the time more energy will be provided than could be used and for a portion of the time more energy will be required than could be provided. Jalalzadeh-Azar [6] investigated this situation and presented a comparison of electrical- and thermal-load-following CHP systems. In his investigation he included in a parametric analysis addressing the influence of the subsystem efficiencies on the total primary energy consumption as well as an economic analysis of these systems. He found that an increase in the efficiencies of the on-site power generation and electrical equipment reduced the total monthly import of electricity. A methodology for calculating performance characteristics of different micro-CHP system components will be introduced in this article. Thermodynamic cycles are used to model each individual prime mover. The prime movers modeled in this article are a spark-ignition internal combustion engine (Otto cycle) and a diesel engine (Diesel cycle). Calculations for heat exchanger, absorption chiller, and boiler modeling are also presented. The individual component models are then linked together to calculate total system performance values. Performance characteristics that will be observed for each system include maximum fuel flow rate, total monthly fuel consumption, and system energy (electrical, thermal, and total) efficiencies. Also, whether or not both the required electrical and thermal loads can sufficiently be accounted for within the system specifications is observed. Case study data for various micro-CHP system configurations have been discussed and compared. Comparisons are made of the different prime mover/fuel combinations. Also, micro- CHP monthly energy cost results are compared for each system configuration to conventional monthly utility costs for equivalent monthly building power, heating, and cooling requirements.

Louay Chamra

2008-09-26T23:59:59.000Z

Note: This page contains sample records for the topic "fuel assemblies cooling" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


181

Preliminary Design Study of Medium Sized Gas Cooled Fast Reactor with Natural Uranium as Fuel Cycle Input  

SciTech Connect

In this study a feasibility design study of medium sized (1000 MWt) gas cooled fast reactors which can utilize natural uranium as fuel cycle input has been conducted. Gas Cooled Fast Reactor (GFR) is among six types of Generation IV Nuclear Power Plants. GFR with its hard neuron spectrum is superior for closed fuel cycle, and its ability to be operated in high temperature (850 deg. C) makes various options of utilizations become possible. To obtain the capability of consuming natural uranium as fuel cycle input, modified CANDLE burn-up scheme[1-6] is adopted this GFR system by dividing the core into 10 parts of equal volume axially. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. As an optimization results, a design of 1000 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input is discussed. The average discharge burn-up is about 280 GWd/ton HM. Enough margin for criticality was obtained for this reactor.

Meriyanti; Su'ud, Zaki; Rijal, K. [Nuclear Physics and Biophysics Research Division, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia); Zuhair; Ferhat, A. [National Nuclear Energ Agency of Indonesia (BATAN) (Indonesia); Sekimoto, H. [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology (Japan)

2010-06-22T23:59:59.000Z

182

Self-cooling mono-container fuel cell generators and power plants using an array of such generators  

DOE Patents (OSTI)

A mono-container fuel cell generator (10) contains a layer of interior insulation (14), a layer of exterior insulation (16) and a single housing (20) between the insulation layers, where fuel cells, containing electrodes and electrolyte, are surrounded by the interior insulation (14) in the interior (12) of the generator, and the generator is capable of operating at temperatures over about 650.degree. C., where the combination of interior and exterior insulation layers have the ability to control the temperature in the housing (20) below the degradation temperature of the housing material. The housing can also contain integral cooling ducts, and a plurality of these generators can be positioned next to each other to provide a power block array with interior cooling.

Gillett, James E. (Greensburg, PA); Dederer, Jeffrey T. (Valencia, PA); Zafred, Paolo R. (Pittsburgh, PA)

1998-01-01T23:59:59.000Z

183

Self-cooling mono-container fuel cell generators and power plants using an array of such generators  

DOE Patents (OSTI)

A mono-container fuel cell generator contains a layer of interior insulation, a layer of exterior insulation and a single housing between the insulation layers, where fuel cells, containing electrodes and electrolyte, are surrounded by the interior insulation in the interior of the generator, and the generator is capable of operating at temperatures over about 650 C, where the combination of interior and exterior insulation layers have the ability to control the temperature in the housing below the degradation temperature of the housing material. The housing can also contain integral cooling ducts, and a plurality of these generators can be positioned next to each other to provide a power block array with interior cooling. 7 figs.

Gillett, J.E.; Dederer, J.T.; Zafred, P.R.

1998-05-12T23:59:59.000Z

184

Development of Correlations for Pressure Loss/Drop Coefficients Obtained From Flow Testing of Fuel Assemblies in Framatome ANP's PHTF  

SciTech Connect

Thermal-hydraulic analyses of pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies are generally performed for either assembly thermal-hydraulic design, thermal-hydraulic compatibility evaluation, or cycle licensing thermal-hydraulic characterization. A key issue in all cases is the hydraulic resistance characterization of the assembly in which the assembly, its components and support plates, etc., are represented by their respective pressure loss and pressure drop coefficients. These hydraulic coefficients can be determined by single-phase flow testing in an experimental facility such as the Framatome ANP Portable Hydraulic Test Facility (PHTF) located at Richland Test Facilities (RTF) in Richland, WA. The goal of this paper is to present a uniform and consistent methodology for the development of coefficient correlations from data obtained from single phase pressure drop testing of PWR and BWR fuel assemblies and their components performed in the PHTF. This methodology reflects the years of accumulated experience from an existing facility with an ongoing test program. (authors)

Madni, Imtiaz K.; Stephens, Lance G.; Turner, Dave M. [Framatome ANP Inc. (United States)

2002-07-01T23:59:59.000Z

185

Air-Cooled Stack Freeze Tolerance - DOE Hydrogen and Fuel Cells Program FY 2012 Annual Progress Report  

NLE Websites -- All DOE Office Websites (Extended Search)

6 6 DOE Hydrogen and Fuel Cells Program FY 2012 Annual Progress Report Dave Hancock Plug Power Inc. 968 Albany Shaker Rd Latham, NY 12110 Phone: (518) 782-7700 Email: david_hancock@plugpower.com DOE Managers HQ: Donna Ho Phone: (202) 586-8000 Email: Donna.Ho@ee.doe.gov GO: Reginald Tyler Phone: (720) 356-1805 Email: Reginald.Tyler@go.doe.gov Technical Advisor Walt Podolski Phone: (630) 252-7558 Email: podolski@anl.gov Contract Number: DE-EE0000473 Subcontractor: Ballard Power Systems, Burnaby, British Columbia, Canada Project Start Date: June 1, 2009 Project End Date: November 15, 2011 Fiscal Year (FY) 2012 Objectives Advance the state of the art in technology for air-cooled * proton exchange membrane (PEM) fuel cell stacks and related GenDrive(tm) material handling application fuel

186

Assessment of the impacts of spent fuel disassembly alternatives on the Nuclear Waste Isolation System. [Preparing and packaging spent fuel assemblies for geologic disposal  

SciTech Connect

The objective of this report was to evaluate four possible alternative methods of preparing and packaging spent fuel assemblies for geologic disposal against the Reference Process of unmodified spent fuel. The four alternative processes were: (1) End fitting removal, (2) Fission gas venting and resealing, (3) Fuel bundle disassembly and close packing of fuel pins, and (4) Fuel shearing and immobilization. Systems analysis was used to develop a basis of comparison of the alternatives. Conceptual processes and facility layouts were devised for each of the alternatives, based on technology deemed feasible for the purpose. Assessments were made of 15 principal attributes from the technical, operational, safety/risk, and economic considerations related to each of the alternatives, including both the surface packaging and underground repository operations. Specific attributes of the alternative processes were evaluated by assigning a number for each that expressed its merit relative to the corresponding attribute of the Reference Process. Each alternative process was then ranked by summing the numbers for attributes in each of the four assessment areas and collectively. Fuel bundle disassembly and close packing of fuel pins was ranked the preferred method of disposal of spent fuel. 63 references, 46 figures, 46 tables.

Not Available

1984-07-01T23:59:59.000Z

187

Fuels  

NLE Websites -- All DOE Office Websites (Extended Search)

Goals > Fuels Goals > Fuels XMAT for nuclear fuels XMAT is ideally suited to explore all of the radiation processes experienced by nuclear fuels.The high energy, heavy ion accleration capability (e.g., 250 MeV U) can produce bulk damage deep in the sample, achieving neutron type depths (~10 microns), beyond the range of surface sputtering effects. The APS X-rays are well matched to the ion beams, and are able to probe individual grains at similar penetrations depths. Damage rates to 25 displacements per atom per hour (DPA/hr), and doses >2500 DPA can be achieved. MORE» Fuels in LWRs are subjected to ~1 DPA per day High burn-up fuel can experience >2000 DPA. Traditional reactor tests by neutron irradiation require 3 years in a reactor and 1 year cool down. Conventional accelerators (>1 MeV/ion) are limited to <200-400 DPAs, and

188

Stator and method of assembly  

DOE Patents (OSTI)

The present application provides a stator. The stator may include a number of poles and a stator tip and cooling assembly. The stator tip and cooling assembly may include a number of stator tips with a number of cooling tubes adjacent thereto such that the stator tips align with the poles and the cooling tubes cool the poles.

Alexander, James Pellegrino; El-Refaie, Ayman Mohamed Fawzi; Shen, Xiaochun

2013-06-18T23:59:59.000Z

189

Critical Configuration and Physics Measurements for Assemblies of U(93.15)O2 Fuel Rods (1.506-cm Pitch)  

SciTech Connect

A series of critical experiments were completed from 1962–1965 at Oak Ridge National Laboratory’s (ORNL’s) Critical Experiments Facility (CEF) in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950s, efforts were made to study “power plants for the production of electrical power in space vehicles.”(a) The MPRE program was a part of those efforts and studied the feasibility of a stainless-steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in [fiscal years] 1964, 1965, and 1966. A summary of the program’s effort was compiled in 1967.a The delayed critical experiments were a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods. Initial experiments, performed in November and December of 1962, consisted of a core of unmoderated stainless-steel tubes, each containing 26 UO2 fuel pellets, surrounded by a graphite reflector. Measurements were performed to determine critical reflector arrangements, relative fission-rate distributions, and cadmium ratio distributions. Subsequent experiments used beryllium reflectors and also measured the reactivity for various materials placed in the core. “The [assemblies were built] on [a] vertical assembly machine so that the movable part was the core and bottom reflector” (see Reference 1). The experiment studied in this evaluation was the second of the series and had the fuel rods in a 1.506-cm-triangular pitch. One critical configuration was found (see Reference 3). Once the critical configuration had been achieved, various measurements of reactivity, relative axial and radial activation rates of 235U,bc and cadmium ratios were performed. The cadmium ratio, reactivity, and activation rate measurements performed on the critical configuration are described in Sections 1.3, 1.4, and 1.7, respectively.

Margaret A. Marshall

2013-03-01T23:59:59.000Z

190

Critical Configuration and Physics Mesaurements for Graphite Reflected Assemblies of U(93.15)O2 Fuel Rods (1.27-CM Pitch)  

SciTech Connect

A series of critical experiments were completed in 1962-1965 at Oak Ridge National Laboratory's Critical Experiments Facility in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950's efforts were made to study 'power plants for the production of electrical power in space vehicles'. The MPRE program was a part of those efforts and studied the feasibility of a stainless steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in FY 1964, 1965, and 1966. A summary of the program's effort was compiled in 1967. The delayed critical experiments served as a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods. Initial experiments, performed in November and December of 1962, consisted of a core of unmoderated 253 stainless steel tubes, each containing 26 UO2 fuel pellets, surrounded by a graphite reflector. Measurements were made to determine critical reflector arrangements, fission-rate distributions, and cadmium ratio distributions. Subsequent experiments used beryllium reflectors and also measured the reactivity for various materials placed in the core. 'The [assemblies were built] on [a] vertical assembly machine so that the movable part was the core and bottom reflector.' The experiment studied within this evaluation was the first of the series and had the 253 fuel tubes packed tightly into a 22.87 cm outside diameter (OD) core tank. Two critical configurations were found by varying the amount of graphite reflector (References 1 and 2). Information for this evaluation was compiled from Reference 1 and 2, reports on subsequent experiments in the series, and the experimental logbook as well as from communication with the experimenter, John T. Mihalczo.

Margaret A. Marshall

2011-09-01T23:59:59.000Z

191

REACTOR COOLING  

DOE Patents (OSTI)

A nuclear reactor with provisions for selectively cooling the fuel elements is described. The reactor has a plurality of tubes extending throughout. Cylindrical fuel elements are disposed within the tubes and the coolant flows through the tubes and around the fuel elements. The fuel elements within the central portion of the reactor are provided with roughened surfaces of material. The fuel elements in the end portions of the tubes within the reactor are provlded with low conduction jackets and the fuel elements in the region between the central portion and the end portions are provided with smooth surfaces of high heat conduction material.

Quackenbush, C.F.

1959-09-29T23:59:59.000Z

192

Pulse superimposition calculational methodology for estimating the subcritcality level of nuclear fuel assemblies.  

Science Conference Proceedings (OSTI)

One of the most reliable experimental methods for measuring the subcriticality level of a nuclear fuel assembly is the Sjoestrand method applied to the reaction rate generated from a pulsed neutron source. This study developed a new analytical methodology simulating the Sjoestrand method, which allows comparing the experimental and analytical reaction rates and the obtained subcriticality levels. In this methodology, the reaction rate is calculated due to a single neutron pulse using MCNP/MCNPX computer code or any other neutron transport code that explicitly simulates the delayed fission neutrons. The calculation simulates a single neutron pulse over a long time period until the delayed neutron contribution to the reaction rate is vanished. The obtained reaction rate is then superimposed to itself, with respect to the time, to simulate the repeated pulse operation until the asymptotic level of the reaction rate, set by the delayed neutrons, is achieved. The superimposition of the pulse to itself was calculated by a simple C computer program. A parallel version of the C program is used due to the large amount of data being processed, e.g. by the Message Passing Interface (MPI). The analytical results of this new calculation methodology have shown an excellent agreement with the experimental data available from the YALINA-Booster facility of Belarus. This methodology can be used to calculate Bell and Glasstone spatial correction factor.

Talamo, A.; Gohar, Y.; Rabiti, C.; Aliberti, G.; Kondev, F.; Smith, D.; Zhong, Z.; Kiyavitskaya, H.; Bournos, V; Fokov, Y.; Routkovskaya, C.; Serafimovich, I. (Nuclear Engineering Division); (INL); (Joint Institute for Power and Nuclear Research-Sosny)

2009-05-01T23:59:59.000Z

193

Neutronics and Depletion Methods for Parametric Studies of Fluoride Salt Cooled High Temperature Reactors with Slab Fuel Geometry and Multi-Batch Fuel Management Schemes  

SciTech Connect

The Advanced High Temperature Reactor (AHTR) is a 3400 MWth fluoride salt cooled high temperature reactor (FHR) that uses TRISO particle fuel compacted into slabs rather than spherical fuel pebbles or cylindrical fuel compacts. Simplified methods are required for parametric design studies such that analyzing the entire feasible design space for an AHTR is tractable. These simplifications include fuel homogenization techniques to increase the speed of neutron transport calculations in depletion analysis and equilibrium depletion analysis methods to analyze systems with multi-batch fuel management schemes. This paper presents three elements of significant novelty. First, the reactivity-equivalent physical transformation (RPT) methodology usually applied in systems with coated particle fuel in cylindrical and spherical geometries was extended to slab geometries. Secondly, based on this newly developed RPT method for slab geometries, a methodology that uses Monte Carlo depletion approaches was further developed to search for the maximum discharge burnup in a multi-batch system by iteratively estimating the beginning of equilibrium cycle composition and sampling different discharge burnups. This iterative equilibrium depletion search (IEDS) method fully defines an equilibrium fuel cycle (keff, power, flux and composition evolutions across space and time), but is computationally demanding, although feasible on single-processor workstations. Finally, an analytical method, the non-linear reactivity model, was developed by expanding the linear reactivity model to include an arbitrary number of higher order terms to extrapolate single-batch depletion results to estimate the maximum discharge burnup and BOEC keff in systems with multi-batch fuel management schemes. Results from this method were benchmarked against equilibrium depletion analysis results using the IEDS.

Cisneros, Anselmo T. [University of California, Berkeley; Ilas, Dan [ORNL

2012-01-01T23:59:59.000Z

194

TRUMP-BD: A computer code for the analysis of nuclear fuel assemblies under severe accident conditions  

Science Conference Proceedings (OSTI)

TRUMP-BD (Boil Down) is an extension of the TRUMP (Edwards 1972) computer program for the analysis of nuclear fuel assemblies under severe accident conditions. This extension allows prediction of the heat transfer rates, metal-water oxidation rates, fission product release rates, steam generation and consumption rates, and temperature distributions for nuclear fuel assemblies under core uncovery conditions. The heat transfer processes include conduction in solid structures, convection across fluid-solid boundaries, and radiation between interacting surfaces. Metal-water reaction kinetics are modeled with empirical relationships to predict the oxidation rates of steam-exposed Zircaloy and uranium metal. The metal-water oxidation models are parabolic in form with an Arrhenius temperature dependence. Uranium oxidation begins when fuel cladding failure occurs; Zircaloy oxidation occurs continuously at temperatures above 13000{degree}F when metal and steam are available. From the metal-water reactions, the hydrogen generation rate, total hydrogen release, and temporal and spatial distribution of oxide formations are computed. Consumption of steam from the oxidation reactions and the effect of hydrogen on the coolant properties is modeled for independent coolant flow channels. Fission product release from exposed uranium metal Zircaloy-clad fuel is modeled using empirical time and temperature relationships that consider the release to be subject to oxidation and volitization/diffusion ( bake-out'') release mechanisms. Release of the volatile species of iodine (I), tellurium (Te), cesium (Ce), ruthenium (Ru), strontium (Sr), zirconium (Zr), cerium (Cr), and barium (Ba) from uranium metal fuel may be modeled.

Lombardo, N.J.; Marseille, T.J.; White, M.D.; Lowery, P.S.

1990-06-01T23:59:59.000Z

195

An integrated performance model for high temperature gas cooled reactor coated particle fuel  

E-Print Network (OSTI)

The performance of coated fuel particles is essential for the development and deployment of High Temperature Gas Reactor (HTGR) systems for future power generation. Fuel performance modeling is indispensable for understanding ...

Wang, Jing, 1976-

2004-01-01T23:59:59.000Z

196

Fabrication development of full-sized components for GCFR core assemblies  

Science Conference Proceedings (OSTI)

This paper presents the status of the development of full-sized components for gas-cooled fast reactor (GCFR) core assemblies. Methods for ribbing of the fuel rod cladding, fabrication of grid spacers of two different designs, drawing of assembly flow ducts, and fabrication of fission gas collection manifolds by several methods are discussed.

Lindgren, J.R.; Flynn, P.W.; Foster, L.C.

1980-05-01T23:59:59.000Z

197

Evaluation and Analysis of an Integrated PEM Fuel Cell with Absorption Cooling and Water Heating System for Sustainable Building Operation  

E-Print Network (OSTI)

In this paper, a parametric study of a PEM fuel cell integrated with a double effect absorption system is carried out in order to study the effect of different operating conditions on the efficiency of the PEM fuel cell, utilization factor of the over all system, COPs of the double effect cooling and heating system, and power and heat output of the PEM fuel cell. It is found that the efficiency of the cell decreases, ranging from 46.2% to 24.4% with increase in membrane thickness and current density, and at the same time the COP increases ranging from 0.65 to 1.52. The heat and power output of the fuel cell decreases from 10.54 kW to 5.12 kW, and 9.12 kW to 6.99 kW, respectively for the increase in membrane thickness. However, when the temperature of the cell is increased the heat and power output increases from 5.12 kW to 10.54 kW, and 6.9 kW to 7.02 kW, respectively. The COP is found to be decreasing ranging from 1.53 to 0.33 with the increase in temperature of the cell and heat input to the HTG. As for the utilization factor, it increases ranging from 17% to 87% with increase in the temperature of the cell and heat input to the HTG. This study reveals that an integrated PEM fuel cell with a double effect absorption cooling systems has a very high potential to be an economical and environmental solution as compared with conventional systems of high electricity and natural gas prices which emit lots of harmful gasses and are not that efficient.

Gadalla, M.; Ratlamwala, T.; Dincer, I.

2010-01-01T23:59:59.000Z

198

CRITICAL CONFIGURATION FOR BERYLLIUM REFLECTED ASSEMBLIES OF U(93.15)O2 FUEL RODS (1.506-CM PITCH AND 7-TUBE CLUSTERS)  

Science Conference Proceedings (OSTI)

A series of critical experiments were completed in 1962-1965 at Oak Ridge National Laboratory’s Critical Experiments Facility in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950’s efforts were made to study “power plants for the production of electrical power in space vehicles”. The MPRE program was a part of those efforts and studied the feasibility of a stainless steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in [fiscal years] 1964, 1965, and 1966. A summary of the program’s effort was compiled in 1967. The delayed critical experiments were a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods. Initial experiments, performed in November and December of 1962, consisted of a core of 253 unmoderated stainless steel tubes, each containing 26 UO2 fuel pellets, surrounded by a graphite reflector. Measurements were made to determine critical reflector arrangements, fission-rate distributions, and cadmium ratio distributions. “The [assemblies were built] on [a] vertical assembly machine so that the movable part was the core and bottom reflector.” The first two experiments in the series were evaluated in HEU-COMP-FAST-001 (SCCA-FUND-EXP-001) and HEU-COMP-FAST-002 (SCCA-FUND-EXP-002). The first experiment had the 253 fuel tubes packed tightly into a 22.87 cm outside diameter (OD) core tank (References 1 and 2). The second experiment in the series, performed in early 1963, had the 253 fuel tubes at a 1.506-cm triangular lattice in a 25.96 cm OD core tank and graphite reflectors on all sides. The third set of experiments in the series, performed in mid-1963, which is studied in this evaluation, used beryllium reflectors. The beryllium reflected system was the preferred reactor configuration for this application because of the small thickness of the reflector. The two core configurations had the 253 fuel tubes at a 1.506-cm triangular lattice and arranged in 7-tube clusters. The experiments have been determined to represent acceptable benchmark experiments. Information for this evaluation was compiled from published reports on all three parts of the experimental series (Reference 1-5) and the experimental logbook as well as from communication with the experimenter, John T. Mihalczo.

Margaret A. Marshall

2012-05-01T23:59:59.000Z

199

Fuel cell integral bundle assembly including ceramic open end seal and vertical and horizontal thermal expansion control  

Science Conference Proceedings (OSTI)

A plurality of integral bundle assemblies contain a top portion with an inlet fuel plenum and a bottom portion containing a base support, the base supports a dense, ceramic air exhaust manifold having four supporting legs, the manifold is below and connects to air feed tubes located in a recuperator zone, the air feed tubes passing into the center of inverted, tubular, elongated, hollow electrically connected solid oxide fuel cells having an open end above a combustion zone into which the air feed tubes pass and a closed end near the inlet fuel plenum, where the open end of the fuel cells rest upon and within a separate combination ceramic seal and bundle support contained in a ceramic support casting, where at least one flexible cushion ceramic band seal located between the recuperator and fuel cells protects and controls horizontal thermal expansion, and where the fuel cells operate in the fuel cell mode and where the base support and bottom ceramic air exhaust manifolds carry from 85% to all of the weight of the generator.

Zafred, Paolo R. (Murrysville, PA); Gillett, James E. (Greensburg, PA)

2012-04-24T23:59:59.000Z

200

Neutronics and Depletion Methods for Parametric Studies of Fluoride Salt Cooled High Temperature Reactors with Slab Fuel Geometry and Multi-Batch Fuel Management Schemes  

SciTech Connect

The Advanced High-Temperature Reactor (AHTR) is a 3400 MWth fluoride-salt-cooled high-temperature reactor (FHR) that uses TRISO particle fuel compacted into slabs rather than spherical or cylindrical fuel compacts. Simplified methods are required for parametric design studies such that analyzing the entire feasible design space for an AHTR is tractable. These simplifications include fuel homogenization techniques to increase the speed of neutron transport calculations in depletion analysis and equilibrium depletion analysis methods to analyze systems with multi-batch fuel management schemes. This paper presents three elements of significant novelty. First, the Reactivity-Equivalent Physical Transformation (RPT) methodology usually applied in systems with coated-particle fuel in cylindrical and spherical geometries has been extended to slab geometries. Secondly, based on this newly developed RPT method for slab geometries, a methodology that uses Monte Carlo depletion approaches was further developed to search for the maximum discharge burnup in a multi-batch system by iteratively estimating the beginning of equilibrium cycle (BOEC) composition and sampling different discharge burnups. This Iterative Equilibrium Depletion Search (IEDS) method fully defines an equilibrium fuel cycle (keff, power, flux, and composition evolutions) but is computationally demanding, although feasible on single-processor workstations. Finally, an analytical method, the Non-Linear Reactivity Model, was developed by expanding the linear reactivity model to include an arbitrary number of higher order terms so that single-batch depletion results could be extrapolated to estimate the maximum discharge burnup and BOEC keff in systems with multi-batch fuel management schemes. Results from this method were benchmarked against equilibrium depletion analysis results using the IEDS.

Cisneros, Anselmo T. [University of California, Berkeley; Ilas, Dan [ORNL

2013-01-01T23:59:59.000Z

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201

Liquid metal cooled nuclear reactors with passive cooling system  

SciTech Connect

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of cooling medium flow circuits which cooperate to remove and carry heat away from the fuel core upon loss of the normal cooling flow circuit to areas external thereto.

Hunsbedt, Anstein (Los Gatos, CA); Fanning, Alan W. (San Jose, CA)

1991-01-01T23:59:59.000Z

202

Development of the JAERI (Japan Atomic Energy Research Institute) fuel cleanup system for tests at the Tritium Systems Test Assembly  

Science Conference Proceedings (OSTI)

Tritium Process Laboratory (TPL) at the Japan Atomic Energy Research Institute (JAERI) has developed the Fuel Cleanup System (FCU) which accepts simulated fusion reactor exhaust and produces pure hydrogen isotopes and tritium-free waste. The major components are: a palladium diffuser, a catalytic reactor, cold traps, a ceramic electrolysis cell, and zirconium-cobalt beds. In 1988, an integrated loop of the FCU process was installed in the TPL and a number of hot'' runs were performed to study the system characteristics and improve system performance. Under the US-Japan collaboration program, the JAERI Fuel Cleanup System'' (JFCU) was designed and fabricated by JAERI/TPL for testing at the Tritium Systems Test Assembly (TSTA) in Los Alamos National Laboratory as a major subsystem of the simulated fusion fuel cycle. The JFCU was installed in the TSTA in early 1990.

Konishi, S.; Inoue, M.; Hayashi, T.; Okuno, K.; Naruse, Y. (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)); Barnes, J.W.; Anderson, J.L. (Los Alamos National Lab., NM (USA))

1990-01-01T23:59:59.000Z

203

Bioethanol - the Climate-Cool Fuel: Biofuels For the Global Environment Fact Sheet  

DOE Green Energy (OSTI)

Bioethanol is alcohol fuel made from cellulosic biomass-renewable resources such as trees, grasses, much of the material in municipal solid waste, and forestry and agricultural residues. Compared to the fossil fuels it will displace, bioethanol contributes little or no net CO2 to the earth's atmosphere.

NONE

1997-11-17T23:59:59.000Z

204

Bioethanol - the Climate-Cool Fuel: Biofuels For the Global Environment Fact Sheet  

SciTech Connect

Bioethanol is alcohol fuel made from cellulosic biomass-renewable resources such as trees, grasses, much of the material in municipal solid waste, and forestry and agricultural residues. Compared to the fossil fuels it will displace, bioethanol contributes little or no net CO2 to the earth's atmosphere.

1997-11-17T23:59:59.000Z

205

Cool Roofs and Heat Islands  

NLE Websites -- All DOE Office Websites (Extended Search)

(510) 486-7494 Links Heat Island Group The Cool Colors Project Batteries and Fuel Cells Buildings Energy Efficiency Applications Commercial Buildings Cool Roofs and...

206

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

DOE Patents (OSTI)

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.

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

1982-05-13T23:59:59.000Z

207

CRITICAL CONFIGURATION AND PHYSICS MEASUREMENTS FOR GRAPHITE REFLECTED ASSEMBLIES OF U(93.15)O2 FUEL RODS (1.506-CM PITCH)  

SciTech Connect

A series of critical experiments were completed in 1962-1965 at Oak Ridge National Laboratory’s Critical Experiments Facility in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950’s efforts were made to study “power plants for the production of electrical power in space vehicles”. The MPRE program was a part of those efforts and studied the feasibility of a stainless steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in [fiscal years] 1964, 1965, and 1966. A summary of the program’s effort was compiled in 1967. The delayed critical experiments were a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods. Initial experiments, performed in November and December of 1962, consisted of a core of 253 unmoderated stainless steel tubes, each containing 26 UO2 fuel pellets, surrounded by a graphite reflector. Measurements were made to determine critical reflector arrangements, fission-rate distributions, and cadmium ratio distributions. Subsequent experiments used beryllium reflectors and also measured the reactivity for various materials placed in the core. “The [assemblies were built] on [a] vertical assembly machine so that the movable part was the core and bottom reflector.” The first experiment in the series was evaluated in HEU-COMP-FAST-001. It had the 253 fuel tubes packed tightly into a 22.87 cm outside diameter (OD) core tank (References 1 and 2). The second experiment in the series, performed in early 1963, which is studied in this evaluation, had the 253 fuel tubes at a 1.506-cm triangular lattice in a 25.96 cm OD core tank and graphite reflectors on all sides. The experiment has been determined to represent an acceptable benchmark experiment. Information for this evaluation was compiled from published reports on all three parts of the experimental series (Reference 1-5) and the experimental logbook as well as from communication with the experimenter, John T. Mihalczo.

Margaret A. Marshall

2012-03-01T23:59:59.000Z

208

Fabrication of gas impervious edge seal for a bipolar gas distribution assembly for use in a fuel cell  

DOE Patents (OSTI)

A bipolar gas reactant distribution assembly for use in a fuel cell is disclosed, the assembly having a solid edge seal to prevent leakage of gaseous reactants wherein a pair of porous plates are provided with peripheral slits generally parallel to, and spaced apart from two edges of the plate, the slit being filled with a solid, fusible, gas impervious edge sealing compound. The plates are assembled with opposite faces adjacent one another with a layer of a fusible sealant material therebetween the slits in the individual plates being approximately perpendicular to one another. The plates are bonded to each other by the simultaneous application of heat and pressure to cause a redistribution of the sealant into the pores of the adjacent plate surfaces and to cause the edge sealing compound to flow and impregnate the region of the plates adjacent the slits and comingle with the sealant layer material to form a continuous layer of sealant along the edges of the assembled plates.

Kaufman, Arthur (West Orange, NJ); Werth, John (Princeton, NJ)

1986-01-01T23:59:59.000Z

209

Fossil Fuel and Biomass Burning Effect on Climate—Heating or Cooling?  

Science Conference Proceedings (OSTI)

Emission from burning of fossil fuels and biomass (associated with deforestation) generates a radiative forcing on the atmosphere and a possible climate chaw. Emitted trace gases heat the atmosphere through their greenhouse effect, while ...

Yoram J. Kaufman; Robert S. Fraser; Robert L. Mahoney

1991-06-01T23:59:59.000Z

210

Assessing the Feasibility of Using Neutron Resonance Transmission Analysis (NRTA) for Assaying Plutonium in Spent Fuel Assemblies  

SciTech Connect

Neutron resonance transmission analysis (NRTA) is an active-interrogation nondestructive assay (NDA) technique capable of assaying spent nuclear fuel to determine plutonium content. Prior experimental work has definitively shown the technique capable of assaying plutonium isotope composition in spent-fuel pins to a precision of approximately 3%, with a spatial resolution of a few millimeters. As a Grand Challenge to investigate NDA options for assaying spent fuel assemblies (SFAs) in the commercial fuel cycle, Idaho National Laboratory has explored the feasibility of using NRTA to assay plutonium in a whole SFA. The goal is to achieve a Pu assay precision of 1%. The NRTA technique uses low-energy neutrons from 0.1-40 eV, at the bottom end of the actinide-resonance range, in a time-of-flight arrangement. Isotopic composition is determined by relating absorption of the incident neutrons to the macroscopic cross-section of the actinides of interest in the material, and then using this information to determine the areal density of the isotopes in the SFA. The neutrons used for NRTA are produced using a pulsed, accelerator-based neutron source. Distinguishable resonances exist for both the plutonium (239,240,241,242Pu) and uranium (235,236,238U) isotopes of interest in spent fuel. Additionally, in this energy range resonances exists for six important fission products (99Tc, 103Rh, 131Xe, 133Cs, 145Nd, and 152Sm) which provide additional information to support spent fuel plutonium assay determinations. Based on extensive modeling of the problem using Monte Carlo-based simulation codes, our preliminary results suggest that by rotating an SFA to acquire four symmetric views, sufficient neutron transmission can be achieved to assay a SFA. In this approach multiple scan information for the same pins may also be unfolded to potentially allow the determination of plutonium for sub-regions of the assembly. For a 17 ? 17 pressurized water reactor SFA, a simplistic preliminary analysis indicates the mass of 239Pu may be determined with a precision on the order of 5%, without the need for operator-supplied fuel information or operational histories. This paper will present our work to date on this topic, indicate our preliminary findings for a conceptual assay approach, discuss resilience against spoofing, and outline our future plans for evaluating the NRTA technique for SFA plutonium determination.

D. L. Chichester; J. W. Sterbentz

2012-07-01T23:59:59.000Z

211

ANL-W MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement  

SciTech Connect

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 (EIS). 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. The paper describes the following: Site map and the LA facility; process descriptions; resource needs; employment requirements; wastes, emissions, and exposures; accident analysis; transportation; qualitative decontamination and decommissioning; post-irradiation examination; LA fuel bundle fabrication; LA EIS data report assumptions; and LA EIS data report supplement.

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

1997-08-01T23:59:59.000Z

212

Alternative Liquid Fuel Effects on Cooled Silicon Nitride Marine Gas Turbine Airfoils  

Science Conference Proceedings (OSTI)

With prior support from the Office of Naval Research, DARPA, and U.S. Department of Energy, United Technologies is developing and engine environment testing what we believe to be the first internally cooled silicon nitride ceramic turbine vane in the United States. The vanes are being developed for the FT8, an aeroderivative stationary/marine gas turbine. The current effort resulted in further manufacturing and development and prototyping by two U.S. based gas turbine grade silicon nitride component manufacturers, preliminary development of both alumina, and YTRIA based environmental barrier coatings (EBC's) and testing or ceramic vanes with an EBC coating.

Holowczak, J.

2002-03-01T23:59:59.000Z

213

Neutronic fuel element fabrication  

SciTech Connect

This disclosure describes a method for metallurgically bonding a complete leak-tight enclosure to a matrix-type fuel element penetrated longitudinally by a multiplicity of coolant channels. Coolant tubes containing solid filler pins are disposed in the coolant channels. A leak-tight metal enclosure is then formed about the entire assembly of fuel matrix, coolant tubes and pins. The completely enclosed and sealed assembly is exposed to a high temperature and pressure gas environment to effect a metallurgical bond between all contacting surfaces therein. The ends of the assembly are then machined away to expose the pin ends which are chemically leached from the coolant tubes to leave the coolant tubes with internal coolant passageways. The invention described herein was made in the course of, or under, a contract with the U.S. Atomic Energy Commission. It relates generally to fuel elements for neutronic reactors and more particularly to a method for providing a leak-tight metal enclosure for a high-performance matrix-type fuel element penetrated longitudinally by a multiplicity of coolant tubes. The planned utilization of nuclear energy in high-performance, compact-propulsion and mobile power-generation systems has necessitated the development of fuel elements capable of operating at high power densities. High power densities in turn require fuel elements having high thermal conductivities and good fuel retention capabilities at high temperatures. A metal clad fuel element containing a ceramic phase of fuel intimately mixed with and bonded to a continuous refractory metal matrix has been found to satisfy the above requirements. Metal coolant tubes penetrate the matrix to afford internal cooling to the fuel element while providing positive fuel retention and containment of fission products generated within the fuel matrix. Metal header plates are bonded to the coolant tubes at each end of the fuel element and a metal cladding or can completes the fuel-matrix enclosure by encompassing the sides of the fuel element between the header plates.

Korton, George (Cincinnati, OH)

2004-02-24T23:59:59.000Z

214

Failed fuel monitoring and surveillance techniques for liquid metal cooled fast reactors  

Science Conference Proceedings (OSTI)

The Experimental Breeder Reactor II (EBR-II) has been used as a facility for irradiation of LMR fuels and components for thirty years. During this time many tests of experimental fuel were continued to cladding breach in order to study modes of element failure; the methods used to identify such failures are described in a parallel paper. This paper summarizes experience of monitoring the delayed-neutron (DN) and fission-gas (FG) release behavior of a smaller number of elements that continued operation in the run-beyond-cladding-breach (RBCB) mode. The scope of RBCB testing, the methods developed to characterize failures on-line, and examples of DN/FG behavior are described.

Lambert, J.D.B.; Mikaili, R.; Gross, K.C.; Strain, R.V. [Argonne National Lab., IL (United States); Aoyama, T.; Ukai, S.; Nomura, S.; Nakae, N. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan)

1995-05-01T23:59:59.000Z

215

Operator decision aid for breached fuel operation in liquid metal cooled nuclear reactors  

SciTech Connect

The purpose of this paper is to report the development of an expert system that provides continuous assessment of the safety significance and technical specification conformance of Delayed Neutron (DN) signals during breached fuel operation. The completed expert system has been parallelized on an innovative distributed-memory network-computing system that enables the computationally intensive kernel of the expert system to run in parallel on a group of low-cost Unix workstations. 1 ref.

Gross, K.C.; Hawkins, R.E.; Nickless, W.K.

1991-01-01T23:59:59.000Z

216

Effects of a potential drop of a shipping cask, a waste container, and a bare fuel assembly during waste-handling operations; Yucca Mountain Site Characterization Project  

Science Conference Proceedings (OSTI)

This study investigates the effects of potential drops of a typical shipping cask, waste container, and bare fuel assembly during waste-handling operations at the prospective Yucca Mountain Repository. The waste-handling process (one stage, no consolidation configuration) is examined to estimate the maximum loads imposed on typical casks and containers as they are handled by various pieces of equipment during waste-handling operations. Maximum potential drop heights for casks and containers are also evaluated for different operations. A nonlinear finite-element model is employed to represent a hybrid spent fuel container subject to drop heights of up to 30 ft onto a reinforced concrete floor. The impact stress, strain, and deformation are calculated, and compared to the failure criteria to estimate the limiting (maximum permissible) drop height for the waste container. A typical Westinghouse 17 {times} 17 PWR fuel assembly is analyzed by a simplified model to estimate the energy absorption by various parts of the fuel assembly during a 30 ft drop, and to determine the amount of kinetic energy in a fuel pin at impact. A nonlinear finite-element analysis of an individual fuel pin is also performed to estimate the amount of fuel pellet fracture due to impact. This work was completed on May 1990.

Wu, C.L.; Lee, J.; Lu, D.L.; Jardine, L.J. [Bechtel National, Inc., San Francisco, CA (United States)

1991-12-01T23:59:59.000Z

217

PEM fuel cellstack development based on membrane-electrode assemblies of ultra-low platinum loadings  

DOE Green Energy (OSTI)

Attempt is made to scale-up single cell technology, based on ultra-low platinum loadings, to develop a polymer electrolyte membrane fuel cell stack for stationary power generation.

Zawodzinski, C.; Wilson, M.S.; Gottesfeld, S.

1995-09-01T23:59:59.000Z

218

Criticality safety considerations in the geologic disposal of spent nuclear fuel assemblies  

SciTech Connect

Features of geologic disposal which hamper the demonstration that criticality cannot occur therein include possible changes of shape and form, intrusion of water as a neutron moderator, and selective leaching of spent fuel constituents. If the criticality safety of spent fuel disposal depends on burnup, independent measurements verifying the burnup should be performed prior to disposal. The status of nondestructive analysis method which might provide such verification is discussed. Calculations were performed to assess the potential for increasing the allowed size of a spent fuel disposal canister if potential water intrusion were limited by close-packing the enclosed rods. Several factors were identified which severely limited the potential of this application. The theoretical limit of hexagonal close-packing cannot be achieved due to fuel rod bowing. It is concluded that disposal canisters should be sized on the basis of assumed optimum moderation. Several topics for additional research were identified during this limited study.

Gore, B.F.; McNair, G.W.; Heaberlin, S.W.

1980-05-01T23:59:59.000Z

219

A Feasibility Study to Determine Cooling Time and Burnup of ATR Fuel Using a Nondestructive Technique and Three Types of Gamma-ray Detectors  

SciTech Connect

A Feasibility Study to Determine Cooling Time and Burnup of ATR Fuel Using a Nondestructive Technique1 Rahmat Aryaeinejad, Jorge Navarro, and David W Nigg Idaho National Laboratory Abstract Effective and efficient Advanced Test Reactor (ATR) fuel management require state of the art core modeling tools. These new tools will need isotopic and burnup validation data before they are put into production. To create isotopic, burn up validation libraries and to determine the setup for permanent fuel scanner system a feasibility study was perform. The study consisted in measuring short and long cooling time fuel elements at the ATR canal. Three gamma spectroscopy detectors (HPGe, LaBr3, and HPXe) and two system configurations (above and under water) were used in the feasibility study. The first stage of the study was to investigate which detector and system configuration would be better suited for different scenarios. The second stage of the feasibility study was to create burnup and cooling time calibrations using experimental isotopic data collected and ORIGEN 2.2 burnup data. The results of the study establish that a better spectra resolution is achieve with an above the water configuration and that three detectors can be used in the permanent fuel scanner system for different situations. In addition it was conclude that a number of isotopic ratios and absolute measurements could be used to predict ATR fuel burnup and cooling times. 1This work was supported by the U.S. Depart¬ment of Energy (DOE) under Battelle Energy Alliance, LLC Contract No. DE-AC07-05ID14517.

Jorge Navarro; Rahmat Aryaeinejad,; David W. Nigg

2011-05-01T23:59:59.000Z

220

The design and evaluation of a water delivery system for evaporative cooling of a proton exchange membrane fuel cell  

E-Print Network (OSTI)

An investigation was performed to demonstrate system design for the delivery of water required for evaporative cooling of a proton exchange membrane fuel cell (PEMFC). The water delivery system uses spray nozzles capable of injecting water directly and uniformly to the nickel metal foam flow-field (element for distributing the reactant gases over the surface of the electrodes) on the anode side from which water can migrate to the cathode side of the cell via electroosmotic drag. For an effective overall cooling, water distribution over the surface of the nickel foam has to be uniform to avoid creation of hotspots within the cell. A prototype PEMFC structure was constructed modeled after a 35 kW electrical output PEMFC stack. Water was sprayed on the nickel metal foam flow-field using two types of nozzle spray, giving conical fog type flow and flat fan type flow. A detailed investigation of the distribution pattern of water over the surface of the nickel metal flow field was conducted. The motive behind the investigation was to determine if design parameters such as type of water flow from nozzles, vertical location of the water nozzles above the flowfield, area of the nozzles, or operating variables such as reactant gas flow had any effect on water distribution over the surface of the Ni-metal foam flow field. It was found that the design parameters (types of flow, area and location of the nozzle) had a direct impact on the distribution of water in the nickel metal foam. However, the operating variable, reactant gas flow, showed no effect on the water distribution pattern in the Ni-foam.

Al-Asad, Dawood Khaled Abdullah

2006-08-01T23:59:59.000Z

Note: This page contains sample records for the topic "fuel assemblies cooling" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


221

Tritium experiments on components for fusion fuel processing at the Tritium Systems Test Assembly  

SciTech Connect

Under a collaborative agreement between US and Japan, two tritium processing components, a palladium diffuser and a ceramic electrolysis cell have been tested with tritium for application to a Fuel Cleanup System (FCU) for plasma exhaust processing at the Los Alamos National Laboratory. The fundamental characteristics, compatibility with tritium, impurities effects with tritium, and long-term behavior of the components, were studied over a three year period. Based on these studies, an integrated process loop, JAERI Fuel Cleanup System'' equipped with above components was installed at the TSTA for full scale demonstration of the plasma exhaust reprocessing.

Konishi, S.; Yoshida, H.; Naruse, Y. (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)); Carlson, R.V.; Binning, K.E.; Bartlit, J.R.; Anderson, J.L. (Los Alamos National Lab., NM (USA))

1990-01-01T23:59:59.000Z

222

Membrane module assembly  

DOE Patents (OSTI)

A membrane module assembly is described which is adapted to provide a flow path for the incoming feed stream that forces it into prolonged heat-exchanging contact with a heating or cooling mechanism. Membrane separation processes employing the module assembly are also disclosed. The assembly is particularly useful for gas separation or pervaporation. 2 figures.

Kaschemekat, J.

1994-03-15T23:59:59.000Z

223

Fuel  

E-Print Network (OSTI)

heavy-water-moderated, light-water-moderated and liquid-metal cooled fast breeder reactors fueled with natural or low-enriched uranium and containing thorium mixed with the uranium or in separate target channels. U-232 decays with a 69-year half-life through 1.9-year half-life Th-228 to Tl-208, which emits a 2.6 MeV gamma ray upon decay. We find that pressurized light-water-reactors fueled with LEU-thorium fuel at high burnup (70 MWd/kg) produce U-233 with U-232 contamination levels of about 0.4 percent. At this contamination level, a 5 kg sphere of U-233 would produce a gammaray dose rate of 13 and 38 rem/hr at 1 meter one and ten years after chemical purification respectively. The associated plutonium contains 7.5 percent of the undesirable heat-generating 88-year half-life isotope Pu-238. However, just as it is possible to produce weapon-grade plutonium in low-burnup fuel, it is also practical to use heavy-water reactors to produce U-233 containing only a few ppm of U-232 if the thorium is segregated in “target ” channels and discharged a few times more frequently than the natural-uranium “driver ” fuel. The dose rate from a 5-kg solid sphere of U-233 containing 5 ppm U-232 could be reduced by a further factor of 30, to about 2 mrem/hr, with a close-fitting lead sphere weighing about 100 kg. Thus the proliferation resistance of thorium fuel cycles depends very much upon how they are implemented. The original version of this manuscript was received by Science & Global Security on

Jungmin Kang A

2001-01-01T23:59:59.000Z

224

Vehicle Cooling Systems - Energy Innovation Portal  

Hydrogen and Fuel Cell; Hydropower, Wave and ... The cabin cooling system includes at least one fan to draw the hot air into the cooling duct at a ...

225

Inlet nozzle assembly  

DOE Patents (OSTI)

An inlet nozzle assembly for directing coolant into the duct tube of a fuel assembly attached thereto. The nozzle assembly includes a shell for housing separable components including an orifice plate assembly, a neutron shield block, a neutron shield plug, and a diffuser block. The orifice plate assembly includes a plurality of stacked plates of differently configurated and sized openings for directing coolant therethrough in a predesigned flow pattern.

Christiansen, D.W.; Karnesky, R.A.; Knight, R.C.; Precechtel, D.R.; Smith, B.G.

1985-09-09T23:59:59.000Z

226

Inlet nozzle assembly  

DOE Patents (OSTI)

An inlet nozzle assembly for directing coolant into the duct tube of a fuel assembly attached thereto. The nozzle assembly includes a shell for housing separable components including an orifice plate assembly, a neutron shield block, a neutron shield plug, and a diffuser block. The orifice plate assembly includes a plurality of stacked plates of differently configurated and sized openings for directing coolant therethrough in a predesigned flow pattern.

Christiansen, David W. (Kennewick, WA); Karnesky, Richard A. (Richland, WA); Precechtel, Donald R. (Richland, WA); Smith, Bob G. (Richland, WA); Knight, Ronald C. (Richland, WA)

1987-01-01T23:59:59.000Z

227

Assessment of Startup Fuel Options for the GNEP Advanced Burner Reactor (ABR)  

SciTech Connect

The Global Nuclear Energy Program (GNEP) includes a program element for the development and construction of an advanced sodium cooled fast reactor to demonstrate the burning (transmutation) of significant quantities of minor actinides obtained from a separations process and fabricated into a transuranic bearing fuel assembly. To demonstrate and qualify transuranic (TRU) fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype is needed. The ABR would necessarily be started up using conventional metal alloy or oxide (U or U, Pu) fuel. Startup fuel is needed for the ABR for the first 2 to 4 core loads of fuel in the ABR. Following start up, a series of advanced TRU bearing fuel assemblies will be irradiated in qualification lead test assemblies in the ABR. There are multiple options for this startup fuel. This report provides a description of the possible startup fuel options as well as possible fabrication alternatives available to the program in the current domestic and international facilities and infrastructure.

Jon Carmack (062056); Kemal O. Pasamehmetoglu (103171); David Alberstein

2008-02-01T23:59:59.000Z

228

Experimental and Analytical Simulation of MFCI (Molten Fuel Coolant Interaction) during CDA (Core Disruptive Accident) in Sodium Cooled Fast Reactor.  

E-Print Network (OSTI)

??With increasing demand for understanding Severe Accident Scenario in Sodium Cooled Fast Reactors, there is an urgent need of enhancing numerical and experimental simulation techniques.… (more)

Natarajan, Venkataraman

2011-01-01T23:59:59.000Z

229

The feasibility study of small long-life gas cooled fast reactor with mixed natural Uranium/Thorium as fuel cycle input  

Science Conference Proceedings (OSTI)

A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE burn-up scheme has been performed. In this study, design GCFR with Helium coolant which can be continuously operated by supplying mixed Natural Uranium/Thorium without fuel enrichment plant or fuel reprocessing plant. The active reactor cores are divided into two region, Thorium fuel region and Uranium fuel region. Each fuel core regions are subdivided into ten parts (region-1 until region-10) with the same volume in the axial direction. The fresh Natural Uranium and Thorium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh natural Uranium/Thorium fuel. This concept is basically applied to all regions in both cores area, i.e. shifted the core of i{sup th} region into i+1 region after the end of 10 years burn-up cycle. For the next cycles, we will add only Natural Uranium and Thorium on each region-1. The calculation results show the reactivity reached by mixed Natural Uranium/Thorium with volume ratio is 4.7:1. This reactor can results power thermal 550 MWth. After reactor start-up the operation, furthermore reactor only needs Natural Uranium/Thorium supply for continue operation along 100 years.

Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Khairurrijal,; Monado, Fiber; Sekimoto, Hiroshi [Department of Physics Bandung Institute of Technology Jl. Ganesha 10, Bandung 40134, Physics Department, Sriwijaya University, Kampus Indralaya, Ogan Ilir, Sumatera Selatan (Indonesia); Department of Physics Bandung Institute of Technology Jl. Ganesha 10, Bandung 40134 (Indonesia); Department of Physics Bandung Institute of Technology Jl. Ganesha 10, Bandung 40134, Physics Department, Sriwijaya University, Kampus Indralaya, Ogan Ilir, Sumatera Selatan (Indonesia); Reserach of Laboratory for Nuclear Reactors, Tokyo Institute of Technology O-okayama, Meguro-ku, Tokyo 152 (Japan)

2012-06-06T23:59:59.000Z

230

Evaluation of the Sub-Channel Code COBRA-TF for Prediction of BWR Fuel Assembly Void Fraction Distribution  

SciTech Connect

Good quality experimental data is needed to refine the thermal hydraulic models for the prediction of rod bundle void distribution and critical heat flux (CHF) or dry-out. The Nuclear Power Engineering Corporation (NUPEC) has provided a valuable database to evaluate the thermal hydraulic codes [1]. Part of this database was selected for the NUPEC BWR Full-size Fine-Mesh Bundle Tests (BFBT) benchmark sponsored by US NRC, METI-Japan, NEA/OECD and Nuclear Engineering Program of the Pennsylvania State University (PSU). Twenty-five organizations from ten countries have confirmed their intention to participate and will provide code predictions to be compared to the measured data for a series of defined exercises within the framework of the BFBT benchmark. This benchmark data includes both the fine-mesh high quality sub-channel void fraction and critical power data. Using a full BWR rod bundle test facility, the void distribution was measured at mesh sizes smaller than the sub-channel by using a state-of-the-art computer tomography (CT) technology [1]. Experiments were performed for different pressures, flow rates, exit qualities, inlet sub-cooling, power distributions, spacer types and assembly designs. There are microscopic and sub-channel averaged void fraction data from the CT scanner at the bundle exit as well as X-ray densitometer void distribution data at different elevation levels in the rod bundle. Each sub-channel's loss coefficient was calculated with using the Rehme method [2,3], and a COBRA-TF sub-channel model was developed for the NUPEC facility. The BWR assembly that was modeled with COBRA-TF includes two water rods at the center. The predicted sub-channel void fraction values from COBRA-TF are compared with the bundle exit void fraction values measured using the CT-scanner void fraction from the BFBT benchmark data. Different plots are used to examine the code prediction of the void distribution at a sub-channel level for the different sub-channels within the bundle. (authors)

Aydogan, Fatih; Hochreiter, Lawrence E.; Ivanov, Kostadin [The Pennsylvania State University, 302 Walker Building, University Park, PA 16802 (United States); Rhee, Gene [U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001 (United States); Sartori, Enrico [OECD/NEA, Le Seine St. Germain, 12 boulevard des Iles, 92130 Issy-les-Moulineaux (France); Utsuno, Hideaki [Japan Nuclear Energy Safety Organization (Japan)

2006-07-01T23:59:59.000Z

231

Benchmarking of the MIT High Temperature Gas-cooled Reactor TRISO-coated particle fuel performance model  

E-Print Network (OSTI)

MIT has developed a Coated Particle Fuel Performance Model to study the behavior of TRISO nuclear fuels. The code, TIMCOAT, is designed to assess the mechanical and chemical condition of populations of coated particles and ...

Stawicki, Michael A

2006-01-01T23:59:59.000Z

232

Power electronics cooling apparatus  

DOE Patents (OSTI)

A semiconductor cooling arrangement wherein a semiconductor is affixed to a thermally and electrically conducting carrier such as by brazing. The coefficient of thermal expansion of the semiconductor and carrier are closely matched to one another so that during operation they will not be overstressed mechanically due to thermal cycling. Electrical connection is made to the semiconductor and carrier, and a porous metal heat exchanger is thermally connected to the carrier. The heat exchanger is positioned within an electrically insulating cooling assembly having cooling oil flowing therethrough. The arrangement is particularly well adapted for the cooling of high power switching elements in a power bridge.

Sanger, Philip Albert (Monroeville, PA); Lindberg, Frank A. (Baltimore, MD); Garcen, Walter (Glen Burnie, MD)

2000-01-01T23:59:59.000Z

233

Materials testing and development of functionally graded composite fuel cladding and piping for the Lead-Bismuth cooled nuclear reactor  

E-Print Network (OSTI)

This study has extended the development of an exciting technology which promises to enable the Pb-Bi eutectic cooled reactors to operate at temperatures up to 650-700°C. This new technology is a functionally graded composite ...

Fray, Elliott Shepard

2013-01-01T23:59:59.000Z

234

Low thermal resistance power module assembly - Energy Innovation ...  

A power module assembly (400) with low thermal resistance and enhanced heat dissipation to a cooling medium. The assembly includes a heat sink or spreader plate (410 ...

235

Impacts Associated with Transfer of Spent Nuclear Fuel from Spent Fuel Storage Pools to Dry Storage After Five Years of Cooling, Revision 1  

Science Conference Proceedings (OSTI)

In 2010, EPRI performed a study of the accelerated transfer of spent fuel from pools to dry storage in response to the threat of terrorist activities at nuclear power plants (report 1021049). Following the March 2011 Great East Japan Earthquake and the subsequent accident at the Fukushima Daiichi nuclear power plant, some organizations issued a renewed call for accelerated transfer of used fuel from spent fuel ...

2012-08-31T23:59:59.000Z

236

Energy Efficiency Improvement and Cost Saving Opportunities for the Vehicle Assembly Industry: An ENERGY STAR Guide for Energy and Plant Managers  

E-Print Network (OSTI)

with absorption cooling District heating Alternative fuelswith absorption cooling District heating Alternative fuels

Galitsky, Christina

2008-01-01T23:59:59.000Z

237

ZPR-6 assembly 7 high {sup 240} PU core : a cylindrical assemby with mixed (PU, U)-oxide fuel and a central high {sup 240} PU zone.  

Science Conference Proceedings (OSTI)

Over a period of 30 years more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited to form the basis for criticality safety benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. The term 'benchmark' in a ZPR program connotes a particularly simple loading aimed at gaining basic reactor physics insight, as opposed to studying a reactor design. In fact, the ZPR-6/7 Benchmark Assembly (Reference 1) had a very simple core unit cell assembled from plates of depleted uranium, sodium, iron oxide, U3O8, and plutonium. The ZPR-6/7 core cell-average composition is typical of the interior region of liquid-metal fast breeder reactors (LMFBRs) of the era. It was one part of the Demonstration Reactor Benchmark Program,a which provided integral experiments characterizing the important features of demonstration-size LMFBRs. As a benchmark, ZPR-6/7 was devoid of many 'real' reactor features, such as simulated control rods and multiple enrichment zones, in its reference form. Those kinds of features were investigated experimentally in variants of the reference ZPR-6/7 or in other critical assemblies in the Demonstration Reactor Benchmark Program.

Lell, R. M.; Schaefer, R. W.; McKnight, R. D.; Tsiboulia, A.; Rozhikhin, Y.; Nuclear Engineering Division; Inst. of Physics and Power Engineering

2007-10-01T23:59:59.000Z

238

Fuel cell with metal screen flow-field  

DOE Patents (OSTI)

A polymer electrolyte membrane (PEM) fuel cell is provided with electrodes supplied with a reactant on each side of a catalyzed membrane assembly (CMA). The fuel cell includes a metal mesh defining a rectangular flow-field pattern having an inlet at a first corner and an outlet at a second corner located on a diagonal from the first corner, wherein all flow paths from the inlet to the outlet through the square flow field pattern are equivalent to uniformly distribute the reactant over the CMA. In a preferred form of metal mesh, a square weave screen forms the flow-field pattern. In a particular characterization of the present invention, a bipolar plate electrically connects adjacent fuel cells, where the bipolar plate includes a thin metal foil having an anode side and a cathode side; a first metal mesh on the anode side of the thin metal foil; and a second metal mesh on the cathode side of the thin metal foil. In another characterization of the present invention, a cooling plate assembly cools adjacent fuel cells, where the cooling plate assembly includes an anode electrode and a cathode electrode formed of thin conducting foils; and a metal mesh flow field there between for distributing cooling water flow over the electrodes to remove heat generated by the fuel cells. 11 figs.

Wilson, M.S.; Zawodzinski, C.

1998-08-25T23:59:59.000Z

239

Safety and core design of large liquid-metal cooled fast breeder reactors  

E-Print Network (OSTI)

with burnup of a depleted-uranium fueled sodium-cooled B&Bwith burnup of a depleted-uranium fueled sodium-cooled B&Bbalance integral of a depleted-uranium fueled sodium-cooled

Qvist, Staffan Alexander

2013-01-01T23:59:59.000Z

240

Fuel cell cooler-humidifier plate  

DOE Patents (OSTI)

A cooler-humidifier plate for use in a proton exchange membrane (PEM) fuel cell stack assembly is provided. The cooler-humidifier plate combines functions of cooling and humidification within the fuel cell stack assembly, thereby providing a more compact structure, simpler manifolding, and reduced reject heat from the fuel cell. Coolant on the cooler side of the plate removes heat generated within the fuel cell assembly. Heat is also removed by the humidifier side of the plate for use in evaporating the humidification water. On the humidifier side of the plate, evaporating water humidifies reactant gas flowing over a moistened wick. After exiting the humidifier side of the plate, humidified reactant gas provides needed moisture to the proton exchange membranes used in the fuel cell stack assembly. The invention also provides a fuel cell plate that maximizes structural support within the fuel cell by ensuring that the ribs that form the boundaries of channels on one side of the plate have ends at locations that substantially correspond to the locations of ribs on the opposite side of the plate.

Vitale, Nicholas G. (Albany, NY); Jones, Daniel O. (Glenville, NY)

2000-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "fuel assemblies cooling" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


241

RETORT ASSEMBLY  

DOE Patents (OSTI)

An improved retort assembly useful in the thermal reduction of volatilizable metals such as magnesium and calcium is described. In this process a high vacuum is maintained in the retort, however the retort must be heated to very high temperatures while at the same time the unloading end must bo cooled to condense the metal vapors, therefore the retention of the vacuum is frequently difficult due to the thermal stresses involved. This apparatus provides an extended condenser sleeve enclosed by the retort cover which forms the vacuum seal. Therefore, the seal is cooled by the fluid in the condenser sleeve and the extreme thermal stresses found in previous designs together with the deterioration of the sealing gasket caused by the high temperatures are avoided.

Loomis, C.C.; Ash, W.J.

1957-11-26T23:59:59.000Z

242

Advanced water-cooled phosphoric acid fuel cell development. Quarterly technical progress report No. 23, January, February, March, 1990  

DOE Green Energy (OSTI)

Fabrication of repeat parts for the small area short stack is nearing completion and assembly activities are being initiated. Electrolyte reservoir plates (ERPs) were completed and processed into integral separator plates, and acid fill of parts was initiated. Fabrication of electrodes was also completed, including catalyzation and applications of seals and matrices.

Not Available

1990-12-31T23:59:59.000Z

243

A strategy for transition from a uranium fueled, open cycle SFR to a transuranic fueled, closed cycle sodium cooled fast reactor  

E-Print Network (OSTI)

Reactors utilizing a highly energetic neutron spectrum, often termed fast reactors, offer large fuel utilization improvements over the thermal reactors currently used for nuclear energy generation. Conventional fast reactor ...

Richard, Joshua (Joshua Glenn)

2012-01-01T23:59:59.000Z

244

Conceptual design for a re-entrant type fuel channel for supercritical water-cooled nuclear reactors.  

E-Print Network (OSTI)

??Current CANDU-type nuclear reactors use a once-through fuel-channel with an annulus gas insulating it from the moderator. The current reference design for a CANDU-type SuperCritical… (more)

Samuel, Jeffrey

2011-01-01T23:59:59.000Z

245

ZPR-6 assembly 7 high {sup 240}Pu core experiments : a fast reactor core with mixed (Pu,U)-oxide fuel and a centeral high{sup 240}Pu zone.  

Science Conference Proceedings (OSTI)

ZPR-6 Assembly 7 (ZPR-6/7) encompasses a series of experiments performed at the ZPR-6 facility at Argonne National Laboratory in 1970 and 1971 as part of the Demonstration Reactor Benchmark Program (Reference 1). Assembly 7 simulated a large sodium-cooled LMFBR with mixed oxide fuel, depleted uranium radial and axial blankets, and a core H/D near unity. ZPR-6/7 was designed to test fast reactor physics data and methods, so configurations in the Assembly 7 program were as simple as possible in terms of geometry and composition. ZPR-6/7 had a very uniform core assembled from small plates of depleted uranium, sodium, iron oxide, U{sub 3}O{sub 8} and Pu-U-Mo alloy loaded into stainless steel drawers. The steel drawers were placed in square stainless steel tubes in the two halves of a split table machine. ZPR-6/7 had a simple, symmetric core unit cell whose neutronic characteristics were dominated by plutonium and {sup 238}U. The core was surrounded by thick radial and axial regions of depleted uranium to simulate radial and axial blankets and to isolate the core from the surrounding room. The ZPR-6/7 program encompassed 139 separate core loadings which include the initial approach to critical and all subsequent core loading changes required to perform specific experiments and measurements. In this context a loading refers to a particular configuration of fueled drawers, radial blanket drawers and experimental equipment (if present) in the matrix of steel tubes. Two principal core configurations were established. The uniform core (Loadings 1-84) had a relatively uniform core composition. The high {sup 240}Pu core (Loadings 85-139) was a variant on the uniform core. The plutonium in the Pu-U-Mo fuel plates in the uniform core contains 11% {sup 240}Pu. In the high {sup 240}Pu core, all Pu-U-Mo plates in the inner core region (central 61 matrix locations per half of the split table machine) were replaced by Pu-U-Mo plates containing 27% {sup 240}Pu in the plutonium component to construct a central core zone with a composition closer to that in an LMFBR core with high burnup. The high {sup 240}Pu configuration was constructed for two reasons. First, the composition of the high {sup 240}Pu zone more closely matched the composition of LMFBR cores anticipated in design work in 1970. Second, comparison of measurements in the ZPR-6/7 uniform core with corresponding measurements in the high {sup 240}Pu zone provided an assessment of some of the effects of long-term {sup 240}Pu buildup in LMFBR cores. The uniform core version of ZPR-6/7 is evaluated in ZPR-LMFR-EXP-001. This document only addresses measurements in the high {sup 240}Pu core version of ZPR-6/7. Many types of measurements were performed as part of the ZPR-6/7 program. Measurements of criticality, sodium void worth, control rod worth and reaction rate distributions in the high {sup 240}Pu core configuration are evaluated here. For each category of measurements, the uncertainties are evaluated, and benchmark model data are provided.

Lell, R. M.; Morman, J. A.; Schaefer, R.W.; McKnight, R.D.; Nuclear Engineering Division

2009-02-23T23:59:59.000Z

246

A technical review of non-destructive assay research for the characterization of spent nuclear fuel assemblies being conducted under the US DOE NGSI  

SciTech Connect

There is a growing belief that expansion of nuclear energy generation will be needed in the coming decades as part of a mixed supply chain to meet global energy demand. At stake is the health of the economic engine that delivers human prosperity. As a consequence renewed interest is being paid to the safe management of spent nuclear fuel (SNF) and the plutonium it contains. In addition to being an economically valuable resource because it can be used to construct explosive devices, Pu must be placed on an inventory and handled securely. A multiinstitutional team of diverse specialists has been assembled under a project funded by the US Department of Energy (DOE) Next Generation Safeguards Initiative (NGSI) to address ways to nondestructively quantify the plutonium content of spent nuclear fuel assemblies, and to also detect the potential diversion of pins from those assemblies. Studies are underway using mostly Monte Carlo tools to assess the feasibility, individual and collective performance capability of some fourteen nondestructive assay methods. Some of the methods are familiar but are being applied in a new way against a challenging target which is being represented with a higher degree of realism in simulation space than has been done before, while other methods are novel. In this work we provide a brief review of the techniques being studied and highlight the main achievements to date. We also draw attention to the deficiencies identified in for example modeling capability and available basic nuclear data. We conclude that this is an exciting time to be working in the NDA field and that much work, both fundamental and applied, remains ahead if we are to advance the state of the practice to meet the challenges posed to domestic and international safeguards by the expansion of nuclear energy together with the emergence of alternative fuel cycles.

Croft, Stephen [Los Alamos National Laboratory; Tobin, Stephen J [Los Alamos National Laboratory

2010-12-06T23:59:59.000Z

247

Progress in the R and D Project on Oxide Dispersion Strengthened and Precipitation Hardened Ferritic Steels for Sodium Cooled Fast Breeder Reactor Fuels  

Science Conference Proceedings (OSTI)

High burnup capability of sodium cooled fast breeder reactor (SFR) fuels depends significantly on irradiation performance of their component materials. Japan Atomic Energy Agency (JAEA) has been developing oxide dispersion strengthened (ODS) ferritic steels and a precipitation hardened (PH) ferritic steel as the most prospective materials for fuel pin cladding and duct tubes, respectively. Technology for small-scale manufacturing is already established, and several hundreds of ODS steel cladding tubes and dozens of PH steel duct tubes were successfully produced. We will step forward to develop manufacturing technology for mass production to supply these steels for future SFR fuels. Mechanical properties of the products were examined by out-of-pile and in-pile tests including material irradiation tests in the experimental fast reactor JOYO and foreign fast reactors. The material strength standards (MSSs) were tentatively compiled in 2005 for ODS steels and in 1993 for PH steel. In order to upgrade the MSSs and to demonstrate high burnup capability of the materials, we will perform a series of irradiation tests in BOR-60 and JOYO until 2015 and contribute to design study for a demonstration SFR of which operation is expected after 2025. (authors)

Kaito, Takeji; Ohtsuka, Satoshi; Inoue, Masaki [Advanced Nuclear System Research and Development Directorate, Japan Atomic Energy Agency - JAEA, 4002 Narita-cho, Oarai-machi, Higashi-ibaraki-gun, Ibaraki-ken, Zip code 311-1393 (Japan)

2007-07-01T23:59:59.000Z

248

THE ORNL GCR-3, A 750-Mw(e) GAS-COOLED CLAD-FUEL REACTOR POWER PLANT. A JOINT DESIGN STUDY  

SciTech Connect

ABS>An advanced, gas-cooled, clad-fuel reactor power plant to generate 750 Mw of electricity was designed as a study of the potential capability of that system. The graphitemoderated reactor generates 1908 Mw of heat in 1062 fuel channels 21 ft long for a power density of 5.5 kw/liter. Gas temperatures entering and leaving the reactor are 574 and 1150 deg F, respectively, operating at 420 psia. Steam at 2415 psia and 950 deg F with reheat to 1000 deg F drives a 763-Mw(e) turbogenerator and also four 31,000-hp blower drive turbines and the boiler feed pumps. Net thermal efficiency of the plant is 39.4%. Estimated direct cost of construction is 0,267,000, or 7 per kilowatt net electric output. Fuel-cycle costs at 20,000 Mwd per metric ton of uranium are 1.46 mills/ kwhr, operating and maintenance costs are 0.39 mill, and fixed charges range from 1.80 to 4.65 mills, depending on method of financing. Total power generation costs at an 80% load factor range from 3.65 to 6.50 mills/kwhr. (auth)

1963-02-01T23:59:59.000Z

249

High Speed, Low Cost Fabrication of Gas Diffusion Electrodes for Membrane Electrode Assemblies - DOE Hydrogen and Fuel Cells Program FY 2012 Annual Progress Report  

NLE Websites -- All DOE Office Websites (Extended Search)

8 8 DOE Hydrogen and Fuel Cells Program FY 2012 Annual Progress Report Emory S. De Castro BASF Fuel Cell, Inc. 39 Veronica Avenue Somerset, NJ 08873 Phone: (732) 545-5100 ext 4114 Email: Emory.DeCastro@BASF.com DOE Managers HQ: Nancy Garland Phone: (202) 586-5673 Email: Nancy.Garland@ee.doe.gov GO: Jesse Adams Phone: (720) 356-1421 Email: Jesse.Adams@go.doe.gov Contract Number: DE-EE0000384 Subcontractor: Dr. Vladimir Gurau Case Western Reserve University, Cleveland, Ohio Project Start Date: July 1, 2009 Project End Date: June 30, 2013 Fiscal Year (FY) 2012 Objectives Reduce cost in fabricating gas diffusion electrodes * through the introduction of high speed coating technology, with a focus on materials used for the high- temperature membrane electrode assemblies (MEAs)

250

SOLERAS solar cooling project  

Science Conference Proceedings (OSTI)

In view of the increasing demand for cooling in both the United States and Saudi Arabia, solar cooling systems are being considered as serious alternatives to the energy intensive conventional systems, especially when confronted with rising fossil fuel costs. Saudi Arabia and the hot, southern regions of the United States, having abundant sunshine and high cooling demand, are obvious candidates for solar active cooling systems and passive cooling design. Solar active cooling has yet to be shown to be either technologically mature or economically feasible, but efforts have been, and are presently being made within the United States National Solar Cooling Program to develop reliable systems which can compete economically with conventional cooling systems. Currently, the program is funding research and development projects in the areas of absorption, Rankine, dessicant, and advanced technologies. Saudi Arabia has a long and successful tradition of building cooling using passive architectural designs. Combining these past achievements with a program of research and development in both active and passive solar cooling should permit an early economical introduction of entirely solar cooled buildings to Saudi Arabia and the southern United States.

Corcoleotes, G.; Williamson, J.S.

1982-01-01T23:59:59.000Z

251

Cool Magnetic Molecules  

NLE Websites -- All DOE Office Websites (Extended Search)

Cool Magnetic Molecules Cool Magnetic Molecules Cool Magnetic Molecules Print Wednesday, 25 May 2011 00:00 Certain materials are known to heat up or cool down when they are exposed to a changing magnetic field. This is known as the magnetocaloric effect. All magnetic materials exhibit this effect, but in most cases, it is too small to be technologically useful. Recently, however, the search for special molecules with a surprisingly large capacity to keep cool has heated up, driven by environmental and cost considerations as well as by recent improvements in our ability to design, assemble, and probe the structure and chemistry of small molecules. An international collaboration of researchers from Spain, Scotland, and the U.S. has utilized ALS Beamline 11.3.1 (small-molecule crystallography) to characterize the design of such "molecular coolers." The work targets the synthesis of molecular cluster compounds containing many unpaired electrons ("nanomagnets") for applications involving enhanced magnetic refrigeration at very low temperatures.

252

THE HGCR-1, A DESIGN STUDY OF A NUCLEAR POWER STATION EMPLOYING A HIGH- TEMPERATURE GAS-COOLED REACTOR WITH GRAPHITE-UO$sub 2$ FUEL ELEMENTS  

SciTech Connect

The preliminary design of a 3095-Mw(thermal), helium-cooled, graphite- moderated reactor employing sign conditions, 1500 deg F reactor outlet gas would be circulated to eight steam generators to produce 1050 deg F, 1450-psi steam which would be converted to electrical power in eight 157-Mw(electrical) turbine- generators. The over-all efficiency of this nuclear power station is 36.5%. The significant activities released from the unclad graphite-UO/sub 2/ fuel appear to be less than 0.2% of those produced and would be equivalent to 0.002 curie/ cm/ sup 3/ in the primary helium circuit. The maintenance problems associated with this contamination level are discussed. A cost analysis indicates that the capital cost of this nuclear station per electrical kilowatt would be around 0, and that the production cost of electrical power would be 7.8 mills/kwhr. (auth)

Cottrell, W.B.; Copenhaver, C.M.; Culver, H.N.; Fontana, M.H.; Kelleghan, V.J.; Samuels, G.

1959-07-28T23:59:59.000Z

253

Solar energy as an alternate energy source to mixed oxide fuels in light-water cooled reactors  

DOE Green Energy (OSTI)

Supplemental information pertaining to the generic environmental impact statement on the Pu recycling process for mixed oxide light-water cooled reactors (GESMO) was requested from several sources. In particular, the role of alternate sources of energy was to be explored and the implications of these alternate sources to the question of Pu recycle in LWRs were to be investigated. In this vein, solar energy as an alternate source is the main subject of this report, along with other information related to solar energy. The general conclusion is that solar energy should have little effect on the decisions concerning GESMO.

Bertini, H.W.

1977-06-30T23:59:59.000Z

254

Neutronics, steady-state, and transient analyses for the Poland MARIA reactor for irradiation testing of LEU lead test fuel assemblies from CERCA : ANL independent verification results.  

Science Conference Proceedings (OSTI)

The MARIA reactor at the Institute of Atomic Energy (IAE) in Swierk (30 km SE of Warsaw) in the Republic of Poland is considering conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel assemblies (FA). The FA design in MARIA is rather unique; a suitable LEU FA has never been designed or tested. IAE has contracted with CERCA (the fuel supply portion of AREVA in France) to supply 2 lead test assemblies (LTA). The LTAs will be irradiated in MARIA to burnup level of at least 40% for both LTAs and to 60% for one LTA. IAE may decide to purchase additional LEU FAs for a full core conversion after the test irradiation. The Reactor Safety Committee within IAE and the National Atomic Energy Agency in Poland (PAA) must approve the LTA irradiation process. The approval will be based, in part, on IAE submitting revisions to portions of the Safety Analysis Report (SAR) which are affected by the insertion of the LTAs. (A similar process will be required for the full core conversion to LEU fuel.) The analysis required was established during working meetings between Argonne National Laboratory (ANL) and IAE staff during August 2006, subsequent email correspondence, and subsequent staff visits. The analysis needs to consider the current high-enriched uranium (HEU) core and 4 core configurations containing 1 and 2 LEU LTAs in various core positions. Calculations have been performed at ANL in support of the LTA irradiation. These calculations are summarized in this report and include criticality, burn-up, neutronics parameters, steady-state thermal hydraulics, and postulated transients. These calculations have been performed at the request of the IAE staff, who are performing similar calculations to be used in their SAR amendment submittal to the PAA. The ANL analysis has been performed independently from that being performed by IAE and should only be used as one step in the verification process.

Garner, P. L.; Hanan, N. A. (Nuclear Engineering Division)

2011-06-07T23:59:59.000Z

255

Cool Magnetic Molecules  

NLE Websites -- All DOE Office Websites (Extended Search)

Cool Magnetic Molecules Print Cool Magnetic Molecules Print Certain materials are known to heat up or cool down when they are exposed to a changing magnetic field. This is known as the magnetocaloric effect. All magnetic materials exhibit this effect, but in most cases, it is too small to be technologically useful. Recently, however, the search for special molecules with a surprisingly large capacity to keep cool has heated up, driven by environmental and cost considerations as well as by recent improvements in our ability to design, assemble, and probe the structure and chemistry of small molecules. An international collaboration of researchers from Spain, Scotland, and the U.S. has utilized ALS Beamline 11.3.1 (small-molecule crystallography) to characterize the design of such "molecular coolers." The work targets the synthesis of molecular cluster compounds containing many unpaired electrons ("nanomagnets") for applications involving enhanced magnetic refrigeration at very low temperatures.

256

Cool Magnetic Molecules  

NLE Websites -- All DOE Office Websites (Extended Search)

Cool Magnetic Molecules Print Cool Magnetic Molecules Print Certain materials are known to heat up or cool down when they are exposed to a changing magnetic field. This is known as the magnetocaloric effect. All magnetic materials exhibit this effect, but in most cases, it is too small to be technologically useful. Recently, however, the search for special molecules with a surprisingly large capacity to keep cool has heated up, driven by environmental and cost considerations as well as by recent improvements in our ability to design, assemble, and probe the structure and chemistry of small molecules. An international collaboration of researchers from Spain, Scotland, and the U.S. has utilized ALS Beamline 11.3.1 (small-molecule crystallography) to characterize the design of such "molecular coolers." The work targets the synthesis of molecular cluster compounds containing many unpaired electrons ("nanomagnets") for applications involving enhanced magnetic refrigeration at very low temperatures.

257

Cool Magnetic Molecules  

NLE Websites -- All DOE Office Websites (Extended Search)

Cool Magnetic Molecules Print Cool Magnetic Molecules Print Certain materials are known to heat up or cool down when they are exposed to a changing magnetic field. This is known as the magnetocaloric effect. All magnetic materials exhibit this effect, but in most cases, it is too small to be technologically useful. Recently, however, the search for special molecules with a surprisingly large capacity to keep cool has heated up, driven by environmental and cost considerations as well as by recent improvements in our ability to design, assemble, and probe the structure and chemistry of small molecules. An international collaboration of researchers from Spain, Scotland, and the U.S. has utilized ALS Beamline 11.3.1 (small-molecule crystallography) to characterize the design of such "molecular coolers." The work targets the synthesis of molecular cluster compounds containing many unpaired electrons ("nanomagnets") for applications involving enhanced magnetic refrigeration at very low temperatures.

258

Cool Magnetic Molecules  

NLE Websites -- All DOE Office Websites (Extended Search)

Cool Magnetic Molecules Print Cool Magnetic Molecules Print Certain materials are known to heat up or cool down when they are exposed to a changing magnetic field. This is known as the magnetocaloric effect. All magnetic materials exhibit this effect, but in most cases, it is too small to be technologically useful. Recently, however, the search for special molecules with a surprisingly large capacity to keep cool has heated up, driven by environmental and cost considerations as well as by recent improvements in our ability to design, assemble, and probe the structure and chemistry of small molecules. An international collaboration of researchers from Spain, Scotland, and the U.S. has utilized ALS Beamline 11.3.1 (small-molecule crystallography) to characterize the design of such "molecular coolers." The work targets the synthesis of molecular cluster compounds containing many unpaired electrons ("nanomagnets") for applications involving enhanced magnetic refrigeration at very low temperatures.

259

Cool Magnetic Molecules  

NLE Websites -- All DOE Office Websites (Extended Search)

Cool Magnetic Molecules Print Cool Magnetic Molecules Print Certain materials are known to heat up or cool down when they are exposed to a changing magnetic field. This is known as the magnetocaloric effect. All magnetic materials exhibit this effect, but in most cases, it is too small to be technologically useful. Recently, however, the search for special molecules with a surprisingly large capacity to keep cool has heated up, driven by environmental and cost considerations as well as by recent improvements in our ability to design, assemble, and probe the structure and chemistry of small molecules. An international collaboration of researchers from Spain, Scotland, and the U.S. has utilized ALS Beamline 11.3.1 (small-molecule crystallography) to characterize the design of such "molecular coolers." The work targets the synthesis of molecular cluster compounds containing many unpaired electrons ("nanomagnets") for applications involving enhanced magnetic refrigeration at very low temperatures.

260

Sustained Recycle in Light Water and Sodium-Cooled Reactors  

Science Conference Proceedings (OSTI)

From a physics standpoint, it is feasible to sustain recycle of used fuel in either thermal or fast reactors. This paper examines multi-recycle potential performance by considering three recycling approaches and calculating several fuel cycle parameters, including heat, gamma, and neutron emission of fresh fuel; radiotoxicity of waste; and uranium utilization. The first recycle approach is homogeneous mixed oxide (MOX) fuel assemblies in a light water reactor (LWR). The transuranic portion of the MOX was varied among Pu, NpPu, NpPuAm, or all-TRU. (All-TRU means all isotopes through Cf-252.) The Pu case was allowed to go to 10% Pu in fresh fuel, but when the minor actinides were included, the transuranic enrichment was kept below 8% to satisfy the expected void reactivity constraint. The uranium portion of the MOX was enriched uranium. That enrichment was increased (to as much as 6.5%) to keep the fuel critical for a typical LWR irradiation. The second approach uses heterogeneous inert matrix fuel (IMF) assemblies in an LWR - a mix of IMF and traditional UOX pins. The uranium-free IMF fuel pins were Pu, NpPu, NpPuAm, or all-TRU. The UOX pins were limited to 4.95% U-235 enrichment. The number of IMF pins was set so that the amount of TRU in discharged fuel from recycle N (from both IMF and UOX pins) was made into the new IMF pins for recycle N+1. Up to 60 of the 264 pins in a fuel assembly were IMF. The assembly-average TRU content was 1-6%. The third approach uses fast reactor oxide fuel in a sodium-cooled fast reactor with transuranic conversion ratio of 0.50 and 1.00. The transuranic conversion ratio is the production of transuranics divided by destruction of transuranics. The FR at CR=0.50 is similar to the CR for the MOX case. The fast reactor cases had a transuranic content of 33-38%, higher than IMF or MOX.

Steven J. Piet; Samuel E. Bays; Michael A. Pope; Gilles J. Youinou

2010-11-01T23:59:59.000Z

Note: This page contains sample records for the topic "fuel assemblies cooling" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


261

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

Science Conference Proceedings (OSTI)

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 %.

Henzl, Vladimir [Los Alamos National Laboratory; Croft, Stephen [Los Alamos National Laboratory; Swinhoe, Martyn T. [Los Alamos National Laboratory; Tobin, Stephen J. [Los Alamos National Laboratory

2012-07-18T23:59:59.000Z

262

Evaluation of Alternate Materials for Coated Particle Fuels for the Gas-Cooled Fast Reactor. Laboratory Directed Research and Development Program FY 2006 Final Report  

Science Conference Proceedings (OSTI)

Candidate ceramic materials were studied to determine their suitability as Gas-Cooled Fast Reactor particle fuel coatings. The ceramics examined in this work were: TiC, TiN, ZrC, ZrN, AlN, and SiC. The studies focused on (i) chemical reactivity of the ceramics with fission products palladium and rhodium, (ii) the thermomechanical stresses that develop in the fuel coatings from a variety of causes during burnup, and (iii) the radiation resiliency of the materials. The chemical reactivity of TiC, TiN, ZrC, and ZrN with Pd and Rh were all found to be much lower than that of SiC. A number of important chemical behaviors were observed at the ceramic-metal interfaces, including the formation of specific intermetallic phases and a variation in reaction rates for the different ceramics investigated. Based on the data collected in this work, the nitride ceramics (TiN and ZrN) exhibit chemical behavior that is characterized by lower reaction rates with Pd and Rh than the carbides TiC and ZrC. The thermomechanical stresses in spherical fuel particle ceramic coatings were modeled using finite element analysis, and included contributions from differential thermal expansion, fission gas pressure, fuel kernel swelling, and thermal creep. In general the tangential stresses in the coatings during full reactor operation are tensile, with ZrC showing the lowest values among TiC, ZrC, and SiC (TiN and ZrN were excluded from the comprehensive calculations due to a lack of available materials data). The work has highlighted the fact that thermal creep plays a critical role in the development of the stress state of the coatings by relaxing many of the stresses at high temperatures. To perform ion irradiations of sample materials, an irradiation beamline and high-temperature sample irradiation stage was constructed at the University of Wisconsin’s 1.7MV Tandem Accelerator Facility. This facility is now capable of irradiating of materials to high dose while controlling sample temperature up to 800ºC.

Paul A. Demkowicz; Karen Wright; Jian Gan; David Petti; Todd Allen; Jake Blanchard

2006-09-01T23:59:59.000Z

263

Deep Burn: Development of Transuranic Fuel for High-Temperature Helium-Cooled Reactors- Monthly Highlights November 2010  

SciTech Connect

During FY 2011 the DB Program will report Highlights on a monthly basis, but will no longer produce Quarterly Progress Reports. Technical details that were previously included in the quarterly reports will be included in the appropriate Milestone Reports that are submitted to FCRD Program Management. These reports will also be uploaded to the Deep Burn website. The Monthly Highlights report for October 2010, ORNL/TM-2010/300, was distributed to program participants on November 29, 2010. This report discusses the following: (1) Thermochemical Data and Model Development; (2) TRU (transuranic elements) TRISO (tri-structural isotropic) Development - (a) TRU Kernel Development, (b) Coating Development; (3) LWR Fully Ceramic Fuel - (a) FCM Fabrication Development, (b) FCM Irradiation Testing.

Snead, Lance Lewis [ORNL; Bell, Gary L [ORNL; Besmann, Theodore M [ORNL

2010-12-01T23:59:59.000Z

264

Stochastic Cooling  

Science Conference Proceedings (OSTI)

Stochastic Cooling was invented by Simon van der Meer and was demonstrated at the CERN ISR and ICE (Initial Cooling Experiment). Operational systems were developed at Fermilab and CERN. A complete theory of cooling of unbunched beams was developed, and was applied at CERN and Fermilab. Several new and existing rings employ coasting beam cooling. Bunched beam cooling was demonstrated in ICE and has been observed in several rings designed for coasting beam cooling. High energy bunched beams have proven more difficult. Signal suppression was achieved in the Tevatron, though operational cooling was not pursued at Fermilab. Longitudinal cooling was achieved in the RHIC collider. More recently a vertical cooling system in RHIC cooled both transverse dimensions via betatron coupling.

Blaskiewicz, M.

2011-01-01T23:59:59.000Z

265

Direct cooled power electronics substrate  

DOE Patents (OSTI)

The disclosure describes directly cooling a three-dimensional, direct metallization (DM) layer in a power electronics device. To enable sufficient cooling, coolant flow channels are formed within the ceramic substrate. The direct metallization layer (typically copper) may be bonded to the ceramic substrate, and semiconductor chips (such as IGBT and diodes) may be soldered or sintered onto the direct metallization layer to form a power electronics module. Multiple modules may be attached to cooling headers that provide in-flow and out-flow of coolant through the channels in the ceramic substrate. The modules and cooling header assembly are preferably sized to fit inside the core of a toroidal shaped capacitor.

Wiles, Randy H. (Powell, TN), Wereszczak, Andrew A. (Oak Ridge, TN); Ayers, Curtis W. (Kingston, TN); Lowe, Kirk T. (Knoxville, TN)

2010-09-14T23:59:59.000Z

266

Indirect passive cooling system for liquid metal cooled nuclear reactors  

SciTech Connect

This patent describes a passive cooling system. It is for liquid metal cooled nuclear reactors having a pool of liquid metal coolant with the heat generating fissionable fuel core substantially immersed in the pool of liquid metal coolant. The passive cooling system including a combination of spaced apart side-by-side partitions in generally concentric arrangement and providing for intermediate fluid circulation and heat transfer therebetween.

Hunsbedt, A.; Boardman, C.E.

1990-09-25T23:59:59.000Z

267

Downhole steam generator with improved preheating/cooling features. [Patent application  

DOE Patents (OSTI)

An apparatus is described for downhole steam generation employing dual-stage preheaters for liquid fuel and for the water. A first heat exchange jacket for the fuel surrounds the fuel/oxidant mixing section of the combustor assembly downstream of the fuel nozzle and contacts the top of the combustor unit of the combustor assembly, thereby receiving heat directly from the combustion of the fuel/oxidant. A second stage heat exchange jacket surrounds an upper portion of the oxidant supply line adjacent the fuel nozzle receiving further heat from the compression heat which results from pressurization of the oxidant. The combustor unit includes an inner combustor sleeve whose inner wall defines the combustion zone. The inner combustor sleeve is surrounded by two concentric water channels, one defined by the space between the inner combustor sleeve and an intermediate sleeve, and the second defined by the space between the intermediate sleeve and an outer cylindrical housing. The channels are connected by an annular passage adjacent the top of the combustor assembly and the countercurrent nature of the water flow provides efficient cooling of the inner combustor sleeve. An annular water ejector with a plurality of nozzles is provided to direct water downwardly into the combustor unit at the boundary of the combustion zone and along the lower section of the intermediate sleeve.

Donaldson, A.B.; Hoke, D.E.; Mulac, A.J.

1980-10-10T23:59:59.000Z

268

Air Breathing Direct Methanol Fuel Cell  

DOE Patents (OSTI)

A method for activating a membrane electrode assembly for a direct methanol fuel cell is disclosed. The method comprises operating the fuel cell with humidified hydrogen as the fuel followed by running the fuel cell with methanol as the fuel.

Ren; Xiaoming (Los Alamos, NM)

2003-07-22T23:59:59.000Z

269

Evaporative Roof Cooling - A Simple Solution to Cut Cooling Costs  

E-Print Network (OSTI)

Since the "Energy Crisis" Evaporative Roof Cooling Systems have gained increased acceptance as a cost effective method to reduce the high cost of air conditioning. Documented case histories in retrofit installations show direct energy savings and paybacks from twelve to thirty months. The main operating cost of an Evaporative Roof Cooling System is water. One thousand gallons of water, completely evaporated, will produce over 700 tons of cooling capability. Water usage seldom averages over 100 gallons per 1000 ft^2 of roof area per day or 10 oz. of water per 100 ft^2 every six minutes. Roof Cooling Systems, when planned in new construction, return 1-1/2 times the investment the first year in equipment savings and operating costs. Roof sprays are a low cost cooling solution for warehouses, distribution centers and light manufacturing or assembly areas with light internal loads. See text "Flywheel Cooling."

Abernethy, D.

1985-01-01T23:59:59.000Z

270

Fuel Cell Membrane Electrode Assembly Manufacturing R&D - DOE Hydrogen and Fuel Cells Program FY 2012 Annual Progress Report  

NLE Websites -- All DOE Office Websites (Extended Search)

7 7 FY 2012 Annual Progress Report DOE Hydrogen and Fuel Cells Program Michael Ulsh (Primary Contact), Guido Bender, Niccolo Aieta, Huyen Dinh National Renewable Energy Laboratory 15013 Denver West Parkway Golden, CO 80401 Phone: (303) 275-3842 Email: michael.ulsh@nrel.gov DOE Manager HQ: Nancy Garland Phone: (202) 586-5673 Email: Nancy.Garland@ee.doe.gov Partners: * Lawrence Berkeley National Laboratory (LBNL), Berkeley, CA * Colorado School of Mines, Golden, CO * University of Hawaii, Hawaii Natural Energy Institute, Honolulu, HI * Rensselaer Polytechnic Institute, Troy, NY

271

EXTENDING SODIUM FAST REACTOR DRIVER FUEL USE TO HIGHER TEMPERATURES  

Science Conference Proceedings (OSTI)

Calculations of potential sodium-cooled fast reactor fuel temperatures were performed to estimate the effects of increasing the outlet temperature of a given fast reactor design by increasing pin power, decreasing assembly flow, or increasing inlet temperature. Based upon experience in the U.S., both metal and mixed oxide (MOX) fuel types are discussed in terms of potential performance effects created by the increased operating temperatures. Assembly outlet temperatures of 600, 650 and 700 °C were used as goal temperatures. Fuel/cladding chemical interaction (FCCI) and fuel melting, as well as challenges to the mechanical integrity of the cladding material, were identified as the limiting phenomena. For example, starting with a recent 1000 MWth fast reactor design, raising the outlet temperature to 650 °C through pin power increase increased the MOX centerline temperature to more than 3300 °C and the metal fuel peak cladding temperature to more than 700 °C. These exceeded limitations to fuel performance; fuel melting was limiting for MOX and FCCI for metal fuel. Both could be alleviated by design ‘fixes’, such as using a barrier inside the cladding to minimize FCCI in the metal fuel, or using annular fuel in the case of MOX. Both would also require an advanced cladding material with improved stress rupture properties. While some of these are costly, the benefits of having a high-temperature reactor which can support hydrogen production, or other missions requiring high process heat may make the extra costs justified.

Douglas L. Porter

2011-02-01T23:59:59.000Z

272

An Evaluation of the Annular Fuel and Bottle-Shaped Fuel Concepts for Sodium Fast Reactors  

E-Print Network (OSTI)

Two innovative fuel concepts, the internally and externally cooled annular fuel and the bottle-shaped fuel, were investigated with the goal of increasing the power density and reduce the pressure drop in the sodium-cooled ...

Memmott, Matthew

273

Passive cooling safety system for liquid metal cooled nuclear reactors  

DOE Patents (OSTI)

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA); Hui, Marvin M. (Sunnyvale, CA); Berglund, Robert C. (Saratoga, CA)

1991-01-01T23:59:59.000Z

274

Indirect passive cooling system for liquid metal cooled nuclear reactors  

DOE Patents (OSTI)

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA)

1990-01-01T23:59:59.000Z

275

Computation of neutron fluxes in clusters of fuel pins arranged in hexagonal assemblies (2D and 3D)  

Science Conference Proceedings (OSTI)

For computations of fluxes, we have used Carvik's method of collision probabilities. This method requires tracking algorithms. An algorithm to compute tracks (in 2D and 3D) has been developed for seven hexagonal geometries with cluster of fuel pins. This has been implemented in the NXT module of the code DRAGON. The flux distribution in cluster of pins has been computed by using this code. For testing the results, they are compared when possible with the EXCELT module of the code DRAGON. Tracks are plotted in the NXT module by using MATLAB, these plots are also presented here. Results are presented with increasing number of lines to show the convergence of these results. We have numerically computed volumes, surface areas and the percentage errors in these computations. These results show that 2D results converge faster than 3D results. The accuracy on the computation of fluxes up to second decimal is achieved with fewer lines. (authors)

Prabha, H.; Marleau, G. [Institut de Genie Nucleaire, Ecole Polytechnique de Montreal, Stn. CV, P.O. Box 6079, Montreal, QC H3C 3A7 (Canada)

2012-07-01T23:59:59.000Z

276

On the Criticality Safety of Transuranic Sodium Fast Reactor Fuel Transport Casks  

SciTech Connect

This work addresses the neutronic performance and criticality safety issues of transport casks for fuel pertaining to low conversion ratio sodium cooled fast reactors, conventionally known as Advanced Burner Reactors. The criticality of a one, three, seven and 19-assembly cask capacity is presented. Both dry “helium” and flooded “water” filled casks are considered. No credit for fuel burnup or fission products was assumed. As many as possible of the conservatisms used in licensing light water reactor universal transport casks were incorporated into this SFR cask criticality design and analysis. It was found that at 7-assemblies or more, adding moderator to the SFR cask increases criticality margin. Also, removal of MAs from the fuel increases criticality margin of dry casks and takes a slight amount of margin away for wet casks. Assuming credit for borated fuel tube liners, this design analysis suggests that as many as 19 assemblies can be loaded in a cask if limited purely by criticality safety. If no credit for boron is assumed, the cask could possibly hold seven assemblies if low conversion ratio fast reactor grade fuel and not breeder reactor grade fuel is assumed. The analysis showed that there is a need for new cask designs for fast reactors spent fuel transportation. There is a potential of modifying existing transportation cask design as the starting point for fast reactor spent fuel transportation.

Samuel Bays; Ayodeji Alajo

2010-05-01T23:59:59.000Z

277

A Parametric Study of the DUPIC Fuel Cycle to Reflect Pressurized Water Reactor Fuel Management Strategy  

SciTech Connect

For both pressurized water reactor (PWR) and Canada deuterium uranium (CANDU) tandem analysis, the Direct Use of spent PWR fuel In CANDU reactor (DUPIC) fuel cycle in a CANDU 6 reactor is studied using the DRAGON/DONJON chain of codes with the ENDF/B-V and ENDF/B-VI libraries. The reference feed material is a 17 x 17 French standard 900-MW(electric) PWR fuel. The PWR spent-fuel composition is obtained from two-dimensional DRAGON assembly transport and depletion calculations. After a number of years of cooling, this defines the initial fuel nuclide field in the CANDU unit cell calculations in DRAGON, where it is further depleted with the same neutron group structure. The resulting macroscopic cross sections are condensed and tabulated to be used in a full-core model of a CANDU 6 reactor to find an optimized channel fueling rate distribution on a time-average basis. Assuming equilibrium refueling conditions and a particular refueling sequence, instantaneous full-core diffusion calculations are finally performed with the DONJON code, from which both the channel power peaking factors and local parameter effects are estimated. A generic study of the DUPIC fuel cycle is carried out using the linear reactivity model for initial enrichments ranging from 3.2 to 4.5 wt% in a PWR. Because of the uneven power histories of the spent PWR assemblies, the spent PWR fuel composition is expected to differ from one assembly to the next. Uneven mixing of the powder during DUPIC fuel fabrication may lead to uncertainties in the composition of the fuel bundle and larger peaking factors in CANDU. A mixing method for reducing composition uncertainties is discussed.

Rozon, Daniel; Shen Wei [Institut de Genie Nucleaire (Canada)

2001-05-15T23:59:59.000Z

278

BWR Assembly Optimization for Minor Actinide Recycling  

Science Conference Proceedings (OSTI)

The Primary objective of the proposed project is to apply and extend the latest advancements in LWR fuel management optimization to the design of advanced boiling water reactor (BWR) fuel assemblies specifically for the recycling of minor actinides (MAs).

G. Ivan Maldonado; John M. Christenson; J.P. Renier; T.F. Marcille; J. Casal

2010-03-22T23:59:59.000Z

279

AIR COOLED NEUTRONIC REACTOR  

DOE Patents (OSTI)

A nuclear reactor of the air-cooled, graphite moderated type is described. The active core consists of a cubicle mass of graphite, approximately 25 feet in each dimension, having horizontal channels of square cross section extending between two of the opposite faces, a plurality of cylindrical uranium slugs disposed in end to end abutting relationship within said channels providing a space in the channels through which air may be circulated, and a cadmium control rod extending within a channel provided in the moderator. Suitable shielding is provlded around the core, as are also provided a fuel element loading and discharge means, and a means to circulate air through the coolant channels through the fuel charels to cool the reactor.

Fermi, E.; Szilard, L.

1958-05-27T23:59:59.000Z

280

Sodium fast reactor fuels and materials : research needs.  

SciTech Connect

An expert panel was assembled to identify gaps in fuels and materials research prior to licensing sodium cooled fast reactor (SFR) design. The expert panel considered both metal and oxide fuels, various cladding and duct materials, structural materials, fuel performance codes, fabrication capability and records, and transient behavior of fuel types. A methodology was developed to rate the relative importance of phenomena and properties both as to importance to a regulatory body and the maturity of the technology base. The technology base for fuels and cladding was divided into three regimes: information of high maturity under conservative operating conditions, information of low maturity under more aggressive operating conditions, and future design expectations where meager data exist.

Denman, Matthew R.; Porter, Douglas (Idaho National Laboratory, Idaho Falls, ID); Wright, Art (Argonne National Laboratory Argonne, IL); Lambert, John (Argonne National Laboratory Argonne, IL); Hayes, Steven (Idaho National Laboratory, Idaho Falls, ID); Natesan, Ken (Argonne National Laboratory Argonne, IL); Ott, Larry J. (Oak Ridge National Laboratory, Oak Ridge, TN); Garner, Frank (Radiation Effects Consulting. Richland, WA); Walters, Leon (Advanced Reactor Concepts, Idaho Falls, ID); Yacout, Abdellatif (Argonne National Laboratory Argonne, IL)

2011-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "fuel assemblies cooling" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


281

Fuel Reliability Project: Boiling Water Fuel Performance at Kernkraftwerk Leibstadt  

Science Conference Proceedings (OSTI)

The Kernkraftwerk Leibstadt (KKL) boiling water reactor (BWR), a General Electric BWR/6, performed a lead use assembly (LUA) program with fuel from three fuel suppliers. This program presented a unique opportunity to evaluate fuel performance on advanced 10x10 designs of AREVA, Global Nuclear Fuel (GNF), and Westinghouse Electric Company (Westinghouse). Fuel assemblies from each supplier (vendor) were loaded into the KKL core in 1997 and 1998. A number of fuel inspections have been performed during annua...

2007-05-16T23:59:59.000Z

282

Inherent Prevention and Mitigation of Severe Accident Consequences in Sodium-Cooled Fast Reactors  

SciTech Connect

Safety challenges for sodium-cooled fast reactors include maintaining core temperatures within design limits and assuring the geometry and integrity of the reactor core. Due to the high power density in the reactor core, heat removal requirements encourage the use of high-heat-transfer coolants such as liquid sodium. The variation of power across the core requires ducted assemblies to control fuel and coolant temperatures, which are also used to constrain core geometry. In a fast reactor, the fuel is not in the most neutronically reactive configuration during normal operation. Accidents leading to fuel melting, fuel pin failure, and fuel relocation can result in positive reactivity, increasing power, and possibly resulting in severe accident consequences including recriticalities that could threaten reactor and containment integrity. Inherent safety concepts, including favorable reactivity feedback, natural circulation cooling, and design choices resulting in favorable dispersive characteristics for failed fuel, can be used to increase the level of safety to the point where it is highly unlikely, or perhaps even not credible, for such severe accident consequences to occur.

Roald A. Wigeland; James E. Cahalan

2011-04-01T23:59:59.000Z

283

Reactor physics behavior of transuranic-bearing TRISO-particle fuel in a pressurized water reactor  

SciTech Connect

Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU) - only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is retained. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint. (authors)

Pope, M. A.; Sen, R. S.; Ougouag, A. M.; Youinou, G. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Boer, B. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); SCK-CEN, Boertang 200, BE-2400 Mol (Belgium)

2012-07-01T23:59:59.000Z

284

Reactor Physics Behavior of Transuranic-Bearing TRISO-Particle Fuel in a Pressurized Water Reactor  

SciTech Connect

Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU)-only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint.

Michael A. Pope; R. Sonat Sen; Abderrafi M. Ougouag; Gilles Youinou; Brian Boer

2012-04-01T23:59:59.000Z

285

Miniature ceramic fuel cell  

DOE Patents (OSTI)

A miniature power source assembly capable of providing portable electricity is provided. A preferred embodiment of the power source assembly employing a fuel tank, fuel pump and control, air pump, heat management system, power chamber, power conditioning and power storage. The power chamber utilizes a ceramic fuel cell to produce the electricity. Incoming hydro carbon fuel is automatically reformed within the power chamber. Electrochemical combustion of hydrogen then produces electricity.

Lessing, Paul A. (Idaho Falls, ID); Zuppero, Anthony C. (Idaho Falls, ID)

1997-06-24T23:59:59.000Z

286

Automatically closing swing gate closure assembly  

DOE Patents (OSTI)

A swing gate closure assembly for nuclear reactor tipoff assembly wherein the swing gate is cammed open by a fuel element or spacer but is reliably closed at a desired closing rate primarily by hydraulic forces in the absence of a fuel charge.

Chang, Shih-Chih (Richland, WA); Schuck, William J. (Richland, WA); Gilmore, Richard F. (Kennewick, WA)

1988-01-01T23:59:59.000Z

287

Method for passive cooling liquid metal cooled nuclear reactors, and system thereof  

DOE Patents (OSTI)

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel.

Hunsbedt, Anstein (Los Gatos, CA); Busboom, Herbert J. (San Jose, CA)

1991-01-01T23:59:59.000Z

288

Performance of Transuranic-Loaded Fully Ceramic Micro-Encapsulated Fuel in LWRs Final Report, Including Void Reactivity Evaluation  

SciTech Connect

The current focus of the Deep Burn Project is on once-through burning of transuranics (TRU) in light-water reactors (LWRs). The fuel form is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the tri-isotropic (TRISO) fuel particle design from high-temperature reactor technology. In the Deep Burn LWR (DB-LWR) concept, these fuel particles are pressed into compacts using SiC matrix material and loaded into fuel pins for use in conventional LWRs. The TRU loading comes from the spent fuel of a conventional LWR after 5 years of cooling. Unit cell and assembly calculations have been performed using the DRAGON-4 code to assess the physics attributes of TRU-only FCM fuel in an LWR lattice. Depletion calculations assuming an infinite lattice condition were performed with calculations of various reactivity coefficients performed at each step. Unit cells and assemblies containing typical UO2 and mixed oxide (MOX) fuel were analyzed in the same way to provide a baseline against which to compare the TRU-only FCM fuel. Then, assembly calculations were performed evaluating the performance of heterogeneous arrangements of TRU-only FCM fuel pins along with UO2 pins.

Michael A. Pope; R. Sonat Sen; Brian Boer; Abderrafi M. Ougouag; Gilles Youinou

2011-09-01T23:59:59.000Z

289

Gas-cooled reactors  

SciTech Connect

Experience to date with operation of high-temperature gas-cooled reactors has been quite favorable. Despite problems in completion of construction and startup, three high-temperature gas-cooled reactor (HTGR) units have operated well. The Windscale Advanced Gas-Cooled Reactor (AGR) in the United Kingdom has had an excellent operating history, and initial operation of commercial AGRs shows them to be satisfactory. The latter reactors provide direct experience in scale-up from the Windscale experiment to fullscale commercial units. The Colorado Fort St. Vrain 330-MWe prototype helium-cooled HTGR is now in the approach-to-power phase while the 300-MWe Pebble Bed THTR prototype in the Federal Republic of Germany is scheduled for completion of construction by late 1978. THTR will be the first nuclear power plant which uses a dry cooling tower. Fuel reprocessing and refabrication have been developed in the laboratory and are now entering a pilot-plant scale development. Several commercial HTGR power station orders were placed in the U.S. prior to 1975 with similar plans for stations in the FRG. However, the combined effects of inflation, reduced electric power demand, regulatory uncertainties, and pricing problems led to cancellation of the 12 reactors which were in various stages of planning, design, and licensing.

Schulten, R.; Trauger, D.B.

1976-01-01T23:59:59.000Z

290

Cooled railplug  

DOE Patents (OSTI)

The railplug is a plasma ignitor capable of injecting a high energy plasma jet into a combustion chamber of an internal combustion engine or continuous combustion system. An improved railplug is provided which has dual coaxial chambers (either internal or external to the center electrode) that provide for forced convective cooling of the electrodes using the normal pressure changes occurring in an internal combustion engine. This convective cooling reduces the temperature of the hot spot associated with the plasma initiation point, particularly in coaxial railplug configurations, and extends the useful life of the railplug. The convective cooling technique may also be employed in a railplug having parallel dual rails using dual, coaxial chambers.

Weldon, William F. (Austin, TX)

1996-01-01T23:59:59.000Z

291

Analysis of experiments in the Phase III GCFR benchmark critical assembly  

SciTech Connect

Experiments carried out in the third gas-cooled fast breeder reactor (GCFR) benchmark critical assembly on the Zero Power Reactor-9 at Argonne National Laboratory were analyzed using methods and computer codes employed routinely for design and performance evaluations on power-plant GCFR cores. The program for the Phase III GCFR assembly, with a 1900-liter, three-enrichment zone core, included measurements of reaction-rate profiles in a typical power-flattened design, studies of material reactivity coefficients, reaction ratio and breeding parameter determinations, and comparison of pin with plate fuel loadings. Calculated parameters to compare with all of the measured results were obtained using 10-group cross sections based on ENDF/B-4 and two-dimensional diffusion theory, with adjustments for fuel-cell heterogeneity and void-lattice streaming effects.

Hess, A.L.; Baylor, K.J.

1980-04-01T23:59:59.000Z

292

Low enrichment fuel conversion for Iowa State University  

SciTech Connect

Work during the reported period was centered primarily in preparation for receiving the LEU fuel and the shipping of the HEU fuel. The LEU fuel has not been received. The HEU fuel assemblies for the UTR-10 reactor will not fit into any current research reactor shipping containers; therefore, the fuel assemblies must be disassembled and the fuel shipped as fuel plates. Procedures and practices have been developed so that the fuel assemblies will be disassembled in a shielded environment.

Rohach, A.F.; Hendrickson, R.A.

1990-08-01T23:59:59.000Z

293

FOCUS COOLING  

NLE Websites -- All DOE Office Websites (Extended Search)

www.datacenterdynamics.com www.datacenterdynamics.com FOCUS COOLING Issue 28, March/April 2013 LBNL'S NOVEL APPROACH TO COOLING Lawrence Berkeley National Laboratory and APC by Schneider Electric test a unique double-exchanger cooling system LBNL program manager Henry Coles says can cut energy use by half A s part of a demonstration sponsored by the California Energy Commission in support of the Silicon Valley Leadership Group's data center summit, Lawrence Berkeley National Laboratory (LBNL) collaborated with APC by Schneider Electric to demonstrate a novel prototype data center cooling device. The device was installed at an LBNL data center in Berkeley, California. It included two air-to-water heat exchangers. Unlike common single-heat-exchanger configurations, one of these was supplied with

294

Ventilative cooling  

E-Print Network (OSTI)

This thesis evaluates the performance of daytime and nighttime passive ventilation cooling strategies for Beijing, Shanghai and Tokyo. A new simulation method for cross-ventilated wind driven airflow is presented . This ...

Graça, Guilherme Carrilho da, 1972-

1999-01-01T23:59:59.000Z

295

Impact of Fuel Failure on Criticality Safety of Used Nuclear Fuel  

SciTech Connect

Commercial used nuclear fuel (UNF) in the United States is expected to remain in storage for considerably longer periods than originally intended (e.g., <40 years). Extended storage (ES) time and irradiation of nuclear fuel to high-burnup values (>45 GWd/t) may increase the potential for fuel failure during normal and accident conditions involving storage and transportation. Fuel failure, depending on the severity, can result in changes to the geometric configuration of the fuel, which has safety and regulatory implications. The likelihood and extent of fuel reconfiguration and its impact on the safety of the UNF is not well understood. The objective of this work is to assess and quantify the impact of fuel reconfiguration due to fuel failure on criticality safety of UNF in storage and transportation casks. This effort is primarily motivated by concerns related to the potential for fuel degradation during ES periods and transportation following ES. The criticality analyses consider representative UNF designs and cask systems and a range of fuel enrichments, burnups, and cooling times. The various failed-fuel configurations considered are designed to bound the anticipated effects of individual rod and general cladding failure, fuel rod deformation, loss of neutron absorber materials, degradation of canister internals, and gross assembly failure. The results quantify the potential impact on criticality safety associated with fuel reconfiguration and may be used to guide future research, design, and regulatory activities. Although it can be concluded that the criticality safety impacts of fuel reconfiguration during transportation subsequent to ES are manageable, the results indicate that certain configurations can result in a large increase in the effective neutron multiplication factor, k{sub eff}. Future work to inform decision making relative to which configurations are credible, and therefore need to be considered in a safety evaluation, is recommended.

Marshall, William BJ J [ORNL; Wagner, John C [ORNL

2012-01-01T23:59:59.000Z

296

Determining the quality and quantity of heat produced by proton exchange membrane fuel cells with application to air-cooled stacks for combined heat and power  

E-Print Network (OSTI)

Determining the quality and quantity of heat produced by proton exchange membrane fuel cells Determining the quality and quantity of heat produced by proton exchange membrane fuel cells with application, the coolant is pumped to a heat recovery system. A water-to-air heat exchange system or water-to-water heat

Victoria, University of

297

Passive cooling system for liquid metal cooled nuclear reactors with backup coolant flow path  

DOE Patents (OSTI)

A liquid metal cooled nuclear fission reactor plant having a passive auxiliary safety cooling system for removing residual heat resulting from fuel decay during reactor shutdown, or heat produced during a mishap. This reactor plant is enhanced by a backup or secondary passive safety cooling system which augments the primary passive auxiliary cooling system when in operation, and replaces the primary system when rendered inoperable.

Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA)

1993-01-01T23:59:59.000Z

298

Nuclear Fuels | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Nuclear Fuels Nuclear Fuels Nuclear Fuels A reactor's ability to produce power efficiently is significantly affected by the composition and configuration of its fuel system. A nuclear fuel assembly consists of hundreds of thousands of uranium pellets, stacked and encapsulated within tubes called fuel rods or fuel pins which are then bundled together in various geometric arrangements. There are many design considerations for the material composition and geometric configuration of the various components comprising a nuclear fuel system. Future designs for the fuel and the assembly or packaging of fuel will contribute to cleaner, cheaper and safer nuclear energy. Today's process for developing and testing new fuel systems is resource and time intensive. The process to manufacture the fuel, build an assembly,

299

Reducing home heating and cooling costs  

SciTech Connect

This report is in response to a request from the House Committee on Energy and Commerce that the Energy Information Administration (EIA) undertake a neutral, unbiased analysis of the cost, safety, and health and environmental effects of the three major heating fuels: heating oil, natural gas, and electricity. The Committee also asked EIA to examine the role of conservation in the choice of heating and cooling fuel. To accommodate a wide audience, EIA decided to respond to the Committee`s request in the context of a report on reducing home heating and cooling costs. Accordingly, this report discusses ways to weatherize the home, compares the features of the three major heating and cooling fuels, and comments on the types of heating and cooling systems on the market. The report also includes a worksheet and supporting tables that will help in the selection of a heating and/or cooling system.

Not Available

1994-07-01T23:59:59.000Z

300

Cooled railplug  

DOE Patents (OSTI)

The railplug is a plasma ignitor capable of injecting a high energy plasma jet into a combustion chamber of an internal combustion engine or continuous combustion system. An improved railplug is provided which has dual coaxial chambers (either internal or external to the center electrode) that provide for forced convective cooling of the electrodes using the normal pressure changes occurring in an internal combustion engine. This convective cooling reduces the temperature of the hot spot associated with the plasma initiation point, particularly in coaxial railplug configurations, and extends the useful life of the railplug. The convective cooling technique may also be employed in a railplug having parallel dual rails using dual, coaxial chambers. 10 figs.

Weldon, W.F.

1996-05-07T23:59:59.000Z

Note: This page contains sample records for the topic "fuel assemblies cooling" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


301

FUEL ELEMENTS FOR THE ARGONNE ADVANCED RESEARCH REACTOR  

SciTech Connect

The core design and the fuel element concept for the high-flux Argonne Advanced Research Reactor are presented. The core is cooled and moderated by light water and utilizes beryllium as a reflector. The fuel element assembly is rhomboidal in cross section and consists of 27 plates fastened together at their edges by dovetailed locking keys, and at each end by end fittings. Each fuel plate is 40 mils thick and contains a uniform dispersion of highly enriched UO/ sub 2/ particles, up to a maximum of 37 wt%, in a matrix of sintered stainless steel powder. A 5 mil thick stainless steel cladding is metallurgically bonded to each side of the fueled matrix. (N.W.R.)

Adolph, N.R.; Silberstein, M.S.; Weinstein, A.

1962-01-01T23:59:59.000Z

302

Transient Testing of Nuclear Fuels and Materials in United States  

Science Conference Proceedings (OSTI)

The US Department of Energy (DOE) has been engaged in an effort to develop and qualify next generation LWR fuel with enhanced performance and safety and reduced waste generation since 2010. This program, which has emphasized collaboration between the DOE, U.S. national laboratories and nuclear industry, was refocused from enhanced performance to enhanced accident tolerance following the events at Fukushima in 2011. Accident tolerant fuels have been specifically described as fuels that, in comparison with standard UO2-Zircaloy, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, operational transients, as well as design-basis and beyond design-basis events. The program maintains an ambitious goal to insert a lead test assembly (LTA) of the new design into a commercial power reactor by 2022 .

Daniel M. Wachs

2012-12-01T23:59:59.000Z

303

Tips: Heating and Cooling | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Tips: Heating and Cooling Tips: Heating and Cooling Tips: Heating and Cooling May 30, 2012 - 7:38pm Addthis Household Heating Systems: Although several different types of fuels are available to heat our homes, more than half of us use natural gas. | Source: Buildings Energy Data Book 2010, 2.1.1 Residential Primary Energy Consumption, by Year and Fuel Type (Quadrillion Btu and Percent of Total). Household Heating Systems: Although several different types of fuels are available to heat our homes, more than half of us use natural gas. | Source: Buildings Energy Data Book 2010, 2.1.1 Residential Primary Energy Consumption, by Year and Fuel Type (Quadrillion Btu and Percent of Total). Heating and cooling your home uses more energy and costs more money than any other system in your home -- typically making up about 54% of your

304

Design of an Online Fission Gas Monitoring System for Post-irradiation Examination Heating Tests of Coated Fuel Particles for High-Temperature Gas-Cooled Reactors  

Science Conference Proceedings (OSTI)

A new Fission Gas Monitoring System (FGMS) has been designed at the Idaho National Laboratory (INL) for use of monitoring online fission gas-released during fuel heating tests. The FGMS will be used with the Fuel Accident Condition Simulator (FACS) at the Hot Fuels Examination Facility (HFEF) located at the Materials and Fuels Complex (MFC) within the INL campus. Preselected Advanced Gas Reactor (AGR) TRISO (Tri-isotropic) fuel compacts will undergo testing to assess the fission product retention characteristics under high temperature accident conditions. The FACS furnace will heat the fuel to temperatures up to 2,000ºC in a helium atmosphere. Released fission products such as Kr and Xe isotopes will be transported downstream to the FGMS where they will accumulate in cryogenically cooledcollection traps and monitored with High Purity Germanium (HPGe) detectors during the heating process. Special INL developed software will be used to monitor the accumulated fission products and will report data in near real-time. These data will then be reported in a form that can be readily available to the INL reporting database. This paper describes the details of the FGMS design, the control and acqusition software, system calibration, and the expected performance of the FGMS. Preliminary online data may be available for presentation at the High Temperature Reactor (HTR) conference.

Dawn Scates

2010-10-01T23:59:59.000Z

305

Performance of Trasuranic-Loaded Fully Ceramic Micro-Encapsulated Fuel in LWRs Interim Report, Including Void Reactivity Evaluation  

Science Conference Proceedings (OSTI)

The current focus of the Deep Burn Project is on once-through burning of transuranice (TRU) in light water reactors (LWRs). The fuel form is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the tri-isotropic (TRISO) fuel particle design from high-temperature reactor technology. In the Deep Burn LWR (DB-LWR) concept, these fuel particles would be pressed into compacts using SiC matrix material and loaded into fuel pins for use in conventional LWRs. The TRU loading comes from the spent fuel of a conventional LWR after 5 years of cooling. Unit cell calculations have been performed using the DRAGON-4 code in order assess the physics attributes of TRU-only FCM fuel in an LWR lattice. Depletion calculations assuming an infinite lattice condition were performed with calculations of various reactivity coefficients performed at each step. Unit cells containing typical UO2 and MOX fuel were analyzed in the same way to provide a baseline against which to compare the TRU-only FCM fuel. Loading of TRU-only FCM fuel into a pin without significant quantities of uranium challenges the design from the standpoint of several key reactivity parameters, particularly void reactivity, and to some degree, the Doppler coefficient. These unit cells, while providing an indication of how a whole core of similar fuel would behave, also provide information of how individual pins of TRU-only FCM fuel would influence the reactivity behavior of a heterogeneous assembly. If these FCM fuel pins are included in a heterogeneous assembly with LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance of the TRU-only FCM fuel pins may be preserved. A configuration such as this would be similar to CONFU assemblies analyzed in previous studies. Analogous to the plutonium content limits imposed on MOX fuel, some amount of TRU-only FCM pins in an otherwise-uranium fuel assembly may give acceptable reactivity performance. Assembly calculations will be performed in future work to explore the design options for heterogeneous assemblies of this type and their impact on reactivity coefficients.

Michael A. Pope; Brian Boer; Gilles Youinou; Abderrafi M. Ougouag

2011-03-01T23:59:59.000Z

306

Liquid metal cooled nuclear reactor plant system  

DOE Patents (OSTI)

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting for fuel decay during reactor shutdown, or heat produced during a mishap. The reactor system is enhanced with sealing means for excluding external air from contact with the liquid metal coolant leaking from the reactor vessel during an accident. The invention also includes a silo structure which resists attack by leaking liquid metal coolant, and an added unique cooling means.

Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA)

1993-01-01T23:59:59.000Z

307

A Technical Review of Non-Destructive Assay Research for the Characterization of Spent Nuclear Fuel Assemblies Being Conducted Under the US DOE NGSI - 11544  

E-Print Network (OSTI)

NA-24), National Nuclear Security Administration, U.S.nuclear energy together with the emergence of alternative fuel cycles. Introduction Energy security

Croft, S.

2012-01-01T23:59:59.000Z

308

NEUTRONIC REACTOR WITH ACCESSIBLE THIMBLE AND EMERGENCY COOLING FEATURES  

DOE Patents (OSTI)

BS>A safety system for a water-moderated reactor is described. The invention comprises a reservoir system for spraying the fuel elements within a fuel assembly with coolant and keeping them in a continuous bath even if the coolant moderator is lost from the reactor vessel. A reservoir gravity feeds one or more nozzels positioned within each fuel assembly which continually forces water past the fuel elements.

McCorkle, W.H.

1960-02-23T23:59:59.000Z

309

Oil cooled, hermetic refrigerant compressor  

DOE Patents (OSTI)

A hermetic refrigerant compressor having an electric motor and compressor assembly in a hermetic shell is cooled by oil which is first cooled in an external cooler and is then delivered through the shell to the top of the motor rotor where most of it is flung radially outwardly within the confined space provided by the cap which channels the flow of most of the oil around the top of the stator and then out to a multiplicity of holes to flow down to the sump and provide further cooling of the motor and compressor. Part of the oil descends internally of the motor to the annular chamber to provide oil cooling of the lower part of the motor, with this oil exiting through vent hole also to the sump. Suction gas with entrained oil and liquid refrigerant therein is delivered to an oil separator from which the suction gas passes by a confined path in pipe to the suction plenum and the separated oil drops from the separator to the sump. By providing the oil cooling of the parts, the suction gas is not used for cooling purposes and accordingly increase in superheat is substantially avoided in the passage of the suction gas through the shell to the suction plenum. 3 figs.

English, W.A.; Young, R.R.

1985-05-14T23:59:59.000Z

310

Oil cooled, hermetic refrigerant compressor  

DOE Patents (OSTI)

A hermetic refrigerant compressor having an electric motor and compressor assembly in a hermetic shell is cooled by oil which is first cooled in an external cooler 18 and is then delivered through the shell to the top of the motor rotor 24 where most of it is flung radially outwardly within the confined space provided by the cap 50 which channels the flow of most of the oil around the top of the stator 26 and then out to a multiplicity of holes 52 to flow down to the sump and provide further cooling of the motor and compressor. Part of the oil descends internally of the motor to the annular chamber 58 to provide oil cooling of the lower part of the motor, with this oil exiting through vent hole 62 also to the sump. Suction gas with entrained oil and liquid refrigerant therein is delivered to an oil separator 68 from which the suction gas passes by a confined path in pipe 66 to the suction plenum 64 and the separated oil drops from the separator to the sump. By providing the oil cooling of the parts, the suction gas is not used for cooling purposes and accordingly increase in superheat is substantially avoided in the passage of the suction gas through the shell to the suction plenum 64.

English, William A. (Murrysville, PA); Young, Robert R. (Murrysville, PA)

1985-01-01T23:59:59.000Z

311

Determining plutonium mass in spent fuel using Cf-252 interrogation with prompt neutron detection  

SciTech Connect

{sup 252}Cf Interrogation with Prompt Neutron (CIPN) detection is proposed as one of 14 NDA techniques to determine Pu mass in spent fuel assemblies (FAs). CIPN is a low-cost and portable instrument, and it looks like a modified fork detector combined with an active interrogation source. Fission chamber (FC) is chosen as neutron detector because of its insensitivity to {gamma} radiation. The CIPN assay is comprised of two measurements, a background count and an active count, without and with the {sup 252}Cf source next to the fuel respectively. The net signal above background is primarily due to the multiplication of Cf source neutrons caused by the fissile content. The capability of CIPN to detect diversion and to determine fissile content was quantified using MCNPX simulations. New schemes were proposed (such as burnup and cooling time correction, etc.) and the results show that the fissile content of a target spent fuel assembly can be determined using CIPN signal.

Hu, Jianwei [Los Alamos National Laboratory; Tobin, Stephen J [Los Alamos National Laboratory; Menlove, Howard O [Los Alamos National Laboratory; Croft, Stephen [Los Alamos National Laboratory

2010-01-01T23:59:59.000Z

312

Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop  

SciTech Connect

The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm/sup 2/, 1000/sup 0/C cladding temperature, and (2) 40 h at 40 W/cm/sup 2/, 1200/sup 0/C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through clad melting at 1370/sup 0/C.

McCulloch, R.W.; MacPherson, R.E.

1983-03-01T23:59:59.000Z

313

FUEL ELEMENT  

DOE Patents (OSTI)

A ceramic fuel element for a nuclear reactor that has improved structural stability as well as improved cooling and fission product retention characteristics is presented. The fuel element includes a plurality of stacked hollow ceramic moderator blocks arranged along a tubular raetallic shroud that encloses a series of axially apertured moderator cylinders spaced inwardly of the shroud. A plurality of ceramic nuclear fuel rods are arranged in the annular space between the shroud and cylinders of moderator and appropriate support means and means for directing gas coolant through the annular space are also provided. (AEC)

Bean, R.W.

1963-11-19T23:59:59.000Z

314

"Hot" for Warm Water Cooling  

E-Print Network (OSTI)

liquid cooling, dry cooler, cooling tower 1. INTRODUCTIONsolutions for cooling. Substituting cooling towers,hybrid cooling towers, or dry coolers that provide warmer

Coles, Henry

2012-01-01T23:59:59.000Z

315

Neutronic Analysis of the Burning of Transuranics in Fully Ceramic Micro-Encapsulated Tri-Isotropic Particle-Fuel in a PWR  

SciTech Connect

Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU) – only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO2 and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO2 and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior is dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint.

Michael A. Pope; R. Sonat Sen; Abderrafi M. Ougouag; Gilles Youinou; Brian Boer

2012-11-01T23:59:59.000Z

316

CONSTRUCTION OF NUCLEAR FUEL ELEMENTS  

DOE Patents (OSTI)

>A rib arrangement and an end construction for nuclearfuel elements laid end to end in a coolant tube are described. The rib arrangement is such that each fuel element, when separated from other fuel elements, fits loosely in the coolant tube and so can easily be inserted or withdrawn from the tube. The end construction of the fuel elements is such that the fuel elements when assembled end to end are keyed against relative rotation, and the ribs of each fuel element cooperate with the ribs of the adjacent fuel elements to give the assembled fuel elements a tight fit with the coolant tube. (AEC)

Weems, S.J.

1963-09-24T23:59:59.000Z

317

Car Parts Self-Assembled From  

E-Print Network (OSTI)

Car Parts Self-Assembled From DNA Next: Navigation Helmet Cr SoundTech, Clay Dillow, clean energy, desali desalinization, electricity production, micro cells, wastewater Not bad for a microbe Microbial fuel cell desalinates water while generating electricity: This microbial

318

Liquid Fuels Market Model of the National Energy Modeling ...  

U.S. Energy Information Administration (EIA)

The outside battery-limit (OSBL) costs include the cost of cooling water, steam and electric power generation and distribution, fuel oil and fuel gas ...

319

Safety aspects of the IFR pyroprocess fuel cycle  

SciTech Connect

This paper addresses the important safety considerations related to the unique Integral Fast Reactor (IFR) fuel cycle technology, the pyroprocess. Argonne has been developing the IFR since 1984. It is a liquid metal cooled reactor, with a unique metal alloy fuel, and it utilizes a radically new fuel cycle. An existing facility, the Hot Fuel Examination Facility-South (HFEF/S) is being modified and equipped to provide a complete demonstration of the fuel cycle. This paper will concentrate on safety aspects of the future HFEF/S operation, slated to begin late next year. HFEF/S is part of Argonne's complex of reactor test facilities located on the Idaho National Engineering Laboratory. HFEF/S was originally put into operation in 1964 as the EBR-II Fuel Cycle Facility (FCF) (Stevenson, 1987). From 1964--69 FCF operated to demonstrate an earlier and incomplete form of today's pyroprocess, recycling some 400 fuel assemblies back to EBR-II. The FCF mission was then changed to one of an irradiated fuels and materials examination facility, hence the name change to HFEF/S. The modifications consist of activities to bring the facility into conformance with today's much more stringent safety standards, and, of course, providing the new process equipment. The pyroprocess and the modifications themselves are described more fully elsewhere (Lineberry, 1987; Chang, 1987). 18 refs., 5 figs., 2 tabs.

Forrester, R.J.; Lineberry, M.J.; Charak, I.; Tessier, J.H.; Solbrig, C.W.; Gabor, J.D.

1989-01-01T23:59:59.000Z

320

Accelerator-Driven Transmutation Of Spent Fuel Elements  

NLE Websites -- All DOE Office Websites (Extended Search)

Fuel Elements An apparatus and method is described for transmuting higher actinides, plutonium and selected fission products in a liquid-fuel subcritical assembly. Available for...

Note: This page contains sample records for the topic "fuel assemblies cooling" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


321

Regenerator cross arm seal assembly  

SciTech Connect

A seal assembly for disposition between a cross arm on a gas turbine engine block and a regenerator disc, the seal assembly including a platform coextensive with the cross arm, a seal and wear layer sealingly and slidingly engaging the regenerator disc, a porous and compliant support layer between the platform and the seal and wear layer porous enough to permit flow of cooling air therethrough and compliant to accommodate relative thermal growth and distortion, a dike between the seal and wear layer and the platform for preventing cross flow through the support layer between engine exhaust and pressurized air passages, and air diversion passages for directing unregenerated pressurized air through the support layer to cool the seal and wear layer and then back into the flow of regenerated pressurized air.

Jackman, Anthony V. (Indianapolis, IN)

1988-01-01T23:59:59.000Z

322

Fuel rod retention device for a nuclear reactor  

DOE Patents (OSTI)

A device is described for supporting a nuclear fuel rod in a fuel rod assembly which allows the rod to be removed without disturbing other rods in the assembly. A fuel rod cap connects the rod to a bolt which is supported in the assembly end fitting by means of a locking assembly. The device is designed so that the bolt is held securely during normal reactor operation yet may be easily disengaged and the fuel rod removed when desired.

Hylton, Charles L. (Madison Heights, VA)

1984-01-01T23:59:59.000Z

323

Best Management Practices Manual for Preventing Cooling Water Intake Blockages  

Science Conference Proceedings (OSTI)

Blockage of the cooling water intake structure (CWIS) occurs frequently at nuclear and fossil power facilities worldwide, regardless of fuel type or cooling water body source. The loss of cooling water impacts facility safety and reliabilityprincipally at nuclear facilitiesand results in a loss of revenue. This Best Management Practices Manual for Preventing Cooling Water Intake Blockages presents a review of debris management at existing facilities and provides procedural and operation and maintenance (...

2009-12-22T23:59:59.000Z

324

BWR Fuel Performance at KKL: 10 x 10 Fuel Irradiated to 54 GWd/MTU  

Science Conference Proceedings (OSTI)

The Kernkraftwerk Leibstadt (KKL) boiling water reactor (BWR) currently has a lead use assembly (LUA) program underway with fuel from three different fuel suppliers. This program presents a unique opportunity to simultaneously measure fuel performance on the advanced 10x10 designs of Framatome-ANP (F-ANP), Global Nuclear Fuel (GNF), and Westinghouse (W). Fuel assemblies from each supplier initially went into KKL in 1997 and 1998. A number of fuel inspections have been performed at various burnup levels a...

2003-12-08T23:59:59.000Z

325

Fuel cell powered irrigation system  

SciTech Connect

Set out herein is a fuel cell power plant for use with irrigation systems wherein the fuel cell is utilized to generate electric current to drive a pump motor. This pump motor drives a first water pump which receives water for distribution through a traveling irrigation system, the output of the first pump first conveyed into a condenser heat exchanger connected to a steam engine or turbine cycle. The fuel cell itself is contained within a boiler assembly and the heat of production of the electric power is used to generate steam which is sent to the steam engine. In the course of cooling the condenser gases of the steam engine the irrigating water is passed through a second pump driven by the steam engine and it is through this second pump that the pressure is raised sufficiently to allow for the necessary spraying fans. To improve the condenser efficiency part of the condensate or the ullage thereof is connected to one of the spray heads on the irrigation system in a venturi nozzle which thereby lowers the back pressure thereof. The lower portion of the condenser or the liquid part thereof is fed back through yet another condenser pump to the boiler to be regenerated into steam.

Jacobi, E.F.; Madden, M.R.

1982-01-12T23:59:59.000Z

326

Cooling system for a bearing of a turbine rotor  

SciTech Connect

In a gas turbine, a bore tube assembly radially inwardly of an aft bearing conveys cooling steam to the buckets of the turbine and returns the cooling steam to a return. To cool the bearing and thermally insulate the bearing from the cooling steam paths, a radiation shield is spaced from the bore tube assembly by a dead air gap. Additionally, an air passageway is provided between the radiation shield and the inner surface of an aft shaft forming part of the rotor. Air is supplied from an inlet for flow along the passage and radially outwardly through bores in the aft shaft disk to cool the bearing and insulate it from transfer of heat from the cooling steam.

Schmidt, Mark Christopher (Niskayuna, NY)

2002-01-01T23:59:59.000Z

327

Latch assembly  

DOE Patents (OSTI)

A latch assembly for releasably securing an article in the form of a canister within a container housing. The assembly includes a cam pivotally mounted on the housing wall and biased into the housing interior. The cam is urged into a disabled position by the canister as it enters the housing and a latch release plate maintains the cam disabled when the canister is properly seated in the housing. Upon displacement of the release plate, the cam snaps into latching engagement against the canister for securing the same within the housing. 2 figs.

Frederickson, J.R.; Harper, W.H.; Perez, R.

1984-08-17T23:59:59.000Z

328

Latch assembly  

DOE Patents (OSTI)

A latch assembly for releasably securing an article in the form of a canister within a container housing. The assembly includes a cam pivotally mounted on the housing wall and biased into the housing interior. The cam is urged into a disabled position by the canister as it enters the housing and a latch release plate maintains the cam disabled when the canister is properly seated in the housing. Upon displacement of the release plate, the cam snaps into latching engagement against the canister for securing the same within the housing.

Frederickson, James R. (Richland, WA); Harper, William H. (Richland, WA); Perez, Raymond (Lynnwood, WA)

1986-01-01T23:59:59.000Z

329

Measurement of Spent Fuel Assemblies - Overview of the Status of the Technology for Initiating Discussion at NATIONAL RESEARCH CENTRE KURCHATOV INSTITUTE June 2013  

SciTech Connect

This presentation provides an overview of the status of the technology for the measurement of the fissile material content of spent nuclear reactor fuel. The emphasis is on how the needs of the U.S. Nuclear Regulatory Commission and the International Atomic Energy Agency are met by the available technology and what more needs to be done in this area.

SISKIND B.; N /A

2013-06-03T23:59:59.000Z

330

Reactor Physics Assessment of the Inclusion of Unseparated Neptunium in MOX Reactor Fuel  

Science Conference Proceedings (OSTI)

Reducing the number of actinide separation streams in a spent fuel recovery process would reduce the cost and complexity of the process, and lower the quantity and numbers of solvents needed. It is more difficult and costly to separate Np and recombine it with Am-Cm prior to co-conversion than to simply co-strip it with the U-Pu-Np. Inclusion of the Np in mixed oxide (MOX) fuel for light water reactor (LWR) applications should not seriously affect the operating behavior of the reactor, nor should it pose insurmountable fuel design issues. In this work, the U, Pu, and Np from typical discharged and cooled PWR spent nuclear fuel are assumed to be used together in the preparation of MOX fuel for use in a pressurized water reactor (PWR). The reactor grade Pu isotopic vector is used in the model and the relative mass ratio of the Pu and Np content (Np/Pu mass is 0.061) from the cooled spent fuel is maintained but the overall Pu-Np MOX wt% is adjusted with respect to the U content (assumed to be at 0.25 wt% 235U enrichment) to offset reactivity and cycle length effects. The SCALE 5.1 scientific package (especially modules TRITON, NEWT, ORIGEN-S, ORIGEN-ARP) was used for the calculations presented in this paper. A typical Westinghouse 17x17 fuel assembly design was modeled at nominal PWR operating conditions. It was seen that U-Pu-Np MOX fuel with NpO2 and PuO2 representing 11.5wt% of the total MOX fuel would be similar to standard MOX fuel in which PuO2 is 9wt% of the fuel. The reactivity, isotopic composition, and neutron and ? sources, and the decay heat details for the discharged MOX fuel are presented and discussed in this paper.

Ellis, Ronald James [ORNL

2009-01-01T23:59:59.000Z

331

Furnace assembly  

DOE Patents (OSTI)

A method of and apparatus for heating test specimens to desired elevated temperatures for irradiation by a high energy neutron source. A furnace assembly is provided for heating two separate groups of specimens to substantially different, elevated, isothermal temperatures in a high vacuum environment while positioning the two specimen groups symmetrically at equivalent neutron irradiating positions.

Panayotou, Nicholas F. (Kennewick, WA); Green, Donald R. (Richland, WA); Price, Larry S. (Pittsburg, CA)

1985-01-01T23:59:59.000Z

332

Shipper/receiver difference verification of spent fuel by use of PDET  

Science Conference Proceedings (OSTI)

Spent fuel storage pools in most countries are rapidly approaching their design limits with the discharge of over 10,000 metric tons of heavy metal from global reactors. Countries like UK, France or Japan have adopted a closed fuel cycle by reprocessing spent fuel and recycling MOX fuel while many other countries opted for above ground interim dry storage for their spent fuel management strategy. Some countries like Finland and Sweden are already well on the way to setting up a conditioning plant and a deep geological repository for spent fuel. For all these situations, shipments of spent fuel are needed and the number of these shipments is expected to increase significantly. Although shipper/receiver difference (SRD) verification measurements are needed by IAEA when the recipient facility receives spent fuel, these are not being practiced to the level that IAEA has desired due to lack of a credible measurement methodology and instrument that can reliably perform these measurements to verify non-diversion of spent fuel during shipment and confirm facility operator declarations on the spent fuel. In this paper, we describe a new safeguards method and an associated instrument, Partial Defect Tester (PDET), which can detect pin diversion from Pressurized Water Reactor (PWR) Spent Fuel Assemblies in an in-situ condition. The PDET uses multiple tiny neutron and gamma detectors in the form of a cluster and a simple, yet highly precise, gravity-driven system to obtain underwater radiation measurements inside a Pressurized Water Reactor (PWR) spent fuel assembly. The method takes advantage of the PWR fuel design which contains multiple guide tubes which can be accessed from the top. The data obtained in such a manner can provide spatial distribution of neutron and gamma flux within a spent fuel assembly. Our simulation study as well as validation measurements indicated that the ratio of the gamma signal to the thermal neutron signal at each detector location normalized to the peak ratio of all the detector locations gives a unique signature that is sensitive to missing pins. The signature is principally dependent on the geometry of the detector locations, and little sensitive to enrichment or burn-up variations. A small variation in the fuel bundle, such as a few missing pins, changes the shape of the signature to enable detection. After verification of the non-diversion of spent fuel pins, the neutron signal and gamma signal are subsequently used to verify the consistency of the operator declaration on the fuel burn-up and cooling time. (authors)

Ham, Y. S.; Sitaraman, S. [Global Security Directorate, Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States)

2011-07-01T23:59:59.000Z

333

RADIATIVE AND PASSIVE COOLING  

E-Print Network (OSTI)

at the 3rd Annual Solar Heating and Cooling R&D Contractors'been supported by the Solar Heating and Cooling Research andof Energy. 3rd Annual Solar Heating and Cooling R&D

Martin, M.

2011-01-01T23:59:59.000Z

334

Closed Cooling Water Chemistry Guideline: Revision 2  

Science Conference Proceedings (OSTI)

This Closed Cooling Water Chemistry Guideline addresses the use of chemicals and monitoring methods to mitigate corrosion, fouling, and microbiological growth in the closed cooling-water (CCW) systems of nuclear and fossil-fueled power plants. The chemical additives used for these purposes depend on plant-design characteristics, water quality, operating parameters, and the specifications of the Nuclear Steam Supply System (NSSS) suppliers. The list of chemicals is not as extensive as that ...

2013-12-09T23:59:59.000Z

335

Air Breathing Direct Methanol Fuel Cell  

NLE Websites -- All DOE Office Websites (Extended Search)

Air Breathing Direct Methanol Fuel Cell Air Breathing Direct Methanol Fuel Cell Air Breathing Direct Methanol Fuel Cell An air breathing direct methanol fuel cell is provided with a membrane electrode assembly, a conductive anode assembly that is permeable to air and directly open to atmospheric air, and a conductive cathode assembly that is permeable to methanol and directly contacting a liquid methanol source. Available for thumbnail of Feynman Center (505) 665-9090 Email Air Breathing Direct Methanol Fuel Cell An air breathing direct methanol fuel cell is provided with a membrane electrode assembly, a conductive anode assembly that is permeable to air and directly open to atmospheric air, and a conductive cathode assembly that is permeable to methanol and directly contacting a liquid methanol

336

Energy Basics: Evaporative Cooling  

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

absorbent material. Evaporative cooling uses evaporated water to naturally and energy-efficiently cool. How Evaporative Coolers Work There are two types of evaporative...

337

Heat-activated cooling devices: A guidebook for general audiences  

DOE Green Energy (OSTI)

Heat-activated cooling is refrigeration or air conditioning driven by heat instead of electricity. A mill or processing facility can us its waste fuel to air condition its offices or plant; using waste fuel in this way can save money. The four basic types of heat-activated cooling systems available today are absorption cycle, desiccant system, steam jet ejector, and steam turbine drive. Each is discussed, along with cool storage and biomass boilers. Steps in determining the feasibility of heat-activated cooling are discussed, as are biomass conversion, system cost and integration, permits, and contractor selection. Case studies are given.

Wiltsee, G.

1994-02-01T23:59:59.000Z

338

Fuel cell gas management system  

SciTech Connect

A fuel cell gas management system including a cathode humidification system for transferring latent and sensible heat from an exhaust stream to the cathode inlet stream of the fuel cell; an anode humidity retention system for maintaining the total enthalpy of the anode stream exiting the fuel cell equal to the total enthalpy of the anode inlet stream; and a cooling water management system having segregated deionized water and cooling water loops interconnected by means of a brazed plate heat exchanger.

DuBose, Ronald Arthur (Marietta, GA)

2000-01-11T23:59:59.000Z

339

pH Adjustment of Power Plant Cooling Water with Flue Gas/Fly Ash  

to fossil fuel burning power plants to control mineral precipitation in cooling water. Flue gas, which is 10% CO2, could be diverted into a plant’s cooling water

340

Seal assembly  

DOE Patents (OSTI)

A seal assembly comprises a tube rotatable about its longitudinal axis and having two longitudinally spaced flanges projecting radially outwardly from the outer surface thereof. Slidably positioned against one of the flanges is a seal ring, and disposed between this seal ring and the other flange are two rings that are forced apart by springs, one of the latter rings being attached to a flexible wall.

Morgan, J.G.; Rennich, M.J.; Whatley, M.E.

1981-03-13T23:59:59.000Z

Note: This page contains sample records for the topic "fuel assemblies cooling" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


341

Dump assembly  

DOE Patents (OSTI)

A dump assembly having a fixed conduit and a rotatable conduit provided with overlapping plates, respectively, at their adjacent ends. The plates are formed with openings, respectively, normally offset from each other to block flow. The other end of the rotatable conduit is provided with means for securing the open end of a filled container thereto. Rotation of the rotatable conduit raises and inverts the container to empty the contents while concurrently aligning the conduit openings to permit flow of material therethrough.

Goldmann, Louis H. (Benton City, WA)

1986-01-01T23:59:59.000Z

342

Solar space cooling | Open Energy Information  

Open Energy Info (EERE)

cooling cooling Jump to: navigation, search Solarcooling.jpg Contents 1 Introduction 2 Solar Absorption Technology 3 Solar Desiccant Technology 4 Passive Solar Cooling 5 References Introduction There are many benefits to Solar Cooling systems. For one the sun is a clean energy resource that we should be using more often. It also produces no emissions and is replenished naturally, it reduces greenhouse gases, it saves the release of 1.6 lbs. of carbon dioxide (CO2) for each kilowatt-hour (kWh) produced, it saves the use of one-half gallon of water for each kWh of solar energy produced, it saves the release of other emissions that result from the burning of fossil fuels such as nitrogen oxides, sulfur dioxide or mercury and it provides customers with options to reduce their electric bills. But up to this point Solar Cooling systems are

343

Positioning and locking device for fuel pin to grid attachment  

DOE Patents (OSTI)

A positioning and locking device for fuel pin to grid attachment provides an inexpensive means of positively positioning and locking the individual fuel pins which make up the driver fuel assemblies used in nuclear reactors. The device can be adapted for use with a currently used attachment grid assembly design and insures that the pins remain in their proper position throughout the in-reactor life of the assembly. This device also simplifies fuel bundle assembly in that a complete row of fuel pins can be added to the bundle during each step of assembly.

Frick, Thomas M. (Irwin, PA); Wineman, Arthur L. (Greensburg, PA)

1976-01-01T23:59:59.000Z

344

Present status and development of PWR fuel in China  

Science Conference Proceedings (OSTI)

This paper describes the nuclear power plant fuel design, manufacture and R and D capability in Republic of China; the progress condition of Qinshan fuel assemblies and preparatory work for Guandong and other PWR's fuel assemblies; and the program of research and development of high performance fuel.

Shouhui, D.; Yinian, Z; Dingcang, T.

1988-01-01T23:59:59.000Z

345

NEUTRONIC REACTOR FUEL ELEMENT  

DOE Patents (OSTI)

A nuclear fuel element comprising a plurality of nuclear fuel bearing strips is presented. The strips are folded along their longitudinal axes to an angle of about 60 deg and are secured at each end by ferrule to form an elongated assembly suitable for occupying a cylindrical coolant channel.

Gurinsky, D.H.; Powell, R.W.; Fox, M.

1959-11-24T23:59:59.000Z

346

'Radio Wave Cooling' Offers New Twist on Laser Cooling  

Science Conference Proceedings (OSTI)

'Radio Wave Cooling' Offers New Twist on Laser Cooling. From NIST Tech Beat: September 13, 2007. ...

2013-07-08T23:59:59.000Z

347

Proceedings: Cooling Tower and Advanced Cooling Systems Conference  

Science Conference Proceedings (OSTI)

Cooling towers and associated systems performance strongly affects availability and heat rate in fossil and nuclear power plants. Papers presented at EPRI's 1994 Cooling Tower and Advanced Cooling Systems Conference discuss research results, industry experience, and case histories of cooling tower problems and solutions. Specific topics include cooling tower upgrades and retrofits, cooling tower performance, cooling tower fouling, and dry and hybrid cooling systems.

1995-03-09T23:59:59.000Z

348

Microchannel heat sink assembly  

DOE Patents (OSTI)

The present invention provides a microchannel heat sink with a thermal range from cryogenic temperatures to several hundred degrees centigrade. The heat sink can be used with a variety of fluids, such as cryogenic or corrosive fluids, and can be operated at a high pressure. The heat sink comprises a microchannel layer preferably formed of silicon, and a manifold layer preferably formed of glass. The manifold layer comprises an inlet groove and outlet groove which define an inlet manifold and an outlet manifold. The inlet manifold delivers coolant to the inlet section of the microchannels, and the outlet manifold receives coolant from the outlet section of the microchannels. In one embodiment, the manifold layer comprises an inlet hole extending through the manifold layer to the inlet manifold, and an outlet hole extending through the manifold layer to the outlet manifold. Coolant is supplied to the heat sink through a conduit assembly connected to the heat sink. A resilient seal, such as a gasket or an O-ring, is disposed between the conduit and the hole in the heat sink in order to provide a watertight seal. In other embodiments, the conduit assembly may comprise a metal tube which is connected to the heat sink by a soft solder. In still other embodiments, the heat sink may comprise inlet and outlet nipples. The present invention has application in supercomputers, integrated circuits and other electronic devices, and is suitable for cooling materials to superconducting temperatures. 13 figs.

Bonde, W.L.; Contolini, R.J.

1992-03-24T23:59:59.000Z

349

Methods of Conditioning Direct Methanol Fuel Cells  

while the catalyst on the anode surface is reduced. Surface oxides on the direct methanol fuel cell anode catalyst of the membrane electrode assembly are thereby reduced.

350

Review of Transmutation Fuel Studies  

SciTech Connect

The technology demonstration element of the Global Nuclear Energy Partnership (GNEP) program is aimed at demonstrating the closure of the fuel cycle by destroying the transuranic (TRU) elements separated from spent nuclear fuel (SNF). Multiple recycle through fast reactors is used for burning the TRU initially separated from light-water reactor (LWR) spent nuclear fuel. For the initial technology demonstration, the preferred option to demonstrate the closed fuel cycle destruction of TRU materials is a sodium-cooled fast reactor (FR) used as burner reactor. The sodium-cooled fast reactor represents the most mature sodium reactor technology available today. This report provides a review of the current state of development of fuel systems relevant to the sodium-cooled fast reactor. This report also provides a review of research and development of TRU-metal alloy and TRU-oxide composition fuels. Experiments providing data supporting the understanding of minor actinide (MA)-bearing fuel systems are summarized and referenced.

Jon Carmack (062056); Kemal O. Pasamehmetoglu (103171)

2008-01-01T23:59:59.000Z

351

OCCUPATIONAL COOLING TOWERS  

E-Print Network (OSTI)

HEALTH SCIENCES LIBRARY COOLING TOWERS EMPLOYEE HEALTH B C D F E CHILDREN'S ELEVATORS MEDICAL SCHOOL

Crews, Stephen

352

Computational fluid dynamics analyses of lateral heat conduction, coolant azimuthal mixing and heat transfer predictions in a BR2 fuel assembly geometry.  

SciTech Connect

To support the analyses related to the conversion of the BR2 core from highly-enriched (HEU) to low-enriched (LEU) fuel, the thermal-hydraulics codes PLTEMP and RELAP-3D are used to evaluate the safety margins during steady-state operation (PLTEMP), as well as after a loss-of-flow, loss-of-pressure, or a loss of coolant event (RELAP). In the 1-D PLTEMP and RELAP simulations, conduction in the azimuthal and axial directions is not accounted. The very good thermal conductivity of the cladding and the fuel meat and significant temperature gradients in the lateral directions (axial and azimuthal directions) could lead to a heat flux distribution that is significantly different than the power distribution. To evaluate the significance of the lateral heat conduction, 3-D computational fluid dynamics (CFD) simulations, using the CFD code STAR-CD, were performed. Safety margin calculations are typically performed for a hot stripe, i.e., an azimuthal region of the fuel plates/coolant channel containing the power peak. In a RELAP model, for example, a channel between two plates could be divided into a number of RELAP channels (stripes) in the azimuthal direction. In a PLTEMP model, the effect of azimuthal power peaking could be taken into account by using engineering factors. However, if the thermal mixing in the azimuthal direction of a coolant channel is significant, a stripping approach could be overly conservative by not taking into account this mixing. STAR-CD simulations were also performed to study the thermal mixing in the coolant. Section II of this document presents the results of the analyses of the lateral heat conduction and azimuthal thermal mixing in a coolant channel. Finally, PLTEMP and RELAP simulations rely on the use of correlations to determine heat transfer coefficients. Previous analyses showed that the Dittus-Boelter correlation gives significantly more conservative (lower) predictions than the correlations of Sieder-Tate and Petukhov. STAR-CD 3-D simulations were performed to compare heat transfer predictions from CFD and the correlations. Section III of this document presents the results of this analysis.

Tzanos, C. P.; Dionne, B. (Nuclear Engineering Division)

2011-05-23T23:59:59.000Z

353

The design of high power density annular fuel for LWRs  

E-Print Network (OSTI)

Fuel performance models have been developed to assess the performance of internally and externally cooled LWR annular fuel. Such fuel may be operated at 30-50% higher core power density than the current operating LWRs, and ...

Yuan, Yi, 1975-

2004-01-01T23:59:59.000Z

354

Nuclear fuel cycles for mid-century development  

E-Print Network (OSTI)

A comparative analysis of nuclear fuel cycles was carried out. Fuel cycles reviewed include: once-through fuel cycles in LWRs, PHWRs, HTGRs, and fast gas cooled breed and burn reactors; single-pass recycle schemes: plutonium ...

Parent, Etienne, 1977-

2003-01-01T23:59:59.000Z

355

Manufacturing of Low-Cost, Durable Membrane Electrode Assemblies Engineered for Rapid Conditioning - DOE Hydrogen and Fuel Cells Program FY 2012 Annual Progress Report  

NLE Websites -- All DOE Office Websites (Extended Search)

1 1 FY 2012 Annual Progress Report DOE Hydrogen and Fuel Cells Program F. Colin Busby W.L. Gore & Associates, Inc (Gore) Gore Electrochemical Technologies Team 201 Airport Road Elkton, MD 21921 Phone: (410) 392-3200 Email: CBusby@WLGore.com DOE Managers HQ: Nancy Garland Phone: (202) 586-5673 Email: Nancy.Garland@ee.doe.gov GO: Jesse Adams Phone: (720) 356-1421 Email: Jesse.Adams@go.doe.gov Contract Number: DE-FСЗ6-08G018052 Subcontractors: * UTC Power, South Windsor, CT * University of Delaware, Newark, DE (UD) * University of Tennessee, Knoxville, TN (UTK) Project Start Date: October 1, 2008 Project End Date: June 30, 2014

356

NREL: Advanced Power Electronics - Modeling of Cooling Technologies  

NLE Websites -- All DOE Office Websites (Extended Search)

Modeling of Cooling Technologies Improves Performance Modeling of Cooling Technologies Improves Performance Thermal modeling image of spray cooling of inverter chip surface shows the liquid breaking up into fine droplets that impinge on the liquid wall, which enhances the spacial uniformity of heat removal. Modeling Cooling Technologies-Spray Cooling The NREL advanced power electronics team is modeling cooling technologies that would enhance performance of the inverters and motors in hybrid-electric and fuel cell vehicles. The team is modeling two-phase spray cooling, jet impingement, and mini- and micro-channel cooling, and has successfully used Fluent software to show a good comparison between numerical models and published experimental data. Currently, the team is conducting modeling to simulate real life conditions such as those that

357

Pushrod assembly  

DOE Patents (OSTI)

A pushrod assembly including a carriage mounted on a shaft for movement therealong and carrying a pushrod engageable with a load to be moved is described. A magnet is mounted on a supporting bracket for movement along such shaft. Means are provided for adjustably spacing magnet away from the carriage to obtain a selected magnetic attractive or coupling force therebetween. Movement of the supporting bracket and the magnet carried thereby pulls the carriage along with it until the selected magnetic force is exceeded by a resistance load acting on the carriage.

Potter, J.D.

1984-03-30T23:59:59.000Z

358

Pushrod assembly  

DOE Patents (OSTI)

A pushrod assembly including a carriage mounted on a shaft for movement therealong and carrying a pushrod engageable with a load to be moved. A magnet is mounted on a supporting bracket for movement along such shaft. Means are provided for adjustably spacing said magnet away from said carriage to obtain a selected magnetic attractive or coupling force therebetween. Movement of the supporting bracket and the magnet carried thereby pulls the carriage along with it until the selected magnetic force is exceeded by a resistance load acting on the carriage.

Potter, Jerry D. (Kennewick, WA)

1987-01-01T23:59:59.000Z

359

Dump assembly  

DOE Patents (OSTI)

This is a claim for a dump assembly having a fixed conduit and a rotatable conduit provided with overlapping plates, respectively, at their adjacent ends. The plates are formed with openings, respectively, normally offset from each other to block flow. The other end of the rotatable conduit is provided with means for securing the open end of a filled container thereto. Rotation of the rotatable conduit raises and inverts the container to empty the contents while concurrently aligning the conduit openings to permit flow of material therethrough. 4 figs.

Goldmann, L.H.

1984-12-06T23:59:59.000Z

360

Hybrid Cooling Systems  

Science Conference Proceedings (OSTI)

Water consumption by power plants has become an increasingly contentious siting issue. In nearly all fossil-fired and nuclear plants, water for plant cooling is by far the greatest water requirement. Therefore, the use of water-conserving cooling systems such as dry or hybrid cooling is receiving increasing attention. This technology overview from the Electric Power Research Institute (EPRI) provides a brief introduction to hybrid cooling systems. As defined in the report, the term "hybrid cooling" refer...

2011-11-23T23:59:59.000Z

Note: This page contains sample records for the topic "fuel assemblies cooling" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


361

Cooling Plant Optimization Guide  

Science Conference Proceedings (OSTI)

Central cooling plants or district cooling systems account for 22 percent of energy costs for cooling commercial buildings. Improving the efficiency of central cooling plants will significantly impact peak demand and energy usage for both building owners and utilities. This guide identifies opportunities for optimizing a central cooling plant and provides a simplified optimization procedure. The guide focuses on plant optimization from the standpoint of minimizing energy costs and maximizing efficiencies...

1998-09-29T23:59:59.000Z

362

Secondary fuel delivery system  

SciTech Connect

A secondary fuel delivery system for delivering a secondary stream of fuel and/or diluent to a secondary combustion zone located in the transition piece of a combustion engine, downstream of the engine primary combustion region is disclosed. The system includes a manifold formed integral to, and surrounding a portion of, the transition piece, a manifold inlet port, and a collection of injection nozzles. A flowsleeve augments fuel/diluent flow velocity and improves the system cooling effectiveness. Passive cooling elements, including effusion cooling holes located within the transition boundary and thermal-stress-dissipating gaps that resist thermal stress accumulation, provide supplemental heat dissipation in key areas. The system delivers a secondary fuel/diluent mixture to a secondary combustion zone located along the length of the transition piece, while reducing the impact of elevated vibration levels found within the transition piece and avoiding the heat dissipation difficulties often associated with traditional vibration reduction methods.

Parker, David M. (Oviedo, FL); Cai, Weidong (Oviedo, FL); Garan, Daniel W. (Orlando, FL); Harris, Arthur J. (Orlando, FL)

2010-02-23T23:59:59.000Z

363

cooling degree-days - U.S. Energy Information Administration (EIA)  

U.S. Energy Information Administration (EIA)

Alternative Fuels. Includes hydropower, solar, wind, geothermal, biomass and ethanol. ... Cooling degree-days. There are no record(s) that match your search criteria.

364

pH Adjustment of Power Plant Cooling Water with Flue Gas/ Fly ...  

The discovery represents a cost-effective way to utilize materials indigenous to fossil fuel burning power platns to control mineral precipitation is cooling water.

365

Experience with non-fuel-bearing components in LWR (light-water reactor) fuel systems  

SciTech Connect

Many non-fuel-bearing components are so closely associated with the spent fuel assemblies that their integrity and behavior must be taken into consideration with the fuel assemblies, when handling spent fuel of planning waste management activities. Presented herein is some of the experience that has been gained over the past two decades from non-fuel-bearing components in light-water reactors (LWRs), both pressurized-water reactors (PWRs) and boiling-water reactors (BWRs). Among the most important of these components are the control rod systems, the absorber and burnable poison rods, and the fuel assembly channels. 15 refs., 5 figs., 2 tabs.

Bailey, W.J.; Berting, F.M.

1990-12-01T23:59:59.000Z

366

FUEL ELEMENT  

DOE Patents (OSTI)

A fuel element was developed for a gas cooled nuclear reactor. The element is constructed in the form of a compacted fuel slug including carbides of fissionable material in some cases with a breeder material carbide and a moderator which slug is disposed in a canning jacket of relatively impermeable moderator material. Such canned fuel slugs are disposed in an elongated shell of moderator having greater gas permeability than the canning material wherefore application of reduced pressure to the space therebetween causes gas diffusing through the exterior shell to sweep fission products from the system. Integral fission product traps and/or exterior traps as well as a fission product monitoring system may be employed therewith. (AEC)

Fortescue, P.; Zumwalt, L.R.

1961-11-28T23:59:59.000Z

367

Hallam fuel decladding. Program summary report  

Science Conference Proceedings (OSTI)

This report summarizes the program of decladding the 150 Hallam fuel assemblies, removal of the sodium, and the packaging and shipment of the recovered fuel to Savannah River for eventual reprocessing.

Dennison, W.F.

1980-04-01T23:59:59.000Z

368

Performance analysis of a Pb-Bi cooled fast reactor - PEACER-300 in proliferation resistance and transmutation aspects  

SciTech Connect

A design study of 850 MWt lead-bismuth cooled reactor cores is performed to maximize the transmutation of both TRU nuclides in homogeneous fuel pin and long-lived fission products in separate target pins. Transmutation of minor actinide under a closed recycling was analyzed with assumption that decontamination factors in pyro-reprocessing plant data be reasonably high. The optimized design parameter were chosen as of a flat core shape with 50 cm in active core height and 5 m in core diameter, loaded with 17 x 17 arrayed fuel assemblies. A pitch to diameter ratio is 2.2, operating coolant temperature range is 300 deg. C-400 deg. C, and core consists of 3 different enrichment zones with one year cycle length. In safety aspects, this core design satisfied large negative temperature feedback coefficients, and sufficient shutdown margin by primary shutdown system with 20 B{sub 4}C control assemblies and by secondary shutdown system with 40 w/o enriched 12 B{sub 4}C control assemblies. Performance of designed core showed a high transmutation capability with support ratio of 2.085 and less TEX values than other reactor types. Better proliferation resistance could be achieved than other reactor types. (authors)

Lim, J. Y.; Kim, M. H. [Dept. of Nuclear Engineering, Kyung Hee Univ., Yongin-shi, Gyeonggi-do, 449-701 (Korea, Republic of)

2006-07-01T23:59:59.000Z

369

Temperature and Burnup Correlated FCCI in U-10Zr Metallic Fuel  

SciTech Connect

Metallic fuels are proposed for use in advanced sodium cooled fast reactors. The experience basis for metallic fuels is extensive and includes development and qualification of fuels for the Experimental Breeder Reactor I, the Experimental Breeder Reactor II, FERMI-I, and the Fast Flux Test Facility (FFTF) reactors. Metallic fuels provide a number of advantages over other fuel types in terms of fabricability, performance, recyclability, and safety. Key to the performance of all nuclear fuel systems is the resistance to “breach” and subsequent release of fission products and fuel constituents to the primary coolant system of the nuclear power plant. In metallic fuel, the experience is that significant fuel-cladding chemical (FCCI) interaction occurs and becomes prevalent at high power-high temperature operation and ultimately leads to fuel pin breach and failure. Empirical relationships for metallic fuel pin failure have been developed from a large body of in-pile and out of pile research, development, and experimentation. It has been found that significant in-pile acceleration of the FCCI rate is experienced over similar condition out-of-pile experiments. The study of FCCI in metallic fuels has led to the quantification of in-pile failure rates to establish an empirical time and temperature dependent failure limit for fuel elements. Up until now the understanding of FCCI layer formation has been limited to data generated in EBR-II experiments. This dissertation provides new FCCI data extracted from the MFF-series of metallic fuel irradiations performed in the FFTF. These fuel assemblies contain valuable information on the formation of FCCI in metallic fuels at a variety of temperature and burnup conditions and in fuel with axial fuel height three times longer than EBR-II experiments. The longer fuel column in the FFTF and the fuel pins examined have significantly different flux, power, temperature, and FCCI profiles than that found in similar tests conducted in the EBR-II and study of the differences between the two fuel systems is critical for design of large advanced sodium cooled fast reactor systems. Comparing FCCI layer formation data between FFTF and EBR-II indicates that the same diffusion model can be used to represent the two systems when considering time, temperature, burnup history, and axial temperature and power profiles. This dissertation shows that FCCI formation peaks further below the top of the fuel column in FFTF experiments than has been observed in EBR-II experiments. The work provided in this dissertation will help forward the design of advanced metallic fuel systems for advanced sodium cooled fast reactors by allowing the prediction of FCCI layer formation in full length reactor designs. This will allow the accurate lifetime prediction of fuel performance capability for new advanced sodium cooled fast reactors with extended core designs.

William J. Carmack

2012-05-01T23:59:59.000Z

370

Non-Contact Sensor Evaluation for Bipolar Plate Manufacturing Process Control and Smart Assembly of Fuel Cell Stacks - DOE Hydrogen and Fuel Cells Program FY 2012 Annual Progress Report  

NLE Websites -- All DOE Office Websites (Extended Search)

4 4 DOE Hydrogen and Fuel Cells Program FY 2012 Annual Progress Report Eric Stanfield National Institute of Standards and Technology 100 Bureau Dr., MS 8211-8211 Gaithersburg, MD 20899 Phone: (301) 975-4882 Email: eric.stanfield@nist.gov DOE Managers HQ: Nancy Garland Phone: (202) 586-5673 Email: Nancy.Garland@ee.doe.gov GO: Jesse Adams Phone: (720) 356-1421 Email: Jesse.Adams@go.doe.gov Contract Number: DE-EE0001047 Project Start Date: October 1, 2009

371

Housing assembly for electric vehicle transaxle  

DOE Patents (OSTI)

Disclosed is a drive assembly (10) for an electrically powered vehicle (12). The assembly includes a transaxle (16) having a two-speed transmission (40) and a drive axle differential (46) disposed in a unitary housing assembly (38), an oil-cooled prime mover or electric motor (14) for driving the transmission input shaft (42), an adapter assembly (24) for supporting the prime mover on the transaxle housing assembly, and a hydraulic system (172) providing pressurized oil flow for cooling and lubricating the electric motor and transaxle and for operating a clutch (84) and a brake (86) in the transmission to shift between the two-speed ratios of the transmission. The adapter assembly allows the prime mover to be supported in several positions on the transaxle housing. The brake is spring-applied and locks the transmission in its low-speed ratio should the hydraulic system fail. The hydraulic system pump is driven by an electric motor (212) independent of the prime mover and transaxle.

Kalns, Ilmars (Northville, MI)

1981-01-01T23:59:59.000Z

372

Mox fuel arrangement for nuclear core  

DOE Patents (OSTI)

In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion. characteristics of the assembly.

Kantrowitz, Mark L. (Portland, CT); Rosenstein, Richard G. (Windsor, CT)

2001-05-15T23:59:59.000Z

373

MOX fuel arrangement for nuclear core  

DOE Patents (OSTI)

In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly. 38 figs.

Kantrowitz, M.L.; Rosenstein, R.G.

1998-10-13T23:59:59.000Z

374

MOX fuel arrangement for nuclear core  

DOE Patents (OSTI)

In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly.

Kantrowitz, Mark L. (Portland, CT); Rosenstein, Richard G. (Windsor, CT)

1998-01-01T23:59:59.000Z

375

MOX fuel arrangement for nuclear core  

DOE Patents (OSTI)

In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly.

Kantrowitz, Mark L. (Portland, CT); Rosenstein, Richard G. (Windsor, CT)

2001-07-17T23:59:59.000Z

376

Restaurateur designs and installs passive solar heating/cooling system  

SciTech Connect

An example of the use of passive solar heating and cooling systems by a Wisconsin restaurateur is discussed. The greenhouse effect is used on three sides of the restaurant's exterior walls. A dozen water-to-air electric heat pumps handle the restaurant's heating and cooling chores. The system doesn't require any fossil fuel for heating or cooling.

1983-04-01T23:59:59.000Z

377

Thermocouple assembly  

DOE Patents (OSTI)

A thermocouple assembly includes a thermocouple; a plurality of lead wires extending from the thermocouple; an insulating jacket extending along and enclosing the plurality of leads; and at least one internally sealed area within the insulating jacket to prevent fluid leakage along and within the insulating jacket. The invention also provides a method of preventing leakage of a fluid along and through an insulating jacket of a thermocouple including the steps of a) attaching a plurality of lead wires to a thermocouple; b) adding a heat sensitive pseudo-wire to extend along the plurality of lead wires; c) enclosing the lead wires and pseudo-wire inside an insulating jacket; d) locally heating axially spaced portions of the insulating jacket to a temperature which melts the pseudo-wire and fuses it with an interior surface of the jacket.

Thermos, Anthony Constantine (Greer, SC); Rahal, Fadi Elias (Easley, SC)

2002-01-01T23:59:59.000Z

378

Fuel cycle design and analysis of SABR: subrcritical advanced burner reactor.  

E-Print Network (OSTI)

??Various fuel cycles for a sodium-cooled, subcritical, fast reactor with a fusion neutron source for the transmutation of light water reactor spent fuel have been… (more)

Sommer, Christopher

2008-01-01T23:59:59.000Z

379

Experimental validation of the DARWIN2.3 package for fuel cycle applications  

Science Conference Proceedings (OSTI)

The DARWIN package, developed by the CEA and its French partners (AREVA and EDF) provides the required parameters for fuel cycle applications: fuel inventory, decay heat, activity, neutron, {gamma}, {alpha}, {beta} sources and spectrum, radiotoxicity. This paper presents the DARWIN2.3 experimental validation for fuel inventory and decay heat calculations on Pressurized Water Reactor (PWR). In order to validate this code system for spent fuel inventory a large program has been undertaken, based on spent fuel chemical assays. This paper deals with the experimental validation of DARWIN2.3 for the Pressurized Water Reactor (PWR) Uranium Oxide (UOX) and Mixed Oxide (MOX) fuel inventory calculation, focused on the isotopes involved in Burn-Up Credit (BUC) applications and decay heat computations. The calculation - experiment (C/E-1) discrepancies are calculated with the latest European evaluation file JEFF-3.1.1 associated with the SHEM energy mesh. An overview of the tendencies is obtained on a complete range of burn-up from 10 to 85 GWd/t (10 to 60 GWcVt for MOX fuel). The experimental validation of the DARWIN2.3 package for decay heat calculation is performed using calorimetric measurements carried out at the Swedish Interim Spent Fuel Storage Facility for Pressurized Water Reactor (PWR) assemblies, covering a large burn-up (20 to 50 GWd/t) and cooling time range (10 to 30 years). (authors)

San-Felice, L.; Eschbach, R.; Bourdot, P. [DEN, DER, CEA-Cadarache, F-13108 ST Paul-Lez-Durance (France); Tsilanizara, A.; Huynh, T. D. [DEN, DM2S, CEA-Saclay, F-91191 Gif sur Yvette (France); Ourly, H. [EDF, R and D, 1 av. General de Gaulle, F-92131 Clamart Cedex (France); Thro, J. F. [AREVA, Tour AREVA, F-92084 Paris la Defense (France)

2012-07-01T23:59:59.000Z

380

DOE Science Showcase - Cool roofs, cool research, at DOE | OSTI...  

Office of Scientific and Technical Information (OSTI)

Accelerator returns cool roof documents from 6 DOE Databases Executive Order on Sustainability Secretary Chu Announces Steps to Implement One Cool Roof Cool Roofs Lead to Cooler...

Note: This page contains sample records for the topic "fuel assemblies cooling" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


381

Actinide destruction and power peaking analysis in a 1000 MWt advanced burner reactor using moderated heterogeneous target assemblies  

SciTech Connect

The purpose of this research was to determine the effect of moderated heterogeneous subassemblies located in the core of a sodium-cooled fast reactor on minor actinide (MA) destruction rates over the lifecycle of the core. Additionally, particular emphasis was placed on the power peaking of the pins and the assemblies with the moderated targets as compared to standard unmoderated heterogeneous targets and a core without MA targets present. Power peaking analysis was performed on the target assemblies and on the fuel assemblies adjacent to the targets. The moderated subassemblies had a marked improvement in the overall destruction of heavy metals in the targets. The design with acceptable power peaking results had a 12.25% greater destruction of heavy metals than a similar ex-core unmoderated assembly. The increase in minor actinide destruction was most evident with americium where the moderated assemblies reduced the initial amount to less than 3% of the initial loading over a period of five years core residency. In order to take advantage of the high minor actinide destruction and minimize the power peaking effects, a hybrid scenario was devised where the targets resided ex-core in a moderated assembly for the first 506.9 effective full power days (EFPDs) and were moved to an in-core arrangement with the moderated targets removed for the remainder of the lifecycle. The hybrid model had an assembly and pin power peaking of less than 2.0 and a higher heavy metal and minor actinide destruction rate than the standard unmoderated heterogeneous targets either in-core or ex-core. The hybrid model has a 54.5% greater Am reduction over the standard ex-core model. It also had a 27.8% greater production of Cm and a 41.5% greater production of Pu than the standard ex-core model. The radiotoxicity of the targets in the hybrid design was 20% less than the discharged standard ex-core targets.

Kenneth Allen; Travis Knight; Samuel Bays

2011-05-01T23:59:59.000Z

382

Assessment of helical-cruciform fuel rods for high power density LWRs  

E-Print Network (OSTI)

In order to significantly increase the power density of Light Water Reactors (LWRs), the helical-cruciform (HC) fuel rod assembly has been proposed as an alternative to traditional fuel geometry. The HC assembly is a ...

Conboy, Thomas M

2010-01-01T23:59:59.000Z

383

Assessment of Evaporative Cooling Enhancement Methods for Air-Cooled Geothermal Power Plants: Preprint  

DOE Green Energy (OSTI)

Many binary-cycle geothermal power plants are air cooled because insufficient water is available to provide year-round water cooling. The performance of air-cooled geothermal plants is highly dependent on the dry bulb temperature of the air (much more so than fossil fuel plants that operate at higher boiler temperatures), and plant electric output can drop by 50% or more on hot summer days, compared to winter performance. This problem of reduced summer performance is exacerbated by the fact that electricity has a higher value in the summer. This paper describes a spreadsheet model that was developed to assess the cost and performance of four methods for using supplemental evaporative cooling to boost summer performance: (1) pre-cooling with spray nozzles, (2) pre-cooling with Munters media, (3) a hybrid combination of nozzles and Munters media, and (4) direct deluge cooling of the air-cooled condenser tubes. Although all four options show significant benefit, deluge cooling has the potential to be the most economic. However, issues of scaling and corrosion would need to be addressed.

Kutscher, C.; Costenaro, D.

2002-08-01T23:59:59.000Z

384

Self assembling magnetic tiles  

E-Print Network (OSTI)

Self assembly is an emerging technology in the field of manufacturing. Inspired by nature's ability to self assembly proteins from amino acids, this thesis attempts to demonstrate self assembly on the macro-scale. The ...

Rabl, Jessica A. (Jessica Ann)

2006-01-01T23:59:59.000Z

385

Space Heating and Cooling  

Energy.gov (U.S. Department of Energy (DOE))

A wide variety of technologies are available for heating and cooling homes and other buildings. In addition, many heating and cooling systems have certain supporting equipment in common, such as...

386

Methods of Conditioning Direct Methanol Fuel Cells  

NLE Websites -- All DOE Office Websites (Extended Search)

Methods of Conditioning Direct Methanol Fuel Cells Methods of Conditioning Direct Methanol Fuel Cells Methods of Conditioning Direct Methanol Fuel Cells Methods for conditioning the membrane electrode assembly of a direct methanol fuel cell ("DMFC") are disclosed. Available for thumbnail of Feynman Center (505) 665-9090 Email Methods of Conditioning Direct Methanol Fuel Cells Methods for conditioning the membrane electrode assembly of a direct methanol fuel cell ("DMFC") are disclosed. In a first method, an electrical current of polarity opposite to that used in a functioning direct methanol fuel cell is passed through the anode surface of the membrane electrode assembly. In a second method, methanol is supplied to an anode surface of the membrane electrode assembly, allowed to cross over the polymer

387

Fuel Cell with Metal Screen Flow-Field - Energy Innovation Portal  

The invention provides a simpler and more effective reactant and cooling flow distribution in biopolar plate fuel cells.

388

Cooling System Functions  

Science Conference Proceedings (OSTI)

...size Flow restrictions Heat exchanger size and design All of these factors must be considered. Every component in the cooling

389

Cooling Systems | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

technologies used in homes and buildings include ventilation, evaporative cooling, air conditioning, absorption cooling, and radiant cooling. Learn more about how these...

390

Cooling load estimation methods  

DOE Green Energy (OSTI)

Ongoing research on quantifying the cooling loads in residential buildings, particularly buildings with passive solar heating systems, is described. Correlations are described that permit auxiliary cooling estimates from monthly average insolation and weather data. The objective of the research is to develop a simple analysis method, useful early in design, to estimate the annual cooling energy required of a given building.

McFarland, R.D.

1984-01-01T23:59:59.000Z

391

Cooling Water System Optimization  

E-Print Network (OSTI)

During summer months, many manufacturing plants have to cut back in rates because the cooling water system is not providing sufficient cooling to support higher production rates. There are many low/no-cost techniques available to improve tower performance. To understand the importance of the optimization techniques, cooling tower theory will be discussed first.

Aegerter, R.

2005-01-01T23:59:59.000Z

392

Stochastic cooling in RHIC  

SciTech Connect

After the success of longitudinal stochastic cooling of bunched heavy ion beam in RHIC, transverse stochastic cooling in the vertical plane of Yellow ring was installed and is being commissioned with proton beam. This report presents the status of the effort and gives an estimate, based on simulation, of the RHIC luminosity with stochastic cooling in all planes.

Brennan,J.M.; Blaskiewicz, M. M.; Severino, F.

2009-05-04T23:59:59.000Z

393

Natural Cooling Retrofit  

E-Print Network (OSTI)

Substantial numbers of existing plants and buildings are found to depend solely upon Mechanical Cooling even though Natural Cooling techniques could be employed utilizing ambient air. Most of these facilities were constructed without Natural Cooling capability due to 'first cost' budget constraints when the cost and availability of energy were of little concern.

Fenster, L. C.; Grantier, A. J.

1981-01-01T23:59:59.000Z

394

Public Assembly Buildings  

U.S. Energy Information Administration (EIA) Indexed Site

Assembly Assembly Characteristics by Activity... Public Assembly Public assembly buildings are those in which people gather for social or recreational activities, whether in private or non-private meeting halls. Basic Characteristics [ See also: Equipment | Activity Subcategories | Energy Use ] Public Assembly Buildings... Most public assembly buildings were not large convention centers or entertainment arenas; about two-fifths fell into the smallest size category. About one-fifth of public assembly buildings were government-owned, mostly by local governments; examples of these types of public assembly buildings are libraries and community recreational facilities. Tables: Buildings and Size Data by Basic Characteristics Establishment, Employment, and Age Data by Characteristics

395

High-temperature gas-cooled reactors: preliminary safety and environmental information document. Volume IV  

SciTech Connect

Information is presented concerning medium-enriched uranium/thorium once-through fuel cycle; medium-enrichment uranium-233/thorium recycle fuel; high-enrichment uranium-235/thorium recycle (spiked) fuel cycle; high-enrichment uranium-233/thorium recycle (spiked) fuel cycle; and gas-turbine high-temperature gas-cooled reactor.

Not Available

1980-01-01T23:59:59.000Z

396

Fuel cell system with coolant flow reversal  

DOE Patents (OSTI)

Method and apparatus for cooling electrochemical fuel cell system components. Periodic reversal of the direction of flow of cooling fluid through a fuel cell stack provides greater uniformity and cell operational temperatures. Flow direction through a recirculating coolant fluid circuit is reversed through a two position valve, without requiring modulation of the pumping component.

Kothmann, Richard E. (Pittsburgh, PA)

1986-01-01T23:59:59.000Z

397

Guide to efficient unitary cooling equipment  

SciTech Connect

The universe of unitary cooling equipment is a large one; these systems are used in nearly forty percent of the residential and commercial buildings in the United States. Unitary cooling equipment is made up of off-the-shelf units: factory-assembled single or split systems, including air-source heat pumps and air conditioners. The efficiency of this class of cooling equipment has increased steadily in recent years, driven primarily by government standards. Although most of the units have efficiencies near the minimum federal standards, a significant number of models beat the standards by 10 to 30 percent. However, the larger the system, the narrower the range of efficiencies available and the fewer models available in the most efficient categories. For the buyer and the utility, this report reveals where to get efficiency information on current products, and a recommended purchasing process. It also examines the ratings, standards, and programs that can expand the number of high-efficiency models available.

Gregerson, J.; George, K.L.

1995-07-01T23:59:59.000Z

398

Filling Knowledge Gaps with Five Fuel Cycle Studies  

SciTech Connect

During FY 2010, five studies were conducted of technology families’ applicability to various fuel cycle strategies to fill in knowledge gaps in option space and to better understand trends and patterns. Here, a “technology family” is considered to be defined by a type of reactor and by selection of which actinides provide fuel. This report summarizes the higher-level findings; the detailed analyses and results are documented in five individual reports, as follows: • Advanced once through with uranium fuel in fast reactors (SFR), • Advanced once through (uranium fuel) or single recycle (TRU fuel) in high temperature gas cooled reactors (HTGR), • Sustained recycle with Th/U-233 in light water reactors (LWRs), • Sustained recycle with Th/U-233 in molten salt reactors (MSR), and • Several fuel cycle missions with Fusion-Fission Hybrid (FFH). Each study examined how the designated technology family could serve one or more designated fuel cycle missions, filling in gaps in overall option space. Each study contains one or more illustrative cases that show how the technology family could be used to meet a fuel cycle mission, as well as broader information on the technology family such as other potential fuel cycle missions for which insufficient information was available to include with an illustrative case. None of the illustrative cases can be considered as a reference, baseline, or nominal set of parameters for judging performance; the assessments were designed to assess areas of option space and were not meant to be optimized. There is no implication that any of the cases or technology families are necessarily the best way to meet a given fuel cycle mission. The studies provide five examples of 1-year fuel cycle assessments of technology families. There is reasonable coverage in the five studies of the performance areas of waste management and uranium utilization. The coverage of economics, safety, and proliferation resistance and physical protection in the five studies was spotty. Some studies did not have existing or past work to draw on in one or more of these areas. Resource constraints limited the amount of new analyses that could be performed. Little or no assessment was done of how soon any of the technologies could be deployed and therefore how quickly they could impact domestic or international fuel cycle performance. There were six common R&D needs, such as the value of advanced fuels, cladding, coating, and structure that would survive high neutron fluence. When a technology family is considered for use in a new fuel cycle mission, fuel cycle performance characteristics are dependent on both the design choices and the fuel cycle approach. For example, the use of the sodium-cooled fast reactor to provide recycle in either breeder or burner mode has been studied for decades, but the SFR could be considered for once-through fuel cycle with the physical reactor design and fuel management parameters changed. In addition, the sustained recycle with Th/U-233 in LWR could be achieved with a heterogeneous assembly and derated power density. Therefore, it may or may not be adjustable for other fuel cycle missions although a reactor intended for one fuel cycle mission is built. Simple parameter adjustment in applying a technology family to a new fuel cycle mission should be avoided and, if observed, the results viewed with caution.

Steven J. Piet; Jess Gehin; William Halsey; Temitope Taiwo

2010-11-01T23:59:59.000Z

399

PEM/SPE fuel cell  

DOE Patents (OSTI)

A PEM/SPE fuel cell is described including a membrane-electrode assembly (MEA) having a plurality of oriented filament embedded the face thereof for supporting the MEA and conducting current therefrom to contiguous electrode plates. 4 figs.

Grot, S.A.

1998-01-13T23:59:59.000Z

400

PEM/SPE fuel cell  

DOE Patents (OSTI)

A PEM/SPE fuel cell including a membrane-electrode assembly (MEA) having a plurality of oriented filament embedded the face thereof for supporting the MEA and conducting current therefrom to contiguous electrode plates.

Grot, Stephen Andreas (Henrietta, NY)

1998-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "fuel assemblies cooling" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


401

Air breathing direct methanol fuel cell  

DOE Patents (OSTI)

An air breathing direct methanol fuel cell is provided with a membrane electrode assembly, a conductive anode assembly that is permeable to air and directly open to atmospheric air, and a conductive cathode assembly that is permeable to methanol and directly contacting a liquid methanol source.

Ren, Xiaoming (Los Alamos, NM)

2002-01-01T23:59:59.000Z

402

Fuel cell stack arrangements  

DOE Patents (OSTI)

Arrangements of stacks of fuel cells and ducts, for fuel cells operating with separate fuel, oxidant and coolant streams. An even number of stacks are arranged generally end-to-end in a loop. Ducts located at the juncture of consecutive stacks of the loop feed oxidant or fuel to or from the two consecutive stacks, each individual duct communicating with two stacks. A coolant fluid flows from outside the loop, into and through cooling channels of the stack, and is discharged into an enclosure duct formed within the loop by the stacks and seals at the junctures at the stacks.

Kothmann, Richard E. (Churchill Boro, PA); Somers, Edward V. (Murrysville, PA)

1982-01-01T23:59:59.000Z

403

Cooling water distribution system  

DOE Patents (OSTI)

A passive containment cooling system for a nuclear reactor containment vessel. Disclosed is a cooling water distribution system for introducing cooling water by gravity uniformly over the outer surface of a steel containment vessel using an interconnected series of radial guide elements, a plurality of circumferential collector elements and collector boxes to collect and feed the cooling water into distribution channels extending along the curved surface of the steel containment vessel. The cooling water is uniformly distributed over the curved surface by a plurality of weirs in the distribution channels.

Orr, Richard (Pittsburgh, PA)

1994-01-01T23:59:59.000Z

404

Spent fuel integrity during transportation  

SciTech Connect

The conditions of recent shipments of light water reactor spent fuel were surveyed. The radioactivity level of cask coolant was examined in an attempt to find the effects of transportation on LWR fuel assemblies. Discussion included potential cladding integrity loss mechanisms, canning requirements, changes of radioactivity levels, and comparison of transportation in wet or dry media. Although integrity loss or degradation has not been identified, radioactivity levels usually increase during transportation, especially for leaking assemblies.

Funk, C.W.; Jacobson, L.D.

1980-01-01T23:59:59.000Z

405

CoolEarth formerly Cool Earth Solar | Open Energy Information  

Open Energy Info (EERE)

CoolEarth formerly Cool Earth Solar CoolEarth formerly Cool Earth Solar Jump to: navigation, search Name CoolEarth (formerly Cool Earth Solar) Place Livermore, California Zip 94550 Product CoolEarth is a concentrated PV developer using inflatable concentrators to focus light onto triple-junction cells. References CoolEarth (formerly Cool Earth Solar)[1] LinkedIn Connections CrunchBase Profile No CrunchBase profile. Create one now! This article is a stub. You can help OpenEI by expanding it. CoolEarth (formerly Cool Earth Solar) is a company located in Livermore, California . References ↑ "CoolEarth (formerly Cool Earth Solar)" Retrieved from "http://en.openei.org/w/index.php?title=CoolEarth_formerly_Cool_Earth_Solar&oldid=343892" Categories: Clean Energy Organizations

406

NREL: Vehicles and Fuels Research - Fuel Combustion Lab  

NLE Websites -- All DOE Office Websites (Extended Search)

Fuel Combustion Lab Fuel Combustion Lab NREL's Fuel Combustion Laboratory focuses on characterizing fuels at the molecular level. This information can then be used to understand and predict the fuel's effect on engine performance and emissions. By understanding the effects of fuel chemistry on ignition we can develop fuels that enable more efficient engine designs, using both today's technology and future advanced combustion concepts. This lab supports the distributed Renewable Fuels and Lubricants (ReFUEL) Laboratory, and the Biofuels activity. Photo of assembled IQT. Ignition Quality Tester The central piece of equipment in the Fuel Combustion Laboratory is the Ignition Quality Tester (IQT(tm)). The IQT(tm) is a constant volume combustion vessel that is used to study ignition properties of liquid

407

Vented Cavity Radiant Barrier Assembly And Method  

DOE Patents (OSTI)

A vented cavity radiant barrier assembly (2) includes a barrier (12), typically a PV module, having inner and outer surfaces (18, 22). A support assembly (14) is secured to the barrier and extends inwardly from the inner surface of the barrier to a building surface (14) creating a vented cavity (24) between the building surface and the barrier inner surface. A low emissivity element (20) is mounted at or between the building surface and the barrier inner surface. At least part of the cavity exit (30) is higher than the cavity entrance (28) to promote cooling air flow through the cavity.

Dinwoodie, Thomas L. (Piedmont, CA); Jackaway, Adam D. (Berkeley, CA)

2000-05-16T23:59:59.000Z

408

Gas-cooled reactor programs annual progress report for period ending December 31, 1972  

SciTech Connect

Information on the gas-cooled reactor development programs is presented concerning HTGR head-end fuel reprocessing development; fuel microsphere preparation development; fuel fabrication process development; HTGR fuel recycle pilot-plant studies; studies and evaluation of commercial HTGR fuel recycle plants; HTGR fuel element development; HTGR fuel irradiations and postirradiation evaluations; HTGR fuel chemistry, fuel integrity, and fission product behavior; reactions of HTGR core materials with steam; fission product behavior in HTGR coolant circuits; HTGR safety program plan and safety analysis; prestressed concrete pressure vessel development; GCFR irradiation experiments; and GCFR steam generator modeling studies. (DCC)

1974-03-01T23:59:59.000Z

409

Demonstration of Emitted-Neutron Computed Tomography to Count Fuel Pins  

SciTech Connect

In this paper, we report demonstration of emitted-neutron computed tomography using fast fission neutrons to infer the geometry of sources of special nuclear material (SNM) such as fuel pins. In a proof-of-concept measurement at the Idaho National Laboratory s (INL s) Zero Power Physics Reactor (ZPPR) facility, an array of unirradiated Pu MOX fuel rodlets in a soup can were imaged, and a bias defect consisting of a single rodlet containing Pu replaced by one containing depleted uranium (DU) was detected. The imaging system employed in the demonstration is based on a newly constructed array of pixelated neutron detectors that are suitable for arrangement in a close-packed imaging array and whose active volume consists of liquid scintillator EJ-309 which allows neutron-gamma discrimination via pulse shape to enable pure fast-neutron imaging. The imaging array was used along with a radial collimator aperture in order to perform high quality fast-neutron imaging where tomographic reconstruction of slices through an object resolve neutron sources similar in dimension to a fuel pellet, or about 1 cm. Measurements were performed at Oak Ridge National Laboratory (ORNL) with neutron sources in addition to those performed at the INL s ZPPR facility with Pu MOX fuel rodlets. An analogous capability to detect single-pin defects in spent fuel assemblies would be desirable, such as for safeguards verification measurements of spent fuel assemblies just prior to transferring them from the spent fuel cooling pool to long term dry cask storage. This paper describes the design and construction of the present imager, characterization measurements with neutron sources at ORNL, measurements with SNM at INL s ZPPR facility, and feasibility of building an analogous imager for spent fuel measurements.

Hausladen, Paul [ORNL; Blackston, Matthew A [ORNL; Brubaker, E. [Sandia National Laboratories (SNL); Chichester, David [Idaho National Laboratory (INL); Marleau, P. [Sandia National Laboratories (SNL); Newby, Robert Jason [ORNL

2012-01-01T23:59:59.000Z

410

Demonstration of Emitted-Neutron Computed Tomography to Count Fuel Pins  

Science Conference Proceedings (OSTI)

In this paper, we report demonstration of emitted-neutron computed tomography using fast fission neutrons to infer the geometry of sources of special nuclear material (SNM) such as fuel pins. In a proof-of-concept measurement at the Idaho National Laboratory’s (INL’s) Zero Power Physics Reactor (ZPPR) facility, an array of unirradiated Pu MOX fuel rodlets in a soup can were imaged, and a bias defect consisting of a single rodlet containing Pu replaced by one containing depleted uranium (DU) was detected. The imaging system employed in the demonstration is based on a newly constructed array of pixelated neutron detectors that are suitable for arrangement in a close-packed imaging array and whose active volume consists of liquid scintillator EJ-309 which allows neutron-gamma discrimination via pulse shape to enable pure fast-neutron imaging. The imaging array was used along with a radial collimator aperture in order to perform high quality fast-neutron imaging where tomographic reconstruction of slices through an object resolve neutron sources similar in dimension to a fuel pellet, or about 1 cm. Measurements were performed at Oak Ridge National Laboratory (ORNL) with neutron sources in addition to those performed at the INL’s ZPPR facility with Pu MOX fuel rodlets. An analogous capability to detect single-pin defects in spent fuel assemblies would be desirable, such as for safeguards verification measurements of spent fuel assemblies just prior to transferring them from the spent fuel cooling pool to long term dry cask storage. This paper describes the design and construction of the present imager, characterization measurements with neutron sources at ORNL, measurements with SNM at INL’s ZPPR facility, and feasibility of building an analogous imager for spent fuel measurements.

P. A. Hausladen; M. A. Blackston; E. Brubaker; D. L. Chichester; P. Marleau; R. J. Newby

2012-07-01T23:59:59.000Z

411

Stochastic cooling in RHIC  

SciTech Connect

The full 6-dimensional [x,x'; y,y'; z,z'] stochastic cooling system for RHIC was completed and operational for the FY12 Uranium-Uranium collider run. Cooling enhances the integrated luminosity of the Uranium collisions by a factor of 5, primarily by reducing the transverse emittances but also by cooling in the longitudinal plane to preserve the bunch length. The components have been deployed incrementally over the past several runs, beginning with longitudinal cooling, then cooling in the vertical planes but multiplexed between the Yellow and Blue rings, next cooling both rings simultaneously in vertical (the horizontal plane was cooled by betatron coupling), and now simultaneous horizontal cooling has been commissioned. The system operated between 5 and 9 GHz and with 3 x 10{sup 8} Uranium ions per bunch and produces a cooling half-time of approximately 20 minutes. The ultimate emittance is determined by the balance between cooling and emittance growth from Intra-Beam Scattering. Specific details of the apparatus and mathematical techniques for calculating its performance have been published elsewhere. Here we report on: the method of operation, results with beam, and comparison of results to simulations.

Brennan J. M.; Blaskiewicz, M.; Mernick, K.

2012-05-20T23:59:59.000Z

412

Fuel Reliability Guidelines: PWR Grid-to-Rod Fretting  

Science Conference Proceedings (OSTI)

Relative motion between the fuel rods and fuel assembly spacer grids can lead to excessive fuel rod wear and, in some cases, to fuel rod failure. Based on industry data, grid-to-rod-fretting (GTRF) has been the number one cause of fuel failures within the U.S. pressurized water reactor (PWR) fleet, accounting for more than 70% of all PWR leaking fuel assemblies. Most of the GTRF failures have occurred during an assembly's final cycle of operation when it is located on the periphery of the core adjacent t...

2008-07-28T23:59:59.000Z

413

A Blueprint for GNEP Advanced Burner Reactor Startup Fuel Fabrication Facility  

SciTech Connect

The purpose of this article is to identify the requirements and issues associated with design of GNEP Advanced Burner Reactor Fuel Facility. The report was prepared in support of providing data for preparation of a NEPA Environmental Impact Statement in support the U. S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives was to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu) -239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn these actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept was proposed to achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR was proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu was assumed to be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) was being considered for fabrication of WG Pu fuel for the ABR. It was estimated that the facility will provide the startup fuel for 10-15 years and would take 3 to 5 years to construct.

S. Khericha

2010-12-01T23:59:59.000Z

414

PRELIMINARY DATA CALL REPORT ADVANCED BURNER REACTOR START UP FUEL FABRICATION FACILITY  

SciTech Connect

The purpose of this report is to provide data for preparation of a NEPA Environmental Impact Statement in support the U. S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives is to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu) -239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn these actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept has been proposed to achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR is proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu will be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) is being considered for fabrication of WG Pu fuel for the ABR. This report is provided in response to ‘Data Call’ for the construction of startup fuel fabrication facility. It is anticipated that the facility will provide the startup fuel for 10-15 years and will take to 3 to 5 years to construct.

S. T. Khericha

2007-04-01T23:59:59.000Z

415

Spacer grid assembly and locking mechanism  

DOE Patents (OSTI)

A spacer grid assembly is disclosed for retaining a plurality of fuel rods in substantially parallel spaced relation, the spacer grids being formed with rhombic openings defining contact means for engaging from one to four fuel rods arranged in each opening, the spacer grids being of symmetric configuration with their rhombic openings being asymmetrically offset to permit inversion and relative rotation of the similar spacer grids for improved support of the fuel rods. An improved locking mechanism includes tie bars having chordal surfaces to facilitate their installation in slotted circular openings of the spacer grids, the tie rods being rotatable into locking engagement with the slotted openings.

Snyder, Jr., Harold J. (Rancho Santa Fe, CA); Veca, Anthony R. (San Diego, CA); Donck, Harry A. (San Diego, CA)

1982-01-01T23:59:59.000Z

416

Gas turbine cooling system  

SciTech Connect

A gas turbine engine (10) having a closed-loop cooling circuit (39) for transferring heat from the hot turbine section (16) to the compressed air (24) produced by the compressor section (12). The closed-loop cooling system (39) includes a heat exchanger (40) disposed in the flow path of the compressed air (24) between the outlet of the compressor section (12) and the inlet of the combustor (14). A cooling fluid (50) may be driven by a pump (52) located outside of the engine casing (53) or a pump (54) mounted on the rotor shaft (17). The cooling circuit (39) may include an orifice (60) for causing the cooling fluid (50) to change from a liquid state to a gaseous state, thereby increasing the heat transfer capacity of the cooling circuit (39).

Bancalari, Eduardo E. (Orlando, FL)

2001-01-01T23:59:59.000Z

417

HTGR spent fuel storage study  

SciTech Connect

This report documents a study of alternate methods of storing high-temperature gas-cooled reactor (HTGR) spent fuel. General requirements and design considerations are defined for a storage facility integral to a fuel recycle plant. Requirements for stand-alone storage are briefly considered. Three alternate water-cooled storage conceptual designs (plug well, portable well, and monolith) are considered and compared to a previous air-cooled design. A concept using portable storage wells in racks appears to be the most favorable, subject to seismic analysis and economic evaluation verification.

Burgoyne, R.M.; Holder, N.D.

1979-04-01T23:59:59.000Z

418

Integrated Radiation Transport and Thermo-Mechanics Simulation of a PWR Assembly  

SciTech Connect

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 towards incorporating an improved predictive nuclear fuel assembly modeling capability to accurately account for source terms, such as the neutron flux distribution, coolant conditions, and assembly mechanical stresses, of traditional (single-pin) nuclear fuel performance simulation. AMPFuel was used to model an entire 17 x 17 Pressurized Water Reactor (PWR) 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, top and bottom structural regions, and the upper and lower (neutron) reflector regions. The final full-assembly calculation was executed on Jaguar (Cray XT5) at the Oak Ridge Leadership Computing Facility using 40,000 cores in under 10 hours to model over 162 billion degrees of freedom for 10 loading steps.

Clarno, Kevin T [ORNL; Hamilton, Steven P [ORNL; Philip, Bobby [ORNL; Sampath, Rahul S [ORNL; Allu, Srikanth [ORNL; Berrill, Mark A [ORNL; Barai, Pallab [ORNL; Banfield, James E [ORNL

2012-01-01T23:59:59.000Z

419

Energy Basics: Absorption Cooling  

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

cooling. Other potential heat sources include propane, solar-heated water, or geothermal-heated water. Although mainly used in industrial or commercial settings, absorption...

420

Passive containment cooling system  

DOE Patents (OSTI)

A containment cooling system utilizes a naturally induced air flow and a gravity flow of water over the containment shell which encloses a reactor core to cool reactor core decay heat in two stages. When core decay heat is greatest, the water and air flow combine to provide adequate evaporative cooling as heat from within the containment is transferred to the water flowing over the same. The water is heated by heat transfer and then evaporated and removed by the air flow. After an initial period of about three to four days when core decay heat is greatest, air flow alone is sufficient to cool the containment.

Conway, Lawrence E. (Robinson Township, Allegheny County, PA); Stewart, William A. (Penn Hills Township, Allegheny County, PA)

1991-01-01T23:59:59.000Z

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