Sample records for jacketed fuel elements

  1. JACKETED FUEL ELEMENTS FOR GRAPHITE MODERATED REACTORS

    DOEpatents

    Szilard, L.; Wigner, E.P.; Creutz, E.C.

    1959-05-12

    Fuel elements for a heterogeneous, fluid cooled, graphite moderated reactor are described. The fuel elements are comprised of a body of natural uranium hermetically sealed in a jacket of corrosion resistant material. The jacket, which may be aluminum or some other material which is non-fissionable and of a type having a low neutron capture cross-section, acts as a barrier between the fissioning isotope and the coolant or moderator or both. The jacket minimizes the tendency of the moderator and coolant to become radioactive and/or contaminated by fission fragments from the fissioning isotope.

  2. JACKETED FUEL ELEMENT

    DOEpatents

    Wigner, E.P.; Szilard, L.; Creutz, E.C.

    1959-02-01

    These fuel elements are comprised of a homogeneous metallic uranium body completely enclosed and sealed in an aluminum cover. The uranium body and aluminum cover are bonded together by a layer of zinc located between them. The bonding layer serves to improve transfer of heat, provides an additional protection against corrosion of the uranium by the coolant, and also localizes any possible corrosion by preventing travel of corrosive material along the surface of the fuel element.

  3. JACKETED URANIUM SLUG

    DOEpatents

    Ohlinger, L.A.; Cooper, C.M.

    1958-10-01

    Fuel elements for nuclear reactors are described. Eacb fuel element is comprised of a solid cylindrical slug containing fissionable material enclosed within a fluid tight jacket of neutron permeable material such as aluminum. The jacket is provided with a flexible end cap and with a sealing member having a substantially fluid-tight fit within the jacket in tight abutment with the end cap and the end of the slug. A fluid passage is provided between the end of the slug and the cap whereby leakage fiuid is principally directed to the end of the slug. In this manner, any reaction between the fissionable material and fiuid which may take place occurs more rapidly at the end of the slug than along the sides between the slug and the jacket, thereby causing longitudinal expansion of the fuel element prior to radial expansion. The longitudinal expansion can be readily detected and the fuel element removed from the coolant tube before radial expansion causes it to become jammed in the tube.

  4. NEUTRONIC REACTOR AND FUEL ELEMENT THEREFOR

    DOEpatents

    Szilard, L.; Young, G.J.

    1958-03-01

    This patent relates to a reactor design of the type which employs solid fuel elements disposed in channels within the moderator through which channels and around the fuel elements is conveyed a coolant fiuid. The coolant channels are comprised of aluminum tubes extending through a solid moderator such as graphite and the fuel elements are comprised of an elongated solid body of natural uranium jacketed in an aluminum jacket with the ends thereof closed by aluminum caps of substantially greater thickness than the jacket was and in good thermal contact with the fuel material to facilitate the conduction of heat from the central portion of said ends to the coolant surrounding the fuel element to prevent overheating of said central portion.

  5. JACKETED FISSIONABLE MEMBER

    DOEpatents

    Boller, E.R.; Robinson, J.W.

    1960-09-13

    A fuel element design for a nuclear reactor is presented. The fuel element comprises a cylindrical fuel body having a portion of smaller diameter at each end thereof with an annular flange at the extreme ends of these portions of smaller diameter. An end cap fits over the ends of the fuel body and has an internal annular groove adapted to receive the flange. The fuel body and end caps are disposed in a cup-shaped jacket, a closure disc completing the enclosure of the fuel body, and tht caps are bonded over their entire periphery to the jacket.

  6. CUTTING AND WEDGING JACKET REMOVER

    DOEpatents

    Freedman, M.; Raynor, S.

    1959-04-01

    A tool is presented for stripping cladded jackets from fissionable fuel elements. The tool is a tube which fits closely around the jacket and which has two cutting edges at opposite sides of one end. These cutting edges are adjusted to penetrate only the jacket so that by moving the edges downward the jacket is cut into two pieces.

  7. METHOD OF JACKETING A FISSIONABLE BODY

    DOEpatents

    Creutz, E.C.

    1960-02-16

    A method for jacketing fuel elements is described. A fissionablc body ts fitted into a steel jacket, and a steel rimmed closure disk is inserted into the open end of the jacket. The jacket is then drawn through a die, aind the rim of the disk is welded to the jacket to form an impervious seal.

  8. CORROSION RESISTANT JACKETED METAL BODY

    DOEpatents

    Brugmann, E.W.

    1958-08-26

    S>Metal jacketed metallic bodies of the type used as feel elements fer nuclear reactors are presented. The fuel element is comprised of a plurality of jacketed cylindrical bodies joined in end to end abutting relationship. The abutting ends of the internal fissionable bodies are provided with a mating screw and thread means for joining the two together. The jacket material is of a corrosion resistant metal and overlaps the abutting ends of the internal bodies, thereby effectively sealing these bodies from contact with exteral reactive gases and liquids.

  9. WELDED JACKETED URANIUM BODY

    DOEpatents

    Gurinsky, D.H.

    1958-08-26

    A fuel element is presented for a neutronic reactor and is comprised of a uranium body, a non-fissionable jacket surrounding sald body, thu jacket including a portion sealed by a weld, and an inclusion in said sealed jacket at said weld of a fiux having a low neutron capture cross-section. The flux is provided by combining chlorine gas and hydrogen in the intense heat of-the arc, in a "Heliarc" welding muthod, to form dry hydrochloric acid gas.

  10. CORROSION RESISTANT JACKETED METAL BODY

    DOEpatents

    Brugmann, E.W.

    1958-08-26

    Jacketed metal bodies of the type used as fuel elements for nuclear reactors, which contain an internal elongated body of fissionable material jacketed in a corrosion resistant metal are described. The ends of the internal bodies are provided with screw threads having a tapered outer end. The jacket material overlaps the ends and extends into the tapered section of the screw threaded opening. Screw caps with a mating tapered section are screwed into the ends of the body to compress the jacket material in the tapered sections to provtde an effective seal against corrosive gases and liquids.

  11. FUEL ELEMENT

    DOEpatents

    Fortescue, P.; Zumwalt, L.R.

    1961-11-28

    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)

  12. REACTOR FUEL ELEMENTS TESTING CONTAINER

    DOEpatents

    Whitham, G.K.; Smith, R.R.

    1963-01-15

    This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

  13. FUEL ELEMENTS FOR NEUTRONIC REACTORS

    DOEpatents

    Foote, F.G.; Jette, E.R.

    1963-05-01

    A fuel element for a nuclear reactor is described that consists of a jacket containing a unitary core of fissionable material and a filling of a metal of the group consisting of sodium and sodium-potassium alloys. (AEC)

  14. NEUTRON REACTOR FUEL ELEMENT UTILIZING ZIRCONIUM-BASE ALLOYS

    DOEpatents

    Saller, H.A.; Keeler, J.R.; Szumachowski, E.R.

    1957-11-12

    This patent relates to clad fuel elements for use in neutronic reactors and is drawn to such a fuel element which consists of a core of fissionable material, comprised of an alloy of zirconium and U/sup 235/ enriched uranium, encased in a jacket of a binary zirconium-tin alloy in which the tin content ranges between 1 and 15% by weight.

  15. NUCLEAR REACTOR FUEL ELEMENT AND METHOD OF MANUFACTURE

    DOEpatents

    Brooks, H.

    1960-04-26

    A description is given for a fuel element comprising a body of uranium metal or an uranium compound dispersed in a matrix material made from magnesium, calcium, or barium and a stainless steel jacket enclosing the body.

  16. CORROSION RESISTANT JACKETED METAL BODY

    DOEpatents

    Brugmann, E.W.

    1958-08-26

    Reactor faul elements of the elongated cylindrical type which are jacketed in a corrosion resistant material are described. Each feel element is comprised of a plurality of jacketed cylinders of fissionable material in end to end abutting relationship, the jackets being welded together at their adjoining ends to retain the individual segments together and seat the interior of the jackets.

  17. Differences in the element contents between gunshot entry wounds with full-jacketed bullet and lead bullet.

    PubMed

    Wunnapuk, Klintean; Durongkadech, Piya; Minami, Takeshi; Ruangyuttikarn, Werawan; Tohno, Setsuko; Vichairat, Karnda; Azuma, Cho; Sribanditmongkol, Pongruk; Tohno, Yoshiyuki

    2007-01-01

    To elucidate characteristics of gunshot residues in gunshot entry wounds with full-jacketed and lead bullets, element contents in entry gunshot wounds and control skins were analyzed by inductively coupled plasma-atomic emission spectrometry (ICP-AES). It was found that a high content of Fe and Zn was deposited in the gunshot entry wounds with full-jacketed bullet, whereas a high content of Pb was deposited in the gunshot entry wounds with lead (unjacked) bullet. It should be noted that the content of Pb was significantly higher in the gunshot entry wounds with lead bullet than in those with full-jacketed bullet. Regarding the relationships among elements, it was found that there were significant direct correlations between Pb and either Sb or Ba contents in both gunshot entry wounds with full-jacketed and lead bullets. As Pb increased in both gunshot entry wounds, Sb and Ba also increased in the wounds.

  18. PROCESS FOR PRODUCING JACKETED BODIES

    DOEpatents

    Saller, H.A.

    1958-01-21

    A method is given for enclosing a metallic core within an outer protective jacket, such as in the production of fuel elements for neutronic reactors. The method comprises the steps of inserting the body of a first metal into an aperture in a frame of a second metal, placing a sheet of the second metal on each of opposite sides of the assembled body and frame, and bonding the sheets to the body and the frame and the body and the frame to one another.

  19. JacketSE: An Offshore Wind Turbine Jacket Sizing Tool; Theory Manual and Sample Usage with Preliminary Validation

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Damiani, Rick

    This manual summarizes the theory and preliminary verifications of the JacketSE module, which is an offshore jacket sizing tool that is part of the Wind-Plant Integrated System Design & Engineering Model toolbox. JacketSE is based on a finite-element formulation and on user-prescribed inputs and design standards' criteria (constraints). The physics are highly simplified, with a primary focus on satisfying ultimate limit states and modal performance requirements. Preliminary validation work included comparing industry data and verification against ANSYS, a commercial finite-element analysis package. The results are encouraging, and future improvements to the code are recommended in this manual.

  20. GEH-4-42, 47; Hot pressed, I and E cooled fuel element irradiation

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Neidner, R.

    1959-11-02

    In our continual effort to improve the present fuel elements which are irradiated in the numerous Hanford reactors, we have made what we believe to be a significant improvement in the hot pressing process for jacketing uranium fuel slugs. We are proposing a large scale evaluation testing program in the Hanford reactors but need the vital and basic information on the operating characteristics of this type slug under known and controlled operating conditions. We, therefore, have prepared two typical fuel slugs and will want them irradiated to about 1000 MWD/T exposure (this will require about four to five total cycles).

  1. Lightweight Vacuum Jacket for Cryogenic Insulation. Volume 1

    NASA Technical Reports Server (NTRS)

    Barclay, D. L.; Bell, J. E.; Brogren, E. W.; Straayer, J. W.

    1975-01-01

    The feasibility of producing a lightweight vacuum jacket using state-of-the-art technology and materials was examined. Design and analytical studies were made on a full-scale, orbital maneuvering system fuel tank. Preliminary design details were made for the tank assembly, including an optimized vacuum jacket and multilayer insulation system. A half-scale LH2 test model was designed and fabricated, and a force/stiffness proof test was conducted on the vacuum jacket. A vacuum leak rate of .000001 atmosphere ml of helium per second was measured, approximately 1500 hours of vacuum pressure were sustained, and 29 vacuum-pressure cycles were experienced prior to failure.

  2. Production test IP-544-A, irradiation of 1.6% enriched thick walled single tube elements in KER-1 and 2

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Kratzer, W.K.; Wise, M.J.

    1962-12-12

    The objective of this production test is to authorize the irradiation of coextruded Zr-2 jacketed thick walled 1.6% enriched tubular elements in KER loops 1 and 2 to evaluate the swelling behavior of fuel elements at high uranium temperatures Coextruded Zr-2 jacketed 1.6% enriched tubular fuel elements 1.79 inch OD, 0.97 inch ID, and 12 inches long will be irradiated KER loops 1 and 2 to exposures no greater than 2500 MWD/T.

  3. The effect on engine performance of change in jacket-water outlet temperature

    NASA Technical Reports Server (NTRS)

    Garlock, E A; Ellis, Greer

    1933-01-01

    Tests made on a Curtiss D-12 engine in the Altitude Laboratory at the Bureau of Standards show the following effects on engine performance of change in jacket-water outlet temperature: 1) Friction at all altitudes is a linear function of the jacket-water temperature, decreasing with increasing temperature. 2) The brake horsepower below an altitude of about 9,000 feet decreases, and at higher altitudes increases, with jacket-water temperature. 3) The brake specific fuel consumption tends to decrease, at all altitudes, with increasing jacket-water temperature. 4) The percentage change in brake power output is roughly equal to the algebraic sum of the percentage change in volumetric efficiency and mechanical efficiency.

  4. Variable pressure thermal insulating jacket

    DOEpatents

    Nelson, Paul A.; Malecha, Richard F.; Chilenskas, Albert A.

    1994-01-01

    A device for controlled insulation of a thermal device. The device includes a thermal jacket with a closed volume able to be evacuated to form an insulating jacket around the thermal source. A getter material is in communcation with the closed volume of the thermal jacket. The getter material can absorb and desorb a control gas to control gas pressure in the volume of the thermal jacket to control thermal conductivity in the thermal jacket.

  5. Lightweight Vacuum Jacket for Cryogenic Insulation - Appendices to Final Report. Volume 2

    NASA Technical Reports Server (NTRS)

    Barclay, D. L.; Bell, J. E.; Brogren, E. W.; Straayer, J. W.

    1975-01-01

    The feasibility is demonstrated of producing a lightweight vacuum jacket using state-of-the-art technology and materials. Design and analytical studies were made on an orbital maneuvering system fuel tank. Preliminary design details were completed for the tank assembly which included an optimized vacuum jacket and multilayered insulation system. A half-scale LH2 test model was designed and fabricated and a force/stiffness proof test was conducted on the vacuum jacket. A vacuum leak rate of 0.00001 was measured, approximately 1500 hours of vacuum pressure was sustained, and 29 vacuum pressure cycles were experienced prior to failure. For vol. 1, see N75-26192.

  6. Variable pressure thermal insulating jacket

    DOEpatents

    Nelson, P.A.; Malecha, R.F.; Chilenskas, A.A.

    1994-09-20

    A device for controlled insulation of a thermal device is disclosed. The device includes a thermal jacket with a closed volume able to be evacuated to form an insulating jacket around the thermal source. A getter material is in communication with the closed volume of the thermal jacket. The getter material can absorb and desorb a control gas to control gas pressure in the volume of the thermal jacket to control thermal conductivity in the thermal jacket. 10 figs.

  7. FUEL ELEMENT SUPPORT

    DOEpatents

    Wyman, W.L.

    1961-06-27

    The described cylindrical fuel element has longitudinally spaced sets of short longitudinal ribs circumferentially spaced from one another. The ribs support the fuel element in a coolant tube so that there is an annular space for coolant flow between the fuel element and the interior of the coolant tube. If the fuel element grows as a result of reactor operation, the circumferential distribution of the ribs maintains the uniformity of the annular space between the coolant tube and the fuel element, and the collapsibility of the ribs prevents the fuel element from becoming jammed in the coolant tube.

  8. 46 CFR 180.71 - Life jackets.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Life jackets. 180.71 Section 180.71 Shipping COAST GUARD... EQUIPMENT AND ARRANGEMENTS Ring Life Buoys and Life Jackets § 180.71 Life jackets. (a) An adult life jacket... life jackets equal to at least 10% of the number of persons permitted on board must be provided, or...

  9. 46 CFR 180.71 - Life jackets.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 7 2011-10-01 2011-10-01 false Life jackets. 180.71 Section 180.71 Shipping COAST GUARD... EQUIPMENT AND ARRANGEMENTS Ring Life Buoys and Life Jackets § 180.71 Life jackets. (a) An adult life jacket... life jackets equal to at least 10% of the number of persons permitted on board must be provided, or...

  10. 46 CFR 180.71 - Life jackets.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 7 2012-10-01 2012-10-01 false Life jackets. 180.71 Section 180.71 Shipping COAST GUARD... EQUIPMENT AND ARRANGEMENTS Ring Life Buoys and Life Jackets § 180.71 Life jackets. (a) An adult life jacket... life jackets equal to at least 10% of the number of persons permitted on board must be provided, or...

  11. 46 CFR 180.71 - Life jackets.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 7 2013-10-01 2013-10-01 false Life jackets. 180.71 Section 180.71 Shipping COAST GUARD... EQUIPMENT AND ARRANGEMENTS Ring Life Buoys and Life Jackets § 180.71 Life jackets. (a) An adult life jacket... life jackets equal to at least 10% of the number of persons permitted on board must be provided, or...

  12. 46 CFR 180.71 - Life jackets.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 7 2014-10-01 2014-10-01 false Life jackets. 180.71 Section 180.71 Shipping COAST GUARD... EQUIPMENT AND ARRANGEMENTS Ring Life Buoys and Life Jackets § 180.71 Life jackets. (a) An adult life jacket... life jackets equal to at least 10% of the number of persons permitted on board must be provided, or...

  13. Regeneratively Cooled Porous Media Jacket

    NASA Technical Reports Server (NTRS)

    Mungas, Greg (Inventor); Fisher, David J. (Inventor); London, Adam Pollok (Inventor); Fryer, Jack Merrill (Inventor)

    2013-01-01

    The fluid and heat transfer theory for regenerative cooling of a rocket combustion chamber with a porous media coolant jacket is presented. This model is used to design a regeneratively cooled rocket or other high temperature engine cooling jacket. Cooling jackets comprising impermeable inner and outer walls, and porous media channels are disclosed. Also disclosed are porous media coolant jackets with additional structures designed to transfer heat directly from the inner wall to the outer wall, and structures designed to direct movement of the coolant fluid from the inner wall to the outer wall. Methods of making such jackets are also disclosed.

  14. Advanced Main Combustion Chamber structural jacket strength analysis

    NASA Astrophysics Data System (ADS)

    Johnston, L. M.; Perkins, L. A.; Denniston, C. L.; Price, J. M.

    1993-04-01

    The structural analysis of the Advanced Main Combustion Chamber (AMCC) is presented. The AMCC is an advanced fabrication concept of the Space Shuttle Main Engine main combustion chamber (MCC). Reduced cost and fabrication time of up to 75 percent were the goals of the AMCC with cast jacket with vacuum plasma sprayed or platelet liner. Since the cast material for the AMCC is much weaker than the wrought material for the MCC, the AMCC is heavier and strength margins much lower in some areas. Proven hand solutions were used to size the manifolds cutout tee areas for combined pressure and applied loads. Detailed finite element strength analyses were used to size the manifolds, longitudinal ribs, and jacket for combined pressure and applied local loads. The design of the gimbal actuator strut attachment lugs were determined by finite element analyses and hand solutions.

  15. 46 CFR 117.71 - Life jackets.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 4 2011-10-01 2011-10-01 false Life jackets. 117.71 Section 117.71 Shipping COAST GUARD... OVERNIGHT ACCOMMODATIONS FOR MORE THAN 49 PASSENGERS LIFESAVING EQUIPMENT AND ARRANGEMENTS Ring Life Buoys and Life Jackets § 117.71 Life jackets. (a) An adult life jacket must be provided for each person...

  16. 46 CFR 117.71 - Life jackets.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Life jackets. 117.71 Section 117.71 Shipping COAST GUARD... OVERNIGHT ACCOMMODATIONS FOR MORE THAN 49 PASSENGERS LIFESAVING EQUIPMENT AND ARRANGEMENTS Ring Life Buoys and Life Jackets § 117.71 Life jackets. (a) An adult life jacket must be provided for each person...

  17. 46 CFR 117.71 - Life jackets.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 4 2012-10-01 2012-10-01 false Life jackets. 117.71 Section 117.71 Shipping COAST GUARD... OVERNIGHT ACCOMMODATIONS FOR MORE THAN 49 PASSENGERS LIFESAVING EQUIPMENT AND ARRANGEMENTS Ring Life Buoys and Life Jackets § 117.71 Life jackets. (a) An adult life jacket must be provided for each person...

  18. 46 CFR 117.71 - Life jackets.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 4 2013-10-01 2013-10-01 false Life jackets. 117.71 Section 117.71 Shipping COAST GUARD... OVERNIGHT ACCOMMODATIONS FOR MORE THAN 49 PASSENGERS LIFESAVING EQUIPMENT AND ARRANGEMENTS Ring Life Buoys and Life Jackets § 117.71 Life jackets. (a) An adult life jacket must be provided for each person...

  19. 46 CFR 117.71 - Life jackets.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 4 2014-10-01 2014-10-01 false Life jackets. 117.71 Section 117.71 Shipping COAST GUARD... OVERNIGHT ACCOMMODATIONS FOR MORE THAN 49 PASSENGERS LIFESAVING EQUIPMENT AND ARRANGEMENTS Ring Life Buoys and Life Jackets § 117.71 Life jackets. (a) An adult life jacket must be provided for each person...

  20. 46 CFR 180.75 - Life jacket lights.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Life jacket lights. 180.75 Section 180.75 Shipping COAST...) LIFESAVING EQUIPMENT AND ARRANGEMENTS Ring Life Buoys and Life Jackets § 180.75 Life jacket lights. (a) Each life jacket carried on a vessel on oceans, coastwise, or Great Lakes route, must have a life jacket...

  1. 46 CFR 117.75 - Life jacket lights.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Life jacket lights. 117.75 Section 117.75 Shipping COAST... Ring Life Buoys and Life Jackets § 117.75 Life jacket lights. (a) Each life jacket carried on a vessel on an oceans, coastwise, or Great Lakes route, must have a life jacket light approved in accordance...

  2. 46 CFR 180.75 - Life jacket lights.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 7 2011-10-01 2011-10-01 false Life jacket lights. 180.75 Section 180.75 Shipping COAST...) LIFESAVING EQUIPMENT AND ARRANGEMENTS Ring Life Buoys and Life Jackets § 180.75 Life jacket lights. (a) Each life jacket carried on a vessel on oceans, coastwise, or Great Lakes route, must have a life jacket...

  3. 46 CFR 117.75 - Life jacket lights.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 4 2011-10-01 2011-10-01 false Life jacket lights. 117.75 Section 117.75 Shipping COAST... Ring Life Buoys and Life Jackets § 117.75 Life jacket lights. (a) Each life jacket carried on a vessel on an oceans, coastwise, or Great Lakes route, must have a life jacket light approved in accordance...

  4. 46 CFR 117.75 - Life jacket lights.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 4 2014-10-01 2014-10-01 false Life jacket lights. 117.75 Section 117.75 Shipping COAST... Ring Life Buoys and Life Jackets § 117.75 Life jacket lights. (a) Each life jacket carried on a vessel on an oceans, coastwise, or Great Lakes route, must have a life jacket light approved in accordance...

  5. 46 CFR 117.75 - Life jacket lights.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 4 2012-10-01 2012-10-01 false Life jacket lights. 117.75 Section 117.75 Shipping COAST... Ring Life Buoys and Life Jackets § 117.75 Life jacket lights. (a) Each life jacket carried on a vessel on an oceans, coastwise, or Great Lakes route, must have a life jacket light approved in accordance...

  6. 46 CFR 117.75 - Life jacket lights.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 4 2013-10-01 2013-10-01 false Life jacket lights. 117.75 Section 117.75 Shipping COAST... Ring Life Buoys and Life Jackets § 117.75 Life jacket lights. (a) Each life jacket carried on a vessel on an oceans, coastwise, or Great Lakes route, must have a life jacket light approved in accordance...

  7. NUCLEAR REACTOR FUEL ELEMENT

    DOEpatents

    Wheelock, C.W.; Baumeister, E.B.

    1961-09-01

    A reactor fuel element utilizing fissionable fuel materials in plate form is described. This fuel element consists of bundles of fuel-bearing plates. The bundles are stacked inside of a tube which forms the shell of the fuel element. The plates each have longitudinal fins running parallel to the direction of coolant flow, and interspersed among and parallel to the fins are ribs which position the plates relative to each other and to the fuel element shell. The plate bundles are held together by thin bands or wires. The ex tended surface increases the heat transfer capabilities of a fuel element by a factor of 3 or more over those of a simple flat plate.

  8. Plating Patches On Heat-Exchanger Jackets

    NASA Technical Reports Server (NTRS)

    Loureiro, Henry; Kubik, Frank

    1989-01-01

    Permanent repairs made without welding. Technique used to repair nickel-alloy nozzle jacket of Space Shuttle main engine. Applicable to other metal heat-exchanger jackets with similar configurations. Does not require welding, brazing, soldering, or other operations involving high temperatures and consequent damage to surrounding areas. Portion of jacket around damaged area removed by grinding and polishing out to edges adjacent to tube/jacket braze bonds. Spaces between tubes filled with wax preventing contamination of spaces during subsequent plating.

  9. NUCLEAR REACTOR FUEL-BREEDER FUEL ELEMENT

    DOEpatents

    Currier, E.L. Jr.; Nicklas, J.H.

    1962-08-14

    A fuel-breeder fuel element was developed for a nuclear reactor wherein discrete particles of fissionable material are dispersed in a matrix of fertile breeder material. The fuel element combines the advantages of a dispersion type and a breeder-type. (AEC)

  10. METHOD AND APPARATUS FOR CONTROLLING NEUTRON DENSITY

    DOEpatents

    Wigner, E.P.; Young, G.J.; Weinberg, A.M.

    1961-06-27

    A neutronic reactor comprising a moderator containing uniformly sized and spaced channels and uniformly dimensioned fuel elements is patented. The fuel elements have a fissionable core and an aluminum jacket. The cores and the jackets of the fuel elements in the central channels of the reactor are respectively thinner and thicker than the cores and jackets of the fuel elements in the remainder of the reactor, producing a flattened flux.

  11. CONSTRUCTION OF NUCLEAR FUEL ELEMENTS

    DOEpatents

    Weems, S.J.

    1963-09-24

    >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)

  12. Neutronic fuel element fabrication

    DOEpatents

    Korton, George

    2004-02-24

    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

  13. Nuclear fuel element

    DOEpatents

    Zocher, Roy W.

    1991-01-01

    A nuclear fuel element and a method of manufacturing the element. The fuel element is comprised of a metal primary container and a fuel pellet which is located inside it and which is often fragmented. The primary container is subjected to elevated pressure and temperature to deform the container such that the container conforms to the fuel pellet, that is, such that the container is in substantial contact with the surface of the pellet. This conformance eliminates clearances which permit rubbing together of fuel pellet fragments and rubbing of fuel pellet fragments against the container, thus reducing the amount of dust inside the fuel container and the amount of dust which may escape in the event of container breach. Also, as a result of the inventive method, fuel pellet fragments tend to adhere to one another to form a coherent non-fragmented mass; this reduces the tendency of a fragment to pierce the container in the event of impact.

  14. 15. VIEW OF DUMMY FUEL ELEMENT ON FUEL ELEMENT HOLDER. ...

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

    15. VIEW OF DUMMY FUEL ELEMENT ON FUEL ELEMENT HOLDER. SHOWS AIR FORCE MAN AT EDGE OF TANK. INEL PHOTO NUMBER 65-6176, TAKEN NOVEMBER 10, 1965. - Idaho National Engineering Laboratory, Advanced Reentry Vehicle Fusing System, Scoville, Butte County, ID

  15. Postexercise Cooling Rates in 2 Cooling Jackets

    PubMed Central

    Brade, Carly; Dawson, Brian; Wallman, Karen; Polglaze, Ted

    2010-01-01

    Abstract Context: Cooling jackets are a common method for removing stored heat accumulated during exercise. To date, the efficiency and practicality of different types of cooling jackets have received minimal investigation. Objective: To examine whether a cooling jacket containing a phase-change material (PC17) results in more rapid postexercise cooling than a gel cooling jacket and a no-jacket (control) condition. Design: Randomized, counterbalanced design with 3 experimental conditions. Setting: Participants exercised at 75% V̇o2max workload in a hot climate chamber (temperature  =  35.0 ± 1.4°C, relative humidity  =  52 ± 4%) for 30 minutes, followed by postexercise cooling for 30 minutes in cool laboratory conditions (ambient temperature  =  24.9 ± 1.8°C, relative humidity  =  39% ± 10%). Patients or Other Participants: Twelve physically active men (age  =  21.3 ± 1.1 years, height  =  182.7 ± 7.1 cm, body mass  =  76.2 ± 9.5 kg, sum of 6 skinfolds  =  50.5 ± 6.9 mm, body surface area  =  1.98 ± 0.14 m2, V̇o2max  =  49.0 ± 7.0 mL·kg−1·min−1) participated. Intervention(s): Three experimental conditions, consisting of a PC17 jacket, a gel jacket, and no jacket. Main Outcome Measure(s): Core temperature (TC), mean skin temperature (TSk), and TC cooling rate (°C/min). Results: Mean peak TC postexercise was 38.49 ± 0.42°C, 38.57 ± 0.41°C, and 38.55 ± 0.40°C for the PC17 jacket, gel jacket, and control conditions, respectively. No differences were observed in peak TC cooling rates among the PC17 jacket (0.038 ± 0.007°C/min), gel jacket (0.040 ± 0.009°C/min), and control (0.034 ± 0.010°C/min, P > .05) conditions. Between trials, no differences were calculated for mean TSk cooling. Conclusions: Similar cooling rates for all 3 conditions indicate that there is no benefit associated with wearing the PC17 or gel jacket. PMID:20210620

  16. FUEL ELEMENT

    DOEpatents

    Bean, R.W.

    1963-11-19

    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)

  17. Fire Protection Jacket

    NASA Technical Reports Server (NTRS)

    1991-01-01

    NERAC, Inc., Tolland, CT, aided Paul Monroe Engineering, Orange, CA, in the development of their PC1200 Series Fire Protection Jacket that protects the oil conduit system on an offshore drilling platform from the intense hydrocarbon fires that cause buckling and could cause structural failure of the platform. The flame-proof jacketing, which can withstand temperatures of 2000 degrees Fahrenheit for four hours or more, was developed from a combination of ceramic cloth (similar to the ceramic in Space Shuttle tiles), and laminates used in space suits.

  18. 46 CFR 117.78 - Stowage of life jackets.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 4 2011-10-01 2011-10-01 false Stowage of life jackets. 117.78 Section 117.78 Shipping... Ring Life Buoys and Life Jackets § 117.78 Stowage of life jackets. (a) General. Unless otherwise stated in this section, life jackets must be stored in convenient places distributed throughout...

  19. 46 CFR 117.78 - Stowage of life jackets.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Stowage of life jackets. 117.78 Section 117.78 Shipping... Ring Life Buoys and Life Jackets § 117.78 Stowage of life jackets. (a) General. Unless otherwise stated in this section, life jackets must be stored in convenient places distributed throughout...

  20. 46 CFR 117.78 - Stowage of life jackets.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 4 2013-10-01 2013-10-01 false Stowage of life jackets. 117.78 Section 117.78 Shipping... Ring Life Buoys and Life Jackets § 117.78 Stowage of life jackets. (a) General. Unless otherwise stated in this section, life jackets must be stored in convenient places distributed throughout...

  1. 46 CFR 117.78 - Stowage of life jackets.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 4 2012-10-01 2012-10-01 false Stowage of life jackets. 117.78 Section 117.78 Shipping... Ring Life Buoys and Life Jackets § 117.78 Stowage of life jackets. (a) General. Unless otherwise stated in this section, life jackets must be stored in convenient places distributed throughout...

  2. 46 CFR 117.78 - Stowage of life jackets.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 4 2014-10-01 2014-10-01 false Stowage of life jackets. 117.78 Section 117.78 Shipping... Ring Life Buoys and Life Jackets § 117.78 Stowage of life jackets. (a) General. Unless otherwise stated in this section, life jackets must be stored in convenient places distributed throughout...

  3. 46 CFR 180.78 - Stowage of life jackets.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Stowage of life jackets. 180.78 Section 180.78 Shipping...) LIFESAVING EQUIPMENT AND ARRANGEMENTS Ring Life Buoys and Life Jackets § 180.78 Stowage of life jackets. (a) General. Unless otherwise stated in this section, life jackets must be stored in convenient places...

  4. 46 CFR 180.78 - Stowage of life jackets.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 7 2011-10-01 2011-10-01 false Stowage of life jackets. 180.78 Section 180.78 Shipping...) LIFESAVING EQUIPMENT AND ARRANGEMENTS Ring Life Buoys and Life Jackets § 180.78 Stowage of life jackets. (a) General. Unless otherwise stated in this section, life jackets must be stored in convenient places...

  5. 46 CFR 180.78 - Stowage of life jackets.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 7 2014-10-01 2014-10-01 false Stowage of life jackets. 180.78 Section 180.78 Shipping...) LIFESAVING EQUIPMENT AND ARRANGEMENTS Ring Life Buoys and Life Jackets § 180.78 Stowage of life jackets. (a) General. Unless otherwise stated in this section, life jackets must be stored in convenient places...

  6. 46 CFR 180.78 - Stowage of life jackets.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 7 2013-10-01 2013-10-01 false Stowage of life jackets. 180.78 Section 180.78 Shipping...) LIFESAVING EQUIPMENT AND ARRANGEMENTS Ring Life Buoys and Life Jackets § 180.78 Stowage of life jackets. (a) General. Unless otherwise stated in this section, life jackets must be stored in convenient places...

  7. 46 CFR 180.78 - Stowage of life jackets.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 7 2012-10-01 2012-10-01 false Stowage of life jackets. 180.78 Section 180.78 Shipping...) LIFESAVING EQUIPMENT AND ARRANGEMENTS Ring Life Buoys and Life Jackets § 180.78 Stowage of life jackets. (a) General. Unless otherwise stated in this section, life jackets must be stored in convenient places...

  8. 46 CFR 185.516 - Life jacket placards.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Life jacket placards. 185.516 Section 185.516 Shipping...) OPERATIONS Preparations for Emergencies § 185.516 Life jacket placards. (a) Placards containing instructions for the donning and use of the life jackets aboard the vessel must be posted in conspicuous places...

  9. 46 CFR 185.516 - Life jacket placards.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 7 2011-10-01 2011-10-01 false Life jacket placards. 185.516 Section 185.516 Shipping...) OPERATIONS Preparations for Emergencies § 185.516 Life jacket placards. (a) Placards containing instructions for the donning and use of the life jackets aboard the vessel must be posted in conspicuous places...

  10. 46 CFR 185.516 - Life jacket placards.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 7 2014-10-01 2014-10-01 false Life jacket placards. 185.516 Section 185.516 Shipping...) OPERATIONS Preparations for Emergencies § 185.516 Life jacket placards. (a) Placards containing instructions for the donning and use of the life jackets aboard the vessel must be posted in conspicuous places...

  11. 46 CFR 185.516 - Life jacket placards.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 7 2012-10-01 2012-10-01 false Life jacket placards. 185.516 Section 185.516 Shipping...) OPERATIONS Preparations for Emergencies § 185.516 Life jacket placards. (a) Placards containing instructions for the donning and use of the life jackets aboard the vessel must be posted in conspicuous places...

  12. 46 CFR 185.516 - Life jacket placards.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 7 2013-10-01 2013-10-01 false Life jacket placards. 185.516 Section 185.516 Shipping...) OPERATIONS Preparations for Emergencies § 185.516 Life jacket placards. (a) Placards containing instructions for the donning and use of the life jackets aboard the vessel must be posted in conspicuous places...

  13. 46 CFR 122.516 - Life jacket placards.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Life jacket placards. 122.516 Section 122.516 Shipping... Emergencies § 122.516 Life jacket placards. (a) Placards containing instructions for the donning and use of the life jackets aboard the vessel must be posted in conspicuous places that are regularly accessible...

  14. 46 CFR 122.516 - Life jacket placards.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 4 2011-10-01 2011-10-01 false Life jacket placards. 122.516 Section 122.516 Shipping... Emergencies § 122.516 Life jacket placards. (a) Placards containing instructions for the donning and use of the life jackets aboard the vessel must be posted in conspicuous places that are regularly accessible...

  15. 46 CFR 122.516 - Life jacket placards.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 4 2012-10-01 2012-10-01 false Life jacket placards. 122.516 Section 122.516 Shipping... Emergencies § 122.516 Life jacket placards. (a) Placards containing instructions for the donning and use of the life jackets aboard the vessel must be posted in conspicuous places that are regularly accessible...

  16. 46 CFR 122.516 - Life jacket placards.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 4 2014-10-01 2014-10-01 false Life jacket placards. 122.516 Section 122.516 Shipping... Emergencies § 122.516 Life jacket placards. (a) Placards containing instructions for the donning and use of the life jackets aboard the vessel must be posted in conspicuous places that are regularly accessible...

  17. 46 CFR 122.516 - Life jacket placards.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 4 2013-10-01 2013-10-01 false Life jacket placards. 122.516 Section 122.516 Shipping... Emergencies § 122.516 Life jacket placards. (a) Placards containing instructions for the donning and use of the life jackets aboard the vessel must be posted in conspicuous places that are regularly accessible...

  18. 46 CFR 180.75 - Life jacket lights.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 7 2013-10-01 2013-10-01 false Life jacket lights. 180.75 Section 180.75 Shipping COAST...) LIFESAVING EQUIPMENT AND ARRANGEMENTS Ring Life Buoys and Life Jackets § 180.75 Life jacket lights. (a) Each... light approved in accordance with § 161.012 in subchapter Q of this chapter, or other standard specified...

  19. 46 CFR 180.75 - Life jacket lights.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 7 2014-10-01 2014-10-01 false Life jacket lights. 180.75 Section 180.75 Shipping COAST...) LIFESAVING EQUIPMENT AND ARRANGEMENTS Ring Life Buoys and Life Jackets § 180.75 Life jacket lights. (a) Each... light approved in accordance with § 161.012 in subchapter Q of this chapter, or other standard specified...

  20. 46 CFR 180.75 - Life jacket lights.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 7 2012-10-01 2012-10-01 false Life jacket lights. 180.75 Section 180.75 Shipping COAST...) LIFESAVING EQUIPMENT AND ARRANGEMENTS Ring Life Buoys and Life Jackets § 180.75 Life jacket lights. (a) Each... light approved in accordance with § 161.012 in subchapter Q of this chapter, or other standard specified...

  1. NEUTRONIC REACTOR FUEL ELEMENT

    DOEpatents

    Shackleford, M.H.

    1958-12-16

    A fuel element possessing good stability and heat conducting properties is described. The fuel element comprises an outer tube formed of material selected from the group consisting of stainhess steel, V, Ti. Mo. or Zr, a fuel tube concentrically fitting within the outer tube and containing an oxide of an isotope selected from the group consisting of U/sup 235/, U/sup 233/, and Pu/sup 239/, and a hollow, porous core concentrically fitting within the fuel tube and formed of an oxide of an element selected from the group consisting of Mg, Be, and Zr.

  2. Water jacket for solid particle solar receiver

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Wasyluk, David T.

    A solar receiver includes: water jacket panels each having a light-receiving side and a back side with a watertight sealed plenum defined in-between; light apertures passing through the watertight sealed plenums to receive light from the light-receiving sides of the water jacket panels; a heat transfer medium gap defined between the back sides of the water jacket panels and a cylindrical back plate; and light channeling tubes optically coupled with the light apertures and extending into the heat transfer medium gap. In some embodiments ends of the light apertures at the light receiving side of the water jacket panel aremore » welded together to define at least a portion of the light-receiving side. A cylindrical solar receiver may be constructed using a plurality of such water jacket panels arranged with their light-receiving sides facing outward.« less

  3. FUEL-BREEDER FUEL ELEMENT FOR NUCLEAR REACTOR

    DOEpatents

    Abbott, W.E.; Balent, R.

    1958-09-16

    A fuel element design to facilitate breeding reactor fuel is described. The fuel element is comprised of a coatainer, a central core of fertile material in the container, a first bonding material surrounding the core, a sheet of fissionable material immediately surrounding the first bonding material, and a second bonding material surrounding the fissionable material and being in coniact with said container.

  4. Jacketed lamp bulb envelope

    DOEpatents

    MacLennan, Donald A.; Turner, Brian P.; Gitsevich, Aleksandr; Bass, Gary K.; Dolan, James T.; Kipling, Kent; Kirkpatrick, Douglas A.; Leng, Yongzhang; Levin, Izrail; Roy, Robert J.; Shanks, Bruce; Smith, Malcolm; Trimble, William C.; Tsai, Peter

    2001-01-01

    A jacketed lamp bulb envelope includes a ceramic cup having an open end and a partially closed end, the partially closed end defining an aperture, a lamp bulb positioned inside the ceramic cup abutting the aperture, and a reflective ceramic material at least partially covering a portion of the bulb not abutting the aperture. The reflective ceramic material may substantially fill an interior volume of the ceramic cup not occupied by the bulb. The ceramic cup may include a structural feature for aiding in alignment of the jacketed lamp bulb envelope in a lamp. The ceramic cup may include an external flange about a periphery thereof. One example of a jacketed lamp bulb envelope includes a ceramic cup having an open end and a closed end, a ceramic washer covering the open end of the ceramic cup, the washer defining an aperture therethrough, a lamp bulb positioned inside the ceramic cup abutting the aperture, and a reflective ceramic material filling an interior volume of the ceramic cup not occupied by the bulb. A method of packing a jacketed lamp bulb envelope of the type comprising a ceramic cup with a lamp bulb disposed therein includes the steps of filling the ceramic cup with a flowable slurry of reflective material, and applying centrifugal force to the cup to pack the reflective material therein.

  5. Ported jacket for use in deformation measurement apparatus

    DOEpatents

    Wagner, L.A.; Senseny, P.E.; Mellegard, K.D.; Olsberg, S.B.

    1990-03-06

    A device for allowing deformation measurement of a jacketed specimen when the specimen is loaded includes an elastomeric specimen container or jacket surrounding a specimen while the specimen is being loaded by a test apparatus. The specimen jacket wall is compressible, and the wall follows and allows deformation of the specimen. The jacket wall of compressible material is provided with at least one opening and a thin layer or shim of substantially non-compressible (metal) material which covers and seals this opening. An extensometer is then positioned with its specimen engaging contact members engaging the substantially non-compressible material to measure the deformation of the specimen when the specimen is loaded, without compressibility effects of the jacket. 9 figs.

  6. FUEL ELEMENT FOR NUCLEAR REACTORS

    DOEpatents

    Dickson, J.J.

    1963-09-24

    A method is described whereby fuel tubes or pins are cut, loaded with fuel pellets and a heat transfer medium, sealed at each end with slotted fittings, and assembled into a rectangular tube bundle to form a fuel element. The tubes comprising the fuel element are laterally connected between their ends by clips and tabs to form a linear group of spaced parallel tubes, which receive their vertical support by resting on a grid. The advantages of this method are that it permits elimination of structural material (e.g., fuel-element cans) within the reactor core, and removal of at least one fuel pin from an element and replacement thereof so that a burnable poison may be utilized during the core lifetime. (AEC)

  7. Combustion system for hybrid solar fossil fuel receiver

    DOEpatents

    Mehos, Mark S.; Anselmo, Kenneth M.; Moreno, James B.; Andraka, Charles E.; Rawlinson, K. Scott; Corey, John; Bohn, Mark S.

    2004-05-25

    A combustion system for a hybrid solar receiver comprises a pre-mixer which combines air and fuel to form an air-fuel mixture. The mixture is introduced tangentially into a cooling jacket. A burner plenum is fluidically connected to the cooling jacket such that the burner plenum and the cooling jacket are arranged in thermal contact with one another. The air-fuel mixture flows through the cooling jacket cooling the burner plenum to reduce pre-ignition of the air-fuel mixture in the burner plenum. A combustion chamber is operatively associated with and open to the burner plenum to receive the air-fuel mixture from the burner plenum. An igniter is operatively positioned in the combustion chamber to combust the air-fuel mixture, releasing heat. A recuperator is operatively associated with the burner plenum and the combustion chamber and pre-heats the air-fuel mixture in the burner plenum with heat from the combustion chamber. A heat-exchanger is operatively associated and in thermal contact with the combustion chamber. The heat-exchanger provides heat for the hybrid solar receiver.

  8. 46 CFR 185.508 - Wearing of life jackets.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 7 2011-10-01 2011-10-01 false Wearing of life jackets. 185.508 Section 185.508... TONS) OPERATIONS Preparations for Emergencies § 185.508 Wearing of life jackets. (a) The master of a vessel shall require passengers to don life jackets when possible hazardous conditions exist, including...

  9. 46 CFR 185.508 - Wearing of life jackets.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Wearing of life jackets. 185.508 Section 185.508... TONS) OPERATIONS Preparations for Emergencies § 185.508 Wearing of life jackets. (a) The master of a vessel shall require passengers to don life jackets when possible hazardous conditions exist, including...

  10. 46 CFR 185.508 - Wearing of life jackets.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 7 2012-10-01 2012-10-01 false Wearing of life jackets. 185.508 Section 185.508... TONS) OPERATIONS Preparations for Emergencies § 185.508 Wearing of life jackets. (a) The master of a vessel shall require passengers to don life jackets when possible hazardous conditions exist, including...

  11. 46 CFR 185.508 - Wearing of life jackets.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 7 2014-10-01 2014-10-01 false Wearing of life jackets. 185.508 Section 185.508... TONS) OPERATIONS Preparations for Emergencies § 185.508 Wearing of life jackets. (a) The master of a vessel shall require passengers to don life jackets when possible hazardous conditions exist, including...

  12. 46 CFR 185.508 - Wearing of life jackets.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 7 2013-10-01 2013-10-01 false Wearing of life jackets. 185.508 Section 185.508... TONS) OPERATIONS Preparations for Emergencies § 185.508 Wearing of life jackets. (a) The master of a vessel shall require passengers to don life jackets when possible hazardous conditions exist, including...

  13. Low cost, lightweight fuel cell elements

    NASA Technical Reports Server (NTRS)

    Kindler, Andrew (Inventor)

    2001-01-01

    New fuel cell elements for use in liquid feed fuel cells are provided. The elements including biplates and endplates are low in cost, light in weight, and allow high efficiency operation. Electrically conductive elements are also a part of the fuel cell elements.

  14. Monitoring arrangement for vented nuclear fuel elements

    DOEpatents

    Campana, Robert J.

    1981-01-01

    In a nuclear fuel reactor core, fuel elements are arranged in a closely packed hexagonal configuration, each fuel element having diametrically opposed vents permitting 180.degree. rotation of the fuel elements to counteract bowing. A grid plate engages the fuel elements and forms passages for communicating sets of three, four or six individual vents with respective monitor lines in order to communicate vented radioactive gases from the fuel elements to suitable monitor means in a manner readily permitting detection of leakage in individual fuel elements.

  15. BARRIERS TO LIFE JACKET USE AMONG ADULT RECREATIONAL BOATERS

    PubMed Central

    Quistberg, D. Alex; Quan, Linda; Ebel, Beth E.; Bennett, Elizabeth E.; Mueller, Beth A.

    2015-01-01

    Objective To identify barriers to life jacket use. Design Cross-sectional survey. Setting Nine public boat ramps in western Washington State, USA, August-November, 2008. Participants 675 adult boaters (>18 years) on motor boats <26 feet long. Main outcome Low or no life jacket use (0–50% of time) versus high life jacket use (51–100% of time). Results Low/no life jacket use (0%–50% of time) was associated with longer boat length (per foot, risk ratio [RR] 1.03, 95% confidence interval [CI] 1.02–1.05), alcohol use (RR 1.11, 95% CI 1.01–1.20), perception of life jackets as “uncomfortable” (RR 1.29, 95%CI 1.09–1.52), perceived greater level of swimming ability (RR 1.25, 95% CI 1.03–1.53 for “expert swimmer”), and possibly with lack of confidence that a life jacket may save one from drowning (RR 1.13, 95%CI 0.96–1.32). Low life jacket use was less likely when a child was onboard (RR 0.88, 95% CI 0.79–0.99), or if the respondent had taken a boating safety class (RR 0.94, 95% CI 0.87–1.01). Conclusions Life jacket use may increase with more comfortable devices, such as inflatable life jackets, and with increased awareness of their efficacy in preventing drowning. Boater education classes may be associated with increased life jacket use among adults. PMID:24686261

  16. 30 CFR 57.15020 - Life jackets and belts.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 30 Mineral Resources 1 2011-07-01 2011-07-01 false Life jackets and belts. 57.15020 Section 57.15020 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR METAL AND NONMETAL... Protection Surface Only § 57.15020 Life jackets and belts. Life jackets or belts shall be worn where there is...

  17. 30 CFR 56.15020 - Life jackets and belts.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Life jackets and belts. 56.15020 Section 56.15020 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR METAL AND NONMETAL... § 56.15020 Life jackets and belts. Life jackets or belts shall be worn where there is danger from...

  18. 30 CFR 57.15020 - Life jackets and belts.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Life jackets and belts. 57.15020 Section 57.15020 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR METAL AND NONMETAL... Protection Surface Only § 57.15020 Life jackets and belts. Life jackets or belts shall be worn where there is...

  19. 46 CFR 122.508 - Wearing of life jackets.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Wearing of life jackets. 122.508 Section 122.508... Preparations for Emergencies § 122.508 Wearing of life jackets. (a) The master of a vessel shall require passengers to don life jackets when possible hazardous conditions exist, including, but not limited to: (1...

  20. 46 CFR 122.508 - Wearing of life jackets.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 4 2011-10-01 2011-10-01 false Wearing of life jackets. 122.508 Section 122.508... Preparations for Emergencies § 122.508 Wearing of life jackets. (a) The master of a vessel shall require passengers to don life jackets when possible hazardous conditions exist, including, but not limited to: (1...

  1. 30 CFR 56.15020 - Life jackets and belts.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 30 Mineral Resources 1 2011-07-01 2011-07-01 false Life jackets and belts. 56.15020 Section 56.15020 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR METAL AND NONMETAL... § 56.15020 Life jackets and belts. Life jackets or belts shall be worn where there is danger from...

  2. 30 CFR 57.15020 - Life jackets and belts.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 30 Mineral Resources 1 2012-07-01 2012-07-01 false Life jackets and belts. 57.15020 Section 57.15020 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR METAL AND NONMETAL... Protection Surface Only § 57.15020 Life jackets and belts. Life jackets or belts shall be worn where there is...

  3. 46 CFR 122.508 - Wearing of life jackets.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 4 2013-10-01 2013-10-01 false Wearing of life jackets. 122.508 Section 122.508... Preparations for Emergencies § 122.508 Wearing of life jackets. (a) The master of a vessel shall require passengers to don life jackets when possible hazardous conditions exist, including, but not limited to: (1...

  4. 30 CFR 56.15020 - Life jackets and belts.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 30 Mineral Resources 1 2014-07-01 2014-07-01 false Life jackets and belts. 56.15020 Section 56.15020 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR METAL AND NONMETAL... § 56.15020 Life jackets and belts. Life jackets or belts shall be worn where there is danger from...

  5. 30 CFR 57.15020 - Life jackets and belts.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 30 Mineral Resources 1 2013-07-01 2013-07-01 false Life jackets and belts. 57.15020 Section 57.15020 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR METAL AND NONMETAL... Protection Surface Only § 57.15020 Life jackets and belts. Life jackets or belts shall be worn where there is...

  6. 30 CFR 56.15020 - Life jackets and belts.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 30 Mineral Resources 1 2013-07-01 2013-07-01 false Life jackets and belts. 56.15020 Section 56.15020 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR METAL AND NONMETAL... § 56.15020 Life jackets and belts. Life jackets or belts shall be worn where there is danger from...

  7. 30 CFR 57.15020 - Life jackets and belts.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 30 Mineral Resources 1 2014-07-01 2014-07-01 false Life jackets and belts. 57.15020 Section 57.15020 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR METAL AND NONMETAL... Protection Surface Only § 57.15020 Life jackets and belts. Life jackets or belts shall be worn where there is...

  8. 46 CFR 122.508 - Wearing of life jackets.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 4 2014-10-01 2014-10-01 false Wearing of life jackets. 122.508 Section 122.508... Preparations for Emergencies § 122.508 Wearing of life jackets. (a) The master of a vessel shall require passengers to don life jackets when possible hazardous conditions exist, including, but not limited to: (1...

  9. 46 CFR 122.508 - Wearing of life jackets.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 4 2012-10-01 2012-10-01 false Wearing of life jackets. 122.508 Section 122.508... Preparations for Emergencies § 122.508 Wearing of life jackets. (a) The master of a vessel shall require passengers to don life jackets when possible hazardous conditions exist, including, but not limited to: (1...

  10. 30 CFR 56.15020 - Life jackets and belts.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 30 Mineral Resources 1 2012-07-01 2012-07-01 false Life jackets and belts. 56.15020 Section 56.15020 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR METAL AND NONMETAL... § 56.15020 Life jackets and belts. Life jackets or belts shall be worn where there is danger from...

  11. NUCLEAR REACTOR FUEL ELEMENT ASSEMBLY

    DOEpatents

    Stengel, F.G.

    1963-12-24

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

  12. Vacuum-jacketed transfer line installation technique

    NASA Technical Reports Server (NTRS)

    Bowers, W. M.

    1968-01-01

    Rolling-type spacers in the form of steel balls retained in appropriate sleeves affixed at intervals to the exterior of the transfer line facilitate the installation of a vacuum-jacketed line. They act as standoffs to position the transfer line concentrically within the vacuum jacket line.

  13. FUEL ELEMENTS FOR NUCLEAR REACTORS

    DOEpatents

    Blainey, A.; Lloyd, H.

    1961-07-11

    A method of sheathing a tubular fuel element for a nuclear reactor is described. A low melting metal core member is centered in a die, a layer of a powdered sheathing substance is placed on the bottom of the die, the tubular fuel element is inserted in the die, the space between the tubular fuel element and the die walls and core member is filled with the same powdered sheathing substance, a layer of the same substance is placed over the fissile material, and the charge within the die is subjected to pressure in the direction of the axis of the fuel element at the sintering temperature of the protective substance.

  14. METHOD OF TESTING FOR LEAKS

    DOEpatents

    Creutz, E.C.; McAdams, Wm.A.; Foss, M.H.

    1958-07-22

    A method is described for detecting minute holes In fuel element jackets. The method comprises submerging the jacketed body in an atmosphere of a radioactive gas under pressure, the radioactive emanations from said gas being sufficientiy penetratIng to penetrate the jacket of the jacketed body. After the jacketed body is removed from the radtoactive gas atmosphere, it is exannined for the presence of emanations from radioactive gas which entered the jacketed body through the minute holes. In this manner, the detectton of radioactive emanations is a positive indication that the fuel element is not perfectly sealed.

  15. FUEL ELEMENT FOR NUCLEAR REACTORS

    DOEpatents

    Bassett, C.H.

    1961-05-16

    A fuel element particularly adapted for use in nuclear reactors of high power density is offered. It has fissionable fuel pellet segments mounted in a tubular housing and defining a central passage in the fuel element. A burnable poison element extends through the central passage, which is designed to contain more poison material at the median portion than at the end portions thereby providing a more uniform hurnup and longer reactivity life.

  16. FUEL ELEMENT FOR NEUTRONIC REACTORS

    DOEpatents

    Evans, T.C.; Beasley, E.G.

    1961-01-17

    A fuel element for neutronic reactors, particularly the gas-cooled type of reactor, is described. The element comprises a fuel-bearing plate rolled to form a cylinder having a spiral passageway passing from its periphery to its center. In operation a coolant is admitted to the passageway at the periphery of the element, is passed through the spiral passageway, and emerges into a central channel defined by the inner turn of the rolled plate. The advantage of the element is that the fully heated coolant (i.e., coolant emerging into the central channel) is separated and thus insulated from the periphery of the element, which may be in contact with a low-temperature moderator, by the intermediate turns of the spiral fuel element.

  17. FUEL ELEMENT CONSTRUCTION

    DOEpatents

    Simnad, M.T.

    1961-08-15

    A method of preventing diffusible and volatile fission products from diffusing through a fuel element container and contaminating reactor coolant is described. More specifically, relatively volatile and diffusible fission products either are adsorbed by or react with magnesium fluoride or difluoride to form stable, less volatile, less diffusible forms. The magnesium fluoride or difluoride is disposed anywhere inwardly from the outer surface of the fuel element container in order to be contacted by the fission products before they reach and contaminate the reactor coolant. (AEC)

  18. Integral Fast Reactor fuel pin processor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Levinskas, D.

    1993-01-01

    This report discusses the pin processor which receives metal alloy pins cast from recycled Integral Fast Reactor (IFR) fuel and prepares them for assembly into new IFR fuel elements. Either full length as-cast or precut pins are fed to the machine from a magazine, cut if necessary, and measured for length, weight, diameter and deviation from straightness. Accepted pins are loaded into cladding jackets located in a magazine, while rejects and cutting scraps are separated into trays. The magazines, trays, and the individual modules that perform the different machine functions are assembled and removed using remote manipulators and master-slaves.

  19. Integral Fast Reactor fuel pin processor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Levinskas, D.

    1993-03-01

    This report discusses the pin processor which receives metal alloy pins cast from recycled Integral Fast Reactor (IFR) fuel and prepares them for assembly into new IFR fuel elements. Either full length as-cast or precut pins are fed to the machine from a magazine, cut if necessary, and measured for length, weight, diameter and deviation from straightness. Accepted pins are loaded into cladding jackets located in a magazine, while rejects and cutting scraps are separated into trays. The magazines, trays, and the individual modules that perform the different machine functions are assembled and removed using remote manipulators and master-slaves.

  20. Thermal-leak analyzer for vacuum-jacketed lines

    NASA Technical Reports Server (NTRS)

    Heisman, R. M.; Iceland, W. F.; Ruppe, E. P.

    1978-01-01

    Technique involves coating suspected area with water-soluble black paint that gives even, infrared emission. Painted area is warmed with heat gun; an infrared scanner is used to detect cooled spot on jacket exterior. Introduction of atmospheric pressure into jacket intensifies leak jet and improves test sensitivity.

  1. FUEL ELEMENT FOR NUCLEAR REACTORS

    DOEpatents

    Bassett, C.H.

    1961-11-21

    A fuel element is designed which is particularly adapted for reactors of high power density used to generate steam for the production of electricity. The fuel element consists of inner and outer concentric tubes forming an annular chamber within which is contained fissionable fuel pellet segments, wedge members interposed between the fuel segments, and a spring which, acting with wedge members, urges said fuel pellets radially into contact against the inner surface of the outer tube. The wedge members may be a fertile material convertible into fissionable fuel material by absorbing neutrons emitted from the fissionable fuel pellet segments. The costly grinding of cylindrical fuel pellets to close tolerances for snug engagement is reduced because the need to finish the exact size is eliminated. (AEC)

  2. Inert matrix fuel in dispersion type fuel elements

    NASA Astrophysics Data System (ADS)

    Savchenko, A. M.; Vatulin, A. V.; Morozov, A. V.; Sirotin, V. L.; Dobrikova, I. V.; Kulakov, G. V.; Ershov, S. A.; Kostomarov, V. P.; Stelyuk, Y. I.

    2006-06-01

    The advantages of using inert matrix fuel (IMF) as a dispersion fuel in an aluminium alloy matrix are considered, in particular, low temperatures in the fuel centre, achievable high burn-ups, serviceability in transients and an environmentally friendly process of fuel rod fabrication. Two main versions of IMF are under development at A.A. Bochvar Institute, i.e. heterogeneous or isolated distribution of plutonium. The out-of-pile results on IMF loaded with uranium dioxide as plutonium simulator are presented. Fuel elements with uranium dioxide composition fabricated at A.A. Bochvar Institute are currently under MIR tests (RIAR, Dimitrovgrad). The fuel elements reached a burn-up of 88 MW d kg-1 (equivalent to the burn up of the standard uranium dioxide pelletized fuel) without loss of leak-tightness of the cladding. The feasibility of fabricating IMF of these particular types with plutonium dioxide is considered with a view to in-pile irradiation.

  3. FUEL ELEMENT CONSTRUCTION

    DOEpatents

    Zumwalt, L.R.

    1961-08-01

    Fuel elements having a solid core of fissionable material encased in a cladding material are described. A conversion material is provided within the cladding to react with the fission products to form stable, relatively non- volatile compounds thereby minimizing the migration of the fission products into the coolant. The conversion material is preferably a metallic fluoride, such as lead difluoride, and may be in the form of a coating on the fuel core or interior of the cladding, or dispersed within the fuel core. (AEC)

  4. CONCENTRIC TUBE FUEL ELEMENT SPRING ALIGNMENT SPACER DEVICE

    DOEpatents

    Weems, S.J.

    1963-09-24

    A rib construction for a nuclear-fuel element is described, in which one of three peripherally spaced ribs adjacent to each end of the fuel element is mounted on a radially yielding spring that embraces the fuel element. This spring enables the fuel element to have a good fit with a coolant tube and yet to be easily inserted in and withdrawn from the tube. (AEC)

  5. NEUTRONIC REACTOR FUEL ELEMENT

    DOEpatents

    Picklesimer, M.L.; Thurber, W.C.

    1961-01-01

    A chemically nonreactive fuel composition for incorporation in aluminum- clad, plate type fuel elements for neutronic reactors is described. The composition comprises a mixture of aluminum and uranium carbide particles, the uranium carbide particles containing at least 80 wt.% UC/sub 2/.

  6. Natural frequency and vibration analysis of jacket type foundation for offshore wind power

    NASA Astrophysics Data System (ADS)

    Hung, Y.-C.; Chang, Y.-Y.; Chen, S.-Y.

    2017-12-01

    There are various types of foundation structure for offshore wind power, engineers may assess the condition of ocean at wind farm, and arrange the transportation, installation of each structure members, furthermore, considering the ability of manufacture steel structure as well, then make an optimum design. To design jacket offshore structure, unlike onshore cases, offshore structure also need to estimate the wave excitation effect. The aim of this paper is to study the difference of natural frequency between different kinds of structural stiffness and discuss the effect of different setting of boundary condition during analysis, besides, compare this value with the natural frequency of sea wave, in order to avoid the resonance effect. In this paper, the finite element analysis software ABAQUS is used to model and analyze the natural vibration behavior of the jacket structure.

  7. Compact Fuel Element Environment Test

    NASA Technical Reports Server (NTRS)

    Bradley, D. E.; Mireles, O. R.; Hickman, R. R.; Broadway, J. W.

    2012-01-01

    Deep space missions with large payloads require high specific impulse (I(sub sp)) and relatively high thrust to achieve mission goals in reasonable time frames. Conventional, storable propellants produce average I(sub sp). Nuclear thermal rockets (NTRs) capable of high I(sub sp) thrust have been proposed. NTR employs heat produced by fission reaction to heat and therefore accelerate hydrogen, which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3,000 K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited. The primary concern is the mechanical failure of fuel elements that employ high melting point metals, ceramics, or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. It is not necessary to include fissile material in test samples intended to explore high-temperature hydrogen exposure of the structural support matrices. A small-scale test bed designed to heat fuel element samples via noncontact radio frequency heating and expose samples to hydrogen for typical mission durations has been developed to assist in optimal material and manufacturing process selection without employing fissile material. This Technical Memorandum details the test bed design and results of testing conducted to date.

  8. Nuclear fuel element with axially aligned fuel pellets and fuel microspheres therein

    DOEpatents

    Sease, J.D.; Harrington, F.E.

    1973-12-11

    Elongated single- and multi-region fuel elements are prepared by replacing within a cladding container a coarse fraction of fuel material which includes plutonium and uranium in the appropriate regions of the fuel element and then infiltrating with vibration a fine-sized fraction of uranium-containing microspheres throughout all interstices in the coarse material in a single loading. The fine, rigid material defines a thin annular layer between the coarse fraction and the cladding to reduce adverse mechanical and chemical interactions. (Official Gazette)

  9. NEUTRONIC REACTOR FUEL ELEMENT

    DOEpatents

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

    1959-11-24

    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.

  10. JACKETING URANIUM

    DOEpatents

    Saller, H.A.; Keeler, J.R.

    1959-07-14

    The bonding to uranium of sheathing of iron or cobalt, or nickel, or alloys thereof is described. The bonding is accomplished by electro-depositing both surfaces to be joined with a coating of silver and amalgamating or alloying the silver layer with mercury or indium. Then the silver alloy is homogenized by exerting pressure on an assembly of the uranium core and the metal jacket, reducing the area of assembly and heating the assembly to homogenize by diffusion.

  11. Rack for storing spent nuclear fuel elements

    DOEpatents

    Rubinstein, Herbert J.; Clark, Philip M.; Gilcrest, James D.

    1978-06-20

    A rack for storing spent nuclear fuel elements in which a plurality of aligned rows of upright enclosures of generally square cross-sectional areas contain vertically disposed fuel elements. The enclosures are fixed at the lower ends thereof to a base. Pockets are formed between confronting walls of adjacent enclosures for receiving high absorption neutron absorbers, such as Boral, cadmium, borated stainless steel and the like for the closer spacing of spent fuel elements.

  12. 35. DETAILS AND SECTIONS OF FUEL ELEMENT SUPPORT PLATFORM, FUEL ...

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

    35. DETAILS AND SECTIONS OF FUEL ELEMENT SUPPORT PLATFORM, FUEL ELEMENT HOLDER, TRIP MECHANISM COVER, AND OTHER DETAILS. F.C. TORKELSON DRAWING NUMBER 842-ARVFS-701-S-3. INEL INDEX CODE NUMBER: 075 0701 60 851 151977. - Idaho National Engineering Laboratory, Advanced Reentry Vehicle Fusing System, Scoville, Butte County, ID

  13. Nuclear fuel element

    DOEpatents

    Meadowcroft, Ronald Ross; Bain, Alastair Stewart

    1977-01-01

    A nuclear fuel element wherein a tubular cladding of zirconium or a zirconium alloy has a fission gas plenum chamber which is held against collapse by the loops of a spacer in the form of a tube which has been deformed inwardly at three equally spaced, circumferential positions to provide three loops. A heat resistant disc of, say, graphite separates nuclear fuel pellets within the cladding from the plenum chamber. The spacer is of zirconium or a zirconium alloy.

  14. MRT fuel element inspection at Dounreay

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Gibson, J.

    1997-08-01

    To ensure that their production and inspection processes are performed in an acceptable manner, ie. auditable and traceable, the MTR Fuel Element Fabrication Plant at Dounreay operates to a documented quality system. This quality system, together with the fuel element manufacturing and inspection operations, has been independently certified to ISO9002-1987, EN29002-1987 and BS5750:Pt2:1987 by Lloyd`s Register Quality Assurance Limited (LRQA). This certification also provides dual accreditation to the relevant German, Dutch and Australian certification bodies. This paper briefly describes the quality system, together with the various inspection stages involved in the manufacture of MTR fuel elements at Dounreay.

  15. Nuclear fuel elements and method of making same

    DOEpatents

    Schweitzer, Donald G.

    1992-01-01

    A nuclear fuel element for a high temperature gas nuclear reactor that has an average operating temperature in excess of 2000.degree. C., and a method of making such a fuel element. The fuel element is characterized by having fissionable fuel material localized and stabilized within pores of a carbon or graphite member by melting the fissionable material to cause it to chemically react with the carbon walls of the pores. The fissionable fuel material is further stabilized and localized within the pores of the graphite member by providing one or more coatings of pyrolytic carbon or diamond surrounding the porous graphite member so that each layer defines a successive barrier against migration of the fissionable fuel from the pores, and so that the outermost layer of pyrolytic carbon or diamond forms a barrier between the fissionable material and the moderating gases used in an associated high temperature gas reactor. The method of the invention provides for making such new elements either as generally spherically elements, or as flexible filaments, or as other relatively small-sized fuel elements that are particularly suited for use in high temperature gas reactors.

  16. DISSOLUTION OF ZIRCONIUM-CONTAINING FUEL ELEMENTS

    DOEpatents

    Horn, F.L.

    1961-12-12

    Uranium is recovered from spent uranium fuel elements containing or clad with zirconium. These fuel elements are placed in an anhydrous solution of hydrogen fluoride and nitrogen dioxide. Within this system uranium forms a soluble complex and zirconium forms an insoluble complex. The uranium can then be separated, treated, and removed from solution as uranium hexafluoride. (AEC)

  17. The confining effectiveness of NiTiNb and NiTi SMA wire jackets for concrete

    NASA Astrophysics Data System (ADS)

    Choi, Eunsoo; Chung, Young-Soo; Choi, Jun-Hyeok; Kim, Hong-Taek; Lee, Hacksoo

    2010-03-01

    The purpose of this study is to assess the confining effectiveness of shape memory alloy (SMA) wire jackets for concrete. The performance of SMA wire jackets was compared to that of steel jackets. A prestrained martensitic SMA wire was wrapped around a concrete cylinder and then heated by a heating jacket. In the process, a confining stress around the cylinder was developed in the SMA wire due to the shape memory effect; this jacketing method can increase the strength and ductility of the cylinder under an axial compressive load. In this study, NiTi and NiTiNb SMA wires of 1.0 mm in diameter were used for the confinement. Recovery tests were conducted on the wires to assess their recovery and residual stress. The confinement by SMA wire jackets increased the strength slightly and greatly increased the ductility compared to the strength and ductility of plain concrete cylinders. The NiTiNb SMA wire jacket showed better performance than that of the NiTi SMA wire jacket. The confining effectiveness of the SMA wire jackets of this study was estimated to be similar to that of the steel jackets. This study showed the potential of the SMA wire jacketing method to retrofit reinforced concrete columns and protect them from seismic risks.

  18. Splice connector with internal heat transfer jacket

    DOEpatents

    Silva, Frank A.; Mayer, Robert W.

    1977-01-01

    A heat transfer jacket is placed over the terminal portions of the conductors of a pair of high voltage cables which are connected in a splice connection wherein a housing surrounds the connected conductor portions, the heat transfer jacket extending longitudinally between the confronting ends of a pair of adaptor sleeves placed upon the insulation of the cables to engage and locate the adaptor sleeves relative to one another, and laterally between the conductors and the housing to provide a path of relatively high thermal conductivity between the connected conductor portions and the housing.

  19. Nuclear fuel elements having a composite cladding

    DOEpatents

    Gordon, Gerald M.; Cowan, II, Robert L.; Davies, John H.

    1983-09-20

    An improved nuclear fuel element is disclosed for use in the core of nuclear reactors. The improved nuclear fuel element has a composite cladding of an outer portion forming a substrate having on the inside surface a metal layer selected from the group consisting of copper, nickel, iron and alloys of the foregoing with a gap between the composite cladding and the core of nuclear fuel. The nuclear fuel element comprises a container of the elongated composite cladding, a central core of a body of nuclear fuel material disposed in and partially filling the container and forming an internal cavity in the container, an enclosure integrally secured and sealed at each end of said container and a nuclear fuel material retaining means positioned in the cavity. The metal layer of the composite cladding prevents perforations or failures in the cladding substrate from stress corrosion cracking or from fuel pellet-cladding interaction or both. The substrate of the composite cladding is selected from conventional cladding materials and preferably is a zirconium alloy.

  20. Means for supporting fuel elements in a nuclear reactor

    DOEpatents

    Andrews, Harry N.; Keller, Herbert W.

    1980-01-01

    A grid structure for a nuclear reactor fuel assembly comprising a plurality of connecting members forming at least one longitudinally extending opening peripheral and inner fuel element openings through each of which openings at least one nuclear fuel element extends, said connecting members forming wall means surrounding said each peripheral and inner fuel element opening, a pair of rigid projections longitudinally spaced from one another extending from a portion of said wall means into said each peripheral and inner opening for rigidly engaging said each fuel element, respectively, yet permit individual longitudinal slippage thereof, and resilient means formed integrally on and from said wall means and positioned in said each peripheral and inner opening in opposed relationship with said projections and located to engage said fuel element to bias the latter into engagement with said rigid projections, respectively

  1. FUEL ELEMENT FOR NUCLEAR REACTORS

    DOEpatents

    Bassett, C.H.

    1961-07-11

    Nuclear reactor fuel elements of the type in which the flssionsble material is in ceramic form, such as uranium dioxide, are described. The fuel element is comprised of elongated inner and outer concentric spaced tubular members providing an annular space therebetween for receiving the fissionable material, the annular space being closed at both ends and the inner tube being open at both ends. The fuel is in the form of compressed pellets of ceramic fissionsble material having the configuration of split bushings formed with wedge surfaces and arranged in seriated inner and outer concentric groups which are urged against the respective tubes in response to relative axial movement of the pellets in the direction toward each other. The pairs of pellets are axially urged together by a resilient means also enclosed within the annulus. This arrangement-permits relative axial displacement of the pellets during use dial stresses on the inner and outer tube members and yet maintains the fuel pellets in good thermal conductive relationship therewith.

  2. VENTED FUEL ELEMENT FOR GAS-COOLED NEUTRONIC REACTORS

    DOEpatents

    Furgerson, W.T.

    1963-12-17

    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)

  3. Expected damages of retrofitted bridges with RC jacketing

    NASA Astrophysics Data System (ADS)

    Montes, O.; Jara, J. M.; Jara, M.; Olmos, B. A.

    2015-07-01

    The bridge infrastructure in many countries of the world consists of medium span length structures built several decades ago and designed for very low seismic forces. Many of them are reinforced concrete structures that according to the current code regulations have to be rehabilitated to increase their seismic capacity. One way to reduce the vulnerability of the bridges is by using retrofitting techniques that increase the strength of the structure or by incorporating devices to reduce the seismic demand. One of the most common retrofit techniques of the bridges substructures is the use of RC jacketing; this research assesses the expected damages of seismically deficient medium length highway bridges retrofitted with reinforced concrete jacketing, by conducting a parametric study. We select a suite of twenty accelerograms of subduction earthquakes recorded close to the Pacific Coast in Mexico. The original structures consist of five 30 m span simple supported bridges with five pier heights of 5 m, 10 m, 15 m 20 and 25 m and the analyses include three different jacket thickness and three steel ratios. The bridges were subjected to the seismic records and non-linear time history analyses were carried out by using the OpenSEEs Plataform. Results allow selecting the reinforced concrete jacketing that better improves the expected seismic behavior of the bridge models.

  4. Mobile communication jacket for people with severe speech impairment.

    PubMed

    Lampe, Renée; Blumenstein, Tobias; Turova, Varvara; Alves-Pinto, Ana

    2018-04-01

    Cerebral palsy is a movement disorder caused by damage to motor control areas of the developing brain during early childhood. Motor disorders can also affect the ability to produce clear speech and to communicate. The aim of this study was to develop and to test a prototype of an assistive tool with an embedded mobile communication device to support patients with severe speech impairments. A prototype was developed by equipping a cycling jacket with a display, a small keyboard, a LED and an alarm system, all controlled by a microcontroller. Functionality of the prototype was tested in six participants (aged 7-20 years) with cerebral palsy and global developmental disorder and three healthy persons. A patient questionnaire consisting of seven items was used as an evaluation tool. A working prototype of the communication jacket was developed and tested. The questionnaire elicited positive responses from participants. Improvements to correct revealed weaknesses were proposed. Enhancements like voice output of pre-selected phrases and enlarged display were implemented. Integration in a jacket makes the system mobile and continuously available to the user. The communication jacket may be of great benefit to patients with motor and speech impairments. Implications for Rehabilitation The communication jacket developed can be easily used by people with movement and speech impairment. All technical components are integrated in a garment and do not have to be held with the hands or transported separately. The system is adaptable to individual use. Both expected and unexpected events can be dealt with, which contributes to the quality of life and self-fulfilment.

  5. FUEL ELEMENT INTERLOCKING ARRANGEMENT

    DOEpatents

    Fortescue, P.; Nicoll, D.

    1963-01-01

    This patent relates to a system for mutually interlocking a multiplicity of elongated, parallel, coextensive, upright reactor fuel elements so as to render a laterally selfsupporting bundle, while admitting of concurrent, selective, vertical withdrawal of a sizeable number of elements without any of the remaining elements toppling, Each element is provided with a generally rectangular end cap. When a rank of caps is aligned in square contact, each free edge centrally defines an outwardly profecting dovetail, and extremitally cooperates with its adjacent cap by defining a juxtaposed half of a dovetail- receptive mortise. Successive ranks are staggered to afford mating of their dovetails and mortises. (AEC)

  6. Upgraded HFIR Fuel Element Welding System

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Sease, John D

    2010-02-01

    The welding of aluminum-clad fuel plates into aluminum alloy 6061 side plate tubing is a unique design feature of the High Flux Isotope Reactor (HFIR) fuel assemblies as 101 full-penetration circumferential gas metal arc welds (GMAW) are required in the fabrication of each assembly. In a HFIR fuel assembly, 540 aluminum-clad fuel plates are assembled into two nested annular fuel elements 610 mm (24-inches) long. The welding process for the HFIR fuel elements was developed in the early 1960 s and about 450 HFIR fuel assemblies have been successfully welded using the GMAW process qualified in the 1960 s. Inmore » recent years because of the degradation of the electronic and mechanical components in the old HFIR welding system, reportable defects in plate attachment or adapter welds have been present in almost all completed fuel assemblies. In October 2008, a contract was awarded to AMET, Inc., of Rexburg, Idaho, to replace the old welding equipment with standard commercially available welding components to the maximum extent possible while maintaining the qualified HFIR welding process. The upgraded HFIR welding system represents a major improvement in the welding system used in welding HFIR fuel elements for the previous 40 years. In this upgrade, the new inner GMAW torch is a significant advancement over the original inner GMAW torch previously used. The innovative breakthrough in the new inner welding torch design is the way the direction of the cast in the 0.762 mm (0.030-inch) diameter aluminum weld wire is changed so that the weld wire emerging from the contact tip is straight in the plane perpendicular to the welding direction without creating any significant drag resistance in the feeding of the weld wire.« less

  7. NEUTRONIC REACTOR FUEL ELEMENT AND CORE SYSTEM

    DOEpatents

    Moore, W.T.

    1958-09-01

    This patent relates to neutronic reactors and in particular to an improved fuel element and a novel reactor core system for facilitating removal of contaminating fission products, as they are fermed, from association with the flssionable fuel, so as to mitigate the interferent effects of such fission products during reactor operation. The fuel elements are comprised of tubular members impervious to fluid and contatning on their interior surfaces a thin layer of fissionable material providing a central void. The core structure is comprised of a plurality of the tubular fuel elements arranged in parallel and a closed manifold connected to their ends. In the reactor the core structure is dispersed in a water moderator and coolant within a pressure vessel, and a means connected to said manifuld is provided for withdrawing and disposing of mobile fission product contamination from the interior of the feel tubes and manifold.

  8. Method of preventing leakage of a fluid along and through an insulating jacket of a thermocouple

    DOEpatents

    Thermos, Anthony Constantine; Rahal, Fadi Elias

    2002-01-01

    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.

  9. PROTECTIVELY COVERED ARTICLE AND METHOD OF MANUFACTURE

    DOEpatents

    Plott, R.F.

    1958-10-28

    A method of casting a protective jacket about a ura nium fuel element that will bond completely to the uranium without the use of stringers or supports that would ordinarily produce gaps in the cast metal coating and bond is presented. Preformed endcaps of alumlnum alloyed with 13% silicon are placed on the ends of the uranium fuel element. These caps will support the fuel element when placed in a mold. The mold is kept at a ing alloy but below that of uranium so the cast metal jacket will fuse with the endcaps forming a complete covering and bond to the fuel element, which would otherwise oxidize at the gaps or discontinuities lefi in the coating by previous casting methods.

  10. FUEL ELEMENT FOR NUCLEAR REACTORS

    DOEpatents

    Bassett, C.H.

    1961-05-01

    A nuclear reactor fuel element comprising high density ceramic fissionable material enclosed in a tubular cladding of corrosion-resistant material is described. The fissionable material is in the form of segments of a tube which have cooperating tapered interfaces which produce outward radial displacement when the segments are urged axially together. A resilient means is provided within the tubular housing to constantly urge the fuel segments axially. This design maintains the fuel material in tight contacting engagement against the inner surface of the outer cladding tube to eliminate any gap therebetween which may be caused by differential thermal expansion between the fuel material and the material of the tube.

  11. Destabilization and intracranial fragmentation of a full metal jacket bullet.

    PubMed

    Farrugia, A; Raul, J S; Geraut, A; Tortel, M C; Ludes, B

    2009-10-01

    We report a case with an atypical entrance wound as a result of a destabilized full metal jacket bullet penetration. The destabilized bullet by an impact with the dorsal hand experiences a yawing to tumbling motion in flight. The large angle of yaw induces a larger presenting profile upon impact that contributes, associated to a rapid deceleration, to a greater mechanical force on the projectile structure and a fragmentation into core and jacket. Forensic pathologists have to be aware that the metal jacket bullet could tend to break up outside or inside the body particularly after a shooting through a target. This phenomenon induces atypical entrance wounds and atypical X-ray presentation. 2009 Elsevier Ltd and Faculty of Forensic and Legal Medicine.

  12. Nuclear reactor fuel element

    DOEpatents

    Johnson, Carl E.; Crouthamel, Carl E.

    1980-01-01

    A nuclear reactor fuel element is described which has an outer cladding, a central core of fissionable or mixed fissionable and fertile fuel material and a layer of oxygen gettering material on the inner surface of the cladding. The gettering material reacts with oxygen released by the fissionable material during irradiation of the core thereby preventing the oxygen from reacting with and corroding the cladding. Also described is an improved method for coating the inner surface of the cladding with a layer of gettering material.

  13. Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

    NASA Astrophysics Data System (ADS)

    Mohammed, Abdul Aziz; Pauzi, Anas Muhamad; Rahman, Shaik Mohmmed Haikhal Abdul; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad

    2016-01-01

    In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 (233U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

  14. Evaluation of the finite element fuel rod analysis code (FRANCO)

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Lee, K.; Feltus, M.A.

    1994-12-31

    Knowledge of temperature distribution in a nuclear fuel rod is required to predict the behavior of fuel elements during operating conditions. The thermal and mechanical properties and performance characteristics are strongly dependent on the temperature, which can vary greatly inside the fuel rod. A detailed model of fuel rod behavior can be described by various numerical methods, including the finite element approach. The finite element method has been successfully used in many engineering applications, including nuclear piping and reactor component analysis. However, fuel pin analysis has traditionally been carried out with finite difference codes, with the exception of Electric Powermore » Research Institute`s FREY code, which was developed for mainframe execution. This report describes FRANCO, a finite element fuel rod analysis code capable of computing temperature disrtibution and mechanical deformation of a single light water reactor fuel rod.« less

  15. TWISTED RIBBON FUEL ELEMENT

    DOEpatents

    Breden, C.R.; Schultz, A.B.

    1961-06-01

    A reactor core formed of bundles of parallel fuel elements in the form of ribbons is patented. The fuel ribbons are twisted about their axes so as to have contact with one another at regions spaced lengthwise of the ribbons and to be out of contact with one another at locations between these spaced regions. The contact between the ribbons is sufficient to allow them to be held together in a stable bundle in a containing tube without intermediate support, while permitting enough space between the ribbon for coolant flowing.

  16. FUEL ELEMENTS FOR THERMAL-FISSION NUCLEAR REACTORS

    DOEpatents

    Flint, O.

    1961-01-10

    Fuel elements for thermal-fission nuclear reactors are described. The fuel element is comprised of a core of alumina, a film of a metal of the class consisting of copper, silver, and nickel on the outer face of the core, and a coating of an oxide of a metal isotope of the class consisting of Un/sup 235/, U/ sup 233/, and Pu/sup 239/ on the metal f ilm.

  17. Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Mohammed, Abdul Aziz, E-mail: azizM@uniten.edu.my; Rahman, Shaik Mohmmed Haikhal Abdul; Pauzi, Anas Muhamad, E-mail: anas@uniten.edu.my

    2016-01-22

    In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 ({sup 233}U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintainingmore » the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.« less

  18. Space reactor fuel element testing in upgraded TREAT

    NASA Astrophysics Data System (ADS)

    Todosow, M.; Bezler, P.; Ludewig, H.; Kato, W. Y.

    The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc.; a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR); NERVA-derivative; and other concepts are discussed. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggest that full-scale PBR elements could be tested at an average energy deposition of approximately 60-80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of approximately 100 MW/L may be achievable.

  19. Space reactor fuel element testing in upgraded TREAT

    NASA Astrophysics Data System (ADS)

    Todosow, Michael; Bezler, Paul; Ludewig, Hans; Kato, Walter Y.

    1993-01-01

    The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggests that full-scale PBR elements could be tested at an average energy deposition of ˜60-80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of ˜100 MW/L may be achievable.

  20. Nuclear fuel elements made from nanophase materials

    DOEpatents

    Heubeck, Norman B.

    1998-01-01

    A nuclear reactor core fuel element is composed of nanophase high temperature materials. An array of the fuel elements in rod form are joined in an open geometry fuel cell that preferably also uses such nanophase materials for the cell structures. The particular high temperature nanophase fuel element material must have the appropriate mechanical characteristics to avoid strain related failure even at high temperatures, in the order of about 3000.degree. F. Preferably, the reactor type is a pressurized or boiling water reactor and the nanophase material is a high temperature ceramic or ceramic composite. Nanophase metals, or nanophase metals with nanophase ceramics in a composite mixture, also have desirable characteristics, although their temperature capability is not as great as with all-ceramic nanophase material. Combinations of conventional or nanophase metals and conventional or nanophase ceramics can be employed as long as there is at least one nanophase material in the composite. The nuclear reactor so constructed has a number of high strength fuel particles, a nanophase structural material for supporting a fuel rod at high temperature, a configuration to allow passive cooling in the event of a primary cooling system failure, an ability to retain a coolable geometry even at high temperatures, an ability to resist generation of hydrogen gas, and a configuration having good nuclear, corrosion, and mechanical characteristics.

  1. Nuclear fuel elements made from nanophase materials

    DOEpatents

    Heubeck, N.B.

    1998-09-08

    A nuclear reactor core fuel element is composed of nanophase high temperature materials. An array of the fuel elements in rod form are joined in an open geometry fuel cell that preferably also uses such nanophase materials for the cell structures. The particular high temperature nanophase fuel element material must have the appropriate mechanical characteristics to avoid strain related failure even at high temperatures, in the order of about 3000 F. Preferably, the reactor type is a pressurized or boiling water reactor and the nanophase material is a high temperature ceramic or ceramic composite. Nanophase metals, or nanophase metals with nanophase ceramics in a composite mixture, also have desirable characteristics, although their temperature capability is not as great as with all-ceramic nanophase material. Combinations of conventional or nanophase metals and conventional or nanophase ceramics can be employed as long as there is at least one nanophase material in the composite. The nuclear reactor so constructed has a number of high strength fuel particles, a nanophase structural material for supporting a fuel rod at high temperature, a configuration to allow passive cooling in the event of a primary cooling system failure, an ability to retain a coolable geometry even at high temperatures, an ability to resist generation of hydrogen gas, and a configuration having good nuclear, corrosion, and mechanical characteristics. 5 figs.

  2. Preparation of high temperature gas-cooled reactor fuel element

    DOEpatents

    Bradley, Ronnie A.; Sease, John D.

    1976-01-01

    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.

  3. Yellow jackets may be an underestimated component of an ant-seed mutualism

    USGS Publications Warehouse

    Bale, M.T.; Zettler, J.A.; Robinson, B.A.; Spira, T.P.; Allen, Craig R.

    2003-01-01

    Yellow jackets (Hymenoptera: Vespidae) are attracted to the typically ant-dispersed seeds of trilliums and will take seeds from ants in the genus Aphaenogaster. To determine if yellow jacket, Vespula maculifrons (Buysson), presence interferes with seed foraging by ants, we presented seeds of Trillium discolor Wray to three species (A. texana carolinensis Wheeler, Formica schaufussi Mayr, and Solenopsis invicta Buren) of seed-carrying ants in areas where vespids were present or excluded. We found that interspecific aggression between yellow jackets and ants is species specific. Vespid presence decreased average foraging time and increased foraging efficiency of two of the three ant species studied, a situation that might reflect competition for a limited food source. We also found that yellow jackets removed more seeds than ants, suggestive that vespids are important, albeit underestimated, components of ant-seed mutualisms.

  4. THE FUEL ELEMENT GRAPHITE. Project DRAGON.

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Graham, L.W.; Price, M.S.T.

    1963-01-15

    The main requirements of a fuel element graphite for reactors based on the Dragon concept are low transmission coefficient for fission products, dimensional stability under service conditions, high strength, high thermal conductivity, high purity, and high resistance to oxidation. Since conclusions reached in early 1960, a considerable amount of information has accumulated concerning the likely behaviour of graphites in high temperature reactor systems, particularly data on dimensional stability under irradiation. The influence of this new knowledge on the development of fuel element graphite with the Dragon Project is discussed in detail in the final section of this paper.

  5. Liquid fuel injection elements for rocket engines

    NASA Technical Reports Server (NTRS)

    Cox, George B., Jr. (Inventor)

    1993-01-01

    Thrust chambers for liquid propellant rocket engines include three principal components. One of these components is an injector which contains a plurality of injection elements to meter the flow of propellants at a predetermined rate, and fuel to oxidizer mixture ratio, to introduce the mixture into the combustion chamber, and to cause them to be atomized within the combustion chamber so that even combustion takes place. Evolving from these injectors are tube injectors. These tube injectors have injection elements for injecting the oxidizer into the combustion chamber. The oxidizer and fuel must be metered at predetermined rates and mixture ratios in order to mix them within the combustion chamber so that combustion takes place smoothly and completely. Hence tube injectors are subject to improvement. An injection element for a liquid propellant rocket engine of the bipropellant type is provided which includes tangential fuel metering orifices, and a plurality of oxidizer tube injection elements whose injection tubes are also provided with tangential oxidizer entry slots and internal reed valves.

  6. Infrared detector Dewars - Increased LN2 hold time and vacuum jacket life spans

    NASA Technical Reports Server (NTRS)

    Jennings, D. E.; Boyd, W. J.; Blass, W. E.

    1976-01-01

    IR detector Dewars commonly suffer from shorter than desired LN2 hold times and insulation jacket vacuum corruption over relatively short time periods. In an attempt to solve this problem for a 9144 detector Dewar, small 1 liter/s appendage ion pumps were selected for continuous pumping of the vacuum jackets. This procedure extended LN2 hold times from 20 to 60 h and virtually eliminated vacuum jacket corruption. Thus the detector systems are usable continuously over periods of 6 months or more.

  7. Nuclear reactor fuel element having improved heat transfer

    DOEpatents

    Garnier, J.E.; Begej, S.; Williford, R.E.; Christensen, J.A.

    1982-03-03

    A nuclear reactor fuel element having improved heat transfer between fuel material and cladding is described. The element consists of an outer cladding tube divided into an upper fuel section containing a central core of fissionable or mixed fissionable and fertile fuel material, slightly smaller in diameter than the inner surface of the cladding tube and a small lower accumulator section, the cladding tube being which is filled with a low molecular weight gas to transfer heat from fuel material to cladding during irradiation. A plurality of essentially vertical grooves in the fuel section extend downward and communicate with the accumulator section. The radial depth of the grooves is sufficient to provide a thermal gradient between the hot fuel surface and the relatively cooler cladding surface to allow thermal segregation to take place between the low molecular weight heat transfer gas and high molecular weight fission product gases produced by the fuel material during irradiation.

  8. Thermal Hydraulic Analysis of a Packed Bed Reactor Fuel Element

    DTIC Science & Technology

    1989-05-25

    Engineer and Master of Science in Nuclear Engineering. ABSTRACT A model of the behavior of a packed bed nuclear reactor fuel element is developed . It...RECOMMENDATIONS FOR FURTHER INVESTIGATION .................... 150 APPENDIX A FUEL ELEMENT MODEL PROGRAM DESIGN AND OPERA- T IO N...follow describe the details of the packed bed reactor and then discuss the development of the mathematical representations of the fuel element. These are

  9. Space reactor fuel element testing in upgraded TREAT

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Todosow, M.; Bezler, P.; Ludewig, H.

    1993-01-14

    The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. initial results suggest that full-scale PBR, elements could be tested at an average energy deposition of {approximately}60--80 MW-s/L in the current TREAT reactor. Ifmore » the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of {approximately}100 MW/L may be achievable.« less

  10. Space reactor fuel element testing in upgraded TREAT

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Todosow, M.; Bezler, P.; Ludewig, H.

    1993-05-01

    The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. initial results suggest that full-scale PBR, elements could be tested at an average energy deposition of {approximately}60--80 MW-s/L in the current TREAT reactor. Ifmore » the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of {approximately}100 MW/L may be achievable.« less

  11. THE MANUFACTURE OF FUEL ELEMENTS OF THE ARGONAUT TYPE

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Kittl, J.; Machado, R.E.; Mazza, J.A.

    1958-06-10

    The conditions required for the manufacture of the RA-1 Argonant type fuel elements are investigated. The fuel elements are in the form of a plate which is manufactured by the extrusion of a presintered mass of U/sub 3/O/sub 8/ (20% enriched) in an aluminum matrix. Steps in the investigation were obtention and specification of U/sub 3/O/sub 8/ and Al in powder form for testing, filling, and extrusion tests, finishing of the fuel elements, and computation of U/sub 3/O/sub 8/ content. (W.D.M.)

  12. Nuclear fuel element

    DOEpatents

    Armijo, Joseph S.; Coffin, Jr., Louis F.

    1980-04-29

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has an improved composite cladding comprised of a moderate purity metal barrier of zirconium metallurgically bonded on the inside surface of a zirconium alloy tube. The metal barrier forms a shield between the alloy tube and a core of nuclear fuel material enclosed in the composite cladding. There is a gap between the cladding and the core. The metal barrier forms about 1 to about 30 percent of the thickness of the composite cladding and has low neutron absorption characteristics. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the alloy tube from contact and reaction with such impurities and fission products. Methods of manufacturing the composite cladding are also disclosed.

  13. Drying results of K-Basin fuel element 1990 (Run 1)

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Marschman, S.C.; Abrefah, J.; Klinger, G.S.

    1998-06-01

    The water-filled K-Basins in the Hanford 100-Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basins have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuels in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtainedmore » from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the first of those tests (Run 1), which was conducted on an N-Reactor inner fuel element (1990) that had been stored underwater in the K-West Basin (see Section 2.0). This fuel element was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The testing was conducted in the Whole Element Furnace Testing System, described in Section 3.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodology are given in Section 4.0, and the experimental results provided in Section 5.0. These results are further discussed in Section 6.0.« less

  14. CANNED SLUG

    DOEpatents

    Burton, M.

    1959-02-17

    Fuel elements of the type comprised of a core of fissionable material enclosed in a jacket of nonfissionable, corrosion resistant material are presented. In this invention the fissionable core is shorter than the jacket member, to provide a void chamber in one end of the assembled element. The fissionable material is separated from the chamber by an inwardly extending portion of the jacket member containing a gas permeable wafer centrally disposed therein. The outer end of the chamber is closed bv the end of the jacket which has a rupture disk centrally disposed therein. Gases formed by the irradiation of the fissionable material pass through the porous wafer into the chanmber thereby causing a gradual increase in the pressure in the chamber. The rupture disk is designed to fail at a lower pressure than that which would rupture the jacket. Upon rupture of the disk, the gases in the chamber escape into the coolant channel and coolant enters the chamber but is prevented from coming into contact with the fissionable material by the action of the gases under pressure passing outwardly through the wafer. The ruptured fuel element may be readily detected by monitoring the reactor coolant system.

  15. Recapturing Graphite-Based Fuel Element Technology for Nuclear Thermal Propulsion

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Trammell, Michael P; Jolly, Brian C; Miller, James Henry

    ORNL is currently recapturing graphite based fuel forms for Nuclear Thermal Propulsion (NTP). This effort involves research and development on materials selection, extrusion, and coating processes to produce fuel elements representative of historical ROVER and NERVA fuel. Initially, lab scale specimens were fabricated using surrogate oxides to develop processing parameters that could be applied to full length NTP fuel elements. Progress toward understanding the effect of these processing parameters on surrogate fuel microstructure is presented.

  16. Piano jacket for perceiving and playing music for patients with cerebral palsy.

    PubMed

    Lampe, Renée; Turova, Varvara; Alves-Pinto, Ana

    2017-12-23

    Patients with cerebral palsy with severe motor disabilities are limited not only in everyday life activities but also in choice of their hobbies. Playing a musical instrument is for the majority not possible, even though music constitutes a central component of many relaxation activities for people with disabilities. To give affected patients the opportunity to make music and to learn piano playing through somatosensory perception, a prototype of piano jacket was developed. A cycling jacket was equipped with boxes, each representing a musical note, incorporated into the sleeves. Each box contains vibration motors and LEDs. These can be used to translate the sequences of piano key presses performed at an external E-piano. An additional operation mode allows the user to actively play a melody himself by touching the same boxes that also incorporate touch sensor components. A working prototype of the piano jacket was developed. The sensory piano jacket provides patients with cerebral palsy and contractures the possibility to develop sensorimotor skills, motor abilities and participation in music-related activities. The jacket is planned to be used in music lessons. Implications for Rehabilitation The system is suitable even for patients with severe motor disabilities and especially joint contractures. Excellent handling through large easily accessible communication buttons. All building blocks are held in a garment, which allows for easy portability and gripping comfort.

  17. Fuel element concept for long life high power nuclear reactors

    NASA Technical Reports Server (NTRS)

    Mcdonald, G. E.; Rom, F. E.

    1969-01-01

    Nuclear reactor fuel elements have burnups that are an order of magnitude higher than can currently be achieved by conventional design practice. Elements have greater time integrated power producing capacity per unit volume. Element design concept capitalizes on known design principles and observed behavior of nuclear fuel.

  18. Nuclear fuel element

    DOEpatents

    Armijo, Joseph S.; Coffin, Jr., Louis F.

    1983-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has a composite cladding having a substrate and a metal barrier metallurgically bonded on the inside surface of the substrate so that the metal barrier forms a shield between the substrate and the nuclear fuel material held within the cladding. The metal barrier forms about 1 to about 30 percent of the thickness of the cladding and is comprised of a low neutron absorption metal of substantially pure zirconium. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the substrate from contact and reaction with such impurities and fission products. The substrate of the composite cladding is selected from conventional cladding materials and preferably is a zirconium alloy. Methods of manufacturing the composite cladding are also disclosed.

  19. Reliability analysis of dispersion nuclear fuel elements

    NASA Astrophysics Data System (ADS)

    Ding, Shurong; Jiang, Xin; Huo, Yongzhong; Li, Lin an

    2008-03-01

    Taking a dispersion fuel element as a special particle composite, the representative volume element is chosen to act as the research object. The fuel swelling is simulated through temperature increase. The large strain elastoplastic analysis is carried out for the mechanical behaviors using FEM. The results indicate that the fission swelling is simulated successfully; the thickness increments grow linearly with burnup; with increasing of burnup: (1) the first principal stresses at fuel particles change from tensile ones to compression ones, (2) the maximum Mises stresses at the particles transfer from the centers of fuel particles to the location close to the interfaces between the matrix and the particles, their values increase with burnup; the maximum Mises stresses at the matrix exist in the middle location between the two particles near the mid-plane along the length (or width) direction, and the maximum plastic strains are also at the above region.

  20. Fuel cell elements with improved water handling capacity

    NASA Technical Reports Server (NTRS)

    Kindler, Andrew (Inventor); Lee, Albany (Inventor)

    2001-01-01

    New fuel cell components for use in liquid feed fuel cell systems are provided. The components include biplates and endplates, having a hydrophilic surface and allow high efficiency operation. Conductive elements and a wicking device also form a part of the fuel cell components of the invention.

  1. Factors associated with life jacket use among cabin sailboat and day sailor boaters in the United States.

    PubMed

    Spitzer, Natalie; Phillips, Maile T; Chow, Wendy; Mangione, Thomas W

    2018-06-01

    In 2015, drowning accounted for 68% of the 626 recreational boating related deaths in the United States. Although life jackets are estimated to prevent between 50% to 80% of boating deaths, approximately 83% of sailboat-related drowning victims were reported to not be wearing life jackets. Life jacket use among adult boaters has remained consistently low across most boat types and may vary by boater, boating, and environmental conditions. Although many risky environmental and boating factors may be associated with a higher risk of boating death, drowning occurs in all situations and it is useful to understand adult life jacket wearing behaviors in differing boating situations. This study uses observational survey data from 61318 adult sailors collected during the summer months of 1999 - 2017 from 124 selected study sites across 30 states in the US. Life jacket use was compared for day sailors and cabin sailboats by multiple boating, boater, and environmental conditions using Chi-square tests for equality of proportions. Results of these tests led to a choice of 3 informative and scientifically compelling variables to summarize variation in life jacket use for each sailboat type. Odds ratios were calculated comparing life jacket use from each boating situation to the lowest risk situation as determined by the 3 selected variables. These variables were represented in a tree diagram, detailing the additive impact of each factor. Following these analyses, all boating conditions were categorized to be risky or non-risky and a count variable was created for each observation based on the number of risks present. Cochran-Armitage trend tests were conducted to test for linearity in life jacket use for both boat types. The overall life jacket wear rate was higher among adults in day sailor sailboats (51.6%) compared to cabin sailboats (13.8%) and in all measured demographic, boating, and environmental circumstances. Comparing high-risk cabin sailboat conditions of low water

  2. Local Burn-Up Effects in the NBSR Fuel Element

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Brown N. R.; Hanson A.; Diamond, D.

    2013-01-31

    This study addresses the over-prediction of local power when the burn-up distribution in each half-element of the NBSR is assumed to be uniform. A single-element model was utilized to quantify the impact of axial and plate-wise burn-up on the power distribution within the NBSR fuel elements for both high-enriched uranium (HEU) and low-enriched uranium (LEU) fuel. To validate this approach, key parameters in the single-element model were compared to parameters from an equilibrium core model, including neutron energy spectrum, power distribution, and integral U-235 vector. The power distribution changes significantly when incorporating local burn-up effects and has lower power peakingmore » relative to the uniform burn-up case. In the uniform burn-up case, the axial relative power peaking is over-predicted by as much as 59% in the HEU single-element and 46% in the LEU single-element with uniform burn-up. In the uniform burn-up case, the plate-wise power peaking is over-predicted by as much as 23% in the HEU single-element and 18% in the LEU single-element. The degree of over-prediction increases as a function of burn-up cycle, with the greatest over-prediction at the end of Cycle 8. The thermal flux peak is always in the mid-plane gap; this causes the local cumulative burn-up near the mid-plane gap to be significantly higher than the fuel element average. Uniform burn-up distribution throughout a half-element also causes a bias in fuel element reactivity worth, due primarily to the neutronic importance of the fissile inventory in the mid-plane gap region.« less

  3. Method of locating a leaking fuel element in a fast breeder power reactor

    DOEpatents

    Honekamp, John R.; Fryer, Richard M.

    1978-01-01

    Leaking fuel elements in a fast reactor are identified by measuring the ratio of .sup.134 Xe to .sup.133 Xe in the reactor cover gas following detection of a fuel element leak, this ratio being indicative of the power and burnup of the failed fuel element. This procedure can be used to identify leaking fuel elements in a power breeder reactor while continuing operation of the reactor since the ratio measured is that of the gases stored in the plenum of the failed fuel element. Thus, use of a cleanup system for the cover gas makes it possible to identify sequentially a multiplicity of leaking fuel elements without shutting the reactor down.

  4. 46 CFR 117.72 - Personal flotation devices carried in addition to life jackets.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Personal flotation devices carried in addition to life jackets. 117.72 Section 117.72 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) SMALL... approved life jackets required to be worn during drills and emergencies. (b) Wearable marine buoyant...

  5. 46 CFR 117.72 - Personal flotation devices carried in addition to life jackets.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 4 2011-10-01 2011-10-01 false Personal flotation devices carried in addition to life jackets. 117.72 Section 117.72 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) SMALL... approved life jackets required to be worn during drills and emergencies. (b) Wearable marine buoyant...

  6. COMPARTMENTED REACTOR FUEL ELEMENT

    DOEpatents

    Cain, F.M. Jr.

    1962-09-11

    A method of making a nuclear reactor fuel element of the elongated red type is given wherein the fissionable fuel material is enclosed within a tubular metal cladding. The method comprises coating the metal cladding tube on its inside wall with a brazing alloy, inserting groups of cylindrical pellets of fissionable fuel material into the tube with spacing members between adjacent groups of pellets, sealing the ends of the tubes to leave a void space therewithin, heating the tube and its contents to an elevated temperature to melt the brazing alloy and to expand the pellets to their maximum dimensions under predetermined operating conditions thereby automatically positioning the spacing members along the tube, and finally cooling the tube to room temperature whereby the spacing disks become permanently fixed at their edges in the brazing alloy and define a hermetically sealed compartment for each fl group of fuel pellets. Upon cooling, the pellets contract thus leaving a space to accommodate thermal expansion of the pellets when in use in a reactor. The spacing members also provide lateral support for the tubular cladding to prevent collapse thereof when subjected to a reactor environment. (AEC)

  7. Arikan and Alamouti matrices based on fast block-wise inverse Jacket transform

    NASA Astrophysics Data System (ADS)

    Lee, Moon Ho; Khan, Md Hashem Ali; Kim, Kyeong Jin

    2013-12-01

    Recently, Lee and Hou (IEEE Signal Process Lett 13: 461-464, 2006) proposed one-dimensional and two-dimensional fast algorithms for block-wise inverse Jacket transforms (BIJTs). Their BIJTs are not real inverse Jacket transforms from mathematical point of view because their inverses do not satisfy the usual condition, i.e., the multiplication of a matrix with its inverse matrix is not equal to the identity matrix. Therefore, we mathematically propose a fast block-wise inverse Jacket transform of orders N = 2 k , 3 k , 5 k , and 6 k , where k is a positive integer. Based on the Kronecker product of the successive lower order Jacket matrices and the basis matrix, the fast algorithms for realizing these transforms are obtained. Due to the simple inverse and fast algorithms of Arikan polar binary and Alamouti multiple-input multiple-output (MIMO) non-binary matrices, which are obtained from BIJTs, they can be applied in areas such as 3GPP physical layer for ultra mobile broadband permutation matrices design, first-order q-ary Reed-Muller code design, diagonal channel design, diagonal subchannel decompose for interference alignment, and 4G MIMO long-term evolution Alamouti precoding design.

  8. Enhanced Low-Enriched Uranium Fuel Element for the Advanced Test Reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Pope, M. A.; DeHart, M. D.; Morrell, S. R.

    2015-03-01

    Under the current US Department of Energy (DOE) policy and planning scenario, the Advanced Test Reactor (ATR) and its associated critical facility (ATRC) will be reconfigured to operate on low-enriched uranium (LEU) fuel. This effort has produced a conceptual design for an Enhanced LEU Fuel (ELF) element. This fuel features monolithic U-10Mo fuel foils and aluminum cladding separated by a thin zirconium barrier. As with previous iterations of the ELF design, radial power peaking is managed using different U-10Mo foil thicknesses in different plates of the element. The lead fuel element design, ELF Mk1A, features only three fuel meat thicknesses,more » a reduction from the previous iterations meant to simplify manufacturing. Evaluation of the ELF Mk1A fuel design against reactor performance requirements is ongoing, as are investigations of the impact of manufacturing uncertainty on safety margins. The element design has been evaluated in what are expected to be the most demanding design basis accident scenarios and has met all initial thermal-hydraulic criteria.« less

  9. An Innovative Cooling Jacket to Combat Heat Intolerance in Children with Anhidrosis.

    PubMed

    Inamadar, Arun C; Palit, Aparna; Khurana, Neha

    2017-07-01

    Hyperthermia and heat intolerance are distressing symptoms in patients with anhidrosis. Body cooling devices are an integral part of management of these patients. A cooling jacket made from easily available materials has been invented for a girl with congenital insensitivity to pain and anhidrosis with severe heat intolerance. This innovative cooling jacket may be helpful for anhidrotic children in resource-poor situations. © 2017 Wiley Periodicals, Inc.

  10. Thermal-Hydraulic Transient Analysis of a Packed Particle Bed Reactor Fuel Element

    DTIC Science & Technology

    1990-06-01

    long fuel elements, arranged to form a core , were analyzed for an up-power transient from 0 MWt to approximately 18 MWt. The simple model significantly...VARIATIONS IN FUEL ELEMENT GEOMETRY ............. 60 4.4 VARIATIONS IN THE MANNER OF TRANSIENT CONTROL ..... 62 4.5 CORE REPRESENTATION BY MULTIPLE FUEL ...the HTGR , however, the PBR packs small fuel particles between inner and outer retention elements, designated as frits. The PBR is appropriate for a

  11. Comparing the cyclic behavior of concrete cylinders confined by shape memory alloy wire or steel jackets

    NASA Astrophysics Data System (ADS)

    Park, Joonam; Choi, Eunsoo; Park, Kyoungsoo; Kim, Hong-Taek

    2011-09-01

    Shape memory alloy (SMA) wire jackets for concrete are distinct from conventional jackets of steel or fiber reinforced polymer (FRP) since they provide active confinement which can be easily achieved due to the shape memory effect of SMAs. This study uses NiTiNb SMA wires of 1.0 mm diameter to confine concrete cylinders with the dimensions of 300 mm × 150 mm (L × D). The NiTiNb SMAs have a relatively wider temperature hysteresis than NiTi SMAs; thus, they are more suitable for the severe temperature-variation environments to which civil structures are exposed. Steel jackets of passive confinement are also prepared in order to compare the cyclic behavior of actively and passively confined concrete cylinders. For this purpose, monotonic and cyclic compressive loading tests are conducted to obtain axial and circumferential strain. Both strains are used to estimate the volumetric strains of concrete cylinders. Plastic strains from cyclic behavior are also estimated. For the cylinders jacketed by NiTiNb SMA wires, the monotonic axial behavior differs from the envelope of cyclic behavior. The plastic strains of the actively confined concrete show a similar trend to those of passive confinement. This study proposed plastic strain models for concrete confined by SMA wire or steel jackets. For the volumetric strain, the active jackets of NiTiNb SMA wires provide more energy dissipation than the passive jacket of steel.

  12. Dart model for irradiation-induced swelling of dispersion fuel elements including aluminum-fuel interaction

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Rest, J.; Hofman, G.L.

    1997-06-01

    The Dispersion Analysis Research Tool (DART) contains models for fission-gas induced fuel swelling, interaction of fuel with the matrix aluminum, resultant reaction-product swelling, and calculation of the stress gradient within the fuel particle. The effects of an aluminide shell on fuel particle swelling are evaluated. Validation of the model is demonstrated by a comparison of DART calculations of fuel swelling of U{sub 3}SiAl-Al and U{sub 3}Si{sub 2}-Al for various dispersion fuel element designs with the data. DART results are compared with data for fuel swelling Of U{sub 3}SiAl-Al in plate, tube, and rod configurations as a function of fission density.more » Plate and tube calculations were performed at a constant fuel temperature of 373 K and 518 K, respectively. An irradiation temperature of 518 K results in a calculated aluminide layer thickness for the Russian tube that is in the center of the measured range (16 {mu}m). Rod calculations were performed with a temperature gradient across the rod characterized by surface and central temperatures of 373 K and 423 K, respectively. The effective yield stress of irradiated Al matrix material and the aluminide was determined by comparing the results of DART calculations with postirradiation immersion volume measurement of U{sub 3}SiAl plates. The values for the effective yield stress were used in all subsequent simulations. The lower calculated fuel swelling in the rod-type element is due to an assumed biaxial stress state. Fuel swelling in plates results in plate thickness increase only. Likewise, in tubes, only the wall thickness increases. Irradiation experiments have shown that plate-type dispersion fuel elements can develop blisters or pillows at high U-235 burnup when fuel compounds exhibiting breakaway swelling are used at moderate to high fuel volume fractions. DART-calculated interaction layer thickness and fuel swelling follows the trends of the observations. 3 refs., 2 figs.« less

  13. Using the HELIOS facility for assessment of bundle-jacket thermal coupling in a CICC

    NASA Astrophysics Data System (ADS)

    Lacroix, B.; Rousset, B.; Cloez, H.; Decool, P.; Duchateau, J. L.; Hoa, C.; Luchier, N.; Nicollet, S.; Topin, F.

    2016-12-01

    In a Cable In Conduit Conductor (CICC) cooled by forced circulation of supercritical helium, the heat exchange in the bundle region can play a significant role for conductor safe operation, while remaining a quite uncertain parameter. Heat exchange between bundle and jacket depends on the relative contributions of convective heat transfer due to the helium flow inside the bundle and of thermal resistance due to the wrappings between the cable and the conduit. In order to qualify this thermal coupling at realistic operating conditions, a dedicated experiment on a 1.2 m sample of ITER Toroidal Field (TF) dummy conductor was designed and performed in the HELIOS test facility at CEA Grenoble. Several methods were envisaged, and the choice was made to assess bundle-jacket heat transfer coefficient by measuring the temperature of a solid copper cylinder inserted over the conductor jacket and submitted to heat deposition on its outer surface. The mock-up was manufactured and tested in spring 2015. Bundle-jacket heat transfer coefficient was found in the range 300-500 W m-2 K-1. Results analysis suggests that the order of magnitude of convective heat transfer coefficient inside bundle is closer to Colburn-Reynolds analogy than to Dittus-Boelter correlation, and that bundle-jacket thermal coupling is mainly limited by thermal resistance due to wrappings. A model based on an equivalent layer of stagnant helium between wraps and jacket was proposed and showed a good consistency with the experiment, with relevant values for the helium layer thickness.

  14. METHOD OF JACKETING URANIUM BODIES

    DOEpatents

    Maloney, J.O.; Haines, E.B.; Tepe, J.B.

    1958-08-26

    An improved process is presented for providing uranium slugs with thin walled aluminum jackets. Since aluminum has a slightiy higher coefficient of thermal expansion than does uraaium, both uranium slugs and aluminum cans are heated to an elevated temperature of about 180 C, and the slug are inserted in the cans at that temperature. During the subsequent cooling of the assembly, the aluminum contracts more than does the uranium and a tight shrink fit is thus assured.

  15. Steam jacket dynamics in underground coal gasification

    NASA Astrophysics Data System (ADS)

    Otto, Christopher; Kempka, Thomas

    2017-04-01

    Underground coal gasification (UCG) has the potential to increase the world-wide hydrocarbon reserves by utilization of deposits not economically mineable by conventional methods. In this context, UCG involves combusting coal in-situ to produce a high-calorific synthesis gas, which can be applied for electricity generation or chemical feedstock production. Apart from high economic potentials, in-situ combustion may cause environmental impacts such as groundwater pollution by by-product leakage. In order to prevent or significantly mitigate these potential environmental concerns, UCG reactors are generally operated below hydrostatic pressure to limit the outflow of UCG process fluids into overburden aquifers. This pressure difference effects groundwater inflow into the reactor and prevents the escape of product gas. In the close reactor vicinity, fluid flow determined by the evolving high reactor temperatures, resulting in the build-up of a steam jacket. Numerical modeling is one of the key components to study coupled processes in in-situ combustion. We employed the thermo-hydraulic numerical simulator MUFITS (BINMIXT module) to address the influence of reactor pressure dynamics as well as hydro-geological coal and caprock parameters on water inflow and steam jacket dynamics. The US field trials Hanna and Hoe Creek (Wyoming) were applied for 3D model validation in terms of water inflow matching, whereby the good agreement between our modeling results and the field data indicates that our model reflects the hydrothermal physics of the process. In summary, our validated model allows a fast prediction of the steam jacket dynamics as well as water in- and outflows, required to avoid aquifer contamination during the entire life cycle of in-situ combustion operations.

  16. Fuel elements of thermionic converters

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Hunter, R.L.; Gontar, A.S.; Nelidov, M.V.

    1997-01-01

    Work on thermionic nuclear power systems has been performed in Russia within the framework of the TOPAZ reactor program since the early 1960s. In the TOPAZ in-core thermionic convertor reactor design, the fuel element`s cladding is also the thermionic convertor`s emitter. Deformation of the emitter can lead to short-circuiting and is the primary cause of premature TRC failure. Such deformation can be the result of fuel swelling, thermocycling, or increased unilateral pressure on the emitter due to the release of gaseous fission products. Much of the work on TRCs has concentrated on preventing or mitigating emitter deformation by improving themore » following materials and structures: nuclear fuel; emitter materials; electrical insulators; moderator and reflector materials; and gas-exhaust device. In addition, considerable effort has been directed toward the development of experimental techniques that accurately mimic operational conditions and toward the creation of analytical and numerical models that allow operational conditions and behavior to be predicted without the expense and time demands of in-pile tests. New and modified materials and structures for the cores of thermionic NPSs and new fabrication processes for the materials have ensured the possibility of creating thermionic NPSs for a wide range of powers, from tens to several hundreds of kilowatts, with life spans of 5 to 10 years.« less

  17. DETECTION OF COATING FAILURES IN A NEUTRONIC REACTOR

    DOEpatents

    Snell, A.H.; Allison, S.K.

    1958-02-11

    This patent relates to water-cooled reactor systems and discloses a means to detect leaks in the jackets of jacketed fuel elements comprising a neutron detector located in the cooling water discharge pipe,the pipe being provided with an enlarged portion for housing the detector so that the latter is completely surrounded by the water in its passage through the pipe, said enlarged portion and detector being shielded from the reactor for the purpose of detecting only those delayed neutrons emitted in the cooling water and due to the latter picking up fission fragments from the defective fuel elements.

  18. Method and apparatus for diagnosing breached fuel elements

    DOEpatents

    Gross, K.C.; Lambert, J.D.B.; Nomura, S.

    1987-03-02

    The invention provides an apparatus and method for diagnosing breached fuel elements in a nuclear reactor. A detection system measures the activity of isotopes from the cover gas in the reactor. A data acquisition and processing system monitors the detection system and corrects for the effects of the cover-gas clean up system on the measured activity and further calculates the derivative curve of the corrected activity as a function of time. A plotting system graphs the derivative curve, which represents the instantaneous release rate of fission gas from a breached fuel element. 8 figs.

  19. PROCESS OF DISSOLVING FUEL ELEMENTS OF NUCLEAR REACTORS

    DOEpatents

    Wall, E.M.V.; Bauer, D.T.; Hahn, H.T.

    1963-09-01

    A process is described for dissolving stainless-steelor zirconium-clad uranium dioxide fuel elements by immersing the elements in molten lead chloride, adding copper, cuprous chloride, or cupric chloride as a catalyst and passing chlorine through the salt mixture. (AEC)

  20. A jacket for assisting sensorimotor-related impairments and spatial perception

    NASA Astrophysics Data System (ADS)

    Blumenstein, Tobias; Turova, Varvara; Alves-Pinto, Ana; Lampe, Renée

    2017-04-01

    A sensorimotor jacket, which is able to measure distances to nearby objects with ultrasonic sensors and to transmit information about distances via vibrating transducers, has been designed with the aim of improving the spatial awareness of patients with cerebral palsy and to facilitate spatial orientation for blind people. The efficiency was tested for patients diagnosed with cerebral palsy, blind participants and healthy people. A positive impact of the sensorimotor jacket on the performance in a spatial task has been established both in patients with cerebral palsy and blind participants. Moreover, for patients with cerebral palsy, the training effect was visible after only three training exercises.

  1. DART model for irradiation-induced swelling of dispersion fuel elements including aluminum-fuel interaction

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Rest, J.; Hofman, G.L.

    1997-12-01

    The Dispersion Analysis Research Tool (DART) contains models for fission-gas-induced fuel swelling, interaction of fuel with the matrix aluminum, for the resultant reaction-product swelling, and for the calculation of the stress gradient within the fuel particle. The effects of an aluminide shell on fuel particle swelling are evaluated. Validation of the model is demonstrated by a comparison of DART calculations of fuel swelling of U{sub 3}SiAl-Al and U{sub 3}Si{sub 2}-Al for various dispersion fuel element designs with the data.

  2. FUEL ELEMENT FOR A NUCLEAR REACTOR

    DOEpatents

    Davidson, J.K.

    1963-11-19

    A fuel element structure particularly useful in high temperature nuclear reactors is presented. Basically, the structure comprises two coaxial graphite sleeves integrally joined together by radial fins. Due to the high structural strength of graphite at high temperatures and the rigidity of this structure, nuclear fuel encased within the inner sleeve in contiguous relation therewith is supported and prevented from expanding radially at high temperatures. Thus, the necessity of relying on the usual cladding materials with relatively low temperature limitations for structural strength is removed. (AEC)

  3. Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

    NASA Technical Reports Server (NTRS)

    Bradley, David E.; Mireles, Omar R.; Hickman, Robert R.

    2011-01-01

    Deep space missions with large payloads require high specific impulse (Isp) and relatively high thrust in order to achieve mission goals in reasonable time frames. Conventional, storable propellants produce average Isp. Nuclear thermal rockets (NTR) capable of high Isp thrust have been proposed. NTR employs heat produced by fission reaction to heat and therefore accelerate hydrogen which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high temperature hydrogen exposure on fuel elements is limited. The primary concern is the mechanical failure of fuel elements which employ high-melting-point metals, ceramics or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. It is not necessary to include fissile material in test samples intended to explore high temperature hydrogen exposure of the structural support matrices. A small-scale test bed designed to heat fuel element samples via non-contact RF heating and expose samples to hydrogen is being developed to assist in optimal material and manufacturing process selection without employing fissile material. This paper details the test bed design and results of testing conducted to date.

  4. METHOD OF JACKETING FISSIONABLE MATERIALS

    DOEpatents

    Foster, L.M.

    1959-02-01

    An improvement is presented in the jacketing of a metal body accomplished by electroplating upon that portion of the metal container to be protected from the bonding material a niatcrial such as Cr which is impermeable to the bonding material. After the bonding operation the electroplate is removed and the metal container surfuce, unimpaired, may be welded to a cap which effects a closure. Generally in such an operation the metal body is U, the metal container is Al and the bonding material is a Zn alloy.

  5. 46 CFR 185.730 - Servicing of inflatable liferafts, inflatable buoyant apparatus, inflatable life jackets, and...

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... apparatus, inflatable life jackets, and inflated rescue boats. 185.730 Section 185.730 Shipping COAST GUARD... liferafts, inflatable buoyant apparatus, inflatable life jackets, and inflated rescue boats. (a) An... inflated rescue boats must be in accordance with the manufacturer's instructions. All repairs must be made...

  6. 46 CFR 185.730 - Servicing of inflatable liferafts, inflatable buoyant apparatus, inflatable life jackets, and...

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... apparatus, inflatable life jackets, and inflated rescue boats. 185.730 Section 185.730 Shipping COAST GUARD... liferafts, inflatable buoyant apparatus, inflatable life jackets, and inflated rescue boats. (a) An... inflated rescue boats must be in accordance with the manufacturer's instructions. All repairs must be made...

  7. 46 CFR 185.730 - Servicing of inflatable liferafts, inflatable buoyant apparatus, inflatable life jackets, and...

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... apparatus, inflatable life jackets, and inflated rescue boats. 185.730 Section 185.730 Shipping COAST GUARD... liferafts, inflatable buoyant apparatus, inflatable life jackets, and inflated rescue boats. (a) An... inflated rescue boats must be in accordance with the manufacturer's instructions. All repairs must be made...

  8. Possible consequences of operation with KIVN fuel elements in K Zircaloy process tubes

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Carlson, P.A.

    1963-08-06

    From considerations of the results of experimental simulations of non-axial placement of fuel elements in process tubes and in-reactor experience, it is concluded that the ultimate outcome of a charging error which results in operation with one or more unsupported fuel elements in a K Zircaloy-2 process tube would be multiple fuel failure and failure of the process tube. The outcome of the accident is determined by the speed with which the fuel failure is detected and the reactor is shut down. The release of fission products would be expected to be no greater than that which has occurred followingmore » severe fuel failure incidents. The highest probability for fission product release occurs during the discharge of failed fuel elements, when a small fraction of the exposed uranium of the fuel element may be oxidized when exposed to air before the element falls into the water-filled discharge chute. The confinement and fog spray facilities were installed to reduce the amount of fission products which might escape from the reactor building after such an event.« less

  9. THE MAN&RSQUO;S JACKET DESIGN FOR DISASSEMBLY: AN IMPLEMENTATION OF C2CAD FRAMEWORK

    EPA Science Inventory

    The C2CAD model served as the basis in the man’s jacket design and production. In man’s jackets, both natural and synthetic materials are commonly used for fabrics, threads, and buttons. To promote disassembly and value retention, we minimized material diversity an...

  10. 46 CFR 122.730 - Servicing of inflatable liferafts, inflatable buoyant apparatus, inflatable life jackets, and...

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... apparatus, inflatable life jackets, and inflated rescue boats. 122.730 Section 122.730 Shipping COAST GUARD..., inflatable life jackets, and inflated rescue boats. (a) An inflatable liferaft or inflatable buoyant... other standard specified by the Commandant. (e) Repair and maintenance of inflated rescue boats must be...

  11. 46 CFR 122.730 - Servicing of inflatable liferafts, inflatable buoyant apparatus, inflatable life jackets, and...

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... apparatus, inflatable life jackets, and inflated rescue boats. 122.730 Section 122.730 Shipping COAST GUARD..., inflatable life jackets, and inflated rescue boats. (a) An inflatable liferaft or inflatable buoyant... other standard specified by the Commandant. (e) Repair and maintenance of inflated rescue boats must be...

  12. 46 CFR 122.730 - Servicing of inflatable liferafts, inflatable buoyant apparatus, inflatable life jackets, and...

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... apparatus, inflatable life jackets, and inflated rescue boats. 122.730 Section 122.730 Shipping COAST GUARD..., inflatable life jackets, and inflated rescue boats. (a) An inflatable liferaft or inflatable buoyant... other standard specified by the Commandant. (e) Repair and maintenance of inflated rescue boats must be...

  13. Systematic determination of perfluoroalkyl and polyfluoroalkyl substances (PFASs) in outdoor jackets.

    PubMed

    Gremmel, Christoph; Frömel, Tobias; Knepper, Thomas P

    2016-10-01

    Sixteen outdoor jackets were purchased in 2011/12 and analyzed for 23 different perfluoroalkyl and polyfluoroalkyl substances (PFASs). The jackets were selected based on their origin of production, price, market, and textile, such as polyester, nylon, polyamide, and content of poly(tetrafluoroethylene) membranes. Two robust analytical methods based on high pressure liquid chromatography combined with tandem mass spectrometry, as well as two liquid extractions, were developed enabling the analysis of PFASs with widely different physico-chemical properties. The jackets were found to contain PFASs in a range between 0.03 and 719 μg/m(2). Perfluorooctanoic acid (PFOA) was omnipresent (0.02-171 μg/m(2)), although at lower concentrations compared to the precursors of perfluoroalkyl carboxylic acids (PFCAs), namely fluorotelomer alcohols (FTOHs) (<0.001-698 μg/m(2)). Perfluoroalkane sulfonic acids and their putative precursors, in particular perfluoroalkane sulfonamides, were detected much less frequently at concentrations up to 5 μg/m(2). To determine the effect of the volatility of FTOHs, four selected jackets were stored in a sealed bag in the dark at room temperature and re-analyzed after 3.5 years. Only 10%-20% of the initial concentration of 8:2-FTOH and 20%-50% of 10:2-FTOH were found, whereas the concentrations of PFOA and perfluorodecanoic acid increased significantly. This supports the hypothesis that PFAS concentrations in textiles are also strongly dependent on age, and conditions of transport and storage. Copyright © 2016 Elsevier Ltd. All rights reserved.

  14. Nuclear Cryogenic Propulsion Stage (NCPS) Fuel Element Testing in the Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

    NASA Technical Reports Server (NTRS)

    Emrich, William J., Jr.

    2017-01-01

    To satisfy the Nuclear Cryogenic Propulsion Stage (NCPS) testing milestone, a graphite composite fuel element using a uranium simulant was received from the Oakridge National Lab and tested in the Nuclear Thermal Rocket Element Environmental Simulator (NTREES) at various operating conditions. The nominal operating conditions required to satisfy the milestone consisted of running the fuel element for a few minutes at a temperature of at least 2000 K with flowing hydrogen. This milestone test was successfully accomplished without incident.

  15. Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Montierth, Leland M.

    2016-07-19

    The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element designmore » for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.« less

  16. Induction Heating Model of Cermet Fuel Element Environmental Test (CFEET)

    NASA Technical Reports Server (NTRS)

    Gomez, Carlos F.; Bradley, D. E.; Cavender, D. P.; Mireles, O. R.; Hickman, R. R.; Trent, D.; Stewart, E.

    2013-01-01

    Deep space missions with large payloads require high specific impulse and relatively high thrust to achieve mission goals in reasonable time frames. Nuclear Thermal Rockets (NTR) are capable of producing a high specific impulse by employing heat produced by a fission reactor to heat and therefore accelerate hydrogen through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000 K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited. The primary concern is the mechanical failure of fuel elements due to large thermal gradients; therefore, high-melting-point ceramics-metallic matrix composites (cermets) are one of the fuels under consideration as part of the Nuclear Cryogenic Propulsion Stage (NCPS) Advance Exploration System (AES) technology project at the Marshall Space Flight Center. The purpose of testing and analytical modeling is to determine their ability to survive and maintain thermal performance in a prototypical NTR reactor environment of exposure to hydrogen at very high temperatures and obtain data to assess the properties of the non-nuclear support materials. The fission process and the resulting heating performance are well known and do not require that active fissile material to be integrated in this testing. A small-scale test bed; Compact Fuel Element Environmental Tester (CFEET), designed to heat fuel element samples via induction heating and expose samples to hydrogen is being developed at MSFC to assist in optimal material and manufacturing process selection without utilizing fissile material. This paper details the analytical approach to help design and optimize the test bed using COMSOL Multiphysics for predicting thermal gradients induced by electromagnetic heating (Induction heating) and Thermal Desktop for radiation calculations.

  17. Methods for making a porous nuclear fuel element

    DOEpatents

    Youchison, Dennis L; Williams, Brian E; Benander, Robert E

    2014-12-30

    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.

  18. Clinical Performance of Pedo Jacket Crowns in Maxillary Anterior Primary Teeth.

    PubMed

    Castro, Aimee; Badr, Sherine B Y; El-Badrawy, Wafa; Kulkarni, Gajanan

    2016-09-15

    To assess the clinical performance of Pedo Jacket crowns for restoration of carious primary anterior teeth. A total of 129 carious primary incisors and canines of 48 children younger than 71 months of age- were restored with Pedo Jacket crowns and resin-modified glass ionomer cementation. They were assessed for: ease of use; presence of recurrent decay; wear; partial or complete loss of the crown; color stability; gingival health; and overall clinical success over a 12-month follow-up. The patient's behavior at the restorative appointment during crown placement was also assessed. An overall clinical success of 89.5 percent of the teeth in 87.3 percent of the children was seen one year later. The crowns were easy to use, even in uncooperative children. The color stability, wear, plaque accumulation, and gingival health were acceptable. Discoloration, wear, or complete loss of the crown were found in 13.1 percent, 5.4 percent, and 7.6 percent of children, respectively. Although not statistically significant, failures were associated with poor patient cooperation at the time of crown placement, poor oral hygiene, or operator error. Pedo Jacket crowns are a viable treatment alternative for carious maxillary primary anterior teeth.

  19. Nuclear fuel element nut retainer cup. [PWR

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Walton, L.A.

    1977-07-19

    A typical embodiment has an end fitting for a nuclear reactor fuel element that is joined to the control rod guide tubes by means of a nut plate assembly. The nut plate assembly has an array of nuts, each engaging the respective threaded end of the control rod guide tubes. The nuts, moreover, are retained on the plate during handling and before fuel element assembly by means of hollow cylindrical locking cups that are brazed to the plate and loosely circumscribe the individual enclosed nuts. After the nuts are threaded onto the respective guide tube ends, the locking cups aremore » partially deformed to prevent one or more of the nuts from working loose during reactor operation. The locking cups also prevent loose or broken end fitting parts from becoming entrained in the reactor coolant.« less

  20. Travel health: a survey of life jacket designs currently in use on commercial aircraft.

    PubMed

    Bauer, Irmgard L

    2002-01-01

    Although any travel harbors potential health problems, travel across water poses the additional risk of involuntary contact with water in the event of an emergency. Studies suggest that fatal boat-related accidents have occurred due to the passengers' inability to don their life jackets. With the dramatic increase of long-haul flights, it was decided to investigate if potential safety hazards can be identified leading to similar problems in relation to the life jackets provided onboard aircraft. It was assumed that the variety of designs combined with a possible lack of attention paid to safety instructions could represent such a problem. Safety instruction cards in passengers' seat pockets usually depict the life jacket model used on a particular aircraft. Ninety-eight such safety cards from 53 airlines were analyzed and categorized to identify the range and variety of designs currently in use. Twelve different categories of life jacket models could be distinguished with some airlines using more than one model. It was also found that not all cards show a clear depiction of the models used, the cards of two airlines show two different designs on the same cards, and two other companies chose to change the model halfway through the pictured donning procedure. There is a wide variety of life jacket designs in use with the safety cards not always giving accurate instructions. This variety also has implications for the safety demonstration at the beginning of a flight. Further research is recommended into a range of topics, one being the quality and content of these instructions, and also into other personal floating devices provided for the safety of travelers in the event of an emergency over water.

  1. FUEL ELEMENT AND METHOD OF PREPARATION

    DOEpatents

    Kingston, W.E.

    1961-04-25

    A nuclear fuel element in the form of a wire is reported. A bar of uranium is enclosed in a thin layer of aluminum and the composite is sheathed in beryllium, zirconium, or stainnless steel. The sheathed article is then drawn to wire form, heated to alloy the aluminum with both uranium and sheath, and finally cold worked.

  2. URANIUM OXIDE-CONTAINING FUEL ELEMENT COMPOSITION AND METHOD OF MAKING SAME

    DOEpatents

    Handwerk, J.H.; Noland, R.A.; Walker, D.E.

    1957-09-10

    In the past, bodies formed of a mixture of uranium dioxide and aluminum powder have been used in fuel elements; however, these mixtures were found not to be suitable when exposed to temperatures of about 600 deg C, because at such high temperatures the fuel elements were distorted. If uranosic oxide, U/sub 3/O/sub 8/, is substituted for UO/sub 2/, the mechanical properties are not impaired when these materials are used at about 600 deg C and no distortion takes place. The uranosic oxide and aluminum, both in powder form, are first mixed, and after a homogeneous mixture has been obtained, are shaped into fuel elements by extrusion at elevated temperature. Magnesium powder may be used in place of the aluminum.

  3. Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

    NASA Technical Reports Server (NTRS)

    Bradley, D. E.; Mireles, O. R.; Hickman, R. R.

    2011-01-01

    Deep space missions with large payloads require high specific impulse and relatively high thrust to achieve mission goals in reasonable time frames.1,2 Conventional storable propellants produce average specific impulse. Nuclear thermal rockets capable of producing high specific impulse are proposed. Nuclear thermal rockets employ heat produced by fission reaction to heat and therefore accelerate hydrogen, which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000 K), and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited.3 The primary concern is the mechanical failure of fuel elements that employ high-melting-point metals, ceramics, or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. The purpose of the testing is to obtain data to assess the properties of the non-nuclear support materials, as-fabricated, and determine their ability to survive and maintain thermal performance in a prototypical NTR reactor environment of exposure to hydrogen at very high temperatures. The fission process of the planned fissile material and the resulting heating performance is well known and does not therefore require that active fissile material be integrated in this testing. A small-scale test bed designed to heat fuel element samples via non-contact radio frequency heating and expose samples to hydrogen is being developed to assist in optimal material and manufacturing process selection without employing fissile material. This paper details the test bed design and results of testing conducted to date.

  4. Review of Rover fuel element protective coating development at Los Alamos

    NASA Technical Reports Server (NTRS)

    Wallace, Terry C.

    1991-01-01

    The Los Alamos Scientific Laboratory (LASL) entered the nuclear propulsion field in 1955 and began work on all aspects of a nuclear propulsion program with a target exhaust temperature of about 2750 K. A very extensive chemical vapor deposition coating technology for preventing catastrophic corrosion of reactor core components by the high temperature, high pressure hydrogen propellant gas was developed. Over the 17-year term of the program, more than 50,000 fuel elements were coated and evaluated. Advances in performance were achieved only through closely coupled interaction between the developing fuel element fabrication and protective coating technologies. The endurance of fuel elements in high temperature, high pressure hydrogen environment increased from several minutes at 2000 K exit gas temperature to 2 hours at 2440 K exit gas temperature in a reactor test and 10 hours at 2350 K exit gas temperature in a hot gas test. The purpose of this paper is to highlight the rationale for selection of coating materials used (NbC and ZrC), identify critical fuel element-coat interactions that had to be modified to increase system performance, and review the evolution of protective coating technology.

  5. The manufacture of LEU fuel elements at Dounreay

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Gibson, J.

    1997-08-01

    Two LEU test elements are being manufactured at Dounreay for test irradiation in the HFR at Petten, The Netherlands. This paper describes the installation of equipment and the development of the fabrication and inspection techniques necessary for the manufacture of LEU fuel plates. The author`s experience in overcoming the technical problems of stray fuel particles, dog-boning, uranium homogeneity and the measurement of uranium distribution is also described.

  6. RECONDITIONING FUEL ELEMENTS

    DOEpatents

    Brandt, H.L.

    1962-02-20

    A process is given for decanning fuel elements that consist of a uranium core, an intermediate section either of bronze, silicon, Al-Si, and uranium silicide layers or of lead, Al-Si, and uranium silicide layers around said core, and an aluminum can bonded to said intermediate section. The aluminum can is dissolved in a solution of sodium hydroxide (9 to 20 wt%) and sodium nitrate (35 to 12 wt %), and the layers of the intermediate section are dissolved in a boiling sodium hydroxide solution of a minimum concentration of 50 wt%. (AEC) A method of selectively reducing plutonium oxides and the rare earth oxides but not uranium oxides is described which comprises placing the oxides in a molten solvent of zinc or cadmium and then adding metallic uranium as a reducing agent. (AEC)

  7. FABRICATION OF TUBE TYPE FUEL ELEMENT FOR NUCLEAR REACTORS

    DOEpatents

    Loeb, E.; Nicklas, J.H.

    1959-02-01

    A method of fabricating a nuclear reactor fuel element is given. It consists essentially of fixing two tubes in concentric relationship with respect to one another to provide an annulus therebetween, filling the annulus with a fissionablematerial-containing powder, compacting the powder material within the annulus and closing the ends thereof. The powder material is further compacted by swaging the inner surface of the inner tube to increase its diameter while maintaining the original size of the outer tube. This process results in reduced fabrication costs of powdered fissionable material type fuel elements and a substantial reduction in the peak core temperatures while materially enhancing the heat removal characteristics.

  8. Nuclear breeder reactor fuel element with silicon carbide getter

    DOEpatents

    Christiansen, David W.; Karnesky, Richard A.

    1987-01-01

    An improved cesium getter 28 is provided in a breeder reactor fuel element or pin in the form of an extended surface area, low density element formed in one embodiment as a helically wound foil 30 located with silicon carbide, and located at the upper end of the fertile material upper blanket 20.

  9. NUCLEAR REACTOR FUEL ELEMENTS AND METHOD OF PREPARATION

    DOEpatents

    Kingston, W.E.; Kopelman, B.; Hausner, H.H.

    1963-07-01

    A fuel element consisting of uranium nitride and uranium carbide in the form of discrete particles in a solid coherent matrix of a metal such as steel, beryllium, uranium, or zirconium and clad with a metal such as steel, aluminum, zirconium, or beryllium is described. The element is made by mixing powdered uranium nitride and uranium carbide with powdered matrix metal, then compacting and sintering the mixture. (AEC)

  10. Viability of Bismuth as a Green Substitute for Lead in Jacketed .357 Magnum Revolver Bullets

    NASA Astrophysics Data System (ADS)

    Jenkins, Joel

    In seeking to develop environmentally friendly lead-free non-toxic bullets, the research ballistically evaluated the performance of copper-jacketed handgun bullets containing a pure bismuth core. The lead was first removed from 140 grain Hornady(TM) XTPRTM bullets of 38 caliber (.357 diameter) by melting. The empty jackets were then refilled with pure bismuth, including the forming of a correctly sized hollow-point cavity. Due to the lower density of bismuth as compared to lead, the bismuth-cored bullets consistently weighed 125 gains. Conveniently this allowed direct comparison to commercially available 125 grain Hornady(TM) XTPRTM lead-cored bullets of 38 caliber. Both bismuth-cored and lead-cored versions of the 125 grain bullets had identical nose dimensions and jacket material, the only dimensional difference being the bullet length below the cannelure. Shooting took place at an outdoor range using a 357 Magnum Ruger(TM) SP101RTM revolver with 3" barrel as the test weapon. FBI protocols were followed when firing through clothing, wallboard, plywood, steel plates and laminated glass. Wound paths and bullets were captured in ballistic gelatin, with data collected for velocity, penetration, expansion, and weight retention. Bismuth compared favorably with lead in all but the laminated glass test, where it under penetrated due to jacket separation.

  11. Multidisciplinary Simulation of Graphite-Composite and Cermet Fuel Elements for NTP Point of Departure Designs

    NASA Technical Reports Server (NTRS)

    Stewart, Mark E.; Schnitzler, Bruce G.

    2015-01-01

    This paper compares the expected performance of two Nuclear Thermal Propulsion fuel types. High fidelity, fluid/thermal/structural + neutronic simulations help predict the performance of graphite-composite and cermet fuel types from point of departure engine designs from the Nuclear Thermal Propulsion project. Materials and nuclear reactivity issues are reviewed for each fuel type. Thermal/structural simulations predict thermal stresses in the fuel and thermal expansion mis-match stresses in the coatings. Fluid/thermal/structural/neutronic simulations provide predictions for full fuel elements. Although NTP engines will utilize many existing chemical engine components and technologies, nuclear fuel elements are a less developed engine component and introduce design uncertainty. Consequently, these fuel element simulations provide important insights into NTP engine performance.

  12. Nuclear breeder reactor fuel element with axial tandem stacking and getter

    DOEpatents

    Gibby, Ronald L.; Lawrence, Leo A.; Woodley, Robert E.; Wilson, Charles N.; Weber, Edward T.; Johnson, Carl E.

    1981-01-01

    A breeder reactor fuel element having a tandem arrangement of fissile and fertile fuel with a getter for fission product cesium disposed between the fissile and fertile sections. The getter is effective at reactor operating temperatures to isolate the cesium generated by the fissile material from reacting with the fertile fuel section.

  13. Determination of trace elements in automotive fuels by filter furnace atomic absorption spectrometry

    NASA Astrophysics Data System (ADS)

    Anselmi, Anna; Tittarelli, Paolo; Katskov, Dmitri A.

    2002-03-01

    The determination of Cd, Cr, Cu, Pb and Ni was performed in gasoline and diesel fuel samples by electrothermal atomic absorption spectrometry using the Transverse Heated Filter Atomizer (THFA). Thermal conditions were experimentally defined for the investigated elements. The elements were analyzed without addition of chemical modifiers, using organometallic standards for the calibration. Forty-microliter samples were injected into the THFA. Gasoline samples were analyzed directly, while diesel fuel samples were diluted 1:4 with n-heptane. The following characteristic masses were obtained: 0.8 pg Cd, 6.4 pg Cr, 12 pg Cu, 17 pg Pb and 27 pg Ni. The limits of determination for gasoline samples were 0.13 μg/kg Cd, 0.4 μg/kg Cr, 0.9 μg/kg Cu, 1.5 μg/kg Pb and 2.5 μg/kg Ni. The corresponding limit of determination for diesel fuel samples was approximately four times higher for all elements. The element recovery was performed using the addition of organometallic compounds to gasoline and diesel fuel samples and was between 85 and 105% for all elements investigated.

  14. METHOD OF MAKING WIRE FUEL ELEMENTS

    DOEpatents

    Zambrow, J.L.

    1960-08-01

    A method is given for making a nuclear reactor fuel element in the form of a uranium-bearing wire clad with zirconium. A uranium bar is enclosed in a zirconium sheath which is coated with an oxide of magnesium, beryllium, or zirconium. The sheathed bar is then placed in a steel tube and reduced to the desired diameter by swaging at 800 to 900 deg C, after which the steel and oxide are removed.

  15. Apollo 12 Mission image - Alan Bean unloads ALSEP RTG fuel element

    NASA Image and Video Library

    1969-11-19

    AS12-46-6790 (19 Nov. 1969) --- Astronaut Alan L. Bean, lunar module pilot, is photographed at quadrant II of the Lunar Module (LM) during the first Apollo 12 extravehicular activity (EVA) on the moon. This picture was taken by astronaut Charles Conrad Jr., commander. Here, Bean is using a fuel transfer tool to remove the fuel element from the fuel cask mounted on the LM's descent stage. The fuel element was then placed in the Radioisotope Thermoelectric Generator (RTG), the power source for the Apollo Lunar Surface Experiments Package (ALSEP) which was deployed on the moon by the two astronauts. The RTG is next to Bean's right leg. While astronauts Conrad and Bean descended in the LM "Intrepid" to explore the Ocean of Storms region of the moon, astronaut Richard F. Gordon Jr., command module pilot, remained with the Command and Service Modules (CSM) "Yankee Clipper" in lunar orbit.

  16. PLUTONIUM FUEL RODS FOR PREPARATION OF TRANSPLUTONIC ELEMENTS

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Bailey, W.J.

    1962-02-01

    Production by coextrusion of metallurgically bonded, Alclad, Al-7.35 wt% Pu alloy fuel rods with integral ends is discussed. The rods had a diameter of 0.94 in., length of, 60 in., and a nominal cladding thickness of 0.070 in. The Pu concentration was maintained at 83.3 g/rod. The coextrusion billets can be assembled with fuel cores in the as-cast condition. The casting hot-tops can be returned to the process stream. The process is useful for preparing transplutonic elements and production of high-exposure Pu. (J.R.D.)

  17. FISSILE MATERIAL AND FUEL ELEMENTS FOR NEUTRONIC REACTORS

    DOEpatents

    Shaner, B.E.

    1961-08-15

    The fissile material consists of about 64 to 70% (weight) zirconium dioxide, 15 to 19% uranium dioxide, and 8 to 17% calcium oxide. The fissile material is formed into sintered composites which are disposed in a compartmented fuel element, comprising essentially a flat filler plate having a plurality of compartments therein, enclosed in cladding plates of the same material as the filler plate. The resultant fuel has good resistance to corrosion in high temperature pressurized water, good dimensional stability to elevated temperatures, and good resistance to thermal shock. (AEC)

  18. Expert system for surveillance and diagnosis of breach fuel elements

    DOEpatents

    Gross, Kenny C.

    1989-01-01

    An apparatus and method are disclosed for surveillance and diagnosis of breached fuel elements in a nuclear reactor. A delayed neutron monitoring system provides output signals indicating the delayed neutron activity and age and the equivalent recoil areas of a breached fuel element. Sensors are used to provide outputs indicating the status of each component of the delayed neutron monitoring system. Detectors also generate output signals indicating the reactor power level and the primary coolant flow rate of the reactor. The outputs from the detectors and sensors are interfaced with an artificial intelligence-based knowledge system which implements predetermined logic and generates output signals indicating the operability of the reactor.

  19. Multiphysics Modeling of a Single Channel in a Nuclear Thermal Propulsion Grooved Ring Fuel Element

    NASA Technical Reports Server (NTRS)

    Kim, Tony; Emrich, William J., Jr.; Barkett, Laura A.; Mathias, Adam D.; Cassibry, Jason T.

    2013-01-01

    In the past, fuel rods have been used in nuclear propulsion applications. A new fuel element concept that reduces weight and increases efficiency uses a stack of grooved discs. Each fuel element is a flat disc with a hole on the interior and grooves across the top. Many grooved ring fuel elements for use in nuclear thermal propulsion systems have been modeled, and a single flow channel for each design has been analyzed. For increased efficiency, a fuel element with a higher surface-area-to-volume ratio is ideal. When grooves are shallower, i.e., they have a lower surface area, the results show that the exit temperature is higher. By coupling the physics of turbulence with those of heat transfer, the effects on the cooler gas flowing through the grooves of the thermally excited solid can be predicted. Parametric studies were done to show how a pressure drop across the axial length of the channels will affect the exit temperatures of the gas. Geometric optimization was done to show the behaviors that result from the manipulation of various parameters. Temperature profiles of the solid and gas showed that more structural optimization is needed to produce the desired results. Keywords: Nuclear Thermal Propulsion, Fuel Element, Heat Transfer, Computational Fluid Dynamics, Coupled Physics Computations, Finite Element Analysis

  20. The technique of numerical research of cooling medium flow in the water jacket of self-lubricated bearing

    NASA Astrophysics Data System (ADS)

    Raikovskiy, N. A.; Tretyakov, A. V.; Abramov, S. A.; Nazmeev, F. G.; Pavlichev, S. V.

    2017-08-01

    The paper presents a numerical study method of the cooling medium flowing in the water jacket of self-lubricating sliding bearing based on ANSYS CFX. The results of numerical calculations have satisfactory convergence with the empirical data obtained on the testbed. Verification data confirm the possibility of applying this numerical technique for the analysis of coolant flowings in the self-lubricating bearing containing the water jacket.

  1. Fossil fuel furnace reactor

    DOEpatents

    Parkinson, William J.

    1987-01-01

    A fossil fuel furnace reactor is provided for simulating a continuous processing plant with a batch reactor. An internal reaction vessel contains a batch of shale oil, with the vessel having a relatively thin wall thickness for a heat transfer rate effective to simulate a process temperature history in the selected continuous processing plant. A heater jacket is disposed about the reactor vessel and defines a number of independent controllable temperature zones axially spaced along the reaction vessel. Each temperature zone can be energized to simulate a time-temperature history of process material through the continuous plant. A pressure vessel contains both the heater jacket and the reaction vessel at an operating pressure functionally selected to simulate the continuous processing plant. The process yield from the oil shale may be used as feedback information to software simulating operation of the continuous plant to provide operating parameters, i.e., temperature profiles, ambient atmosphere, operating pressure, material feed rates, etc., for simulation in the batch reactor.

  2. A comparison study of offshore wind support structures with monopiles and jackets for U.S. waters

    NASA Astrophysics Data System (ADS)

    Damiani, R.; Dykes, K.; Scott, G.

    2016-09-01

    U.S. experience in offshore wind is limited, and high costs are expected unless innovations are introduced in one or multiple aspects of the project, from the installed technology to the balance of system (BOS). The substructure is the main single component responsible for the BOS capital expenditure (CapEx) and thus one that, if improved, could yield significant levelized cost of energy (LCOE) savings. For projects in U.S. waters, multimember lattice structures (also known as jackets) can render required stiffness for transitional water depths at potentially lower costs than monopiles (MPs). In this study, we used a systems engineering approach to evaluate the LCOE of prototypical wind power plants at six locations along the eastern seaboard and the Gulf of Mexico for both types of support structures. Using a reference wind turbine and actual metocean conditions for the selected sites, we calculated loads for a parked and an operational situation, and we optimized the MP- and jacket-based support structures to minimize their overall mass. Using a suite of cost models, we then computed their associated LCOE. For all water depths, the MP-based configurations were heavier than their jacket counterparts, but the overall costs for the MPs were less than they were for jackets up to depths of slightly less than 30m. When the associated manufacturing and installation costs were included, jackets resulted in lower LCOE for depths greater than 40m. These results can be used by U.S. stakeholders to understand the potential for different technologies at different sites, but the methodology illustrated in this study can be further employed to analyze the effects of innovations and design choices throughout wind power plant systems.

  3. Optimal startup control of a jacketed tubular reactor.

    NASA Technical Reports Server (NTRS)

    Hahn, D. R.; Fan, L. T.; Hwang, C. L.

    1971-01-01

    The optimal startup policy of a jacketed tubular reactor, in which a first-order, reversible, exothermic reaction takes place, is presented. A distributed maximum principle is presented for determining weak necessary conditions for optimality of a diffusional distributed parameter system. A numerical technique is developed for practical implementation of the distributed maximum principle. This involves the sequential solution of the state and adjoint equations, in conjunction with a functional gradient technique for iteratively improving the control function.

  4. NUCLEAR REACTOR

    DOEpatents

    Treshow, M.

    1958-08-19

    A neuclear reactor is described of the heterogeneous type and employing replaceable tubular fuel elements and heavy water as a coolant and moderator. A pluraltty of fuel tubesa having their axes parallel, extend through a tank type pressure vessel which contatns the liquid moderator. The fuel elements are disposed within the fuel tubes in the reaetive portion of the pressure vessel during normal operation and the fuel tubes have removable plug members at each end to permit charging and discharging of the fuel elements. The fuel elements are cylindrical strands of jacketed fissionable material having helical exterior ribs. A bundle of fuel elements are held within each fuel tube with their longitudinal axes parallel, the ribs serving to space them apart along their lengths. Coolant liquid is circulated through the fuel tubes between the spaced fuel elements. Suitable control rod and monitoring means are provided for controlling the reactor.

  5. Modeling and Simulation of a Nuclear Fuel Element Test Section

    NASA Technical Reports Server (NTRS)

    Moran, Robert P.; Emrich, William

    2011-01-01

    "The Nuclear Thermal Rocket Element Environmental Simulator" test section closely simulates the internal operating conditions of a thermal nuclear rocket. The purpose of testing is to determine the ideal fuel rod characteristics for optimum thermal heat transfer to their hydrogen cooling/working fluid while still maintaining fuel rod structural integrity. Working fluid exhaust temperatures of up to 5,000 degrees Fahrenheit can be encountered. The exhaust gas is rendered inert and massively reduced in temperature for analysis using a combination of water cooling channels and cool N2 gas injectors in the H2-N2 mixer portion of the test section. An extensive thermal fluid analysis was performed in support of the engineering design of the H2-N2 mixer in order to determine the maximum "mass flow rate"-"operating temperature" curve of the fuel elements hydrogen exhaust gas based on the test facilities available cooling N2 mass flow rate as the limiting factor.

  6. Expert system for surveillance and diagnosis of breach fuel elements

    DOEpatents

    Gross, K.C.

    1988-01-21

    An apparatus and method are disclosed for surveillance and diagnosis of breached fuel elements in a nuclear reactor. A delayed neutron monitoring system provides output signals indicating the delayed neutron activity and age and the equivalent recoil area of a breached fuel element. Sensors are used to provide outputs indicating the status of each component of the delayed neutron monitoring system. Detectors also generate output signals indicating the reactor power level and the primary coolant flow rate of the reactor. The outputs from the detectors and sensors are interfaced with an artificial intelligence-based knowledge system which implements predetermined logic and generates output signals indicating the operability of the reactor. 2 figs.

  7. 77 FR 16868 - Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-03-22

    ... Fuel Elements for Use in Research and Test Reactors AGENCY: Nuclear Regulatory Commission. ACTION... Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test Reactors.'' This guide describes... plate-type uranium-aluminum fuel elements used in research and test reactors (RTRs). DATES: Submit...

  8. Mechanical and Chemical Properties of Harvested Hypalon Cable Jacket Subjected to Accelerated Thermal Aging

    DOE PAGES

    Duckworth, Robert C.; Kidder, Michelle K.; Aytug, Tolga; ...

    2018-02-27

    We report that for nuclear power plants (NPPs) considering second license renewal for operation beyond 60 years, knowledge of long-term operation, condition monitoring, and viability for the reactor components including reactor pressure vessel, concrete structures, and cable systems is essential. Such knowledge will provide NPP owners/operators with a basis for predicting performance and estimating the costs associated with monitoring or replacement programs for the affected systems. For cable systems that encompass a wide variety of materials, manufacturers, and in-plant locations, accelerated aging of harvested cable jacket and insulation can provide insight into a remaining useful life and methods for monitoring.more » Accelerated thermal aging in air at temperatures between 80°C and 120°C was conducted on a multiconductor control rod drive mechanism cable manufactured by Boston Insulated Wire (BIW). The cable, which had been in service for over 30 years, was jacketed with Hypalon and insulated with ethylene propylene rubber. From elongation at break (EAB) measurements and supporting Arrhenius analysis of the jacket material, an activation energy of 97.84 kJ/mol was estimated, and the time to degradation, as represented by 50% EAB at the expected maximum operating temperature of 45°C, was estimated to be 80 years. These values were slightly below previous measurements on similar BIW Hypalon cable jacket and could be attributed to either in-service degradation or variations in material properties from production variations. Lastly, results from indenter modulus measurements and Fourier transform infrared spectroscopy suggest possible markers that could be beneficial in monitoring cable conditions.« less

  9. Mechanical and Chemical Properties of Harvested Hypalon Cable Jacket Subjected to Accelerated Thermal Aging

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Duckworth, Robert C.; Kidder, Michelle K.; Aytug, Tolga

    We report that for nuclear power plants (NPPs) considering second license renewal for operation beyond 60 years, knowledge of long-term operation, condition monitoring, and viability for the reactor components including reactor pressure vessel, concrete structures, and cable systems is essential. Such knowledge will provide NPP owners/operators with a basis for predicting performance and estimating the costs associated with monitoring or replacement programs for the affected systems. For cable systems that encompass a wide variety of materials, manufacturers, and in-plant locations, accelerated aging of harvested cable jacket and insulation can provide insight into a remaining useful life and methods for monitoring.more » Accelerated thermal aging in air at temperatures between 80°C and 120°C was conducted on a multiconductor control rod drive mechanism cable manufactured by Boston Insulated Wire (BIW). The cable, which had been in service for over 30 years, was jacketed with Hypalon and insulated with ethylene propylene rubber. From elongation at break (EAB) measurements and supporting Arrhenius analysis of the jacket material, an activation energy of 97.84 kJ/mol was estimated, and the time to degradation, as represented by 50% EAB at the expected maximum operating temperature of 45°C, was estimated to be 80 years. These values were slightly below previous measurements on similar BIW Hypalon cable jacket and could be attributed to either in-service degradation or variations in material properties from production variations. Lastly, results from indenter modulus measurements and Fourier transform infrared spectroscopy suggest possible markers that could be beneficial in monitoring cable conditions.« less

  10. METHOD OF MAKING FUEL ELEMENTS

    DOEpatents

    Bean, C.H.; Macherey, R.E.

    1959-12-01

    A method is described for fabricating fuel elements, particularly for enclosing a plate of metal with a second metal by inserting the plate into an aperture of a frame of a second plate, placing a sheet of the second metal on each of opposite faces of the assembled plate and frame, purging with an inert gas the air from the space within the frame and the sheets while sealing the seams between the frame and the sheets, exhausting the space, purging the space with air, re-exhausting the spaces, sealing the second aperture, and applying heat and pressure to bond the sheets, the plate, and the frame to one another.

  11. Nuclear reactor fuel element with vanadium getter on cladding

    DOEpatents

    Johnson, Carl E.; Carroll, Kenneth G.

    1977-01-01

    A nuclear reactor fuel element is described which has an outer cladding, a central core of fissionable or mixed fissionable and fertile fuel material and a layer of vanadium as an oxygen getter on the inner surface of the cladding. The vanadium reacts with oxygen released by the fissionable material during irradiation of the core to prevent the oxygen from reacting with and corroding the cladding. Also described is a method for coating the inner surface of small diameter tubes of cladding with a layer of vanadium.

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

    DOEpatents

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

    2012-06-26

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

  13. A Comparison Study of Offshore Wind Support Structures with Monopiles and Jackets for U.S. Waters: Preprint

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Damiani, Rick; Dykes, Katherine; Scott, George

    2016-08-01

    U.S. experience in offshore wind is limited, and high costs are expected unless innovations are introduced in one or multiple aspects of the project, from the installed technology to the balance of system (BOS). The substructure is the main single component responsible for the BOS capital expenditure (CapEx) and thus one that, if improved, could yield significant levelized cost of energy (LCOE) savings. For projects in U.S. waters, multimember, lattice structures (also known as jackets) can render required stiffness for transitional water depths at potentially lower costs than monopiles (MPs). In this study, we used a systems engineering approach tomore » evaluate the LCOE of prototypical wind power plants at six locations along the eastern seaboard and the Gulf of Mexico for both types of support structures. Using a reference wind turbine and actual metocean conditions for the selected sites, we calculated loads for a parked and an operational situation, and we optimized the MP- and jacket-based support structures to minimize their overall mass. Using a suite of cost models, we then computed their associated LCOE. For all water depths, the MP-based configurations were heavier than their jacket counterparts, but the overall costs for the MPs were less than they were for jackets up to depths of slightly less than 30 m. When the associated manufacturing and installation costs were included, jackets resulted in lower LCOE for depths greater than 40 m. These results can be used by U.S. stakeholders to understand the potential for different technologies at different sites, but the methodology illustrated in this study can be further employed to analyze the effects of innovations and design choices throughout wind power plant systems.« less

  14. Thermionic fuel element for the S-prime reactor

    NASA Astrophysics Data System (ADS)

    Van Hagan, Thomas H.; Drees, Elizabeth A.

    1993-01-01

    Technical aspects of the thermionic fuel element (TFE) design proposed for the S-PRIME space nuclear power system are discussed. Topics covered include the rational for selecting a multicell TFE approach, a technical description of the S-PRIME TFE and its estimated performance, and the technology readiness of the design, which emphasizes techology maturity and low risk.

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

    DOEpatents

    Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pacoima, CA; Benander, Robert E [Pacoima, CA

    2011-03-01

    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.

  16. NUCLEAR REACTOR AND THERMIONIC FUEL ELEMENT THEREFOR

    DOEpatents

    Rasor, N.S.; Hirsch, R.L.

    1963-12-01

    The patent relates to the direct conversion of fission heat to electricity by use of thermionic plasma diodes having fissionable material cathodes, said diodes arranged to form a critical mass in a nuclear reactor. The patent describes a fuel element comprising a plurality of diodes each having a fissionable material cathode, an anode around said cathode, and an ionizable gas therebetween. Provision is made for flowing the gas and current serially through the diodes. (AEC)

  17. Countercurrent flow limited (CCFL) heat flux in the high flux isotope reactor (HFIR) fuel element

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Ruggles, A.E.

    1990-10-12

    The countercurrent flow (CCF) performance in the fuel element region of the HFIR is examined experimentally and theoretically. The fuel element consists of two concentric annuli filled with aluminum clad fuel plates of 1.27 mm thickness separated by 1.27 mm flow channels. The plates are curved as they go radially outward to accomplish constant flow channel width and constant metal-to-coolant ratio. A full-scale HFIR fuel element mock-up is studied in an adiabatic air-water CCF experiment. A review of CCF models for narrow channels is presented along with the treatment of CCFs in system of parallel channels. The experimental results aremore » related to the existing models and a mechanistic model for the annular'' CCF in a narrow channel is developed that captures the data trends well. The results of the experiment are used to calculate the CCFL heat flux of the HFIR fuel assembly. It was determined that the HFIR fuel assembly can reject 0.62 Mw of thermal power in the CCFL situation. 31 refs., 17 figs.« less

  18. Research on the interfacial behaviors of plate-type dispersion nuclear fuel elements

    NASA Astrophysics Data System (ADS)

    Wang, Qiming; Yan, Xiaoqing; Ding, Shurong; Huo, Yongzhong

    2010-04-01

    The three-dimensional constitutive relations are constructed, respectively, for the fuel particles, the metal matrix and the cladding of dispersion nuclear fuel elements, allowing for the effects of large deformation and thermal-elastoplasticity. According to the constitutive relations, the method of modeling their irradiation behaviors in ABAQUS is developed and validated. Numerical simulations of the interfacial performances between the fuel meat and the cladding are implemented with the developed finite element models for different micro-structures of the fuel meat. The research results indicate that: (1) the interfacial tensile stresses and shear stresses for some cases will increase with burnup, but the relative stresses will decrease with burnup for some micro-structures; (2) at the lower burnups, the interfacial stresses increase with the particle sizes and the particle volume fractions; however, it is not the case at the higher burnups; (3) the particle distribution characteristics distinctly affect the interfacial stresses, and the face-centered cubic case has the best interfacial performance of the three considered cases.

  19. Advanced Ceramics for Use as Fuel Element Materials in Nuclear Thermal Propulsion Systems

    NASA Technical Reports Server (NTRS)

    Valentine, Peter G.; Allen, Lee R.; Shapiro, Alan P.

    2012-01-01

    With the recent start (October 2011) of the joint National Aeronautics and Space Administration (NASA) and Department of Energy (DOE) Advanced Exploration Systems (AES) Nuclear Cryogenic Propulsion Stage (NCPS) Program, there is renewed interest in developing advanced ceramics for use as fuel element materials in nuclear thermal propulsion (NTP) systems. Three classes of fuel element materials are being considered under the NCPS Program: (a) graphite composites - consisting of coated graphite elements containing uranium carbide (or mixed carbide), (b) cermets (ceramic/metallic composites) - consisting of refractory metal elements containing uranium oxide, and (c) advanced carbides consisting of ceramic elements fabricated from uranium carbide and one or more refractory metal carbides [1]. The current development effort aims to advance the technology originally developed and demonstrated under Project Rover (1955-1973) for the NERVA (Nuclear Engine for Rocket Vehicle Application) [2].

  20. Method for measuring recovery of catalytic elements from fuel cells

    DOEpatents

    Shore, Lawrence [Edison, NJ; Matlin, Ramail [Berkeley, NJ

    2011-03-08

    A method is provided for measuring the concentration of a catalytic clement in a fuel cell powder. The method includes depositing on a porous substrate at least one layer of a powder mixture comprising the fuel cell powder and an internal standard material, ablating a sample of the powder mixture using a laser, and vaporizing the sample using an inductively coupled plasma. A normalized concentration of catalytic element in the sample is determined by quantifying the intensity of a first signal correlated to the amount of catalytic element in the sample, quantifying the intensity of a second signal correlated to the amount of internal standard material in the sample, and using a ratio of the first signal intensity to the second signal intensity to cancel out the effects of sample size.

  1. PROCESS FOR JACKETING A CORE

    DOEpatents

    Last, G.A.

    1960-07-19

    A process is given for enclosing the uranium core of a nuclear fuel element by placing the core in an aluminum cup and closing the open end of the cup over the core. As the metal of the cup is brought together in a weld over the center of the end of the core, it is extruded inwardly as internal projection into a central recess in the core and outwardly as an external projection. Thus oxide inclusions in the weld of the cup are spread out into the internal and external projections and do not interfere with the integrity of the weld.

  2. Control of heat transfer in continuous-feeding Czochralski-silicon crystal growth with a water-cooled jacket

    NASA Astrophysics Data System (ADS)

    Zhao, Wenhan; Liu, Lijun

    2017-01-01

    The continuous-feeding Czochralski method is an effective method to reduce the cost of single crystal silicon. By promoting the crystal growth rate, the cost can be reduced further. However, more latent heat will be released at the melt-crystal interface under a high crystal growth rate. In this study, a water-cooled jacket was applied to enhance the heat transfer at the melt-crystal interface. Quasi-steady-state numerical calculation was employed to investigate the impact of the water-cooled jacket on the heat transfer at the melt-crystal interface. Latent heat released during the crystal growth process at the melt-crystal interface and absorbed during feedstock melting at the feeding zone was modeled in the simulations. The results show that, by using the water-cooled jacket, heat transfer in the growing crystal is enhanced significantly. Melt-crystal interface deflection and thermal stress increase simultaneously due to the increase of radial temperature at the melt-crystal interface. With a modified heat shield design, heat transfer at the melt-crystal interface is well controlled. The crystal growth rate can be increased by 20%.

  3. DART model for irradiation-induced swelling of uranium silicide dispersion fuel elements

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Rest, J.; Hofman, G.L.

    1999-04-01

    Models for the interaction of uranium silicide dispersion fuels with an aluminum matrix, for the resultant reaction product swelling, and for the calculation of the stress gradient within the fuel particles are described within the context of DART fission-gas-induced swelling models. The effects of an aluminide shell on fuel particle swelling are evaluated. Validation of the model is demonstrated by comparing DART calculations with irradiation data for the swelling of U{sub 3}SiAl-Al and U{sub 3}Si{sub 2}-Al in variously designed dispersion fuel elements.

  4. Current status of U{sub 3}Si{sub 2} fuel element fabrication in Brazil

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Durazzo, M.; Carvalho, E.F. Urano de; Saliba-Silva, A.M.

    2008-07-15

    IPEN has been working for increasing radioisotope production in order to supply the expanding demand for radiopharmaceutical medicines requested by the Brazilian welfare. To reach this objective, the IEA-R1 research reactor power capacity was recently increased from 2 MW to 4 MW. Since 1988 IPEN has been manufacturing its own fuel element, initially based on U{sub 3}O{sub 8}-Al dispersion fuel plates with 2.3 gU/cm{sup 3}. To support the reactor power increase, higher uranium density in the fuel plate meat had to be achieved for better irradiation flux and also to minimize the irradiated fuel elements to be stored. Uranium silicidemore » was the chosen option and the fuel fabrication development started with the support of the IAEA BRA/4/047 Technical Cooperation Project. This paper describes the results of this program and the current status of silicide fuel fabrication and its qualification. (author)« less

  5. Electrochemical fuel cell generator having an internal and leak tight hydrocarbon fuel reformer

    DOEpatents

    Dederer, J.T.; Hager, C.A.

    1998-03-31

    An electrochemical fuel cell generator configuration is made having a generator section which contains a plurality of axially elongated fuel cells, each cell containing a fuel electrode, air electrode, and solid oxide electrolyte between the electrodes, in which axially elongated dividers separate portions of the fuel cells from each other, and where at least one divider also reforms a reformable fuel gas mixture prior to electricity generation reactions, the at least one reformer-divider is hollow having a closed end and an open end entrance for a reformable fuel mixture to pass to the closed end of the divider and then reverse flow and pass back along the hollowed walls to be reformed, and then finally to pass as reformed fuel out of the open end of the divider to contact the fuel cells, and further where the reformer-divider is a composite structure having a gas diffusion barrier of metallic foil surrounding the external walls of the reformer-divider except at the entrance to prevent diffusion of the reformable gas mixture through the divider, and further housed in an outer insulating jacket except at the entrance to prevent short-circuiting of the fuel cells by the gas diffusion barrier. 10 figs.

  6. Electrochemical fuel cell generator having an internal and leak tight hydrocarbon fuel reformer

    DOEpatents

    Dederer, Jeffrey T.; Hager, Charles A.

    1998-01-01

    An electrochemical fuel cell generator configuration is made having a generator section which contains a plurality of axially elongated fuel cells, each cell containing a fuel electrode, air electrode, and solid oxide electrolyte between the electrodes, in which axially elongated dividers separate portions of the fuel cells from each other, and where at least one divider also reforms a reformable fuel gas mixture prior to electricity generation reactions, the at least one reformer-divider is hollow having a closed end and an open end entrance for a reformable fuel mixture to pass to the closed end of the divider and then reverse flow and pass back along the hollowed walls to be reformed, and then finally to pass as reformed fuel out of the open end of the divider to contact the fuel cells, and further where the reformer-divider is a composite structure having a gas diffusion barrier of metallic foil surrounding the external walls of the reformer-divider except at the entrance to prevent diffusion of the reformable gas mixture through the divider, and further housed in an outer insulating jacket except at the entrance to prevent short-circuiting of the fuel cells by the gas diffusion barrier.

  7. Triaxial Swirl Injector Element for Liquid-Fueled Engines

    NASA Technical Reports Server (NTRS)

    Muss, Jeff

    2010-01-01

    A triaxial injector is a single bi-propellant injection element located at the center of the injector body. The injector element consists of three nested, hydraulic swirl injectors. A small portion of the total fuel is injected through the central hydraulic injector, all of the oxidizer is injected through the middle concentric hydraulic swirl injector, and the balance of the fuel is injected through an outer concentric injection system. The configuration has been shown to provide good flame stabilization and the desired fuel-rich wall boundary condition. The injector design is well suited for preburner applications. Preburner injectors operate at extreme oxygen-to-fuel mass ratios, either very rich or very lean. The goal of a preburner is to create a uniform drive gas for the turbomachinery, while carefully controlling the temperature so as not to stress or damage turbine blades. The triaxial injector concept permits the lean propellant to be sandwiched between two layers of the rich propellant, while the hydraulic atomization characteristics of the swirl injectors promote interpropellant mixing and, ultimately, good combustion efficiency. This innovation is suited to a wide range of liquid oxidizer and liquid fuels, including hydrogen, methane, and kerosene. Prototype testing with the triaxial swirl injector demonstrated excellent injector and combustion chamber thermal compatibility and good combustion performance, both at levels far superior to a pintle injector. Initial testing with the prototype injector demonstrated over 96-percent combustion efficiency. The design showed excellent high -frequency combustion stability characteristics with oxygen and kerosene propellants. Unlike the more conventional pintle injector, there is not a large bluff body that must be cooled. The absence of a protruding center body enhances the thermal durability of the triaxial swirl injector. The hydraulic atomization characteristics of the innovation allow the design to be

  8. 34. DETAILS AND SECTIONS OF SHIELDING TANK FUEL ELEMENT SUPPORT ...

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

    34. DETAILS AND SECTIONS OF SHIELDING TANK FUEL ELEMENT SUPPORT FRAME. F.C. TORKELSON DRAWING NUMBER 842-ARVFS-701-S-4. INEL INDEX CODE NUMBER: 075 0701 60 851 151978. - Idaho National Engineering Laboratory, Advanced Reentry Vehicle Fusing System, Scoville, Butte County, ID

  9. Calculation of Distribution Dynamics of Inhomogeneous Temperature Field in Range of Fuel Elements by Using FreeFem++

    NASA Astrophysics Data System (ADS)

    Amosova, E. V.; Shishkin, A. V.

    2017-11-01

    This article introduces the result of studying the heat exchange in the fuel element of the nuclear reactor fuel magazine. Fuel assemblies are completed as a bundle of cylindrical fuel elements located at the tops of a regular triangle. Uneven distribution of fuel rods in a nuclear reactor’s core forms the inhomogeneity of temperature fields. This article describes the developed method for heat exchange calculation with the account for impact of an inhomogeneous temperature field on the thermal-physical properties of materials and unsteady effects. The acquired calculation results are used for evaluating the tolerable temperature levels in protective case materials.

  10. METHOD OF PREPARING A FUEL ELEMENT FOR A NUCLEAR REACTOR

    DOEpatents

    Hauth, J.J.; Anicetti, R.J.

    1962-12-01

    A method is described for preparing a fuel element for a nuclear reactor. According to the patent uranium dioxide is compacted in a metal tabe by directlng intense sound waves at the tabe prior to tamp packing or vibration compaction of the powder. (AEC)

  11. AN EVALUATION OF POTENTIAL LINER MATERIALS FOR ELIMINATING FCCI IN IRRADIATED METALLIC NUCLEAR FUEL ELEMENTS

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    D. D. Keiser; J. I. Cole

    2007-09-01

    Metallic nuclear fuels are being looked at as part of the Global Nuclear Energy Program for transmuting longlive transuranic actinide isotopes contained in spent nuclear fuel into shorter-lived fission products. In order to optimize the performance of these fuels, the concept of using liners to eliminate the fuel/cladding chemical interactions that can occur during irradiation of a fuel element has been investigated. The potential liner materials Zr and V have been tested using solid-solid diffusion couples, consisting of liner materials butted against fuel alloys and against cladding materials. The couples were annealed at the relatively high temperature of 700°C. Thismore » temperature would be the absolute maximum temperature present at the fuel/cladding interface for a fuel element in-reactor. Analysis was performed using a scanning electron microscope equipped with energy-dispersive and wavelengthdispersive spectrometers (SEM/EDS/WDS) to evaluate any developed diffusion structures. At 700°C, minimal interaction was observed between the metallic fuels and either Zr or V. Similarly, limited interaction was observed between the Zr and V and the cladding materials. The best performing liner material appeared to be the V, based on amounts of interaction.« less

  12. METHOD OF PREPARING A CERAMIC FUEL ELEMENT

    DOEpatents

    Ross, W.T.; Bloomster, C.H.; Bardsley, R.E.

    1963-09-01

    A method is described for preparing a fuel element from -325 mesh PuO/ sub 2/ and -20 mesh UO/sub 2/, and the steps of screening --325 mesh UO/sub 2/ from the -20 mesh UO/sub 2/, mixing PuO/sub 2/ with the --325 mesh UO/sub 2/, blending this mixture with sufficient --20 mesh UO/sub 2/ to obtain the desired composition, introducing the blend into a metal tube, repeating the procedure until the tube is full, and vibrating the tube to compact the powder are included. (AEC)

  13. STUDIES OF FAST REACTOR FUEL ELEMENT BEHAVIOR UNDER TRANSIENT HEATING TO FAILURE. I. INITIAL EXPERIMENTS ON METALLIC SAMPLES IN THE ABSENCE OF COOLANT

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Dickerman, C. E.; Sowa, E. S.; Okrent, D.

    1961-08-01

    Meltdown tests on single metallic unirradiated fuel elements in TREAT are described. The fuel elements (EBRII Mark I fuel pins, EBR-II fuel pins with retractory Nb or Ta cladding, and Fermi-I fuel pins) are tested in an inert atmosphere, with no coolant. The fuel elements are exposed to reactor power bursts of 200 msec to 25 sec duration, under conditions simulating fast reactor operations. For these tests, the type of power burst, the integrated power, the fuel enrichment, the maximum cladding temperature, and the effects of the test on the fuel element are recorded. ( T.F.H.)

  14. Comparison of Gunshot Entrance Morphologies Caused by .40-Caliber Smith & Wesson, .380-Caliber, and 9-mm Luger Bullets: A Finite Element Analysis Study

    PubMed Central

    Matoso, Rodrigo Ivo; Freire, Alexandre Rodrigues; Santos, Leonardo Soriano de Mello; Daruge Junior, Eduardo; Rossi, Ana Claudia; Prado, Felippe Bevilacqua

    2014-01-01

    Firearms can cause fatal wounds, which can be identified by traces on or around the body. However, there are cases where neither the bullet nor gun is found at the crime scene. Ballistic research involving finite element models can reproduce computational biomechanical conditions, without compromising bioethics, as they involve no direct tests on animals or humans. This study aims to compare the morphologies of gunshot entrance holes caused by.40-caliber Smith & Wesson (S&W), .380-caliber, and 9×19-mm Luger bullets. A fully metal-jacketed.40 S&W projectile, a fully metal-jacketed.380 projectile, and a fully metal-jacketed 9×19-mm Luger projectile were computationally fired at the glabellar region of the finite element model from a distance of 10 cm, at perpendicular incidence. The results show different morphologies in the entrance holes produced by the three bullets, using the same skull at the same shot distance. The results and traits of the entrance holes are discussed. Finite element models allow feasible computational ballistic research, which may be useful to forensic experts when comparing and analyzing data related to gunshot wounds in the forehead. PMID:25343337

  15. METHOD AND MEANS FOR CLOSING TUBES BY SPINNING

    DOEpatents

    Graves, E.E.; Coonfare, R.H.

    1958-08-26

    An improved spinning tool is described for producing a fold-free closed end on an aluminum jacketing tube such as is commonly used to protect a uranium fuel element. The tool will fit the toolholder of a lathe in which the jacket is rotated. The tool has a number of working faces so that the hemispherical end- closure is formed, the folds and wrinkles are smcothed out, and the excess metal is trimmed off in one transverse cutting operation. This tool considerably speeds up the closure process, and eliminates the need for a weld seal.

  16. Space shuttle orbit maneuvering engine, reusable thrust chamber program. Task 6: Data dump hot fuel element investigation

    NASA Technical Reports Server (NTRS)

    Nurick, W. H.

    1974-01-01

    An evaluation of reusable thrust chambers for the space shuttle orbit maneuvering engine was conducted. Tests were conducted using subscale injector hot-fire procedures for the injector configurations designed for a regenerative cooled engine. The effect of operating conditions and fuel temperature on combustion chamber performance was determined. Specific objectives of the evaluation were to examine the optimum like-doublet element geometry for operation at conditions consistent with a fuel regeneratively cooled engine (hot fuel, 200 to 250 F) and the sensitivity of the triplet injector element to hot fuels.

  17. The Finite Element Modelling and Dynamic Characteristics Analysis about One Kind of Armoured Vehicles’ Fuel Tanks

    NASA Astrophysics Data System (ADS)

    Gao, Yang; Ge, Zhishang; Zhai, Weihao; Tan, Shiwang; Zhang, Feng

    2018-01-01

    The static and dynamic characteristics of fuel tank are studied for the armoured vehicle in this paper. The CATIA software is applied to build the CAD model of the armoured vehicles’ fuel tank, and the finite element model is established in ANSYS Workbench. The finite element method is carried out to analyze the static and dynamic mechanical properties of the fuel tank, and the first six orders of mode shapes and their frequencies are also computed and given in the paper, then the stress distribution diagram and the high stress areas are obtained. The results of the research provide some references to the fuel tanks’ design improvement, and give some guidance for the installation of the fuel tanks on armoured vehicles, and help to improve the properties and the service life of this kind of armoured vehicles’ fuel tanks.

  18. Phase characteristics of rare earth elements in metallic fuel for a sodium-cooled fast reactor by injection casting

    NASA Astrophysics Data System (ADS)

    Kuk, Seoung Woo; Kim, Ki Hwan; Kim, Jong Hwan; Song, Hoon; Oh, Seok Jin; Park, Jeong-Yong; Lee, Chan Bock; Youn, Young-Sang; Kim, Jong-Yun

    2017-04-01

    Uranium-zirconium-rare earth (U-Zr-RE) fuel slugs for a sodium-cooled fast reactor were manufactured using a modified injection casting method, and investigated with respect to their uniformity, distribution, composition, and phase behavior according to RE content. Nd, Ce, Pr, and La were chosen as four representative lanthanide elements because they are considered to be major RE components of fuel ingots after pyroprocessing. Immiscible layers were found on the top layers of the melt-residue commensurate with higher fuel slug RE content. Scanning electron microscopy-energy-dispersive X-ray spectroscopy (SEM-EDS) data showed that RE elements in the melt-residue were distributed uniformly throughout the fuel slugs. RE element agglomeration did not contaminate the fuel slugs but strongly affected the RE content of the slugs.

  19. Nuclear Cryogenic Propulsion Stage (NCPS) Fuel Element Testing in the Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

    NASA Technical Reports Server (NTRS)

    Emrich, William J., Jr.

    2017-01-01

    To support the on-going nuclear thermal propulsion effort, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The facility to perform this testing is referred to as the Nuclear Thermal Rocket Element Environment Simulator (NTREES). Last year NTREES was successfully used to satisfy a testing milestone for the Nuclear Cryogenic Propulsion Stage (NCPS) project and met or exceeded all required objectives.

  20. Covalent immobilization of metal organic frameworks onto chemical resistant poly(ether ether ketone) jacket for stir bar extraction.

    PubMed

    Wang, Chenlu; Zhou, Wei; Liao, Xiaoyan; Wang, Xuemei; Chen, Zilin

    2018-09-26

    Preparation of stir bar extraction (SBSE) device with high physical and chemical stability is important and challenging by date. A novel poly (ether ether ketone) (PEEK) tube with excellent mechanical property and chemical stability was firstly used as jacket of metal bar for preparation of stir bar. By employing covalent modification method, the inherent chemical resistant problem of PEEK which restricts the modification of sorbents was well solved. After functionalization, plenty of benzoic acid groups were formed onto the PEEK jacket. Metal organic frameworks of aluminium-based Materials of Institute Lavoisier-68 (MIL-68) was in situ immobilized onto the PEEK surface (MIL-68@PEEK) by the bonding with benzoic acid groups. Afterwards, a facile dumbbell-shaped structure was designed for reducing the friction between sorbents and bottom of container. Due to superior property of the PEEK jacket and the covalent modification method, the MIL-68 modified PEEK jacket SBSE device showed good robustness. After coupling with HPLC-MS/MS, the MIL-68@PEEK-based SBSE device was used to analyse of three parabens including methyl paraben, ethyl paraben and propyl paraben. The method had low limit detection up to 1 pg mL -1 with good linearity (R 2  ≥ 0.9978) and good reproducibility (relative standard deviation ≤ 9.74%). The method has been applied to the detection of parabens in cosmetics and rabbit plasma after painted with cosmetics with recoveries between 73.25% and 104.23%. Copyright © 2018 Elsevier B.V. All rights reserved.

  1. 36. DETAILS AND SECTIONS OF SHIELDING TANK, FUEL ELEMENT SUPPORT ...

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

    36. DETAILS AND SECTIONS OF SHIELDING TANK, FUEL ELEMENT SUPPORT FRAME AND SUPPORT PLATFORM, AND SAFETY MECHANISM ASSEMBLY (SPRING-LOADED HINGE). F.C. TORKELSON DRAWING NUMBER 842-ARVFS-701-S-1. INEL INDEX CODE NUMBER: 075 0701 60 851 151975. - Idaho National Engineering Laboratory, Advanced Reentry Vehicle Fusing System, Scoville, Butte County, ID

  2. High performance fuel element with end seal

    DOEpatents

    Lee, Gary E.; Zogg, Gordon J.

    1987-01-01

    A nuclear fuel element comprising an elongate block of refractory material having a generally regular polygonal cross section. The block includes parallel, spaced, first and second end surfaces. The first end surface has a peripheral sealing flange formed thereon while the second end surface has a peripheral sealing recess sized to receive the flange. A plurality of longitudinal first coolant passages are positioned inwardly of the flange and recess. Elongate fuel holes are separate from the coolant passages and disposed inwardly of the flange and the recess. The block is further provided with a plurality of peripheral second coolant passages in general alignment with the flange and the recess for flowing coolant. The block also includes two bypasses for each second passage. One bypass intersects the second passage adjacent to but spaced from the first end surface and intersects a first passage, while the other bypass intersects the second passage adjacent to but spaced from the second end surface and intersects a first passage so that coolant flowing through the second passages enters and exits the block through the associated first passages.

  3. 46 CFR 180.72 - Personal flotation devices carried in addition to life jackets.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Personal flotation devices carried in addition to life jackets. 180.72 Section 180.72 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) SMALL... emergencies. (b) Wearable marine buoyant devices that include “ski vests,” “boating vests,” and “fishing vests...

  4. 46 CFR 180.72 - Personal flotation devices carried in addition to life jackets.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 7 2011-10-01 2011-10-01 false Personal flotation devices carried in addition to life jackets. 180.72 Section 180.72 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) SMALL... emergencies. (b) Wearable marine buoyant devices that include “ski vests,” “boating vests,” and “fishing vests...

  5. METHOD OF FORMING A FUEL ELEMENT FOR A NUCLEAR REACTOR

    DOEpatents

    Layer, E.H. Jr.; Peet, C.S.

    1962-01-23

    A method is given for preparing a fuel element for a nuclear reactor. The method includes the steps of sandblasting a body of uranium dioxide to roughen the surface thereof, depositing a thin layer of carbon thereon by thermal decomposition of methane, and cladding the uranium dioxide body with zirconium by gas pressure bonding. (AEC)

  6. REMOVAL OF ALUMINUM COATINGS

    DOEpatents

    Peterson, J.H.

    1959-08-25

    A process is presented for dissolving aluminum jackets from uranium fuel elements without attack of the uranium in a boiling nitric acid-mercuric nitrate solution containing up to 50% by weight of nitrtc acid and mercuric nitrate in a concentration of between 0.05 and 1% by weight.

  7. 78 FR 33132 - Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-06-03

    ... Fuel Elements for Use in Research and Test Reactors AGENCY: Nuclear Regulatory Commission. ACTION... Research and Test Reactors.'' This guide describes a method that the staff of the NRC considers acceptable... assurance program for verifying the quality of plate-type uranium-aluminum fuel elements used in research...

  8. Optimisation and evaluation of pre-design models for offshore wind turbines with jacket support structures and their influence on integrated load simulations

    NASA Astrophysics Data System (ADS)

    Schafhirt, S.; Kaufer, D.; Cheng, P. W.

    2014-12-01

    In recent years many advanced load simulation tools, allowing an aero-servo-hydroelastic analyses of an entire offshore wind turbine, have been developed and verified. Nowadays, even an offshore wind turbine with a complex support structure such as a jacket can be analysed. However, the computational effort rises significantly with an increasing level of details. This counts especially for offshore wind turbines with lattice support structures, since those models do naturally have a higher number of nodes and elements than simpler monopile structures. During the design process multiple load simulations are demanded to obtain an optimal solution. In the view of pre-design tasks it is crucial to apply load simulations which keep the simulation quality and the computational effort in balance. The paper will introduce a reference wind turbine model consisting of the REpower5M wind turbine and a jacket support structure with a high level of detail. In total twelve variations of this reference model are derived and presented. Main focus is to simplify the models of the support structure and the foundation. The reference model and the simplified models are simulated with the coupled simulation tool Flex5-Poseidon and analysed regarding frequencies, fatigue loads, and ultimate loads. A model has been found which reaches an adequate increase of simulation speed while holding the results in an acceptable range compared to the reference results.

  9. Vented nuclear fuel element

    DOEpatents

    Grossman, Leonard N.; Kaznoff, Alexis I.

    1979-01-01

    A nuclear fuel cell for use in a thermionic nuclear reactor in which a small conduit extends from the outside surface of the emitter to the center of the fuel mass of the emitter body to permit escape of volatile and gaseous fission products collected in the center thereof by virtue of molecular migration of the gases to the hotter region of the fuel.

  10. Aluminum hydroxide coating thickness measurements and brushing tests on K West Basin fuel elements

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Pitner, A.L.

    1998-09-11

    Aluminum hydroxide coating thicknesses were measured on fuel elements stored in aluminum canisters in K West Basin using specially developed eddy current probes . The results were used to estimate coating inventories for MCO fuel,loading. Brushing tests successfully demonstrated the ability to remove the coating if deemed necessary prior to MCO loading.

  11. METHOD AND APPARATUS FOR EXAMINING FUEL ELEMENTS FOR LEAKAGE

    DOEpatents

    Smith, R.R.; Echo, M.W.; Doe, C.B.

    1963-12-31

    A process and a device for the continuous monitoring of fuel elements while in use in a liquid-metal-cooled, argonblanketed nuclear reactor are presented. A fraction of the argon gas is withdrawn, contacted with a negative electrical charge for attraction of any alkali metal formed from argon by neutron reaction, and recycled into the reactor. The electrical charge is introduced into water, and the water is examined for radioactive alkali metals. (AEC)

  12. FUEL ELEMENTS FOR NUCLEAR REACTORS AND PROCESS OF MAKING

    DOEpatents

    Roake, W.E.

    1958-08-19

    A process is described for producing uranium metal granules for use in reactor fuel elements. The granules are made by suspending powdered uramiunn metal or uranium hydride in a viscous, non-reactive liquid, such as paraffin oil, aad pouring the resulting suspension in droplet, on to a bed of powdered absorbent. In this manner the liquid vehicle is taken up by the sorbent and spherical pellets of uranium metal are obtained. The

  13. Discharge lamp with reflective jacket

    DOEpatents

    MacLennan, Donald A.; Turner, Brian P.; Kipling, Kent

    2001-01-01

    A discharge lamp includes an envelope, a fill which emits light when excited disposed in the envelope, a source of excitation power coupled to the fill to excite the fill and cause the fill to emit light, and a reflector disposed around the envelope and defining an opening, the reflector being configured to reflect some of the light emitted by the fill back into the fill while allowing some light to exit through the opening. The reflector may be made from a material having a similar thermal index of expansion as compared to the envelope and which is closely spaced to the envelope. The envelope material may be quartz and the reflector material may be either silica or alumina. The reflector may be formed as a jacket having a rigid structure which does not adhere to the envelope. The lamp may further include an optical clement spaced from the envelope and configured to reflect an unwanted component of light which exited the envelope back into the envelope through the opening in the reflector. Light which can be beneficially recaptured includes selected wavelength regions, a selected polarization, and selected angular components.

  14. Vacuum plasma spray applications on liquid fuel rocket engines

    NASA Technical Reports Server (NTRS)

    Mckechnie, T. N.; Zimmerman, F. R.; Bryant, M. A.

    1992-01-01

    The vacuum plasma spray process (VPS) has been developed by NASA and Rocketdyne for a variety of applications on liquid fuel rocket engines, including the Space Shuttle Main Engine. These applications encompass thermal barrier coatings which are thermal shock resistant for turbopump blades and nozzles; bond coatings for cryogenic titanium components; wear resistant coatings and materials; high conductivity copper, NaRloy-Z, combustion chamber liners, and structural nickel base material, Inconel 718, for nozzle and combustion chamber support jackets.

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

    DOEpatents

    Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pocoima, CA; Benander, Robert E [Pacoima, CA

    2010-02-23

    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.

  16. Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors

    DOEpatents

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2013-09-03

    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.

  17. 46 CFR 185.730 - Servicing of inflatable liferafts, inflatable buoyant apparatus, inflatable life jackets, and...

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... inflatable liferaft or inflatable buoyant apparatus must be serviced at a facility specifically approved by... apparatus, inflatable life jackets, and inflated rescue boats. 185.730 Section 185.730 Shipping COAST GUARD... Operational Readiness, Maintenance, and Inspection of Lifesaving Equipment § 185.730 Servicing of inflatable...

  18. 46 CFR 185.730 - Servicing of inflatable liferafts, inflatable buoyant apparatus, inflatable life jackets, and...

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... inflatable liferaft or inflatable buoyant apparatus must be serviced at a facility specifically approved by... apparatus, inflatable life jackets, and inflated rescue boats. 185.730 Section 185.730 Shipping COAST GUARD... Operational Readiness, Maintenance, and Inspection of Lifesaving Equipment § 185.730 Servicing of inflatable...

  19. Vacuum jacketed composite propulsion feedlines for cryogenic launch and space vehicles, volume 1. [development of glass fiber composite for strength and protection from handling damage

    NASA Technical Reports Server (NTRS)

    Spond, D. E.; Laintz, D. J.; Hall, C. A.; Dulaigh, D. E.

    1974-01-01

    Thin metallic liners that provide leak-free service in cryogenic propulsion systems are overwrapped with a glass-fiber composite that provides strength and protection from handling damage. The resultant tube is lightweight, strong, and has a low thermal flux. The inside commodity flow line and the outside vacuum jacket were fabricated using this method. Several types of vacuum jackets were fabricated and tested at operating temperatures from 294 to 21 K (+70 to minus 423 F) and operating pressure up to 69 N/cm2 (100 psi). The primary objective of the program was to develop vacuum jacket concepts, using previously developed concepts for the inner line. All major program objectives were met resulting in a design concept that is adaptable to a wide range of aerospace vehicle requirements. Major items of development included convolution of thin metallic sections up to 46 cm (18 in.) in diameter, design and fabrication of an extremely lightweight tension membrane concept for the vacuum jacket, and analytical tools that predict the failure mode and levels.

  20. PROCESS FOR DISSOLVING BINARY URANIUM-ZIRCONIUM OR ZIRCONIUM-BASE ALLOYS

    DOEpatents

    Jonke, A.A.; Barghusen, J.J.; Levitz, N.M.

    1962-08-14

    A process of dissolving uranium-- zirconium and zircaloy alloys, e.g. jackets of fuel elements, with an anhydrous hydrogen fluoride containing from 10 to 32% by weight of hydrogen chloride at between 400 and 450 deg C., preferably while in contact with a fluidized inert powder, such as calcium fluoride is described. (AEC)

  1. METHOD OF PREPARING A FUEL ELEMENT FOR A NUCLEAR REACTOR

    DOEpatents

    Handwerk, J.H.; BAch, R.A.

    1959-08-18

    A method is described for preparing a reactor fuel element by forming a mixture of thorium dioxide and an oxide of uranium, the uranium being present. In an oxidation state at least as high as it is in U/sub 3/O/sub 8/, into a desired shape and firing in air at a temperature siifficiently high to reduce the higher uranium oxide to uranium dioxide.

  2. NEUTRONIC REACTOR FUEL ELEMENT

    DOEpatents

    Horning, W.A.; Lanning, D.D.; Donahue, D.J.

    1959-10-01

    A fuel slug for a reactor which acts as a safety device is described. The fuel slug is an aluminum tube with a foil lining the inside surface of the tube, the foil being fabricated of uranium in a lead matrix.

  3. 46 CFR 122.730 - Servicing of inflatable liferafts, inflatable buoyant apparatus, inflatable life jackets, and...

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... Inspection of Lifesaving Equipment § 122.730 Servicing of inflatable liferafts, inflatable buoyant apparatus... apparatus must be serviced at a facility specifically approved by the Commandant for the particular brand... apparatus, inflatable life jackets, and inflated rescue boats. 122.730 Section 122.730 Shipping COAST GUARD...

  4. 46 CFR 122.730 - Servicing of inflatable liferafts, inflatable buoyant apparatus, inflatable life jackets, and...

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... Inspection of Lifesaving Equipment § 122.730 Servicing of inflatable liferafts, inflatable buoyant apparatus... apparatus must be serviced at a facility specifically approved by the Commandant for the particular brand... apparatus, inflatable life jackets, and inflated rescue boats. 122.730 Section 122.730 Shipping COAST GUARD...

  5. A novel microbial fuel cell sensor with biocathode sensing element.

    PubMed

    Jiang, Yong; Liang, Peng; Liu, Panpan; Wang, Donglin; Miao, Bo; Huang, Xia

    2017-08-15

    The traditional microbial fuel cell (MFC) sensor with bioanode as sensing element delivers limited sensitivity to toxicity monitoring, restricted application to only anaerobic and organic rich water body, and increased potential fault warning to the combined shock of organic matter/toxicity. In this study, the biocathode for oxygen reduction reaction was employed for the first time as the sensing element in MFC sensor for toxicity monitoring. The results shown that the sensitivity of MFC sensor with biocathode sensing element (7.4±2.0 to 67.5±4.0mA% -1 cm -2 ) was much greater than that showed by bioanode sensing element (3.4±1.5 to 5.5±0.7mA% -1 cm -2 ). The biocathode sensing element achieved the lowest detection limit reported to date using MFC sensor for formaldehyde detection (0.0005%), while the bioanode was more applicable for higher concentration (>0.0025%). There was a quicker response of biocathode sensing element with the increase of conductivity and dissolved oxygen (DO). The biocathode sensing element made the MFC sensor directly applied to clean water body monitoring, e.g., drinking water and reclaimed water, without the amending of background organic matter, and it also decreased the warning failure when challenged by a combined shock of organic matter/toxicity. Copyright © 2017 Elsevier B.V. All rights reserved.

  6. Simulation of irradiation hardening of Zircaloy within plate-type dispersion nuclear fuel elements

    NASA Astrophysics Data System (ADS)

    Jiang, Yijie; Wang, Qiming; Cui, Yi; Huo, Yongzhong; Ding, Shurong

    2011-06-01

    Within plate-type dispersion nuclear fuel elements, the metal matrix and cladding attacked continuously by fast neutrons undergo irradiation hardening, which might have remarkable effects upon the mechanical behaviors within fuel elements. In this paper, with the irradiation hardening effect of metal materials mainly considered together with irradiation growth effect of the cladding, the three-dimensional large-deformation constitutive relations for the metal matrix and cladding are developed. The method of virtual temperature increase in the previous studies is further developed to model the irradiation swelling of fuel particles; the method of anisotropic thermal expansion is introduced to model irradiation growth of the cladding; and a method of multi-step-temperature loading is proposed to simulate the coupling features of irradiation-induced swelling of the fuel particles together with irradiation growth of the cladding. Above all, based on the developed relationship between irradiation growth at certain burnup and the loaded virtual temperatures, with considering that certain burnup corresponds to certain fast neutron fluence, the time-dependent constitutive relation due to irradiation hardening effect is replaced by the virtual-temperature-dependent one which is introduced into the commercial software to simulate the irradiation hardening effects of the matrix and cladding. Numerical simulations of the irradiation-induced mechanical behaviors are implemented with the finite element method in consideration of the micro-structure of the fuel meat. The obtained results indicate that when the irradiation hardening effects are introduced into the constitutive relations of the metal matrix and cladding: (1) higher maximum Mises stresses for certain burnup at the matrix exist with the equivalent plastic strains remaining almost the same at lower burnups; (2) the maximum Mises stresses for certain burnup at the cladding are enhanced while the maximum equivalent

  7. Heat Exchanger With Internal Pin Elements

    DOEpatents

    Gerstmann, Joseph; Hannon, Charles L.

    2004-01-13

    A heat exchanger/heater comprising a tubular member having a fluid inlet end, a fluid outlet end and plurality of pins secured to the interior wall of the tube. Various embodiments additionally comprise a blocking member disposed concentrically inside the pins, such as a core plug or a baffle array. Also disclosed is a vapor generator employing an internally pinned tube, and a fluid-heater/heat-exchanger utilizing an outer jacket tube and fluid-side baffle elements, as well as methods for heating a fluid using an internally pinned tube.

  8. Pu-Zr alloy for high-temperature foil-type fuel

    DOEpatents

    McCuaig, Franklin D.

    1977-01-01

    A nuclear reactor fuel alloy consists essentially of from slightly greater than 7 to about 4 w/o zirconium, balance plutonium, and is characterized in that the alloy is castable and is rollable to thin foils. A preferred embodiment of about 7 w/o zirconium, balance plutonium, has a melting point substantially above the melting point of plutonium, is rollable to foils as thin as 0.0005 inch thick, and is compatible with cladding material when repeatedly cycled to temperatures above 650.degree. C. Neutron reflux densities across a reactor core can be determined with a high-temperature activation-measurement foil which consists of a fuel alloy foil core sandwiched and sealed between two cladding material jackets, the fuel alloy foil core being a 7 w/o zirconium, plutonium foil which is from 0.005 to 0.0005 inch thick.

  9. Some Effects of Injection Advance Angle, Engine-Jacket Temperature, and Speed on Combustion in a Compression-Ignition Engine

    NASA Technical Reports Server (NTRS)

    Rothrock, A M; Waldron, C D

    1936-01-01

    An optical indicator and a high-speed motion-picture camera capable of operating at the rate of 2,000 frames per second were used to record simultaneously the pressure development and the flame formation in the combustion chamber of the NACA combustion apparatus. Tests were made at engine speeds of 570 and 1,500 r.p.m. The engine-jacket temperature was varied from 100 degrees to 300 degrees F. And the injection advance angle from 13 degrees after top center to 120 degrees before top center. The results show that the course of the combustion is largely controlled by the temperature and pressure of the air in the chamber from the time the fuel is injected until the time at which combustion starts and by the ignition lag. The conclusion is presented that in a compression-ignition engine with a quiescent combustion chamber the ignition lag should be the longest that can be used without excessive rates of pressure rise; any further shortening of the ignition lag decreased the effective combustion of the engine.

  10. CERAMIC FUEL ELEMENT MATERIAL FOR A NEUTRONIC REACTOR AND METHOD OF FABRICATING SAME

    DOEpatents

    Duckworth, W.H.

    1957-12-01

    This patent relates to ceramic composition, and to neutronic reactor fuel elements formed therefrom. These ceramic elements have high density and excellent strength characteristics and are formed by conventional ceramic casting and sintering at a temperature of about 2700 deg F in a nitrogen atmosphere. The composition consists of silicon carbide, silicon, uranium oxide and a very small percentage of molybdenum. Compositions containing molybdenum are markedly stronger than those lacking this ingredient.

  11. Study of a Tricarbide Grooved Ring Fuel Element for Nuclear Thermal Propulsion

    NASA Technical Reports Server (NTRS)

    Taylor, Brian; Emrich, Bill; Tucker, Dennis; Barnes, Marvin; Donders, Nicolas; Benensky, Kelsa

    2018-01-01

    Deep space exploration, especially that of Mars, is on the horizon as the next big challenge for space exploration. Nuclear propulsion, through which high thrust and efficiency can be achieved, is a promising option for decreasing the cost and logistics of such a mission. Work on nu- clear thermal engines goes back to the days of the NERVA program. Currently, nuclear thermal propulsion is under development again in various forms to provide a superior propulsion system for deep space exploration. The authors have been working to develop a concept nuclear thermal engine that uses a grooved ring fuel element as an alternative to the traditional hexagonal rod design. The authors are also studying the use of carbide fuels. The concept was developed in order to increase surface area and heat transfer to the propellant. The use of carbides would also raise the operating temperature of the reactor. It is hoped that this could lead to a higher thrust to weight nuclear thermal engine. This paper describes the modeling of neutronics, heat transfer, and fluid dynamics of this alternative nuclear fuel element geometry. Fabrication experiments of grooved rings from carbide refractory metals are also presented along with material characterization and interactions with a hot hydrogen environment. Results of experiments and associated analysis are desired densities with some success in material distribution and reaching a solid solution. Future work is needed to improve distribution of material, minimize oxidation during the milling process, and de ne a fabrication process that will serve for constructing grooved ring fuel rods for large system tests.

  12. Engine performance with a hydrogenated safety fuel

    NASA Technical Reports Server (NTRS)

    Schey, Oscar W; Young, Alfred W

    1933-01-01

    This report presents the results of an investigation to determine the engine performance obtained with a hydrogenated safety fuel developed to eliminate fire hazard. The tests were made on a single-cylinder universal test engine at compression ratios of 5.0, 5.5, and 6.0. Most of the tests were made with a fuel-injection system, although one set of runs was made with a carburetor when using gasoline to establish comparative performance. The tests show that the b.m.e.p. obtained with safety fuel when using a fuel-injection system is slightly higher than that obtained with gasoline when using a carburetor, although the fuel consumption with safety fuel is higher. When the fuel-injection system is used with each fuel and with normal engine temperatures the b.m.e.p. with safety fuel is from 2 to 4 percent lower than with gasoline and the fuel consumption about 25 to 30 percent higher. However, a few tests at an engine coolant temperature of 250 F have shown a specific fuel consumption approximating that obtained with gasoline with only a slight reduction in power. The idling of the test engine was satisfactory with the safety fuel. Starting was difficult with a cold engine but could be readily accomplished when the jacket water was hot. It is believed that the use of the safety fuel would practically eliminate crash fires.

  13. NUCLEAR REACTOR FUEL ELEMENT

    DOEpatents

    Currier, E.L. Jr.; Nicklas, J.H.

    1963-06-11

    A fuel plate is designed for incorporation into control rods of the type utilized in high-flux test reactors. The fuel plate is designed so that the portion nearest the poison section of the control rod contains about one-half as much fissionable material as in the rest of the plate, thereby eliminating dangerous flux peaking in that portion. (AEC)

  14. PREIRRADIATION MEASUREMENTS OF PIQUA FUEL ELEMENTS NO. P-1111, P-1113, P- 1114, AND P-1120

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Hubbell, H.J.

    1962-11-01

    Results of preirradiation measurements and tests performed during the processing and assembly of the individual fuel cylinders contained in Piqua Fuel Elements No. P-1111, P-1113, P-1114, and P-1120 are presented. A description of the techniques and equipment used in obtaining the data is also included. (auth)

  15. Study of a Tricarbide Grooved Ring Fuel Element for Nuclear Thermal Propulsion

    NASA Technical Reports Server (NTRS)

    Taylor, Brian; Emrich, Bill; Tucker, Dennis; Barnes, Marvin; Donders, Nicolas; Benensky, Kelsa

    2018-01-01

    Deep space exploration, especially that of Mars, is on the horizon as the next big challenge for space exploration. Nuclear propulsion, through which high thrust and efficiency can be achieved, is a promising option for decreasing the cost and logistics of such a mission. Work on nuclear thermal engines goes back to the days of the NERVA program. Currently, nuclear thermal propulsion is under development again in various forms to provide a superior propulsion system for deep space exploration. The authors have been working to develop a concept nuclear thermal engine that uses a grooved ring fuel element as an alternative to the traditional hexagonal rod design. The authors are also studying the use of carbide fuels. The concept was developed in order to increase surface area and heat transfer to the propellant. The use of carbides would also raise the operating temperature of the reactor. It is hoped that this could lead to a higher thrust to weight nuclear thermal engine. This paper describes the modeling of neutronics, heat transfer, and fluid dynamics of this alternative nuclear fuel element geometry. Fabrication experiments of grooved rings from carbide refractory metals are also presented along with material characterization and interactions with a hot hydrogen environment. Results of experiments and associated analysis are discussed. The authors demonstrated success in reaching desired densities with some success in material distribution and reaching a solid solution. Future work is needed to improve distribution of material, minimize oxidation during the milling process, and define a fabrication process that will serve for constructing grooved ring fuel rods for large system tests.

  16. Apparatus for inspecting fuel elements

    DOEpatents

    Oakley, David J.; Groves, Oliver J.; Kaiser, Bruce J.

    1986-01-01

    Disclosed is an alpha monitor usable in an automated nuclear fuel pin loading and processing unit. Fuel pins or other elongated pins are fed laterally into the alpha monitor in a singular fashion and are translated by a first roller assembly into a weld flare machining and decontamination substation not forming a part of the invention. Pins return and are lifted upwardly and transferred across to a combined pin lifting and electrode operating means which lifts the pins upwardly into a clamshell electrode assembly which is spread open by a combined pin lifting and electrode operating means. Once inserted the clamshell type electrode arrangement closes around the fuel pins so that inspection can occur. Fuel pins are inspected by charging electrodes to a negative potential and measuring the change in charge occurring when positively charged alpha particles strike the negatively charged electrodes. After inspection, the fuel pins are lowered by the pin lifting and electrode operating means into a second roller assembly which longitudinally conveys approved pins from the airtight enclosure in which the alpha monitor is mounted. If the fuel pins are rejected then they are moved laterally by a second transfer means and onto another system for further processing.

  17. Apparatus for inspecting fuel elements

    DOEpatents

    Kaiser, B.J.; Oakley, D.J.; Groves, O.J.

    1984-12-21

    This disclosure describes an alpha monitor usable in an automated nuclear fuel pin loading and processing unit. Fuel pins or other elongated pins are fed laterally into the alpha monitor in a singular fashion and are translated by a first roller assembly into a weld flare machining and decontamination substation not forming a part of the invention. Pins return and are lifted upwardly and transferred across to a combined pin lifting and electrode operating means which lifts the pins upwardly into a clamshell electrode assembly which is spread open by a combined pin lifting and electrode operating means. Once inserted the clamshell type electrode arrangement closes around the fuel pins so that inspection can occur. Fuel pins are inspected by charging electrodes to a negative potential and measuring the change in charge occurring when positively charged alpha particles strike the negatively charged electrodes. After inspection, the fuel pins are lowered by the pin lifting and electrode operating means into a second roller assembly which longitudinally conveys approved pins from the airtight enclosure in which the alpha monitor is mounted. If the fuel pins are rejected then they are moved laterally by a second transfer means and onto another system for further processing.

  18. Evaluation of Military Fuels Using a Ford 6.7L Powerstroke Diesel Engine

    DTIC Science & Technology

    2011-08-01

    natural steady state values during idle testing steps. Engine oil cooler plumbing was factory integrated to the engine water jacket, thus not...Innospec Fuel Specialties DCI-4A. Per QPL-25017, the minimum effective treat rate of DCI-4A required an additive concentration level of 9ppm in the...dynamometer was used to control engine speed and dissipate load. Engine load was manipulated through the actuation of the engine throttle pedal assembly

  19. Experimental evaluation of thermal ratcheting behavior in UO2 fuel elements

    NASA Technical Reports Server (NTRS)

    Phillips, W. M.

    1973-01-01

    The effects of thermal cycling of UO2 at high temperatures has been experimentally evaluated to determine the rates of distortion of UO2/clad fuel elements. Two capsules were rested in the 1500 C range, one with a 50 C thermal cycle, the other with a 100 C thermal cycle. It was observed that eight hours at the lower cycle temperature produced sufficient UO2 redistribution to cause clad distortion. The amount of distortion produced by the 100 C cycle was less than double that produced by the 50 C, indicating smaller thermal cycles would result in clad distortion. An incubation period was observed to occur before the onset of distortion with cycling similar to fuel swelling observed in-pile at these temperatures.

  20. Finite Element Stress Analysis of Spent Nuclear Fuel Disposal Canister in a Deep Geological Repository

    NASA Astrophysics Data System (ADS)

    Kwon, Young Joo; Choi, Jong Won

    This paper presents the finite element stress analysis of a spent nuclear fuel disposal canister to provide basic information for dimensioning the canister and configuration of canister components and consequently to suggest the structural analysis methodology for the disposal canister in a deep geological repository which is nowadays very important in the environmental waste treatment technology. Because of big differences in the pressurized water reactor (PWR) and the Canadian deuterium and uranium reactor (CANDU) fuel properties, two types of canisters are conceived. For manufacturing, operational reasons and standardization, however, both canisters have the same outer diameter and length. The construction type of canisters introduced here is a solid structure with a cast insert and a corrosion resistant overpack. The structural stress analysis is carried out using a finite element analysis code, NISA, and focused on the structural strength of the canister against the expected external pressures due to the swelling of the bentonite buffer and the hydrostatic head. The canister must withstand these large pressure loads. Consequently, canisters presented here contain 4 PWR fuel assemblies and 33×9 CANDU fuel bundles. The outside diameter of the canister for both fuels is 122cm and the cast insert diameter is 112cm. The total length of the canister is 483cm with the lid/bottom and the outer shell of 5cm.

  1. Vertical feed stick wood fuel burning furnace system

    DOEpatents

    Hill, Richard C.

    1982-01-01

    A stove or furnace for efficient combustion of wood fuel includes a vertical feed combustion chamber (15) for receiving and supporting wood fuel in a vertical attitude or stack. A major upper portion of the combustion chamber column comprises a water jacket (14) for coupling to a source of water or heat transfer fluid for convection circulation of the fluid. The locus (31) of wood fuel combustion is thereby confined to the refractory base of the combustion chamber. A flue gas propagation delay channel (34) extending laterally from the base of the chamber affords delayed travel time in a high temperature refractory environment sufficient to assure substantially complete combustion of the gaseous products of wood burning with forced air prior to extraction of heat in heat exchanger (16). Induced draft draws the fuel gas and air mixture laterally through the combustion chamber and refractory high temperature zone to the heat exchanger and flue. Also included are active sources of forced air and induced draft, multiple circuit couplings for the recovered heat, and construction features in the refractory material substructure and metal component superstructure.

  2. Vertical feed stick wood fuel burning furnace system

    DOEpatents

    Hill, Richard C.

    1984-01-01

    A new and improved stove or furnace for efficient combustion of wood fuel including a vertical feed combustion chamber for receiving and supporting wood fuel in a vertical attitude or stack, a major upper portion of the combustion chamber column comprising a water jacket for coupling to a source of water or heat transfer fluid and for convection circulation of the fluid for confining the locus of wood fuel combustion to the bottom of the vertical gravity feed combustion chamber. A flue gas propagation delay channel extending from the laterally directed draft outlet affords delayed travel time in a high temperature environment to assure substantially complete combustion of the gaseous products of wood burning with forced air as an actively induced draft draws the fuel gas and air mixture laterally through the combustion and high temperature zone. Active sources of forced air and induced draft are included, multiple use and circuit couplings for the recovered heat, and construction features in the refractory material substructure and metal component superstructure.

  3. Investigation of a Tricarbide Grooved Ring Fuel Element for a Nuclear Thermal Rocket

    NASA Technical Reports Server (NTRS)

    Taylor, Brian D.; Emrich, Bill; Tucker, Dennis; Barnes, Marvin; Donders, Nicolas; Benensky, Kelsa

    2017-01-01

    Deep space exploration, especially that of Mars, is on the horizon as the next big challenge for space exploration. Nuclear propulsion, through which high thrust and efficiency can be achieved, is a promising option for decreasing the cost and logistics of such a mission. Work on nuclear thermal engines goes back to the days of the NERVA program. Currently, nuclear thermal propulsion is under development again in various forms to provide a superior propulsion system for deep space exploration. The authors have been working to develop a concept nuclear thermal engine that uses a grooved ring fuel element as an alternative to the traditional hexagonal rod design. The authors are also studying the use of carbide fuels. The concept was developed in order to increase surface area and heat transfer to the propellant. The use of carbides would also raise the temperature limitations of the reactor. It is hoped that this could lead to a higher thrust to weight nuclear thermal engine. This paper describes the modeling of neutronics, heat transfer, and fluid dynamics of this alternative nuclear fuel element geometry. Fabrication experiments of grooved rings from carbide refractory metals are also presented along with material characterization and interactions with a hot hydrogen environment.

  4. Modeling of the heat transfer performance of plate-type dispersion nuclear fuel elements

    NASA Astrophysics Data System (ADS)

    Ding, Shurong; Huo, Yongzhong; Yan, XiaoQing

    2009-08-01

    Considering the mutual actions between fuel particles and the metal matrix, the three-dimensional finite element models are developed to simulate the heat transfer behaviors of dispersion nuclear fuel plates. The research results indicate that the temperatures of the fuel plate might rise more distinctly with considering the particle swelling and the degraded surface heat transfer coefficients with increasing burnup; the local heating phenomenon within the particles appears when their thermal conductivities are too low. With rise of the surface heat transfer coefficients, the temperatures within the fuel plate decrease; the temperatures of the fuel plate are sensitive to the variations of the heat transfer coefficients whose values are lower, but their effects are weakened and slight when the heat transfer coefficients increase and reach a certain extent. Increasing the heat generation rate leads to elevating the internal temperatures. The temperatures and the maximum temperature differences within the plate increase along with the particle volume fractions. The surface thermal flux goes up along with particle volume fractions and heat generation rates, but the effects of surface heat transfer coefficients are not evident.

  5. Ionizing radiation effects on ISS ePTFE jacketed cable assembly

    NASA Astrophysics Data System (ADS)

    Koontz, S. L.; Golden, J. L.; Lorenz, M. J.; Pedley, M. D.

    2003-09-01

    Polytetrafluoroethylene (PTFE), which is susceptible to embrittlement by ionizing radiation, is used as a primary material in the Mobile Transporter's (MT) Trailing Umbilical System (TUS) cable on the International Space Station (ISS). The TUS cable provides power and data service between the ISS truss and the MT. The TUS cable is normally stowed in an uptake reel and is fed out to follow the MT as it moves along rails on the ISS truss structure. For reliable electrical and mechanical performance, TUS cable polymeric materials must be capable of >3.5% elongation without cracking or breaking. The MT TUS cable operating temperature on ISS is expected to range between -100°C and +130°C. The on-orbit functional life requirement for the MT TUS cable is 10 years. Analysis and testing were performed to verify that the MT TUS cable would be able to meet full-life mechanical and electrical performance requirements, despite progressive embrittlement by the natural ionizing radiation environment. Energetic radiation belt electrons (trapped electrons) are the principal contributor to TUS cable radiation dose. TUS cable specimens were irradiated, in vacuum, with both energetic electrons and gamma rays. Electron beam energy was chosen to minimize charging effects on the non-conductive ePTFE (expanded PTFE) targets. Tensile testing was then performed, over the expected range of operating temperatures, as a function of radiation dose. When compared to the expected in-flight radiation dose/depth profile, atomic oxygen (AO) erosion of the radiation damaged TUS cable jacket surfaces is more rapid than the development of radiation induced embrittlement of the same surfaces. Additionally, the layered construction of the jacket prevents crack growth propagation, leaving the inner layer material compliant with the design elongation requirements. As a result, the TUS cable insulation design was verified to meet performance life requirements.

  6. Thermal breeder fuel enrichment zoning

    DOEpatents

    Capossela, Harry J.; Dwyer, Joseph R.; Luce, Robert G.; McCoy, Daniel F.; Merriman, Floyd C.

    1992-01-01

    A method and apparatus for improving the performance of a thermal breeder reactor having regions of higher than average moderator concentration are disclosed. The fuel modules of the reactor core contain at least two different types of fuel elements, a high enrichment fuel element and a low enrichment fuel element. The two types of fuel elements are arranged in the fuel module with the low enrichment fuel elements located between the high moderator regions and the high enrichment fuel elements. Preferably, shim rods made of a fertile material are provided in selective regions for controlling the reactivity of the reactor by movement of the shim rods into and out of the reactor core. The moderation of neutrons adjacent the high enrichment fuel elements is preferably minimized as by reducing the spacing of the high enrichment fuel elements and/or using a moderator having a reduced moderating effect.

  7. Inconsistency in 9 mm bullets: correlation of jacket thickness to post-impact geometry measured with non-destructive X-ray computed tomography.

    PubMed

    Thornby, John; Landheer, Dirk; Williams, Tim; Barnes-Warden, Jane; Fenne, Paul; Norman, Danielle G; Attridge, Alex; Williams, Mark A

    2014-01-01

    Fundamental to any ballistic armour standard is the reference projectile to be defeated. Typically, for certification purposes, a consistent and symmetrical bullet geometry is assumed, however variations in bullet jacket dimensions can have far reaching consequences. Traditionally, characteristics and internal dimensions have been analysed by physically sectioning bullets--an approach which is of restricted scope and which precludes subsequent ballistic assessment. The use of a non-destructive X-ray computed tomography (CT) method has been demonstrated and validated (Kumar et al., 2011 [15]); the authors now apply this technique to correlate bullet impact response with jacket thickness variations. A set of 20 bullets (9 mm DM11) were selected for comparison and an image-based analysis method was employed to map jacket thickness and determine the centre of gravity of each specimen. Both intra- and inter-bullet variations were investigated, with thickness variations of the order of 200 μm commonly found along the length of all bullets and angular variations of up to 50 μm in some. The bullets were subsequently impacted against a rigid flat plate under controlled conditions (observed on a high-speed video camera) and the resulting deformed projectiles were re-analysed. The results of the experiments demonstrate a marked difference in ballistic performance between bullets from different manufacturers and an asymmetric thinning of the jacket is observed in regions of pre-impact weakness. The conclusions are relevant for future soft armour standards and provide important quantitative data for numerical model correlation and development. The implications of the findings of the work on the reliability and repeatability of the industry standard V50 ballistic test are also discussed. Copyright © 2013 Elsevier Ireland Ltd. All rights reserved.

  8. Experimental Characterization and Modeling of Thermal Contact Resistance of Electric Machine Stator-to-Cooling Jacket Interface Under Interference Fit Loading

    DOE PAGES

    Cousineau, Justine Emily; Bennion, Kevin S.; Chieduko, Victor; ...

    2018-05-08

    Cooling of electric machines is a key to increasing power density and improving reliability. This paper focuses on the design of a machine using a cooling jacket wrapped around the stator. The thermal contact resistance (TCR) between the electric machine stator and cooling jacket is a significant factor in overall performance and is not well characterized. This interface is typically an interference fit subject to compressive pressure exceeding 5 MPa. An experimental investigation of this interface was carried out using a thermal transmittance setup using pressures between 5 and 10 MPa. Furthermore, the results were compared to currently available modelsmore » for contact resistance, and one model was adapted for prediction of TCR in future motor designs.« less

  9. Experimental Characterization and Modeling of Thermal Contact Resistance of Electric Machine Stator-to-Cooling Jacket Interface Under Interference Fit Loading

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Cousineau, Justine Emily; Bennion, Kevin S.; Chieduko, Victor

    Cooling of electric machines is a key to increasing power density and improving reliability. This paper focuses on the design of a machine using a cooling jacket wrapped around the stator. The thermal contact resistance (TCR) between the electric machine stator and cooling jacket is a significant factor in overall performance and is not well characterized. This interface is typically an interference fit subject to compressive pressure exceeding 5 MPa. An experimental investigation of this interface was carried out using a thermal transmittance setup using pressures between 5 and 10 MPa. Furthermore, the results were compared to currently available modelsmore » for contact resistance, and one model was adapted for prediction of TCR in future motor designs.« less

  10. The development of fuel performance models at the European institute for transuranium elements

    NASA Astrophysics Data System (ADS)

    Lassmann, K.; Ronchi, C.; Small, G. J.

    1989-07-01

    The design and operational performance of fuel rods for nuclear power stations has been the subject of detailed experimental research for over thirty years. In the last two decades the continuous demands for greater economy in conjunction with more stringent safety criteria have led to an increasing reliance on computer simulations. Conditions within a fuel rod must be calculated both for normal operation and for proposed reactor faults. It has thus been necessary to build up a reliable, theoretical understanding of the intricate physical, mechanical and chemical processes occurring under a wide range of conditions to obtain a quantitative insight into the behaviour of the fuel. A prime requirement, which has also proved to be the most taxing, is to predict the conditions under which failure of the cladding might occur, particularly in fuel nearing the end of its useful life. In this paper the general requirements of a fuel performance code are discussed briefly and an account is given of the basic concepts of code construction. An overview is then given of recent progress at the European Institute for Transuranium Elements in the development of a fuel rod performance code for general application and of more detailed mechanistic models for fission product behaviour.

  11. Identification of failed fuel element

    DOEpatents

    Fryer, Richard M.; Matlock, Robert G.

    1976-06-22

    A passive fission product gas trap is provided in the upper portion of each fuel subassembly in a nuclear reactor. The gas trap consists of an inverted funnel of less diameter than the subassembly having a valve at the apex thereof. An actuating rod extends upwardly from the valve through the subassembly to a point where it can be contacted by the fuel handling mechanism for the reactor. Interrogation of the subassembly for the presence of fission products is accomplished by lowering the fuel handling machine onto the subassembly to press down on the actuating rod and open the valve.

  12. Discrete element method study of fuel relocation and dispersal during loss-of-coolant accidents

    NASA Astrophysics Data System (ADS)

    Govers, K.; Verwerft, M.

    2016-09-01

    The fuel fragmentation, relocation and dispersal (FFRD) during LOCA transients today retain the attention of the nuclear safety community. The fine fragmentation observed at high burnup may, indeed, affect the Emergency Core Cooling System performance: accumulation of fuel debris in the cladding ballooned zone leads to a redistribution of the temperature profile, while dispersal of debris might lead to coolant blockage or to debris circulation through the primary circuit. This work presents a contribution, by discrete element method, towards a mechanistic description of the various stages of FFRD. The fuel fragments are described as a set of interacting particles, behaving as a granular medium. The model shows qualitative and quantitative agreement with experimental observations, such as the packing efficiency in the balloon, which is shown to stabilize at about 55%. The model is then applied to study fuel dispersal, for which experimental parametric studies are both difficult and expensive.

  13. Electrolyser and fuel cells, key elements for energy and life support

    NASA Astrophysics Data System (ADS)

    Bockstahler, Klaus; Funke, Helmut; Lucas, Joachim

    Both, Electrolyser and Fuel Cells are key elements for regenerative energy and life support systems. Electrolyser technology is originally intended for oxygen production in manned space habitats and in submarines, through splitting water into hydrogen and oxygen. Fuel cells serve for energy production through the reaction, triggered in the presence of an electrolyte, between a fuel and an oxidant. Now combining both technologies i.e. electrolyser and fuel cell makes it a Regenerative Fuel Cell System (RFCS). In charge mode, i.e. with energy supplied e.g. by solar cells, the electrolyser splits water into hydrogen and oxygen being stored in tanks. In discharge mode, when power is needed but no energy is available, the stored gases are converted in the fuel cell to generate electricity under the formation of water that is stored in tanks. Rerouting the water to the electrolyser makes it a closed-loop i.e. regenerative process. Different electrolyser and fuel cell technologies are being evolved. At Astrium emphasis is put on the development of an RFCS comprised of Fixed Alkaline Electrolyser (FAE) and Fuel Cell (AFC) as such technology offers a high electrical efficiency and thus reduced system weight, which is important in space applications. With increasing power demand and increasing discharge time an RFCS proves to be superior to batteries. Since the early technology development multiple design refinements were done at Astrium, funded by the European Space Agency ESA and the German National Agency DLR as well as based on company internal R and T funding. Today a complete RFCS energy system breadboard is established and the operational behavior of the system is being tested. In parallel the electrolyser itself is subject to design refinement and testing in terms of oxygen production in manned space habitats. In addition essential features and components for process monitoring and control are being developed. The present results and achievements and the dedicated

  14. Determination of neutron multiplication coefficients for fuel elements irradiated by spallation neutrons

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Bhatia, Chitra; Kumar, V.

    2010-02-15

    A neutron multiplication coefficient, k{sub eff}, has been estimated for spallation neutron flux using the data of spectrum average cross sections of all absorption, fission, and nonelastic reaction channels of {sup 232}Th, {sup 238}U, {sup 235}U, and {sup 233}U fuel elements. It has been revealed that in spallation neutron flux (i) nonfission, nonabsorption reactions play an important role in the calculation of k{sub eff}, (ii) one can obtain a high value of k{sub eff} even for fertile {sup 232}Th fuel, which is hardly possible in a conventional fast reactor, and (iii) spectrum average absorption cross sections of neutron poisons ofmore » a conventional reactor are relatively very small.« less

  15. DEFLECTION PRESSURE TESTER

    DOEpatents

    Cooper, C.M.

    1961-01-01

    A method and apparatus for determining whether the jacket of a nuclear- fuel slug has a leak are described. The region of the jacket to be leak-tested is sealed off, and gas under pressure is applied thereto. If there is an imperfection, the gas will enter the jacket and bulge another region of the jacket. The bulge occurring is measured by a gage.

  16. Nuclear reactor composite fuel assembly

    DOEpatents

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

    1980-01-01

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

  17. Pumped lithium loop test to evaluate advanced refractory metal alloys and simulated nuclear fuel elements

    NASA Technical Reports Server (NTRS)

    Brandenburf, G. P.; Hoffman, E. E.; Smith, J. P.

    1974-01-01

    The performance was determined of refractory metal alloys and uranium nitride fuel element specimens in flowing 1900F (1083C) lithium. The results demonstrate the suitability of the selected materials to perform satisfactorily from a chemical compatibility standpoint.

  18. Mesocarbon microbead based graphite for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor

    NASA Astrophysics Data System (ADS)

    Zhong, Yajuan; Zhang, Junpeng; Lin, Jun; Xu, Liujun; Zhang, Feng; Xu, Hongxia; Chen, Yu; Jiang, Haitao; Li, Ziwei; Zhu, Zhiyong; Guo, Quangui

    2017-07-01

    Mesocarbon microbeads (MCMB) and quasi-isostatic pressing method were used to prepare MCMB based graphite (MG) for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor (MSR). Characteristics of mercury infiltration and molten salt infiltration in MG were investigated and compared with A3-3 (graphite for spherical fuel element in high temperature gas cooled reactor) to identify the infiltration behaviors. The results indicated that MG had a low porosity about 14%, and an average pore diameter of 96 nm. Fluoride salt occupation of A3-3 (average pore diameter was 760 nm) was 10 wt% under 6.5 atm, whereas salt gain did not infiltrate in MG even up to 6.5 atm. It demonstrated that MG could inhibit the infiltration of liquid fluoride salt effectively. Coefficient of thermal expansion (CTE) of MG lies in 6.01 × 10-6 K-1 (α∥) and 6.15 × 10-6 K-1 (α⊥) at the temperature range of 25-700 °C. The anisotropy factor of MG calculated by CTE maintained below 1.02, which could meet the requirement of the spherical fuel element (below 1.30). The constant isotropic property of MG is beneficial for the integrity and safety of the graphite used in the spherical fuel element for a MSR.

  19. Experimental duplication of the important physical evidence of the lapel bulge of the jacket worn by Governor Connally when bullet 399 went through him.

    PubMed

    Lattimer, J K; Laidlaw, A; Heneghan, P; Haubner, E J

    1994-05-01

    By duplicating the wound to the neck of President Kennedy, which caused bullet 399 to turn sideways, and having it then hit a Connally-type rib cage with shirt and jacket, we reproduced the right-sided bulge of the jacket worn by Connally, with lapel eversion, which is so significant in frame 224. The extensive damage to his shirtfront was from the hail of rib fragments and soft tissue, exactly as described with his own shirt. Our tumbling bullet then went on to fracture a radius and be recovered intact except that it was somewhat flattened and bent and had lead extruded from the rear, as did bullet 399. Fragments of this lead were scraped off on the ragged bone-ends of some of our fractured radiuses, just as with Governor Connally's radius. It is believed that this duplication of the jacket and lapel bulge of Governor Connally, which occurred dependably, when we reproduced the circumstances at Dallas, confirmed this very important detail in this technical demonstration of the findings in the shooting of President Kennedy and Governor Connally. The bulge and the lapel eversion of the jacket worn by Governor Connally, starting in Zapruder frame 224, does indeed establish, beyond any shadow of a doubt, the exact moment when bullet 399 went through him. The right arms of both men were seen to react simultaneously, immediately thereafter. It also permits us to establish that there was plenty of time (three and one-half seconds) between the first two shots (frames 160 to 224) and even more time (five seconds) between the last two shots (frames 224 to 313), for Oswald to reload, reacquire his target (the head of President Kennedy) plus two full seconds to lock onto it. If the bullet does not traverse the neck of President Kennedy, it does not cause Governor Connally's jacket and lapel to bulge. The lapel bulge is a very important bit of actual physical evidence in establishing the fact that one bullet hit both men and that Oswald had plenty of time to hit the President

  20. Fuel pumping system and method

    DOEpatents

    Shafer, Scott F [Morton, IL; Wang, Lifeng ,

    2006-12-19

    A fuel pumping system that includes a pump drive is provided. A first pumping element is operatively connected to the pump drive and is operable to generate a first flow of pressurized fuel. A second pumping element is operatively connected to the pump drive and is operable to generate a second flow of pressurized fuel. A first solenoid is operatively connected to the first pumping element and is operable to vary at least one of a fuel pressure and a fuel flow rate of the first flow of pressurized fuel. A second solenoid is operatively connected to the second pumping element and is operable to vary at least one of a fuel pressure and a fuel flow rate of the second flow of pressurized fuel.

  1. Fuel Pumping System And Method

    DOEpatents

    Shafer, Scott F.; Wang, Lifeng

    2005-12-13

    A fuel pumping system that includes a pump drive is provided. A first pumping element is operatively connected to the pump drive and is operable to generate a first flow of pressurized fuel. A second pumping element is operatively connected to the pump drive and is operable to generate a second flow of pressurized fuel. A first solenoid is operatively connected to the first pumping element and is operable to vary at least one of a fuel pressure and a fuel flow rate of the first flow of pressurized fuel. A second solenoid is operatively connected to the second pumping element and is operable to vary at least one of a fuel pressure and a fuel flow rate of the second flow of pressurized fuel.

  2. Distribution and leaching characteristics of trace elements in ashes as a function of different waste fuels and incineration technologies.

    PubMed

    Saqib, Naeem; Bäckström, Mattias

    2015-10-01

    Impact of waste fuels (virgin/waste wood, mixed biofuel (peat, bark, wood chips) industrial, household, mixed waste fuel) and incineration technologies on partitioning and leaching behavior of trace elements has been investigated. Study included 4 grate fired and 9 fluidized boilers. Results showed that mixed waste incineration mostly caused increased transfer of trace elements to fly ash; particularly Pb/Zn. Waste wood incineration showed higher transfer of Cr, As and Zn to fly ash as compared to virgin wood. The possible reasons could be high input of trace element in waste fuel/change in volatilization behavior due to addition of certain waste fractions. The concentration of Cd and Zn increased in fly ash with incineration temperature. Total concentration in ashes decreased in order of Zn>Cu>Pb>Cr>Sb>As>Mo. The concentration levels of trace elements were mostly higher in fluidized boilers fly ashes as compared to grate boilers (especially for biofuel incineration). It might be attributed to high combustion efficiency due to pre-treatment of waste in fluidized boilers. Leaching results indicated that water soluble forms of elements in ashes were low with few exceptions. Concentration levels in ash and ash matrix properties (association of elements on ash particles) are crucial parameters affecting leaching. Leached amounts of Pb, Zn and Cr in >50% of fly ashes exceeded regulatory limit for disposal. 87% of chlorine in fly ashes washed out with water at the liquid to solid ratio 10 indicating excessive presence of alkali metal chlorides/alkaline earths. Copyright © 2015. Published by Elsevier B.V.

  3. VANADIUM ALLOYS

    DOEpatents

    Smith, K.F.; Van Thyne, R.J.

    1959-05-12

    This patent deals with vanadium based ternary alloys useful as fuel element jackets. According to the invention the ternary vanadium alloys, prepared in an arc furnace, contain from 2.5 to 15% by weight titanium and from 0.5 to 10% by weight niobium. Characteristics of these alloys are good thermal conductivity, low neutron capture cross section, good corrosion resistance, good welding and fabricating properties, low expansion coefficient, and high strength.

  4. Development of jacket platform tsunami risk rating system in waters offshore North Borneo

    NASA Astrophysics Data System (ADS)

    Lee, H. E.; Liew, M. S.; Mardi, N. H.; Na, K. L.; Toloue, Iraj; Wong, S. K.

    2016-09-01

    This work details the simulation of tsunami waves generated by seaquakes in the Manila Trench and their effect on fixed oil and gas jacket platforms in waters offshore North Borneo. For this study, a four-leg living quarter jacket platform located in a water depth of 63m is modelled in SACS v5.3. Malaysia has traditionally been perceived to be safe from the hazards of earthquakes and tsunamis. Local design practices tend to neglect tsunami waves and include no such provisions. In 2004, a 9.3 M w seaquake occurred off the northwest coast of Aceh, which generated tsunami waves that caused destruction in Malaysia totalling US 25 million and 68 deaths. This event prompted an awareness of the need to study the reliability of fixed offshore platforms scattered throughout Malaysian waters. In this paper, we present a review of research on the seismicity of the Manila Trench, which is perceived to be high risk for Southeast Asia. From the tsunami numerical model TUNA-M2, we extract computer-simulated tsunami waves at prescribed grid points in the vicinity of the platforms in the region. Using wave heights as input, we simulate the tsunami using SACS v5.3 structural analysis software of offshore platforms, which is widely accepted by the industry. We employ the nonlinear solitary wave theory in our tsunami loading calculations for the platforms, and formulate a platform-specific risk quantification system. We then perform an intensive structural sensitivity analysis and derive a corresponding platform-specific risk rating model.

  5. Doing Research in a Professional Development School and Why I Feel like Houdini in a Straight Jacket

    ERIC Educational Resources Information Center

    Neapolitan, Jane

    2004-01-01

    When the author was a little girl, her father told her the story of how he once saw the great Harry Houdini escape from a straight jacket. In the early 1900s, Houdini fascinated thousands of people by performing feats of life-threatening daring and agility. Houdini was a professional in the traditional sense: he acted alone, kept his trade secrets…

  6. 40 CFR 79.56 - Fuel and fuel additive grouping system.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... further testing under the provisions of Tier 3 or to support regulatory decisions affecting that fuel or... elements or classes of compounds other than those permitted in the base fuel for the respective fuel family... all of the following criteria: (1) Contain no elements other than carbon, hydrogen, oxygen, nitrogen...

  7. Physiological Evaluation of A1 (Extreme-Cold-Weather) and A2 (Buoyant, Intermediate-Cold-Weather) Jackets.

    DTIC Science & Technology

    1983-08-01

    the resting metabolic heat will be dissipated through the clothing with the remaining 25% lost through the respiratory tract and insensible sweating...AD-A258 410 PHYSIOLOGICAL EVALUATION OF Al (EXTREME-COLD-WEATHER) AND A2 (BUOYANT, INTERMEDIATE-COLD-WEATHER) JACKETS NAVY CLOTHING AND TEXTILE...Navy Clothing and Textile Research Facility 523-003-30-06 21 Strathmore Road 523-003-30-08 Natick, MA 01760 11. CONTROLLING OFFICE NAME AND ADDRESS

  8. Accelerator-driven transmutation of spent fuel elements

    DOEpatents

    Venneri, Francesco; Williamson, Mark A.; Li, Ning

    2002-01-01

    An apparatus and method is described for transmuting higher actinides, plutonium and selected fission products in a liquid-fuel subcritical assembly. Uranium may also be enriched, thereby providing new fuel for use in conventional nuclear power plants. An accelerator provides the additional neutrons required to perform the processes. The size of the accelerator needed to complete fuel cycle closure depends on the neutron efficiency of the supported reactors and on the neutron spectrum of the actinide transmutation apparatus. Treatment of spent fuel from light water reactors (LWRs) using uranium-based fuel will require the largest accelerator power, whereas neutron-efficient high temperature gas reactors (HTGRs) or CANDU reactors will require the smallest accelerator power, especially if thorium is introduced into the newly generated fuel according to the teachings of the present invention. Fast spectrum actinide transmutation apparatus (based on liquid-metal fuel) will take full advantage of the accelerator-produced source neutrons and provide maximum utilization of the actinide-generated fission neutrons. However, near-thermal transmutation apparatus will require lower standing

  9. A New Innovative Spherical Cermet Nuclear Fuel Element to Achieve an Ultra-Long Core Life for use in Grid-Appropriate LWRs

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Senor, David J.; Painter, Chad L.; Geelhood, Ken J.

    2007-12-01

    Spherical cermet fuel elements are proposed for use in the Atoms For Peace Reactor (AFPR-100) concept. AFPR-100 is a small-scale, inherently safe, proliferation-resistant reactor that would be ideal for deployment to nations with emerging economies that decide to select nuclear power for the generation of carbon-free electricity. The basic concept of the AFPR core is a water-cooled fixed particle bed, randomly packed with spherical fuel elements. The flow of coolant within the particle bed is at such a low rate that the bed does not fluidize. This report summarizes an approach to fuel fabrication, results associated with fuel performance modeling,more » core neutronics and thermal hydraulics analyses demonstrating a ~20 year core life, and a conclusion that the proliferation resistance of the AFPR reactor concept is high.« less

  10. Fuel assembly for nuclear reactors

    DOEpatents

    Creagan, Robert J.; Frisch, Erling

    1977-01-01

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

  11. Effects of defeathering and insulative jackets on production by laying hens at low temperatures.

    PubMed

    Gonyou, H W; Morrison, W D

    1983-07-01

    Exposure to a temperature of 5 degrees C compared with 20 degrees C resulted in a 20.5% increase in food consumption and an 18.8% decrease in efficiency of food utilisation with intermediate values resulting from exposure to 10 degrees C and 15 degrees C. Removal of feathers from the neck, back and (or) breast resulted in a 5 to 6% increase in food consumption. The effects of feather removal and temperature on food consumption were additive. Cloth jackets effectively insulated the back and breast areas when feathers had been removed but also resulted in increased food intake and lower efficiency.

  12. Low temperature chemical processing of graphite-clad nuclear fuels

    DOEpatents

    Pierce, Robert A.

    2017-10-17

    A reduced-temperature method for treatment of a fuel element is described. The method includes molten salt treatment of a fuel element with a nitrate salt. The nitrate salt can oxidize the outer graphite matrix of a fuel element. The method can also include reduced temperature degradation of the carbide layer of a fuel element and low temperature solubilization of the fuel in a kernel of a fuel element.

  13. Design of a reconfigurable liquid hydrogen fuel tank for use in the Genii unmanned aerial vehicle

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Adam, Patrick; Leachman, Jacob

    2014-01-29

    Long endurance flight, on the order of days, is a leading flight performance characteristic for Unmanned Aerial Vehicles (UAVs). Liquid hydrogen (LH2) is well suited to providing multi-day flight times with a specific energy 2.8 times that of conventional kerosene based fuels. However, no such system of LH2 storage, delivery, and use is currently available for commercial UAVs. In this paper, we develop a light weight LH2 dewar for integration and testing in the proton exchange membrane (PEM) fuel cell powered, student designed and constructed, Genii UAV. The fuel tank design is general for scaling to suit various UAV platforms.more » A cylindrical vacuum-jacketed design with removable end caps was chosen to incorporate various fuel level gauging, pressurizing, and slosh mitigation systems. Heat and mechanical loadings were modeled to compare with experimental results. Mass performance of the fuel tank is characterized by the fraction of liquid hydrogen to full tank mass, and the insulation performance was characterized by effective thermal conductivity and boil-off rate.« less

  14. Design of a reconfigurable liquid hydrogen fuel tank for use in the Genii unmanned aerial vehicle

    NASA Astrophysics Data System (ADS)

    Adam, Patrick; Leachman, Jacob

    2014-01-01

    Long endurance flight, on the order of days, is a leading flight performance characteristic for Unmanned Aerial Vehicles (UAVs). Liquid hydrogen (LH2) is well suited to providing multi-day flight times with a specific energy 2.8 times that of conventional kerosene based fuels. However, no such system of LH2 storage, delivery, and use is currently available for commercial UAVs. In this paper, we develop a light weight LH2 dewar for integration and testing in the proton exchange membrane (PEM) fuel cell powered, student designed and constructed, Genii UAV. The fuel tank design is general for scaling to suit various UAV platforms. A cylindrical vacuum-jacketed design with removable end caps was chosen to incorporate various fuel level gauging, pressurizing, and slosh mitigation systems. Heat and mechanical loadings were modeled to compare with experimental results. Mass performance of the fuel tank is characterized by the fraction of liquid hydrogen to full tank mass, and the insulation performance was characterized by effective thermal conductivity and boil-off rate.

  15. NEUTRONIC REACTOR CONTROL ELEMENT

    DOEpatents

    Newson, H.W.

    1960-09-13

    A novel composite neutronic reactor control element is offered. The element comprises a multiplicity of sections arranged in end-to-end relationship, each of the sections having a markedly different neutron-reactive characteristic. For example, a three-section control element could contain absorber, moderator, and fuel sections. By moving such an element longitudinally through a reactor core, reactivity is decreased by the absorber, increased slightly by the moderator, or increased substantially by the fuel. Thus, control over a wide reactivity range is provided.

  16. Trace element partitioning in ashes from boilers firing pure wood or mixtures of solid waste with respect to fuel composition, chlorine content and temperature.

    PubMed

    Saqib, Naeem; Bäckström, Mattias

    2014-12-01

    Trace element partitioning in solid waste (household waste, industrial waste, waste wood chips and waste mixtures) incineration residues was investigated. Samples of fly ash and bottom ash were collected from six incineration facilities across Sweden including two grate fired and four fluidized bed incinerators, to have a variation in the input fuel composition (from pure biofuel to mixture of waste) and different temperature boiler conditions. As trace element concentrations in the input waste at the same facilities have already been analyzed, the present study focuses on the concentration of trace elements in the waste fuel, their distribution in the incineration residues with respect to chlorine content of waste and combustion temperature. Results indicate that Zn, Cu and Pb are dominating trace elements in the waste fuel. Highly volatile elements mercury and cadmium are mainly found in fly ash in all cases; 2/3 of lead also end up in fly ash while Zn, As and Sb show a large variation in distribution with most of them residing in the fly ash. Lithophilic elements such as copper and chromium are mainly found in bottom ash from grate fired facilities while partition mostly into fly ash from fluidized bed incinerators, especially for plants fuelled by waste wood or ordinary wood chips. There is no specific correlation between input concentration of an element in the waste fuel and fraction partitioned to fly ash. Temperature and chlorine content have significant effects on partitioning characteristics by increasing the formation and vaporization of highly volatile metal chlorides. Zinc and cadmium concentrations in fly ash increase with the incineration temperature. Copyright © 2014 Elsevier Ltd. All rights reserved.

  17. High temperature fuel/emitter system for advanced thermionic fuel elements

    NASA Astrophysics Data System (ADS)

    Moeller, Helen H.; Bremser, Albert H.; Gontar, Alexander; Fiviesky, Evgeny

    1997-01-01

    Specialists in space applications are currently focusing on bimodal power systems designed to provide both electric power and thermal propulsion (Kennedy, 1994 and Houts, 1995). Our work showed that thermionics is a viable technology for nuclear bimodal power systems. We demonstrated that materials for a thermionic fuel-emitter combination capable of performing at operating temperatures of 2473 K are not only possible but available. The objective of this work, funded by the US Department of Energy, Office of Space and Defense Power Systems, was to evaluate the compatibility of fuel material consisting of an uranium carbide/tantalum carbide solid solution with an emitter material consisting of a monocrystalline tungsten-niobium alloy. The uranium loading of the fuel material was 70 mole% uranium carbide. The program was successfully accomplished by a B&W/SIA LUTCH team. Its workscope was integrated with tasks being performed at both Babcock & Wilcox, Lynchburg Research Center, Lynchburg, Virginia, and SIA LUTCH, Podolsk, Russia. Samples were fabricated by LUTCH and seven thermal tests were performed in a hydrogen atmosphere. The first preliminary test was performed at 2273 K by LUTCH, and the remaining six tests were performed At B&W. Three tests were performed at 2273 K, two at 2373 K, and the final test at 2473 K. The results showed that the fuel and emitter materials were compatible in the presence of hydrogen. No evidence of liquid formation, dissolution of the uranium carbide from the uranium carbide/tantalum carbide solid solution, or diffusion of the uranium into the monocrystalline tungsten alloy was observed. Among the highlights of the program was the successful export of the fuel samples from Russia and their import into the US by commercial transport. This paper will discuss the technical aspects of this work.

  18. Fuel shipment experience, fuel movements from the BMI-1 transport cask

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Bauer, Thomas L.; Krause, Michael G

    1986-07-01

    The University of Texas at Austin received two shipments of irradiated fuel elements from Northrup Aircraft Corporation on April 11 and 16, 1985. A total of 59 elements consisting of standard and instrumented TRIGA fuel were unloaded from the BMI-1 shipping cask. At the time of shipment, the Northrup core burnup was approximately 50 megawatt days with fuel element radiation levels, after a cooling time of three months, of approximately 1.75 rem/hr at 3 feet. In order to facilitate future planning of fuel shipment at the UT facility and other facilities, a summary of the recent transfer process including severalmore » factors which contributed to its success are presented. Numerous color slides were made of the process for future reference by UT and others involved in fuel transfer and handling of the BMI-1 cask.« less

  19. Thermodynamic and kinetic modelling of fuel oxidation behaviour in operating defective fuel

    NASA Astrophysics Data System (ADS)

    Lewis, operating defective fuel B. J.; Thompson, W. T.; Akbari, F.; Thompson, D. M.; Thurgood, C.; Higgs, J.

    2004-07-01

    A theoretical treatment has been developed to predict the fuel oxidation behaviour in operating defective nuclear fuel elements. The equilibrium stoichiometry deviation in the hyper-stoichiometric fuel has been derived from thermodynamic considerations using a self-consistent set of thermodynamic properties for the U-O system, which emphasizes replication of solubilities and three-phase invariant conditions displayed in the U-O binary phase diagram. The kinetics model accounts for multi-phase transport including interstitial oxygen diffusion in the solid and gas-phase transport of hydrogen and steam in the fuel cracks. The fuel oxidation model is further coupled to a heat conduction model to account for the feedback effect of a reduced thermal conductivity in the hyper-stoichiometric fuel. A numerical solution has been developed using a finite-element technique with the FEMLAB software package. The model has been compared to available data from several in-reactor X-2 loop experiments with defective fuel conducted at the Chalk River Laboratories. The model has also been benchmarked against an O/U profile measurement for a spent defective fuel element discharged from a commercial reactor.

  20. Condition assessment and determination of methods for evaluating corrosion damage in piles encapsulated in protective jackets on the Hampton Roads Bridge-Tunnel.

    DOT National Transportation Integrated Search

    2010-04-01

    The primary purpose of this study was to assess the condition of piles that had been encapsulated in fiberglass and mortar jackets on four bridges that are part of the Hampton Roads Bridge-Tunnel (HRBT). Since these four bridges contain a total of ap...

  1. Nuclear fuel pin scanner

    DOEpatents

    Bramblett, Richard L.; Preskitt, Charles A.

    1987-03-03

    Systems and methods for inspection of nuclear fuel pins to determine fiss loading and uniformity. The system includes infeed mechanisms which stockpile, identify and install nuclear fuel pins into an irradiator. The irradiator provides extended activation times using an approximately cylindrical arrangement of numerous fuel pins. The fuel pins can be arranged in a magazine which is rotated about a longitudinal axis of rotation. A source of activating radiation is positioned equidistant from the fuel pins along the longitudinal axis of rotation. The source of activating radiation is preferably oscillated along the axis to uniformly activate the fuel pins. A detector is provided downstream of the irradiator. The detector uses a plurality of detector elements arranged in an axial array. Each detector element inspects a segment of the fuel pin. The activated fuel pin being inspected in the detector is oscillated repeatedly over a distance equal to the spacing between adjacent detector elements, thereby multiplying the effective time available for detecting radiation emissions from the activated fuel pin.

  2. Protected Nuclear Fuel Element

    DOEpatents

    Kittel, J. H.; Schumar, J. F.

    1962-12-01

    A stainless steel-clad actinide metal fuel rod for use in fast reactors is reported. In order to prevert cladding failures due to alloy formation between the actinide metal and the stainless steel, a mesh-like sleeve of expanded metal is interposed between them, the sleeve metal being of niobium, tantalum, molybdenum, tungsten, zirconium, or vanadium. Liquid alkali metal is added as a heat transfer agent. (AEC)

  3. A numerical investigation of the influence of radiation and moisture content on pyrolysis and ignition of a leaf-like fuel element

    Treesearch

    B.L. Yashwanth; B. Shotorban; S. Mahalingam; C.W. Lautenberger; David Weise

    2016-01-01

    The effects of thermal radiation and moisture content on the pyrolysis and gas phase ignition of a solid fuel element containing high moisture content were investigated using the coupled Gpyro3D/FDS models. The solid fuel has dimensions of a typical Arctostaphylos glandulosa leaf which is modeled as thin cellulose subjected to radiative heating on...

  4. Discrimination of bullet types using analysis of lead isotopes deposited in gunshot entry wounds.

    PubMed

    Wunnapuk, Klintean; Minami, Takeshi; Durongkadech, Piya; Tohno, Setsuko; Ruangyuttikarn, Werawan; Moriwake, Yumi; Vichairat, Karnda; Sribanditmongkol, Pongruk; Tohno, Yoshiyuki

    2009-01-01

    In order to discriminate bullet types used in firearms, of which the victims died, the authors investigated lead isotope ratios in gunshot entry wounds from nine lead (unjacketed) bullets, 15 semi-jacketed bullets, and 14 full-jacketed bullets by inductively coupled plasma-mass spectrometry. It was found that the lead isotope ratio of 207/206 in gunshot entry wounds was the highest with lead bullets, and it decreased in order from full-jacketed to semi-jacketed bullets. Lead isotope ratios of 208/206 or 208/207 to 207/206 at the gunshot entry wound were able to discriminate semi-jacketed bullets from lead and full-jacketed ones, but it was difficult to discriminate between lead and full-jacketed bullets. However, a combination of element and lead isotope ratio analyses in gunshot entry wounds enabled discrimination between lead, semi-jacketed, and full-jacketed bullets.

  5. Grooved Fuel Rings for Nuclear Thermal Rocket Engines

    NASA Technical Reports Server (NTRS)

    Emrich, William

    2009-01-01

    An alternative design concept for nuclear thermal rocket engines for interplanetary spacecraft calls for the use of grooved-ring fuel elements. Beyond spacecraft rocket engines, this concept also has potential for the design of terrestrial and spacecraft nuclear electric-power plants. The grooved ring fuel design attempts to retain the best features of the particle bed fuel element while eliminating most of its design deficiencies. In the grooved ring design, the hydrogen propellant enters the fuel element in a manner similar to that of the Particle Bed Reactor (PBR) fuel element.

  6. Overview of past and current activities on fuels for fast reactors at the Institute for Transuranium Elements

    NASA Astrophysics Data System (ADS)

    Fernandez, A.; McGinley, J.; Somers, J.; Walter, M.

    2009-07-01

    Nuclear energy has the potential to provide a secure and sustainable electricity supply at a competitive price and to make a significant contribution to the reduction of greenhouse gas emissions. The renewal of interest in fast neutron spectra reactors to meet more ambitious sustainable development criteria (i.e., resource maximisation and waste minimisation), opens a favourable framework for R&D activities in this area. The Institute for Transuranium Elements has extensive experience in the fabrication, characterization and irradiation testing (Phénix, Dounreay, Rapsodie) of fast reactor fuels, in oxide, nitride and carbide forms. An overview of these past and current activities on fast reactor fuels is presented.

  7. Evaluation of Erosion of the Dummy “EE” Plate 19 in YA Type ATR Fuel Element During Reactor PALM Cycles

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Brower, Jeffrey O.; Glazoff, Michael V.; Eiden, Thomas J.

    Advanced Test Reactor (ATR) Cycle 153B-1 was a 14-day, high-power, powered axial locator mechanism (PALM) operating cycle that completed on April 12, 2013. Cycle 153B-1 was a typical operating cycle for the ATR, and did not result in any unusual plant transients. ATR was started up and shut down as scheduled. The PALM drive physically moves the selected experiments into and out of the core to simulate reactor startup and heat up, and shutdown and cooldown transients, while the reactor remains in steady-state conditions. However, after the cycle was over, when the fuel elements were removed from the core andmore » inspected, several thousand flow-assisted erosion pits and “horseshoeing” defects were readily observed on the surface of the several YA-type fuel elements (these are aluminum “dummy” plates that contain no fuel). In order to understand these erosion phenomena, a thermal-hydraulic model of coolant channel 20 on a YA-M fuel element was generated. The boundaries of the model were the aluminum EE plate of a YA-M fuel element and a beryllium reflector block with 13 horizontal saw cuts which represented regions of zero flow. The heat generated in fuel plates 1 through 18 was modeled to be passing through the aluminum EE plate. The coolant channel 20 width was set at 0.058 in. (58 mils). It was established that the horizontal saw cuts had a significant effect on the temperature of the coolant. The flow, which was expected to vary linearly with gradual heating of the coolant as it passed through the channel, was extremely turbulent. The temperature rise, which was expected to be a smooth “S” curve, was represented by a series temperature rise “humps,” which occurred at each horizontal saw cut in the beryllium reflector block. Each of the 13 saw cuts had a chamfered edge which resulted in the coolant flow being re-directed as a jet across the coolant channel into the surface of the EE plate, which explained the temperature rise and the observed

  8. A Comparison of Materials Issues for Cermet and Graphite-Based NTP Fuels

    NASA Technical Reports Server (NTRS)

    Stewart, Mark E.; Schnitzler, Bruce G.

    2013-01-01

    This paper compares material issues for cermet and graphite fuel elements. In particular, two issues in NTP fuel element performance are considered here: ductile to brittle transition in relation to crack propagation, and orificing individual coolant channels in fuel elements. Their relevance to fuel element performance is supported by considering material properties, experimental data, and results from multidisciplinary fluid/thermal/structural simulations. Ductile to brittle transition results in a fuel element region prone to brittle fracture under stress, while outside this region, stresses lead to deformation and resilience under stress. Poor coolant distribution between fuel element channels can increase stresses in certain channels. NERVA fuel element experimental results are consistent with this interpretation. An understanding of these mechanisms will help interpret fuel element testing results.

  9. Trace element partitioning in ashes from boilers firing pure wood or mixtures of solid waste with respect to fuel composition, chlorine content and temperature

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Saqib, Naeem, E-mail: naeem.saqib@oru.se; Bäckström, Mattias, E-mail: mattias.backstrom@oru.se

    Highlights: • Different solids waste incineration is discussed in grate fired and fluidized bed boilers. • We explained waste composition, temperature and chlorine effects on metal partitioning. • Excessive chlorine content can change oxide to chloride equilibrium partitioning the trace elements in fly ash. • Volatility increases with temperature due to increase in vapor pressure of metals and compounds. • In Fluidized bed boiler, most metals find themselves in fly ash, especially for wood incineration. - Abstract: Trace element partitioning in solid waste (household waste, industrial waste, waste wood chips and waste mixtures) incineration residues was investigated. Samples of flymore » ash and bottom ash were collected from six incineration facilities across Sweden including two grate fired and four fluidized bed incinerators, to have a variation in the input fuel composition (from pure biofuel to mixture of waste) and different temperature boiler conditions. As trace element concentrations in the input waste at the same facilities have already been analyzed, the present study focuses on the concentration of trace elements in the waste fuel, their distribution in the incineration residues with respect to chlorine content of waste and combustion temperature. Results indicate that Zn, Cu and Pb are dominating trace elements in the waste fuel. Highly volatile elements mercury and cadmium are mainly found in fly ash in all cases; 2/3 of lead also end up in fly ash while Zn, As and Sb show a large variation in distribution with most of them residing in the fly ash. Lithophilic elements such as copper and chromium are mainly found in bottom ash from grate fired facilities while partition mostly into fly ash from fluidized bed incinerators, especially for plants fuelled by waste wood or ordinary wood chips. There is no specific correlation between input concentration of an element in the waste fuel and fraction partitioned to fly ash. Temperature and

  10. Fuel handling apparatus for a nuclear reactor

    DOEpatents

    Hawke, Basil C.

    1987-01-01

    Fuel handling apparatus for transporting fuel elements into and out of a nuclear reactor and transporting them within the reactor vessel extends through a penetration in the side of the reactor vessel. A lateral transport device carries the fuel elements laterally within the vessel and through the opening in the side of the vessel, and a reversible lifting device raises and lowers the fuel elements. In the preferred embodiment, the lifting device is supported by a pair of pivot arms.

  11. Incorporation mechanisms of actinide elements into the structures of U 6+ phases formed during the oxidation of spent nuclear fuel

    NASA Astrophysics Data System (ADS)

    Burns, Peter C.; Ewing, Rodney C.; Miller, Mark L.

    1997-05-01

    Uranyl oxide hydrate and uranyl silicate phases will form due to the corrosion and alteration of spent nuclear fuel under oxidizing conditions in silica-bearing solution. The actinide elements in the spent fuel may be incorporated into the structures of these secondary U6+ phases during the long-term corrosion of the UO 2 in spent fuel. The incorporation of actinide elements into the crystal structures of the alteration products may decrease actinide mobility. The crystal chemistry of the various oxidation states of the actinide elements of environmental concern is examined to identify possible incorporation mechanisms. The substitutions Pu 6+U 6+ and (Pu 5+, Np 5+)U 6+ should readily occur in many U 6+ structures, although structural modification may be required to satisfy local bond-valence requirements. Crystal-chemical characteristics of the U 6+ phases indicate that An 4+ (An: actinide)U 6+ substitution is likely to occur in the sheets of uranyl polyhedra that occur in the structures of the minerals schoepite, [(UO 2) 8O 2(OH) 12](H 2O) 12, ianthinite, [U 24+ (UO 2) 4O 6(OH) 4(H 2O) 4](H 2O) 5, becquerelite, Ca[(UO 2) 3O 2(OH) 3] 2(H 2O) 8, compreignacite, K 2[(UO 2) 3O 2(OH) 3] 2(H 2O) 8, α-uranophane, Ca[(UO 2)(SiO 3OH)] 2(H 2O) 5, and boltwoodite, K(H 2O)[(UO 2)(SiO 4)], all of which are likely to form due to the oxidation and alteration of the UO 2 in spent fuel. The incorporation of An 3+ into the sheets of the structures of α-uranophane and boltwoodite, as well as interlayer sites of various uranyl phases, may occur.

  12. The influence of curvature configuration on the characteristic of alcohol gel insertion jacket of polymer optical fiber liquid level sensor

    NASA Astrophysics Data System (ADS)

    Arumnika, N.; Kuswanto, H.

    2018-04-01

    This study aimed to determine the effect of curvature configuration to sensitivities and linearities of Polymer Optical Fiber (POF) water level sensor. POF type SH-4001-1.3 has been used in this study. The jacket of POF of 20 cm was removed. Transparent piped inserted by alcohol gel has been used to replace the jacket. This is head of a sensor. The head of a sensor is curved with variations of the specified path length, peel length, the width of curvature, the height of curvature and waveform. Configuration A (20 cm, 34 cm, 6 cm, 2 cm, 1 wave), configuration B (20 cm, 34 cm, 8 cm, 2 cm, 1 wave), configuration C (20 cm, 34 cm, 9 cm, 2 cm, ½ wave), configuration D (20 cm, 34 cm, 10 cm, 2 cm, ½ wave). The head of a sensor inserted into the water tank. The light source inserted to one end POF is a He-Ne laser light with a power of 5 mW and a wavelength of 632.8 nm. Power output at the other end received by the Optical Power Meter (OPM). The curvature configuration the head sensor of POF affects the output. Configuration A has good sensitivity, however good linearity given by configuration.

  13. Fuel dissipater for pressurized fuel cell generators

    DOEpatents

    Basel, Richard A.; King, John E.

    2003-11-04

    An apparatus and method are disclosed for eliminating the chemical energy of fuel remaining in a pressurized fuel cell generator (10) when the electrical power output of the fuel cell generator is terminated during transient operation, such as a shutdown; where, two electrically resistive elements (two of 28, 53, 54, 55) at least one of which is connected in parallel, in association with contactors (26, 57, 58, 59), a multi-point settable sensor relay (23) and a circuit breaker (24), are automatically connected across the fuel cell generator terminals (21, 22) at two or more contact points, in order to draw current, thereby depleting the fuel inventory in the generator.

  14. Spent nuclear fuel assembly inspection using neutron computed tomography

    NASA Astrophysics Data System (ADS)

    Pope, Chad Lee

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

  15. Low exchange element for nuclear reactor

    DOEpatents

    Brogli, Rudolf H.; Shamasunder, Bangalore I.; Seth, Shivaji S.

    1985-01-01

    A flow exchange element is presented which lowers temperature gradients in fuel elements and reduces maximum local temperature within high temperature gas-cooled reactors. The flow exchange element is inserted within a column of fuel elements where it serves to redirect coolant flow. Coolant which has been flowing in a hotter region of the column is redirected to a cooler region, and coolant which has been flowing in the cooler region of the column is redirected to the hotter region. The safety, efficiency, and longevity of the high temperature gas-cooled reactor is thereby enhanced.

  16. Experimental detailed power distribution in a fast spectrum thermionic reactor fuel element at the core/BeO reflector interface region

    NASA Technical Reports Server (NTRS)

    Klann, P. G.; Lantz, E.

    1973-01-01

    A zero-power critical assembly was designed, constructed, and operated for the prupose of conducting a series of benchmark experiments dealing with the physics characteristics of a UN-fueled, Li-7-cooled, Mo-reflected, drum-controlled compact fast reactor for use with a space-power conversion system. The critical assembly was modified to simulate a fast spectrum advanced thermionics reactor by: (1) using BeO as a reflector in place of some of the existing molybdenum, (2) substituting Nb-1Zr tubing for some of the existing Ta tubing, and (3) inserting four full-scale mockups of thermionic type fuel elements near the core and BeO reflector boundary. These mockups were surrounded with a buffer zone having the equivalent thermionic core composition. In addition to measuring the critical mass of this thermionic configuration, a detailed power distribution in one of the thermionic element stages in the mixed spectrum region was measured. A power peak to average ratio of two was observed for this fuel stage at the midplane of the core and adjacent to the reflector. Also, the power on the outer surface adjacent to the BeO was slightly more than a factor of two larger than the power on the inside surface of a 5.08 cm (2.0 in.) high annular fuel segment with a 2.52 cm (0.993 in. ) o.d. and a 1.86 cm (0.731 in.) i.d.

  17. Improved nuclear fuel assembly grid spacer

    DOEpatents

    Marshall, John; Kaplan, Samuel

    1977-01-01

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

  18. Direct carbon fuel cell and stack designs

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Gorte, Raymond J.; Oh, Tae-Sik

    Disclosed are novel configurations of Direct Carbon Fuel Cells (DCFCs), which optionally comprise a liquid anode. The liquid anode comprises a molten salt/metal, preferably Sb, and a fuel, which has significant elemental carbon content (coal, bio-mass, etc.). The supply of fuel is continuously replenished in the anode. In addition, a stack configuration is suggested where combining a large number of planar or tubular fuel elements.

  19. IMPROVED TYPE OF FUEL ELEMENT

    DOEpatents

    Monson, H.O.

    1961-01-24

    A radiator-type fuel block assembly is described. It has a hexagonal body of neutron fissionable material having a plurality of longitudinal equal- spaced coolant channels therein aligned in rows parallel to each face of the hexagonal body. Each of these coolant channels is hexagonally shaped with the corners rounded and enlarged and the assembly has a maximum temperature isothermal line around each channel which is approximately straight and equidistant between adjacent channels.

  20. Effect of confining pressure due to external jacket of steel plate or shape memory alloy wire on bond behavior between concrete and steel reinforcing bars.

    PubMed

    Choi, Eunsoo; Kim, Dongkyun; Park, Kyoungsoo

    2014-12-01

    For external jackets of reinforced concrete columns, shape memory alloy (SMA) wires are easy to install, and they provide active and passive confining pressure; steel plates, on the other hand, only provide passive confining pressure, and their installation on concrete is not convenient because of the requirement of a special device. To investigate how SMA wires distinctly impact bond behavior compared with steel plates, this study conducted push-out bond tests of steel reinforcing bars embedded in concrete confined by SMA wires or steel plates. For this purpose, concrete cylinders were prepared with dimensions of 100 mm x 200 mm, and D-22 reinforcing bars were embedded at the center of the concrete cylinders. External jackets of 1.0 mm and 1.5 mm thickness steel plates were used to wrap the concrete cylinders. Additionally, NiTiNb SMA wire with a diameter of 1.0 mm was wound around the concrete cylinders. Slip of the reinforcing bars due to pushing force was measured by using a displacement transducer, while the circumferential deformation of specimens was obtained by using an extensometer. The circumferential deformation was used to calculate the circumferential strains of the specimens. This study assessed the radial confining pressure due to the external jackets on the reinforcing bars at bond strength from bond stress-slip curves and bond stress-circumferential strain curves. Then, the effects of the radial confining pressure on the bond behavior of concrete are investigated, and an equation is suggested to estimate bond strength using the radial confining pressure. Finally, this study focused on how active confining pressure due to recovery stress of the SMA wires influences bond behavior.

  1. Lightweight Tanks for Storing Liquefied Natural Gas

    NASA Technical Reports Server (NTRS)

    DeLay, Tom

    2008-01-01

    Single-walled, jacketed aluminum tanks have been conceived for storing liquefied natural gas (LNG) in LNG-fueled motor vehicles. Heretofore, doublewall steel tanks with vacuum between the inner and outer walls have been used for storing LNG. In comparison with the vacuum- insulated steel tanks, the jacketed aluminum tanks weigh less and can be manufactured at lower cost. Costs of using the jacketed aluminum tanks are further reduced in that there is no need for the vacuum pumps heretofore needed to maintain vacuum in the vacuum-insulated tanks.

  2. 10 CFR Appendix O to Part 110 - Illustrative List of Fuel Element Fabrication Plant Equipment and Components Under NRC's Export...

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... performance and safety during reactor operation. Also, in all cases precise control of processes, procedures... elements include equipment that: (1) Normally comes in direct contact with, or directly processes or... pellets; (2) Automatic welding machines especially designed or prepared for welding end caps onto the fuel...

  3. Support grid for fuel elements in a nuclear reactor

    DOEpatents

    Finch, Lester M.

    1977-01-01

    A support grid is provided for holding nuclear fuel rods in a rectangular array. Intersecting sheet metal strips are interconnected using opposing slots in the strips to form a rectangular cellular grid structure for engaging the sides of a multiplicity of fuel rods. Spring and dimple supports for engaging fuel and guide rods extending through each cell in the support grid are formed in the metal strips with the springs thus formed being characterized by nonlinear spring rates.

  4. Reduced size fuel cell for portable applications

    NASA Technical Reports Server (NTRS)

    Narayanan, Sekharipuram R. (Inventor); Valdez, Thomas I. (Inventor); Clara, Filiberto (Inventor); Frank, Harvey A. (Inventor)

    2004-01-01

    A flat pack type fuel cell includes a plurality of membrane electrode assemblies. Each membrane electrode assembly is formed of an anode, an electrolyte, and an cathode with appropriate catalysts thereon. The anode is directly into contact with fuel via a wicking element. The fuel reservoir may extend along the same axis as the membrane electrode assemblies, so that fuel can be applied to each of the anodes. Each of the fuel cell elements is interconnected together to provide the voltage outputs in series.

  5. Coolant mass flow equalizer for nuclear fuel

    DOEpatents

    Betten, Paul R.

    1978-01-01

    The coolant mass flow distribution in a liquid metal cooled reactor is enhanced by restricting flow in sub-channels defined in part by the peripheral fuel elements of a fuel assembly. This flow restriction, which results in more coolant flow in interior sub-channels, is achieved through the use of a corrugated liner positioned between the bundle of fuel elements and the inner wall of the fuel assembly coolant duct. The corrugated liner is expandable to accommodate irradiation induced growth of fuel assembly components.

  6. Statistical study of EBR-II fuel elements manufactured by the cold line at Argonne-West and by Atomics International

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Harkness, A. L.

    1977-09-01

    Nine elements from each batch of fuel elements manufactured for the EBR-II reactor have been analyzed for /sup 235/U content by NDA methods. These values, together with those of the manufacturer, are used to estimate the product variance and the variances of the two measuring methods. These variances are compared with the variances computed from the stipulations of the contract. A method is derived for resolving the several variances into their within-batch and between-batch components. Some of these variance components have also been estimated by independent and more familiar conventional methods for comparison.

  7. Transition Core Properties during Conversion of the NBSR from HEU to LEU Fuel

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Hanson, A. L.; Diamond, D.

    2013-10-31

    The transition of the NBSR from HEU to LEU fuel is challenging due to reactivity constraints and the need to maintain an uninterrupted science program, the mission of the NBSR. The transition cannot occur with a full change of HEU to LEU fuel elements since the excess reactivity would be large enough that the NBSR would violate the technical specification for shutdown margin. Manufacturing LEU fuel elements to represent irradiated fuel elements would be cost prohibitive since 26 one-of-a-kind fuel elements would need to be manufactured. For this report a gradual transition from the present HEU fuel to the proposedmore » LEU fuel was studied. The gradual change approach would follow the present fuel management scheme and replace four HEU fuel elements with four LEU fuel elements each cycle. This manuscript reports the results of a series of calculations to predict the neutronic characteristics and how the neutronics will change during the transition from HEU to LEU in the NBSR.« less

  8. Fuel cell generator energy dissipator

    DOEpatents

    Veyo, Stephen Emery; Dederer, Jeffrey Todd; Gordon, John Thomas; Shockling, Larry Anthony

    2000-01-01

    An apparatus and method are disclosed for eliminating the chemical energy of fuel remaining in a fuel cell generator when the electrical power output of the fuel cell generator is terminated. During a generator shut down condition, electrically resistive elements are automatically connected across the fuel cell generator terminals in order to draw current, thereby depleting the fuel

  9. Initial Operation of the Nuclear Thermal Rocket Element Environmental Simulator

    NASA Technical Reports Server (NTRS)

    Emrich, William J., Jr.; Pearson, J. Boise; Schoenfeld, Michael P.

    2015-01-01

    The Nuclear Thermal Rocket Element Environmental Simulator (NTREES) facility is designed to perform realistic non-nuclear testing of nuclear thermal rocket (NTR) fuel elements and fuel materials. Although the NTREES facility cannot mimic the neutron and gamma environment of an operating NTR, it can simulate the thermal hydraulic environment within an NTR fuel element to provide critical information on material performance and compatibility. The NTREES facility has recently been upgraded such that the power capabilities of the facility have been increased significantly. At its present 1.2 MW power level, more prototypical fuel element temperatures nay now be reached. The new 1.2 MW induction heater consists of three physical units consisting of a transformer, rectifier, and inverter. This multiunit arrangement facilitated increasing the flexibility of the induction heater by more easily allowing variable frequency operation. Frequency ranges between 20 and 60 kHz can accommodated in the new induction heater allowing more representative power distributions to be generated within the test elements. The water cooling system was also upgraded to so as to be capable of removing 100% of the heat generated during testing In this new higher power configuration, NTREES will be capable of testing fuel elements and fuel materials at near-prototypic power densities. As checkout testing progressed and as higher power levels were achieved, several design deficiencies were discovered and fixed. Most of these design deficiencies were related to stray RF energy causing various components to encounter unexpected heating. Copper shielding around these components largely eliminated these problems. Other problems encountered involved unexpected movement in the coil due to electromagnetic forces and electrical arcing between the coil and a dummy test article. The coil movement and arcing which were encountered during the checkout testing effectively destroyed the induction coil in use at

  10. H∞ memory feedback control with input limitation minimization for offshore jacket platform stabilization

    NASA Astrophysics Data System (ADS)

    Yang, Jia Sheng

    2018-06-01

    In this paper, we investigate a H∞ memory controller with input limitation minimization (HMCIM) for offshore jacket platforms stabilization. The main objective of this study is to reduce the control consumption as well as protect the actuator when satisfying the requirement of the system performance. First, we introduce a dynamic model of offshore platform with low order main modes based on mode reduction method in numerical analysis. Then, based on H∞ control theory and matrix inequality techniques, we develop a novel H∞ memory controller with input limitation. Furthermore, a non-convex optimization model to minimize input energy consumption is proposed. Since it is difficult to solve this non-convex optimization model by optimization algorithm, we use a relaxation method with matrix operations to transform this non-convex optimization model to be a convex optimization model. Thus, it could be solved by a standard convex optimization solver in MATLAB or CPLEX. Finally, several numerical examples are given to validate the proposed models and methods.

  11. Hydrogen Fuel System Design Trades for High-Altitude Long-Endurance Remotely- Operated Aircraft

    NASA Technical Reports Server (NTRS)

    Millis, Marc G.; Tornabene, Robert T.; Jurns, John M.; Guynn, Mark D.; Tomsik, Thomas M.; VanOverbeke, Thomas J.

    2009-01-01

    Preliminary design trades are presented for liquid hydrogen fuel systems for remotely-operated, high-altitude aircraft that accommodate three different propulsion options: internal combustion engines, and electric motors powered by either polymer electrolyte membrane fuel cells or solid oxide fuel cells. Mission goal is sustained cruise at 60,000 ft altitude, with duration-aloft a key parameter. The subject aircraft specifies an engine power of 143 to 148 hp, gross liftoff weight of 9270 to 9450 lb, payload of 440 lb, and a hydrogen fuel capacity of 2650 to 2755 lb stored in two spherical tanks (8.5 ft inside diameter), each with a dry mass goal of 316 lb. Hydrogen schematics for all three propulsion options are provided. Each employs vacuum-jacketed tanks with multilayer insulation, augmented with a helium pressurant system, and using electric motor driven hydrogen pumps. The most significant schematic differences involve the heat exchangers and hydrogen reclamation equipment. Heat balances indicate that mission durations of 10 to 16 days appear achievable. The dry mass for the hydrogen system is estimated to be 1900 lb, including 645 lb for each tank. This tank mass is roughly twice that of the advanced tanks assumed in the initial conceptual vehicle. Control strategies are not addressed, nor are procedures for filling and draining the tanks.

  12. Design of a fuel element for a lead-cooled fast reactor

    NASA Astrophysics Data System (ADS)

    Sobolev, V.; Malambu, E.; Abderrahim, H. Aït

    2009-03-01

    The options of a lead-cooled fast reactor (LFR) of the fourth generation (GEN-IV) reactor with the electric power of 600 MW are investigated in the ELSY Project. The fuel selection, design and optimization are important steps of the project. Three types of fuel are considered as candidates: highly enriched Pu-U mixed oxide (MOX) fuel for the first core, the MOX containing between 2.5% and 5.0% of the minor actinides (MA) for next core and Pu-U-MA nitride fuel as an advanced option. Reference fuel rods with claddings made of T91 ferrite-martensitic steel and two alternative fuel assembly designs (one uses a closed hexagonal wrapper and the other is an open square variant without wrapper) have been assessed. This study focuses on the core variant with the closed hexagonal fuel assemblies. Based on the neutronic parameters provided by Monte-Carlo modeling with MCNP5 and ALEPH codes, simulations have been carried out to assess the long-term thermal-mechanical behaviour of the hottest fuel rods. A modified version of the fuel performance code FEMAXI-SCK-1, adapted for fast neutron spectrum, new fuels, cladding materials and coolant, was utilized for these calculations. The obtained results show that the fuel rods can withstand more than four effective full power years under the normal operation conditions without pellet-cladding mechanical interaction (PCMI). In a variant with solid fuel pellets, a mild PCMI can appear during the fifth year, however, it remains at an acceptable level up to the end of operation when the peak fuel pellet burnup ∼80 MW d kg-1 of heavy metal (HM) and the maximum clad damage of about 82 displacements per atom (dpa) are reached. Annular pellets permit to delay PCMI for about 1 year. Based on the results of this simulation, further steps are envisioned for the optimization of the fuel rod design, aiming at achieving the fuel burnup of 100 MW d kg-1 of HM.

  13. Accuracy of trace element determinations in alternate fuels

    NASA Technical Reports Server (NTRS)

    Greenbauer-Seng, L. A.

    1980-01-01

    NASA-Lewis Research Center's work on accurate measurement of trace level of metals in various fuels is presented. The differences between laboratories and between analytical techniques especially for concentrations below 10 ppm, are discussed, detailing the Atomic Absorption Spectrometry (AAS) and DC Arc Emission Spectrometry (dc arc) techniques used by NASA-Lewis. Also presented is the design of an Interlaboratory Study which is considering the following factors: laboratory, analytical technique, fuel type, concentration and ashing additive.

  14. Toxicity of irradiated advanced heavy water reactor fuels.

    PubMed

    Priest, N D; Richardson, R B; Edwards, G W R

    2013-02-01

    The good neutron economy and online refueling capability of the CANDU® heavy water moderated reactor (HWR) enable it to use many different fuels such as low enriched uranium (LEU), plutonium, or thorium, in addition to its traditional natural uranium (NU) fuel. The toxicity and radiological protection methods for these proposed fuels, unlike those for NU, are not well established. This study uses software to compare the fuel composition and toxicity of irradiated NU fuel against those of two irradiated advanced HWR fuel bundles as a function of post-irradiation time. The first bundle investigated is a CANFLEX® low void reactor fuel (LVRF), of which only the dysprosium-poisoned central element, and not the outer 42 LEU elements, is specifically analyzed. The second bundle investigated is a heterogeneous high-burnup (LEU,Th)O(2) fuelled bundle, whose two components (LEU in the outer 35 elements and thorium in the central eight elements) are analyzed separately. The LVRF central element was estimated to have a much lower toxicity than that of NU at all times after shutdown. Both the high burnup LEU and the thorium fuel had similar toxicity to NU at shutdown, but due to the creation of such inhalation hazards as (238)Pu, (240)Pu, (242)Am, (242)Cm, and (244)Cm (in high burnup LEU), and (232)U and (228)Th (in irradiated thorium), the toxicity of these fuels was almost double that of irradiated NU after 2,700 d of cooling. New urine bioassay methods for higher actinoids and the analysis of thorium in fecal samples are recommended to assess the internal dose from these two fuels.

  15. History of fast reactor fuel development

    NASA Astrophysics Data System (ADS)

    Kittel, J. H.; Frost, B. R. T.; Mustelier, J. P.; Bagley, K. Q.; Crittenden, G. C.; Van Dievoet, J.

    1993-09-01

    The first fast breeder reactors, constructed in the 1945-1960 time period, used metallic fuels composed of uranium, plutonium, or their alloys. They were chosen because most existing reactor operating experience had been obtained on metallic fuels and because they provided the highest breeding ratios. Difficulties in obtaining adequate dimensional stability in metallic fuel elements under conditions of high fuel burnup led in the 1960s to the virtual worldwide choice of ceramic fuels. Although ceramic fuels provide lower breeding performance, this objective is no longer an important consideration in most national programs. Mixed uranium and plutonium dioxide became the ceramic fuel that has received the widest use. The more advanced ceramic fuels, mixed uranium and plutonium carbides and nitrides, continue under development. More recently, metal fuel elements of improved design have joined ceramic fuels in achieving goal burnups of 15 to 20 percent. Low-swelling fuel cladding alloys have also been continuously developed to deal with the unexpected problem of void formation in stainless steels subjected to fast neutron irradiation, a phenomenon first observed in the 1960s.

  16. Basic elements of light water reactor fuel rod design. [FUELROD code

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Weisman, J.; Eckart, R.

    1981-06-01

    Basic design techniques and equations are presented to allow students to understand and perform preliminary fuel design for normal reactor conditions. Each of the important design considerations is presented and discussed in detail. These include the interaction between fuel pellets and cladding and the changes in fuel and cladding that occur during the operating lifetime of the fuel. A simple, student-oriented, fuel rod design computer program, called FUELROD, is described. The FUELROD program models the in-pile pellet cladding interaction and allows a realistic exploration of the effect of various design parameters. By use of FUELROD, the student can gain anmore » appreciation of the fuel rod design process. 34 refs.« less

  17. Modeling 3D PCMI using the Extended Finite Element Method with higher order elements

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Jiang, W.; Spencer, Benjamin W.

    2017-03-31

    This report documents the recent development to enable XFEM to work with higher order elements. It also demonstrates the application of higher order (quadratic) elements to both 2D and 3D models of PCMI problems, where discrete fractures in the fuel are represented using XFEM. The modeling results demonstrate the ability of the higher order XFEM to accurately capture the effects of a crack on the response in the vicinity of the intersecting surfaces of cracked fuel and cladding, as well as represent smooth responses in the regions away from the crack.

  18. Summary of LCRE fuel element design including supporting experimental data

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    None, None

    Declassified 18 Sep 1973. The design basis of the LCRE fuel pin is presented. The fuel pin consists of a Cb-1 Zr alloy cladding tube 0.305 inch diameter, 0.015 inch wall thickness and 35.96 inches long. The active fuel section is 13.5 inches long, with top and bottom reflector rods each 6.9 inches long and with a 4 inch gas accumulation space at each end. The cladding is designed as a pressure vessel to contain the gases released from the fuel and end refiector materials, which results in an internal gas pressure buildup in the pins during reactor operation. (23more » referencea) (auth)« less

  19. Design Evolutuion of Hot Isotatic Press Cans for NTP Cermet Fuel Fabrication

    NASA Technical Reports Server (NTRS)

    Mireles, O. R.; Broadway, J.; Hickman, R.

    2014-01-01

    Nuclear Thermal Propulsion (NTP) is under consideration for potential use in deep space exploration missions due to desirable performance properties such as a high specific impulse (> 850 seconds). Tungsten (W)-60vol%UO2 cermet fuel elements are under development, with efforts emphasizing fabrication, performance testing and process optimization to meet NTP service life requirements [1]. Fuel elements incorporate design features that provide redundant protection from crack initiation, crack propagation potentially resulting in hot hydrogen (H2) reduction of UO2 kernels. Fuel erosion and fission product retention barriers include W coated UO2 fuel kernels, W clad internal flow channels and fuel element external W clad resulting in a fully encapsulated fuel element design as shown.

  20. Madness Decolonized?: Madness as Transnational Identity in Gail Hornstein's Agnes's Jacket.

    PubMed

    Miller, Gavin

    2017-02-13

    The US psychologist Gail Hornstein's monograph, Agnes's Jacket: A Psychologist's Search for the Meanings of Madness (2009), is an important intervention in the identity politics of the mad movement. Hornstein offers a resignified vision of mad identity that embroiders the central trope of an "anti-colonial" struggle to reclaim the experiential world "colonized" by psychiatry. A series of literal and figurative appeals makes recourse to the inner world and (corresponding) cultural world of the mad as well as to the ethno-symbolic cultural materials of dormant nationhood. This rhetoric is augmented by a model in which the mad comprise a diaspora without an origin, coalescing into a single transnational community. The mad are also depicted as persons displaced from their metaphorical homeland, the "inner" world "colonized" by the psychiatric regime. There are a number of difficulties with Hornstein's rhetoric, however. Her "ethnicity-and-rights" response to the oppression of the mad is symptomatic of Western parochialism, while her proposed transmutation of putative psychopathology from limit upon identity to parameter of successful identity is open to contestation. Moreover, unless one accepts Hornstein's porous vision of mad identity, her self-ascribed insider status in relation to the mad community may present a problematic "re-colonization" of mad experience.

  1. Development of variable-width ribbon heating elements for liquid-metal and gas-cooled fast breeder reactor fuel-pin simulators

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    McCulloch, R.W.; Post, D.W.; Lovell, R.T.

    1981-04-01

    Variable-width ribbon heating elements that provide a chopped-cosine variable heat flux profile have been fabricated for fuel pin simulators used in test loops by the Breeder Reactor Program Thermal-Hydraulic Out-of-Reactor Safety test facility and the Gas-Cooled Fast Breeder Reactor-Core Flow Test Loop. Thermal, mechanical, and electrical design considerations are used to derive an analytical expression that precisely describes ribbon contour in terms of the major fabrication parameters. These parameters are used to generate numerical control tapes that control ribbon cutting and winding machines. Infrared scanning techniques are developed to determine the optimum transient thermal profile of the coils and relatemore » this profile to that generated by the coils in completed fuel pin simulators.« less

  2. FUEL ASSAY REACTOR

    DOEpatents

    Spinrad, B.I.; Sandmeier, H.A.; Martens, F.H.

    1962-12-25

    A reactor having maximum sensitivity to perturbations is described comprising a core consisting of a horizontally disposed, rectangular, annular fuel zone containing enriched uranium dioxide dispersed in graphite, the concentration of uranium dioxide increasing from the outside to the inside of the fuel zone, an internal reflector of graphite containing an axial test opening disposed within the fuel zone, an external graphite reflector, means for changing the neutron spectrum in the test opening, and means for measuring perturbations in the neutron flux caused by the introduction of different fuel elements into the test opening. (AEC)

  3. Calculation of Heat-Bearing Agent’s Steady Flow in Fuel Bundle

    NASA Astrophysics Data System (ADS)

    Amosova, E. V.; Guba, G. G.

    2017-11-01

    This paper introduces the result of studying the heat exchange in the fuel bundle of the nuclear reactor’s fuel magazine. The article considers the fuel bundle of the infinite number of fuel elements, fuel elements are considered in the checkerboard fashion (at the tops of a regular triangle a fuel element is a plain round rod. The inhomogeneity of volume energy release in the rod forms the inhomogeneity of temperature and velocity fields, and pressure. Computational methods for studying hydrodynamics in magazines and cores with rod-shape fuel elements are based on a significant simplification of the problem: using basic (averaged) equations, isobaric section hypothesis, porous body model, etc. This could be explained by the complexity of math description of the three-dimensional fluid flow in the multi-connected area with the transfer coefficient anisotropy, curved boundaries and technical computation difficulties. Thus, calculative studying suggests itself as promising and important. There was developed a method for calculating the heat-mass exchange processes of inter-channel fuel element motions, which allows considering the contribution of natural convection to the heat-mass exchange based on the Navier-Stokes equations and Boussinesq approximation.

  4. Elemental balance of SRF production process: solid recovered fuel produced from municipal solid waste.

    PubMed

    Nasrullah, Muhammad; Vainikka, Pasi; Hannula, Janne; Hurme, Markku; Oinas, Pekka

    2016-01-01

    In the production of solid recovered fuel (SRF), certain waste components have excessive influence on the quality of product. The proportion of rubber, plastic (hard) and certain textiles was found to be critical as to the elemental quality of SRF. The mass flow of rubber, plastic (hard) and textiles (to certain extent, especially synthetic textile) components from input waste stream into the output streams of SRF production was found to play the decisive role in defining the elemental quality of SRF. This paper presents the mass flow of polluting and potentially toxic elements (PTEs) in SRF production. The SRF was produced from municipal solid waste (MSW) through mechanical treatment (MT). The results showed that of the total input chlorine content to process, 55% was found in the SRF and 30% in reject material. Of the total input arsenic content, 30% was found in the SRF and 45% in fine fraction. In case of cadmium, lead and mercury, of their total input content to the process, 62%, 38% and 30%, respectively, was found in the SRF. Among the components of MSW, rubber material was identified as potential source of chlorine, containing 8.0 wt.% of chlorine. Plastic (hard) and textile components contained 1.6 and 1.1. wt.% of chlorine, respectively. Plastic (hard) contained higher lead and cadmium content compared with other waste components, i.e. 500 mg kg(-1) and 9.0 mg kg(-1), respectively. © The Author(s) 2015.

  5. Evaluation of Corrosion of the Dummy “EE” Plate 19 in YA Type ATR Fuel Element During Reactor PALM Cycles

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Brower, Jeffrey Owen; Glazoff, Michael Vasily; Eiden, Thomas John

    Advanced Test Reactor (ATR) Cycle 153B-1 was a 14-day, high-power, powered axial locator mechanism (PALM) operating cycle that completed on April 12, 2013. Cycle 153B-1 was a typical operating cycle for the ATR and did not result in any unusual plant transients. ATR was started up and shut down as scheduled. The PALM drive physically moves the selected experiments into and out of the core to simulate reactor startup and heat up, and shutdown and cooldown transients, while the reactor remains in steady state conditions. However, after the cycle was over, several thousand of the flow-assisted corrosion pits and “horseshoeing”more » defects were readily observable on the surface of the several YA-type fuel elements (these are “dummy” plates that contain no fuel). In order understand these corrosion phenomena a thermal-hydraulic model of coolant channel 20 on a YA-M fuel element was generated. The boundaries of the model were the aluminum EE plate of a YA-M fuel element and a beryllium reflector block with 13 horizontal saw cuts which represented regions of zero flow. The heat generated in fuel plates 1 through 18 was modeled to be passing through the aluminum EE plate. The coolant channel 20 width was set at 0.058 in. (58 mils). It was established that the horizontal saw cuts had a significant effect on the temperature of the coolant. The flow, which was expected to vary linearly with gradual heating of the coolant as it passed through the channel, was extremely turbulent. The temperature rise, which was expected to be a smooth “S” curve, was represented by a series temperature rise “humps,” which occurred at each horizontal saw cut in the beryllium reflector block. Each of the 13 saw cuts had a chamfered edge which resulted in the coolant flow being re-directed as a jet across the coolant channel into the surface of the EE plate, which explained the temperature rise and the observed sscalloping and possibly pitting degradation on the YA-M fuel

  6. Current status of the development of high density LEU fuel for Russian research reactors

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Vatulin, A.; Dobrikova, I.; Suprun, V.

    2008-07-15

    One of the main directions of the Russian RERTR program is to develop U-Mo fuel and fuel elements/FA with this fuel. The development is carried out both for existing reactors, and for new advanced designs of reactors. Many organizations in Russia, i.e. 'TVEL', RDIPE, RIAR, IRM, NPCC participate in the work. Two fuels are under development: dispersion and monolithic U-Mo fuel, as well two types of FA to use the dispersion U-Mo fuel: with tubular type fuel elements and with pin type fuel elements. The first stage of works was successfully completed. This stage included out-pile, in-pile and post irradiationmore » examinations of U-Mo dispersion fuel in experimental tubular and pin fuel elements under parameters similar to operation conditions of Russian design pool-type research reactors. The results received both in Russia and abroad enabled to go on to the next stage of development which includes irradiation tests both of full-scale IRT pin-type and tube-type fuel assemblies with U-Mo dispersion fuel and of mini-fuel elements with modified U-Mo dispersion fuel and monolithic fuel. The paper gives a generalized review of the results of U-Mo fuel development accomplished by now. (author)« less

  7. Checking the possibility of controlling fuel element by X-ray computerized tomography

    NASA Astrophysics Data System (ADS)

    Trinh, V. B.; Zhong, Y.; Osipov, S. P.; Batranin, A. V.

    2017-08-01

    The article considers the possibility of checking fuel elements by X-ray computerized tomography. The checking tasks are based on the detection of particles of active material, evaluation of the heterogeneity of the distribution of uranium salts and the detection of clusters of uranium particles. First of all, scheme of scanning improve the performance and quality of the resulting three-dimensional images of the internal structure is determined. Further, the possibility of detecting clusters of uranium particles having the size of 1 mm3 and measuring the coordinates of clusters of uranium particles in the middle layer with the accuracy of within a voxel size (for the considered experiments of about 80 μm) is experimentally proved in the main part. The problem of estimating the heterogeneity of the distribution of the active material in the middle layer and the detection of particles of active material with a nominal diameter of 0.1 mm in the “blank” is solved.

  8. A Multi-Dimensional Heat Transfer Model of a Tie-Tube and Hexagonal Fuel Element for Nuclear Thermal Propulsion

    NASA Technical Reports Server (NTRS)

    Gomez, C. F.; Mireles, O. R.; Stewart, E.

    2016-01-01

    The Space Capable Cryogenic Thermal Engine (SCCTE) effort considers a nuclear thermal rocket design based around a Low-Enriched Uranium (LEU) design fission reactor. The reactor core is comprised of bundled hexagonal fuel elements that directly heat hydrogen for expansion in a thrust chamber and hexagonal tie-tubes that house zirconium hydride moderator mass for the purpose of thermalizing fast neutrons resulting from fission events. Created 3D steady state Hex fuel rod model with 1D flow channels. Hand Calculation were used to set up initial conditions for fluid flow. The Hex Fuel rod uses 1D flow paths to model the channels using empirical correlations for heat transfer in a pipe. Created a 2-D axisymmetric transient to steady state model using the CFD turbulent flow and Heat Transfer module in COMSOL. This model was developed to find and understand the hydrogen flow that might effect the thermal gradients axially and at the end of the tie tube where the flow turns and enters an annulus. The Hex fuel rod and Tie tube models were made based on requirements given to us by CSNR and the SCCTE team. The models helped simplify and understand the physics and assumptions. Using pipe correlations reduced the complexity of the 3-D fuel rod model and is numerically more stable and computationally more time-efficient compared to the CFD approach. The 2-D axisymmetric tie tube model can be used as a reference "Virtual test model" for comparing and improving 3-D Models.

  9. Ion chromatographic determination of sulfur in fuels

    NASA Technical Reports Server (NTRS)

    Mizisin, C. S.; Kuivinen, D. E.; Otterson, D. A.

    1978-01-01

    The sulfur content of fuels was determined using an ion chromatograph to measure the sulfate produced by a modified Parr bomb oxidation. Standard Reference Materials from the National Bureau of Standards, of approximately 0.2 + or - 0.004% sulfur, were analyzed resulting in a standard deviation no greater than 0.008. The ion chromatographic method can be applied to conventional fuels as well as shale-oil derived fuels. Other acid forming elements, such as fluorine, chlorine and nitrogen could be determined at the same time, provided that these elements have reached a suitable ionic state during the oxidation of the fuel.

  10. HOT CELL SYSTEM FOR DETERMINING FISSION GAS RETENTION IN METALLIC FUELS

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Sell, D. A.; Baily, C. E.; Malewitz, T. J.

    2016-09-01

    A system has been developed to perform measurements on irradiated, sodium bonded-metallic fuel elements to determine the amount of fission gas retained in the fuel material after release of the gas to the element plenum. During irradiation of metallic fuel elements, most of the fission gas developed is released from the fuel and captured in the gas plenums of the fuel elements. A significant amount of fission gas, however, remains captured in closed porosities which develop in the fuel during irradiation. Additionally, some gas is trapped in open porosity but sealed off from the plenum by frozen bond sodium aftermore » the element has cooled in the hot cell. The Retained fission Gas (RFG) system has been designed, tested and implemented to capture and measure the quantity of retained fission gas in characterized cut pieces of sodium bonded metallic fuel. Fuel pieces are loaded into the apparatus along with a prescribed amount of iron powder, which is used to create a relatively low melting, eutectic composition as the iron diffuses into the fuel. The apparatus is sealed, evacuated, and then heated to temperatures in excess of the eutectic melting point. Retained fission gas release is monitored by pressure transducers during the heating phase, thus monitoring for release of fission gas as first the bond sodium melts and then the fuel. A separate hot cell system is used to sample the gas in the apparatus and also characterize the volume of the apparatus thus permitting the calculation of the total fission gas release from the fuel element samples along with analysis of the gas composition.« less

  11. Etched poly(ether ether ketone) jacket stir bar with detachable dumbbell-shaped structure for stir bar sorptive extraction.

    PubMed

    Zhou, Wei; Wang, Chenlu; Wang, Xuemei; Chen, Zilin

    2018-06-08

    Development of stir bar sorptive extraction (SBSE) device with high stability and extraction efficiency is critical and challenging by date. In this work, etched poly(ether ether ketone) (PEEK) tube with high mechanical strength and large specific surface area was used as jacket for SBSE device. By etching with concentrated sulfuric acid, the smooth outer surface of PEEK become porous with plenty of micro holes, which was beneficial for coating of sorbents and significantly improved the extraction performance. After functionalized by bio-polydopamine method, strong hydrophobic p-naphtholbenzein molecular was immobilized onto the chemical resistant PEEK surface (PNB@E-PEEK) as stationary phase. We also firstly developed a simple detachable dumbbell-shaped structure for improving the workability of PEEK jacket stir bar. The dumbbell-shaped construction can eliminate the friction between stir bar and container, and the design of detachable structure make elution can be accomplished easier with small amount of organic solvent. It was interesting that the developed detachable dumbbell-shaped PNB@E-PEEK stir bar showed exceptional stability and extraction efficiency for SBSE enrichment of multiple analytes including several Sudan dyes, triazines, polycyclic aromatic hydrocarbons (PAHs), alkaloids and flavonoid. By coupling with high performance liquid chromatography-ultraviolet detection (HPLC-UV), PNB@E-PEEK stir bar based SBSE-HPLC-UV method was applied for the analysis of common Sudan dye pollutants. The method showed low limits of detection (0.02-0.03 ng/mL), good linearity (R 2  ≥ 0.9979) and good reproducibility (relative standard deviation ≤ 7.96%). It has been successfully applied to determine three dye pollutants in tap and lake water. Copyright © 2018 Elsevier B.V. All rights reserved.

  12. Sensitivity Analysis of Fuel Centerline Temperatures in SuperCritical Water-cooled Reactors (SCWRs)

    NASA Astrophysics Data System (ADS)

    Abdalla, Ayman

    SuperCritical Water-cooled Reactors (SCWRs) are one of the six nuclear-reactor concepts currently being developed under the Generation-IV International Forum (GIF). A main advantage of SCW Nuclear Power Plants (NPPs) is that they offer higher thermal efficiencies compared to those of current conventional NPPs. Unlike today's conventional NPPs, which have thermal efficiencies between 30 - 35%, SCW NPPs will have thermal efficiencies within a range of 45 - 50%, owing to high operating temperatures and pressures (i.e., coolant temperatures as high as 625°C at 25 MPa pressure). The use of current fuel bundles with UO2 fuel at the high operating parameters of SCWRs may cause high fuel centerline temperatures, which could lead to fuel failure and fission gas release. Studies have shown that when the Variant-20 (43-element) fuel bundle was examined at SCW conditions, the fuel centerline temperature industry limit of 1850°C for UO2 and the sheath temperature design limit of 850°C might be exceeded. Therefore, new fuel-bundle designs, which comply with the design requirements, are required for future use in SCWRs. The main objective of this study to conduct a sensitivity analysis in order to identify the main factors that leads to fuel centerline temperature reduction. Therefore, a 54-element fuel bundle with smaller diameter of fuel elements compared to that of the 43-element bundle was designed and various nuclear fuels are examined for future use in a generic Pressure Tube (PT) SCWR. The 54-element bundle consists of 53 heated fuel elements with an outer diameter of 9.5 mm and one central unheated element of 20-mm outer diameter which contains burnable poison. The 54-element fuel bundle has an outer diameter of 103.45 mm, which is the same as the outer diameter of the 43-element fuel bundle. After developing the 54-element fuel bundle, one-dimensional heat-transfer analysis was conducted using MATLAB and NIST REFPROP programs. As a result, the Heat Transfer

  13. Flow tests of a single fuel element coolant channel for a compact fast reactor for space power

    NASA Technical Reports Server (NTRS)

    Springborn, R. H.

    1971-01-01

    Water flow tests were conducted on a single-fuel-element cooling channel for a nuclear concept to be used for space power. The tests established a method for measuring coolant flow rate which is applicable to water flow testing of a complete mockup of the reference reactor. The inlet plenum-to-outlet plenum pressure drop, which approximates the overall core pressure drop, was measured and correlated with flow rate. This information can be used for reactor coolant flow and heat transfer calculations. An analytical study of the flow characteristics was also conducted.

  14. ELM - A SIMPLE TOOL FOR THERMAL-HYDRAULIC ANALYSIS OF SOLID-CORE NUCLEAR ROCKET FUEL ELEMENTS

    NASA Technical Reports Server (NTRS)

    Walton, J. T.

    1994-01-01

    ELM is a simple computational tool for modeling the steady-state thermal-hydraulics of propellant flow through fuel element coolant channels in nuclear thermal rockets. Written for the nuclear propulsion project of the Space Exploration Initiative, ELM evaluates the various heat transfer coefficient and friction factor correlations available for turbulent pipe flow with heat addition. In the past, these correlations were found in different reactor analysis codes, but now comparisons are possible within one program. The logic of ELM is based on the one-dimensional conservation of energy in combination with Newton's Law of Cooling to determine the bulk flow temperature and the wall temperature across a control volume. Since the control volume is an incremental length of tube, the corresponding pressure drop is determined by application of the Law of Conservation of Momentum. The size, speed, and accuracy of ELM make it a simple tool for use in fuel element parametric studies. ELM is a machine independent program written in FORTRAN 77. It has been successfully compiled on an IBM PC compatible running MS-DOS using Lahey FORTRAN 77, a DEC VAX series computer running VMS, and a Sun4 series computer running SunOS UNIX. ELM requires 565K of RAM under SunOS 4.1, 360K of RAM under VMS 5.4, and 406K of RAM under MS-DOS. Because this program is machine independent, no executable is provided on the distribution media. The standard distribution medium for ELM is one 5.25 inch 360K MS-DOS format diskette. ELM was developed in 1991. DEC, VAX, and VMS are trademarks of Digital Equipment Corporation. Sun4 and SunOS are trademarks of Sun Microsystems, Inc. IBM PC is a registered trademark of International Business Machines. MS-DOS is a registered trademark of Microsoft Corporation.

  15. FOIL ELEMENT FOR NUCLEAR REACTOR

    DOEpatents

    Noland, R.A.; Walker, D.E.; Spinrad, B.I.

    1963-07-16

    A method of making a foil-type fuel element is described. A foil of fuel metal is perforated in; regular design and sheets of cladding metal are placed on both sides. The cladding metal sheets are then spot-welded to each other through the perforations, and the edges sealed. (AEC)

  16. TRISO-fuel element thermo-mechanical performance modeling for the hybrid LIFE engine with Pu fuel blanket

    NASA Astrophysics Data System (ADS)

    DeMange, P.; Marian, J.; Caro, M.; Caro, A.

    2010-10-01

    A TRISO-coated fuel thermo-mechanical performance study is performed for the fusion-fission hybrid Laser Inertial Fusion Engine (LIFE) to test the viability of TRISO particles to achieve ultra-high burn-up of Pu or transuranic spent nuclear fuel blankets. Our methodology includes full elastic anisotropy, time and temperature varying material properties, and multilayer capabilities. In order to achieve fast fluences up to 30 × 10 25 n m -2 ( E > 0.18 MeV), judicious extrapolations across several orders of magnitude of existing material databases have been carried out. The results of our study indicate that failure of the pyrolytic carbon (PyC) layers occurs within the first 2 years of operation. The particles then behave as a single-SiC-layer particle and the SiC layer maintains reasonably-low tensile stresses until the end-of-life. It is also found that the PyC creep constant, K, has a striking influence on the fuel performance of TRISO-coated particles, whose stresses scale almost inversely proportional to K. Conversely, varying the geometry of the TRISO-coated fuel particles results in little differences in terms of fuel performance.

  17. Neutron source, linear-accelerator fuel enricher and regenerator and associated methods

    DOEpatents

    Steinberg, Meyer; Powell, James R.; Takahashi, Hiroshi; Grand, Pierre; Kouts, Herbert

    1982-01-01

    A device for producing fissile material inside of fabricated nuclear elements so that they can be used to produce power in nuclear power reactors. Fuel elements, for example, of a LWR are placed in pressure tubes in a vessel surrounding a liquid lead-bismuth flowing columnar target. A linear-accelerator proton beam enters the side of the vessel and impinges on the dispersed liquid lead-bismuth columns and produces neutrons which radiate through the surrounding pressure tube assembly or blanket containing the nuclear fuel elements. These neutrons are absorbed by the natural fertile uranium-238 elements and are transformed to fissile plutonium-239. The fertile fuel is thus enriched in fissile material to a concentration whereby they can be used in power reactors. After use in the power reactors, dispensed depleted fuel elements can be reinserted into the pressure tubes surrounding the target and the nuclear fuel regenerated for further burning in the power reactor.

  18. The EPA National Fuels Surveillance Network. I. Trace constituents in gasoline and commercial gasoline fuel additives.

    PubMed Central

    Jungers, R H; Lee, R E; von Lehmden, D J

    1975-01-01

    A National Fuels Surveillance Network has been established to collect gasoline and other fuels through the 10 regional offices of the Environmental Protection Agency. Physical, chemical, and trace element analytical determinations are made on the collected fuel samples to detect components which may present an air pollution hazard or poison exhaust catalytic control devices. A summary of trace elemental constituents in over 50 gasoline samples and 18 commercially marketed consumer purchased gasoline additives is presented. Quantities of Mn, Ni, Cr, Zn, Cu, Fe, Sb, B, Mg, Pb, and S were found in most regular and premium gasoline. Environmental implications of trace constituents in gasoline are discussed. PMID:1157783

  19. Molten tin reprocessing of spent nuclear fuel elements

    DOEpatents

    Heckman, Richard A.

    1983-01-01

    A method and apparatus for reprocessing spent nuclear fuel is described. Within a containment vessel, a solid plug of tin and nitride precipitates supports a circulating bath of liquid tin therein. Spent nuclear fuel is immersed in the liquid tin under an atmosphere of nitrogen, resulting in the formation of nitride precipitates. The layer of liquid tin and nitride precipitates which interfaces the plug is solidified and integrated with the plug. Part of the plug is melted, removing nitride precipitates from the containment vessel, while a portion of the plug remains solidified to support the liquid tin and nitride precipitates remaining in the containment vessel. The process is practiced numerous times until substantially all of the precipitated nitrides are removed from the containment vessel.

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

    DOEpatents

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

    1983-06-10

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

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

    DOEpatents

    Cawley, William E.; Trapp, Turner J.

    1985-01-01

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

  2. NACA Research on Slurry Fuels

    NASA Technical Reports Server (NTRS)

    Pinns, M L; Olson, W T; Barnett, H C; Breitwieser, R

    1958-01-01

    An extensive program was conducted to investigate the use of concentrated slurries of boron and magnesium in liquid hydrocarbon as fuels for afterburners and ramjet engines. Analytical calculations indicated that magnesium fuel would give greater thrust and that boron fuel would give greater range than are obtainable from jet hydrocarbon fuel alone. It was hoped that the use of these solid elements in slurry form would permit the improvement to be obtained without requiring unconventional fuel systems or combustors. Small ramjet vehicles fueled with magnesium slurry were flown successfully, but the test flights indicated that further improvement of combustors and fuel systems was needed.

  3. APPARATUS AND METHOD FOR ARC WELDING

    DOEpatents

    Noland, R.A.; Stone, C.C.

    1960-05-10

    An apparatus and method are given for forming a welding arc which is rotated by a magnetic field very rapidly about an annular electrode so that a weld is produced simultaneously over all points of an annular or closed path. This invention inhibits outgassing from the jacket of a fuel slug which is being welded by adjusting the pressure throughout the welding cycle to establish a balance between the gas pressure within the jacket and that of the atmosphere surrounding the jacket. Furthermore, an improved control of the magnetic field producing rotation of the welding arc is disclosed whereby this rotation is prevented from splashing about the metal being welded as the welding arc makes it molten.

  4. Lightweight evacuated multilayer insulation systems for the space shuttle vehicle

    NASA Technical Reports Server (NTRS)

    Barclay, D. L.; Bell, J. E.; Zimmerman, D. K.

    1973-01-01

    The elements in the evacuated multilayer insulation system were investigated, and the major weight contributors for optimization selected. Outgassing tests were conducted on candidate vacuum jacket materials and experiments were conducted to determine the vacuum and structural integrity of selected vacuum jacket configurations. A nondestructive proof test method, applicable to externally pressurized shells, was validated on this program.

  5. Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Upgrade Activities

    NASA Technical Reports Server (NTRS)

    Emrich, William J., Jr.

    2014-01-01

    Over the past year the Nuclear Thermal Rocket Element Environmental Simulator (NTREES) has been undergoing a significant upgrade beyond its initial configuration. The NTREES facility is designed to perform realistic non-nuclear testing of nuclear thermal rocket (NTR) fuel elements and fuel materials. Although the NTREES facility cannot mimic the neutron and gamma environment of an operating NTR, it can simulate the thermal hydraulic environment within an NTR fuel element to provide critical information on material performance and compatibility. The first phase of the upgrade activities which was completed in 2012 in part consisted of an extensive modification to the hydrogen system to permit computer controlled operations outside the building through the use of pneumatically operated variable position valves. This setup also allows the hydrogen flow rate to be increased to over 200 g/sec and reduced the operation complexity of the system. The second stage of modifications to NTREES which has just been completed expands the capabilities of the facility significantly. In particular, the previous 50 kW induction power supply has been replaced with a 1.2 MW unit which should allow more prototypical fuel element temperatures to be reached. The water cooling system was also upgraded to so as to be capable of removing 100% of the heat generated during. This new setup required that the NTREES vessel be raised onto a platform along with most of its associated gas and vent lines. In this arrangement, the induction heater and water systems are now located underneath the platform. In this new configuration, the 1.2 MW NTREES induction heater will be capable of testing fuel elements and fuel materials in flowing hydrogen at pressures up to 1000 psi at temperatures up to and beyond 3000 K and at near-prototypic reactor channel power densities. NTREES is also capable of testing potential fuel elements with a variety of propellants, including hydrogen with additives to inhibit

  6. An Empirical Analysis of Life Jacket Effectiveness in Recreational Boating.

    PubMed

    Viauroux, Christelle; Gungor, Ali

    2016-02-01

    This article gives a measure of life jacket (LJ) effectiveness in U.S. recreational boating. Using the U.S. Coast Guard's Boating Accident Report Database from 2008 to 2011, we find that LJ wear is one of the most important determinants influencing the number of recreational boating fatalities, together with the number of vessels involved, and the type and engine of the vessel(s). We estimate a decrease in the number of deceased per vessel of about 80% when the operator wears their LJs compared to when they do not. The odds of dying are 86% higher than average if the accident involves a canoe or kayak, but 80% lower than average when more than one vessel is involved in the accident and 34% lower than average when the operator involved in the accident has more than 100 hours of boating experience. Interestingly, we find that LJ effectiveness decreases significantly as the length of the boat increases and decreases slightly as water temperature increases. However, it increases slightly as the operator's age increases. We find that between 2008 and 2011, an LJ regulation that requires all operators to wear their LJs-representing a 20% increase in wear rate-would have saved 1,721 (out of 3,047) boaters or 1,234 out of 2,185 drowning victims. The same policy restricted to boats 16-30 feet in length would have saved approximately 778 victims. Finally, we find that such a policy would reduce the percentage of drowning victims compared to other causes of death. © 2015 Society for Risk Analysis.

  7. Reinforcement of Existing Cast-Iron Structural Elements by Means of Fiber Reinforced Composites / Wzmacnianie Istniejących, Żeliwnych Elementów Konstrukcyjnych za Pomocą Włóknokompozytów

    NASA Astrophysics Data System (ADS)

    Marcinowski, Jakub; Różycki, Zbigniew

    2016-03-01

    The paperdeals with tubular, cast-iron columns which should be reinforced due to the planned new structural function of these elements. According to the requirements of the monument conservator the general appearance of columns should not be altered significantly. Reinforcement with an external, thin coating (sleeve or jacket) made of composite (carbon fibre reinforced polymer - CFRP) was proposed. Details of the proposedtechniquewerepresented. The reinforcementeffect was verifiedin destructivetestsperformed on two columns without reinforcement and the two other columns reinforced with the chosentechnique. Due to the expected very high load capacity of the axially loaded column, the test rig was designed in such a manner that the force could be applied on big eccentricity. For this purpose a specialbase was prepared(comp. Fig. 1). Destructivetests have confirmed the high effectiveness of the adopted strengthening technique.

  8. Yttrium and rare earth stabilized fast reactor metal fuel

    DOEpatents

    Guon, Jerold; Grantham, LeRoy F.; Specht, Eugene R.

    1992-01-01

    To increase the operating temperature of a reactor, the melting point and mechanical properties of the fuel must be increased. For an actinide-rich fuel, yttrium, lanthanum and/or rare earth elements can be added, as stabilizers, to uranium and plutonium and/or a mixture of other actinides to raise the melting point of the fuel and improve its mechanical properties. Since only about 1% of the actinide fuel may be yttrium, lanthanum, or a rare earth element, the neutron penalty is low, the reactor core size can be reduced, the fuel can be burned efficiently, reprocessing requirements are reduced, and the nuclear waste disposal volumes reduced. A further advantage occurs when yttrium, lanthanum, and/or other rare earth elements are exposed to radiation in a reactor, they produce only short half life radioisotopes, which reduce nuclear waste disposal problems through much shorter assured-isolation requirements.

  9. A Study of Alternative Approaches to Produce Obscuring Smoke with JP-8 in Vehicle Engine Exhaust Smoke Systems (VEESS)

    DTIC Science & Technology

    1989-12-01

    fuel. Full D 2887 distillation data are also shown in the Appendix. Fractionation ADparatus - The glass vacuum distillation apparatus used was from a...liter - 3-neck glass round bottom flask 2. 120 cm (4 ft), 50-mm diameter vacuum -jacketed column packed with No. 2918 Helipak coils 12 TABLE 3. ASTM D...swinging bucket for variable reflux ratios and an integral condenser, all of which are vacuum jacketed 4. Product receiver of 1-liter capacity, vacuum

  10. The Transuranium Elements.

    ERIC Educational Resources Information Center

    Seaborg, Glenn T.

    1985-01-01

    Discusses the unusual chemistry of the transuranium elements as well as their impact on the periodic table. Also considers the practical applications of transuranium isotopes, such as their use in nuclear fuel for the large-scale generation of electricity. (JN)

  11. Alternative aircraft fuels technology

    NASA Technical Reports Server (NTRS)

    Grobman, J.

    1976-01-01

    NASA is studying the characteristics of future aircraft fuels produced from either petroleum or nonpetroleum sources such as oil shale or coal. These future hydrocarbon based fuels may have chemical and physical properties that are different from present aviation turbine fuels. This research is aimed at determining what those characteristics may be, how present aircraft and engine components and materials would be affected by fuel specification changes, and what changes in both aircraft and engine design would be required to utilize these future fuels without sacrificing performance, reliability, or safety. This fuels technology program was organized to include both in-house and contract research on the synthesis and characterization of fuels, component evaluations of combustors, turbines, and fuel systems, and, eventually, full-scale engine demonstrations. A review of the various elements of the program and significant results obtained so far are presented.

  12. Axially staggered seed-blanket reactor-fuel-module construction. [LWBR

    DOEpatents

    Cowell, G.K.; DiGuiseppe, C.P.

    1982-10-28

    A heterogeneous nuclear reactor of the seed-blanket type is provided wherein the fissile (seed) and fertile (blanket) nuclear fuels are segregated axially within each fuel element such that fissile and fertile regions occur in an alternating pattern along the length of the fuel element. Further, different axial stacking patterns are used for the fuel elements of at least two module types such that when modules of different types are positioned adjacent to one another, the fertile regions of the modules are offset or staggered. Thus, when a module of one type is surrounded by modules of the second type the fertile regions thereof will be surrounded on all sides by fissile material. This provides enhanced neutron communication both radially and axially, thereby resulting in greater power oscillation stability than other axial arrangements.

  13. Corrosion protected, multi-layer fuel cell interface

    DOEpatents

    Feigenbaum, Haim; Pudick, Sheldon; Wang, Chiu L.

    1986-01-01

    An improved interface configuration for use between adjacent elements of a fuel cell stack. The interface is impervious to gas and liquid and provides resistance to corrosion by the electrolyte of the fuel cell. The multi-layer configuration for the interface comprises a non-cupreous metal-coated metallic element to which is film-bonded a conductive layer by hot pressing a resin therebetween. The multi-layer arrangement provides bridging electrical contact.

  14. Applying Thermodynamics to Fossil Fuels: Heats of Combustion from Elemental Compositions.

    ERIC Educational Resources Information Center

    Lloyd, William G.; Davenport, Derek A.

    1980-01-01

    Discussed are the calculations of heats of combustions of some selected fossil fuel compounds such as some foreign shale oils and United States coals. Heating values for coal- and petroleum-derived fuel oils are also presented. (HM)

  15. Automated fuel pin loading system

    DOEpatents

    Christiansen, David W.; Brown, William F.; Steffen, Jim M.

    1985-01-01

    An automated loading system for nuclear reactor fuel elements utilizes a gravity feed conveyor which permits individual fuel pins to roll along a constrained path perpendicular to their respective lengths. The individual lengths of fuel cladding are directed onto movable transports, where they are aligned coaxially with the axes of associated handling equipment at appropriate production stations. Each fuel pin can be reciprocated axially and/or rotated about its axis as required during handling steps. The fuel pins are inserted as a batch prior to welding of end caps by one of two disclosed welding systems.

  16. Automated fuel pin loading system

    DOEpatents

    Christiansen, D.W.; Brown, W.F.; Steffen, J.M.

    An automated loading system for nuclear reactor fuel elements utilizes a gravity feed conveyor which permits individual fuel pins to roll along a constrained path perpendicular to their respective lengths. The individual lengths of fuel cladding are directed onto movable transports, where they are aligned coaxially with the axes of associated handling equipment at appropriate production stations. Each fuel pin can be be reciprocated axially and/or rotated about its axis as required during handling steps. The fuel pins are inerted as a batch prior to welding of end caps by one of two disclosed welding systems.

  17. Fuel injection and mixing systems having piezoelectric elements and methods of using the same

    DOEpatents

    Mao, Chien-Pei [Clive, IA; Short, John [Norwalk, IA; Klemm, Jim [Des Moines, IA; Abbott, Royce [Des Moines, IA; Overman, Nick [West Des Moines, IA; Pack, Spencer [Urbandale, IA; Winebrenner, Audra [Des Moines, IA

    2011-12-13

    A fuel injection and mixing system is provided that is suitable for use with various types of fuel reformers. Preferably, the system includes a piezoelectric injector for delivering atomized fuel, a gas swirler, such as a steam swirler and/or an air swirler, a mixing chamber and a flow mixing device. The system utilizes ultrasonic vibrations to achieve fuel atomization. The fuel injection and mixing system can be used with a variety of fuel reformers and fuel cells, such as SOFC fuel cells.

  18. Apparatus for testing high pressure injector elements

    NASA Technical Reports Server (NTRS)

    Myers, William Neill (Inventor); Scott, Ewell M. (Inventor); Forbes, John C. (Inventor); Shadoan, Michael D. (Inventor)

    1995-01-01

    An apparatus for testing and evaluating the spray pattern of high pressure fuel injector elements for use in supplying fuel to combustion engines is presented. Prior art fuel injector elements were normally tested by use of low pressure apparatuses which did not provide a purge to prevent mist from obscuring the injector element or to prevent frosting of the view windows; could utilize only one fluid during each test; and had their viewing ports positioned one hundred eighty (180 deg) apart, thus preventing optimum use of laser diagnostics. The high pressure fluid injector test apparatus includes an upper hub, an upper weldment or housing, a first clamp and stud/nut assembly for securing the upper hub to the upper weldment, a standoff assembly within the upper weldment, a pair of window housings having view glasses within the upper weldment, an injector block assembly and purge plate within the upper weldment for holding an injector element to be tested and evaluated, a lower weldment or housing, a second clamp and stud/nut assembly for securing the lower weldment to the upper hub, a third clamp and stud/nut assembly for securing the lower hub to the lower weldment, mechanisms for introducing fluid under high pressure for testing an injector element, and mechanisms for purging the apparatus to prevent frosting of view glasses within the window housings and to permit unobstructed viewing of the injector element.

  19. Apparatus for testing high pressure injector elements

    NASA Technical Reports Server (NTRS)

    Myers, William Neill (Inventor); Scott, Ewell M. (Inventor); Forbes, John C. (Inventor); Shadoan, Michael D. (Inventor)

    1993-01-01

    An apparatus for testing and evaluating the spray pattern of high pressure fuel injector elements for use in supplying fuel to combustion engines is presented. Prior art fuel injector elements were normally tested by use of low pressure apparatuses which did not provide a purge to prevent mist from obscuring the injector element or to prevent frosting of the view windows; could utilize only one fluid during each test; and had their viewing ports positioned one hundred eighty (180 deg) apart, thus preventing optimum use of laser diagnostics. The high pressure fluid injector test apparatus includes an upper hub, an upper weldment or housing, a first clamp and stud/nut assembly for securing the upper hub to the upper weldment, a standoff assembly within the upper weldment, a pair of window housings having view glasses within the upper weldment, an injector block assembly and purge plate within the upper weldment for holding an injector element to be tested and evaluated, a lower weldment or housing, a second clamp and stud/nut assembly for securing the lower weldment to the upper weldment, a lower hub, a third clamp and stud/nut assembly for securing the lower hub to the lower weldment, mechanisms for introducing fluid under high pressure for testing an injector element, and mechanisms for purging the apparatus to prevent frosting of view glasses within the window housings and to permit unobstructed viewing of the injector element.

  20. Evaluation of thermal optical analysis method of elemental carbon for marine fuel exhaust.

    PubMed

    Lappi, Maija K; Ristimäki, Jyrki M

    2017-12-01

    The awareness of black carbon (BC) as the second largest anthropogenic contributor in global warming and an ice melting enhancer has increased. Due to prospected increase in shipping especially in the Arctic reliability of BC emissions and their invented amounts from ships is gaining more attention. The International Maritime Organization (IMO) is actively working toward estimation of quantities and effects of BC especially in the Arctic. IMO has launched work toward constituting a definition for BC and agreeing appropriate methods for its determination from shipping emission sources. In our study we evaluated the suitability of elemental carbon (EC) analysis by a thermal-optical transmittance (TOT) method to marine exhausts and possible measures to overcome the analysis interferences related to the chemically complex emissions. The measures included drying with CaSO 4, evaporation at 40-180ºC, H 2 O treatment, and variation of the sampling method (in-stack and diluted) and its parameters (e.g., dilution ratio, Dr). A reevaluation of the nominal organic carbon (OC)/EC split point was made. Measurement of residual carbon after solvent extraction (TC-C SOF ) was used as a reference, and later also filter smoke number (FSN) measurement, which is dealt with in a forthcoming paper by the authors. Exhaust sources used for collecting the particle sample were mainly four-stroke marine engines operated with variable loads and marine fuels ranging from light to heavy fuel oils (LFO and HFO) with a sulfur content range of <0.1-2.4% S. The results were found to be dependent on many factors, namely, sampling, preparation and analysis method, and fuel quality. It was found that the condensed H 2 SO 4 + H 2 O on the particulate matter (PM) filter had an effect on the measured EC content, and also promoted the formation of pyrolytic carbon (PyC) from OC, affecting the accuracy of EC determination. Thus, uncertainty remained regarding the EC results from HFO fuels. The work

  1. Study of fuel systems for LH2-fueled subsonic transport aircraft, volume 1

    NASA Technical Reports Server (NTRS)

    Brewer, G. D.; Morris, R. E.; Davis, G. W.; Versaw, E. F.; Cunnington, G. R., Jr.; Riple, J. C.; Baerst, C. F.; Garmong, G.

    1978-01-01

    Several engine concepts examined to determine a preferred design which most effectively exploits the characteristics of hydrogen fuel in aircraft tanks received major emphasis. Many candidate designs of tank structure and cryogenic insulation systems were evaluated. Designs of all major elements of the aircraft fuel system including pumps, lines, valves, regulators, and heat exchangers received attention. Selected designs of boost pumps to be mounted in the LH2 tanks, and of a high pressure pump to be mounted on the engine were defined. A final design of LH2-fueled transport aircraft was established which incorporates a preferred design of fuel system. That aircraft was then compared with a conventionally fueled counterpart designed to equivalent technology standards.

  2. Nuclear Cryogenic Propulsion Stage Fuel Design and Fabrication

    NASA Technical Reports Server (NTRS)

    Hickman, Robert; Broadway, Jeramie; Mireles, Omar; Webb, Jon; Qualls, Lou

    2012-01-01

    Nuclear Cryogenic Propulsion Stage (NCPS) is a game changing technology for space exploration. Goal of assessing the affordability and viability of an NCPS includes these overall tasks: (1) Pre-conceptual design of the NCPS and architecture integration (2) NCPS Fuel Design and Testing (3) Nuclear Thermal Rocket Element Environmental Simulator (NTREES) (4) Affordable NCPS Development and Qualification Strategy (5) Second Generation NCPS Concepts. There is a critical need for fuels development. Fuel task objectives are to demonstrate capabilities and critical technologies using full scale element fabrication and testing.

  3. Nuclear Cryogenic Propulsion Stage Fuel Design and Fabrication

    NASA Technical Reports Server (NTRS)

    Hickman, Robert; Broadway, Jeramie; Mireles, Omar; Webb, Jon; Qualls, Lou

    2012-01-01

    Nuclear Cryogenic Propulsion Stage (NCPS) is a game changing technology for space exploration. Goal of assessing the affordability and viability of an NCPS includes thses overall tasks: (1) Pre-conceptual design of the NCPS and architecture integration (2) NCPS Fuel Design and Testing (3) Nuclear Thermal Rocket Element Environmental Simulator (NTREES) (4) Affordable NCPS Development and Qualification Strategy (5) Second Generation NCPS Concepts. There is a critical need for fuels development. Fuel task objectives are to demonstrate capabilities and critical technologies using full scale element fabrication and testing.

  4. NUCLEAR REACTOR ELEMENT

    DOEpatents

    Sanz, M.C.; Scully, C.N.

    1961-06-27

    The patented fuel element is a hexagonal graphite body having an axial channel therethrough. The graphite is impregnated with uranium which is concentrated near the axial channel. Layers of tantalum nitride and tantalum carbide are disposed on the surface of the body confronting the channel.

  5. 14 CFR 31.46 - Pressurized fuel systems.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 14 Aeronautics and Space 1 2014-01-01 2014-01-01 false Pressurized fuel systems. 31.46 Section 31... AIRWORTHINESS STANDARDS: MANNED FREE BALLOONS Design Construction § 31.46 Pressurized fuel systems. For pressurized fuel systems, each element and its connecting fittings and lines must be tested to an ultimate...

  6. 14 CFR 31.46 - Pressurized fuel systems.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Pressurized fuel systems. 31.46 Section 31... AIRWORTHINESS STANDARDS: MANNED FREE BALLOONS Design Construction § 31.46 Pressurized fuel systems. For pressurized fuel systems, each element and its connecting fittings and lines must be tested to an ultimate...

  7. 14 CFR 31.46 - Pressurized fuel systems.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 14 Aeronautics and Space 1 2013-01-01 2013-01-01 false Pressurized fuel systems. 31.46 Section 31... AIRWORTHINESS STANDARDS: MANNED FREE BALLOONS Design Construction § 31.46 Pressurized fuel systems. For pressurized fuel systems, each element and its connecting fittings and lines must be tested to an ultimate...

  8. Acute Dermal Irritation Study of Six Jet Fuels in New Zealand White Rabbits: Comparison of Four Bio-Based Jet Fuels with Two Petroleum JP-8 Fuels

    DTIC Science & Technology

    2014-02-01

    NA 5c. PROGRAM ELEMENT NUMBER 62202F 6. AUTHOR(S) Sterner, Teresa R.1; Hurley, Jonathon M.2; Edwards, James T.3; Shafer, Linda M.4; Mattie , David R... Mattie , D.R. 2014. Acute Dermal Irritation Study of Ten Jet Fuels in New Zealand White Rabbits: Comparison of Synthetic and Bio -Based Jet Fuels with...AFRL-RH-WP-TR-2014-0046 ACUTE DERMAL IRRITATION STUDY OF SIX JET FUELS IN NEW ZEALAND WHITE RABBITS: COMPARISON OF FOUR BIO -BASED JET FUELS

  9. Axially staggered seed-blanket reactor fuel module construction

    DOEpatents

    Cowell, Gary K.; DiGuiseppe, Carl P.

    1985-01-01

    A heterogeneous nuclear reactor of the seed-blanket type is provided wher the fissile (seed) and fertile (blanket) nuclear fuels are segregated axially within each fuel element such that fissile and fertile regions occur in an alternating pattern along the length of the fuel element. Further, different axial stacking patterns are used for the fuel elements of at least two module types such that when modules of different types are positioned adjacent to one another, the fertile regions of the modules are offset or staggered. Thus, when a module of one type is surrounded by modules of the second type the fertile regions thereof will be surrounded on all sides by fissile material. This provides enhanced neutron communication both radially and axially, thereby resulting in greater power oscillation stability than other axial arrangements. The arrangements of the fissile and fertile regions in an alternating axial manner minimizes the radial power peaking factors and provides a more optional thermal-hydraulic design than is afforded by radial arrangements.

  10. Irradiation behavior of U 6Mn-Al dispersion fuel elements

    NASA Astrophysics Data System (ADS)

    Meyer, M. K.; Wiencek, T. C.; Hayes, S. L.; Hofman, G. L.

    2000-02-01

    Irradiation testing of U 6Mn-Al dispersion fuel miniplates was conducted in the Oak Ridge Research Reactor (ORR). Post-irradiation examination showed that U 6Mn in an unrestrained plate configuration performs similarly to U 6Fe under irradiation, forming extensive and interlinked fission gas bubbles at a fission density of approximately 3×10 27 m-3. Fuel plate failure occurs by fission gas pressure driven `pillowing' on continued irradiation.

  11. APPLICATION OF ELECTROLESS-NICKEL BRAZING TO TUBULAR FUEL ELEMENTS FOR THE N.S. SAVANNAH. Status Report

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Lamartine, J T; Thurber, W C

    1959-06-01

    The feasibility of using electroless nickel, a chemical deposit containing about 10 wt.% phosphorous in nickel, as the brazing alloy for assembling tubular stainless steel fuel elements of the type specified in Core I of the N. S. Savannah was investigated. This material was selected primarily because of the ease of braze-metal preplacement by chemical deposition of the alloy on type 304 stainiess steel ferrule spacers, prior to fuelbundle assembly. Brazed joints produced by this method were generally characterized by a relatively ductile solid-solution region at the thinnest portions of the fillet. This ductile zone should minimize the possibility ofmore » complete propagation of hairline cracks, which form in the brittle, eutectic regions of fillet. The microstructural appearance of the electroless-nickel joints was not appreciably affected by variations in the brazing temperature from 1750 to 1900 deg F or the brazing time from 15 to 60 min. Several plating solutions were evaluated and all were found to be capable of producing deposits suitable for brazing applications. Corrosion tests conducted in static 525 deg F water indicated that no significant attack of joints brazed with electroless nickel had occurred after 300-hr exposure. A small fuel bundle was successfully assembled by brazing with electroless nickel. (auth)« less

  12. Fully ceramic nuclear fuel and related methods

    DOEpatents

    Venneri, Francesco; Katoh, Yutai; Snead, Lance Lewis

    2016-03-29

    Various embodiments of a nuclear fuel for use in various types of nuclear reactors and/or waste disposal systems are disclosed. One exemplary embodiment of a nuclear fuel may include a fuel element having a plurality of tristructural-isotropic fuel particles embedded in a silicon carbide matrix. An exemplary method of manufacturing a nuclear fuel is also disclosed. The method may include providing a plurality of tristructural-isotropic fuel particles, mixing the plurality of tristructural-isotropic fuel particles with silicon carbide powder to form a precursor mixture, and compacting the precursor mixture at a predetermined pressure and temperature.

  13. The fuelbed: a key element of the Fuel Characteristic Classification System.

    Treesearch

    Cynthia L. Riccardi; Roger D. Ottmar; David V. Sandberg; Anne Andreu; Ella Elman; Karen Kopper; Jennifer Long

    2007-01-01

    Wildland fuelbed characteristics are temporally and spatially complex and can vary widely across regions. To capture this variability, we designed the Fuel Characteristic Classification System (FCCS), a national system to create fuelbeds and classify those fuelbeds for their capacity to support fire and consume fuels. This paper describes the structure of the fuelbeds...

  14. Laser diagnostics for NTP fuel corrosion studies

    NASA Technical Reports Server (NTRS)

    Wantuck, Paul J.; Butt, D. P.; Sappey, A. D.

    1993-01-01

    Viewgraphs and explanations on laser diagnostics for nuclear thermal propulsion (NTP) fuel corrosion studies are presented. Topics covered include: NTP fuels; U-Zr-C system corrosion products; planar laser-induced fluorescence (PLIF); utilization of PLIF for corrosion product characterization of nuclear thermal rocket fuel elements under test; ZrC emission spectrum; and PLIF imaging of ZrC plume.

  15. FCRD Advanced Reactor (Transmutation) Fuels Handbook

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Janney, Dawn Elizabeth; Papesch, Cynthia Ann

    2016-09-01

    Transmutation of minor actinides such as Np, Am, and Cm in spent nuclear fuel is of international interest because of its potential for reducing the long-term health and safety hazards caused by the radioactivity of the spent fuel. One important approach to transmutation (currently being pursued by the DOE Fuel Cycle Research & Development Advanced Fuels Campaign) involves incorporating the minor actinides into U-Pu-Zr alloys, which can be used as fuel in fast reactors. U-Pu-Zr alloys are well suited for electrolytic refining, which leads to incorporation rare-earth fission products such as La, Ce, Pr, and Nd. It is, therefore, importantmore » to understand not only the properties of U-Pu-Zr alloys but also those of U-Pu-Zr alloys with concentrations of minor actinides (Np, Am) and rare-earth elements (La, Ce, Pr, and Nd) similar to those in reprocessed fuel. In addition to requiring extensive safety precautions, alloys containing U, Pu, and minor actinides (Np and Am) are difficult to study for numerous reasons, including their complex phase transformations, characteristically sluggish phasetransformation kinetics, tendency to produce experimental results that vary depending on the histories of individual samples, rapid oxidation, and sensitivity to contaminants such as oxygen in concentrations below a hundred parts per million. Although less toxic, rare-earth elements such as La, Ce, Pr, and Nd are also difficult to study for similar reasons. Many of the experimental measurements were made before 1980, and the level of documentation for experimental methods and results varies widely. It is, therefore, not surprising that little is known with certainty about U-Pu-Zr alloys, particularly those that also contain minor actinides and rare-earth elements. General acceptance of results commonly indicates that there is only a single measurement for a particular property. This handbook summarizes currently available information about U, Pu, Zr, Np, Am, La, Ce, Pr, and

  16. METHOD AND APPARATUS FOR FABRICATING TUBULAR UNITS

    DOEpatents

    Haldeman, G.W.

    1959-02-24

    A method and apparatus are described for fabricating tubular assemblies such as clad fuel elements for nuclear reactors. According to this method, a plurality of relatively short cylindrical slug-shaped members are inserted in an outer protective tubular jacket, and the assembly is passed through a reducing die to draw the outer tubular member into tight contact with the slug members, the slugs being automatically spaced with respect to each other and helium being inserted during the drawing operation to fill the spaces. The apparatus includes a pusher rod which functions to space the slugelements equidistantly by pushing on them in the direction of drawing but traveling at a slower rate than that of the tubular member.

  17. Apparatus for and method of monitoring for breached fuel elements

    DOEpatents

    Gross, Kenny C.; Strain, Robert V.

    1983-01-01

    This invention teaches improved apparatus for the method of detecting a breach in cladded fuel used in a nuclear reactor. The detector apparatus uses a separate bypass loop for conveying part of the reactor coolant away from the core, and at least three separate delayed-neutron detectors mounted proximate this detector loop. The detectors are spaced apart so that the coolant flow time from the core to each detector is different, and these differences are known. The delayed-neutron activity at the detectors is a function of the dealy time after the reaction in the fuel until the coolant carrying the delayed-neutron emitter passes the respective detector. This time delay is broken down into separate components including an isotopic holdup time required for the emitter to move through the fuel from the reaction to the coolant at the breach, and two transit times required for the emitter now in the coolant to flow from the breach to the detector loop and then via the loop to the detector. At least two of these time components are determined during calibrated operation of the reactor. Thereafter during normal reactor operation, repeated comparisons are made by the method of regression approximation of the third time component for the best-fit line correlating measured delayed-neutron activity against activity that is approximated according to specific equations. The equations use these time-delay components and known parameter values of the fuel and of the part and emitting daughter isotopes.

  18. Fuel clad chemical interactions in fast reactor MOX fuels

    NASA Astrophysics Data System (ADS)

    Viswanathan, R.

    2014-01-01

    Clad corrosion being one of the factors limiting the life of a mixed-oxide fast reactor fuel element pin at high burn-up, some aspects known about the key elements (oxygen, cesium, tellurium, iodine) in the clad-attack are discussed and many Fuel-Clad-Chemical-Interaction (FCCI) models available in the literature are also discussed. Based on its relatively superior predictive ability, the HEDL (Hanford Engineering Development Laboratory) relation is recommended: d/μm = ({0.507 ṡ [B/(at.% fission)] ṡ (T/K-705) ṡ [(O/M)i-1.935]} + 20.5) for (O/M)i ⩽ 1.98. A new model is proposed for (O/M)i ⩾ 1.98: d/μm = [B/(at.% fission)] ṡ (T/K-800)0.5 ṡ [(O/M)i-1.94] ṡ [P/(W cm-1)]0.5. Here, d is the maximum depth of clad attack, B is the burn-up, T is the clad inner surface temperature, (O/M)i is the initial oxygen-to-(uranium + plutonium) ratio, and P is the linear power rating. For fuels with [n(Pu)/n(M = U + Pu)] > 0.25, multiplication factors f are recommended to consider the potential increase in the depth of clad-attack.

  19. Uranium chloride extraction of transuranium elements from LWR fuel

    DOEpatents

    Miller, William E.; Ackerman, John P.; Battles, James E.; Johnson, Terry R.; Pierce, R. Dean

    1992-01-01

    A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels containing rare earth and noble metal fission products as well as other fission products is disclosed. The oxide fuel is reduced with Ca metal in the presence of Ca chloride and a U-Fe alloy which is liquid at about 800.degree. C. to dissolve uranium metal and the noble metal fission product metals and transuranium actinide metals and rare earth fission product metals leaving Ca chloride having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein. The Ca chloride and CaO and the fission products contained therein are separated from the U-Fe alloy and the metal values dissolved therein. The U-Fe alloy having dissolved therein reduced metals from the spent nuclear fuel is contacted with a mixture of one or more alkali metal or alkaline earth metal halides selected from the class consisting of alkali metal or alkaline earth metal and Fe or U halide or a combination thereof to transfer transuranium actinide metals and rare earth metals to the halide salt leaving the uranium and some noble metal fission products in the U-Fe alloy and thereafter separating the halide salt and the transuranium metals dissolved therein from the U-Fe alloy and the metals dissolved therein.

  20. Uranium chloride extraction of transuranium elements from LWR fuel

    DOEpatents

    Miller, W.E.; Ackerman, J.P.; Battles, J.E.; Johnson, T.R.; Pierce, R.D.

    1992-08-25

    A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels containing rare earth and noble metal fission products as well as other fission products is disclosed. The oxide fuel is reduced with Ca metal in the presence of Ca chloride and a U-Fe alloy which is liquid at about 800 C to dissolve uranium metal and the noble metal fission product metals and transuranium actinide metals and rare earth fission product metals leaving Ca chloride having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein. The Ca chloride and CaO and the fission products contained therein are separated from the U-Fe alloy and the metal values dissolved therein. The U-Fe alloy having dissolved therein reduced metals from the spent nuclear fuel is contacted with a mixture of one or more alkali metal or alkaline earth metal halides selected from the class consisting of alkali metal or alkaline earth metal and Fe or U halide or a combination thereof to transfer transuranium actinide metals and rare earth metals to the halide salt leaving the uranium and some noble metal fission products in the U-Fe alloy and thereafter separating the halide salt and the transuranium metals dissolved therein from the U-Fe alloy and the metals dissolved therein. 1 figure.

  1. Validation of MCNP: SPERT-D and BORAX-V fuel

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Crawford, C.; Palmer, B.

    1992-11-01

    This report discusses critical experiments involving SPERT-D{sup 1,2} fuel elements and BORAX-V{sup 3-8} fuel which have been modeled and calculations performed with MCNP. MCNP is a Monte Carlo based transport code. For this study continuous-energy nuclear data from the ENDF/B-V cross section library was used. The SPERT-D experiments consisted of various arrays of fuel elements moderated and reflected with either water or a uranyl nitrate solution. Some SPERT-D experiments used cadmium as a fixed neutron poison, while others were poisoned with various concentrations of boron in the moderating/reflecting solution. ne BORAX-V experiments were arrays of either boiling fuel rod assembliesmore » or superheater assemblies, both types of arrays were moderated and reflected with water. In one boiling fuel experiment, two fuel rods were replaced with borated stainless steel poison rods.« less

  2. Validation of MCNP: SPERT-D and BORAX-V fuel

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Crawford, C.; Palmer, B.

    1992-11-01

    This report discusses critical experiments involving SPERT-D[sup 1,2] fuel elements and BORAX-V[sup 3-8] fuel which have been modeled and calculations performed with MCNP. MCNP is a Monte Carlo based transport code. For this study continuous-energy nuclear data from the ENDF/B-V cross section library was used. The SPERT-D experiments consisted of various arrays of fuel elements moderated and reflected with either water or a uranyl nitrate solution. Some SPERT-D experiments used cadmium as a fixed neutron poison, while others were poisoned with various concentrations of boron in the moderating/reflecting solution. ne BORAX-V experiments were arrays of either boiling fuel rod assembliesmore » or superheater assemblies, both types of arrays were moderated and reflected with water. In one boiling fuel experiment, two fuel rods were replaced with borated stainless steel poison rods.« less

  3. Initial Operation and Shakedown of the Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

    NASA Technical Reports Server (NTRS)

    Emrich, William J., Jr.

    2014-01-01

    To support the on-going nuclear thermal propulsion effort, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The facility to perform this testing is referred to as the Nuclear Thermal Rocket Element Environment Simulator (NTREES). This device can simulate the environmental conditions (minus the radiation) to which nuclear rocket fuel components will be subjected during reactor operation. Prototypical fuel elements mounted in the simulator are inductively heated in such a manner so as to accurately reproduce the temperatures and heat fluxes which would normally occur as a result of nuclear fission in addition to being exposed to flowing hydrogen. Recent upgrades to NTREES now allow power levels 24 times greater than those achievable in the previous facility configuration. This higher power operation will allow near prototypical power densities and flows to finally be achieved in most prototypical fuel elements.

  4. An under-designed RC frame: Seismic assessment through displacement based approach and possible refurbishment with FRP strips and RC jacketing

    NASA Astrophysics Data System (ADS)

    Valente, Marco; Milani, Gabriele

    2017-07-01

    Many existing reinforced concrete buildings in Southern Europe were built (and hence designed) before the introduction of displacement based design in national seismic codes. They are obviously highly vulnerable to seismic actions. In such a situation, simplified methodologies for the seismic assessment and retrofitting of existing structures are required. In this study, a displacement based procedure using non-linear static analyses is applied to a four-story existing RC frame. The aim is to obtain an estimation of its overall structural inadequacy as well as the effectiveness of a specific retrofitting intervention by means of GFRP laminates and RC jacketing. Accurate numerical models are developed within a displacement based approach to reproduce the seismic response of the RC frame in the original configuration and after strengthening.

  5. Apparatus for and method of monitoring for breached fuel elements

    DOEpatents

    Gross, K.C.; Strain, R.V.

    1981-04-28

    This invention teaches improved apparatus for the method of detecting a breach in cladded fuel used in a nuclear reactor. The detector apparatus uses a separate bypass loop for conveying part of the reactor coolant away from the core, and at least three separate delayed-neutron detectors mounted proximate this detector loop. The detectors are spaced apart so that the coolant flow time from the core to each detector is different, and these differences are known. The delayed-neutron activity at the detectors is a function of the delay time after the reaction in the fuel until the coolant carrying the delayed-neutron emitter passes the respective detector. This time delay is broken down into separate components including an isotopic holdup time required for the emitter to move through the fuel from the reaction to the coolant at the breach, and two transit times required for the emitter now in the coolant to flow from the breach to the detector loop and then via the loop to the detector.

  6. Local Heat Flux Measurements with Single Element Coaxial Injectors

    NASA Technical Reports Server (NTRS)

    Jones, Gregg; Protz, Christopher; Bullard, Brad; Hulka, James

    2006-01-01

    To support the mission for the NASA Vision for Space Exploration, the NASA Marshall Space Flight Center conducted a program in 2005 to improve the capability to predict local thermal compatibility and heat transfer in liquid propellant rocket engine combustion devices. The ultimate objective was to predict and hence reduce the local peak heat flux due to injector design, resulting in a significant improvement in overall engine reliability and durability. Such analyses are applicable to combustion devices in booster, upper stage, and in-space engines, as well as for small thrusters with few elements in the injector. In this program, single element and three-element injectors were hot-fire tested with liquid oxygen and ambient temperature gaseous hydrogen propellants at The Pennsylvania State University Cryogenic Combustor Laboratory from May to August 2005. Local heat fluxes were measured in a 1-inch internal diameter heat sink combustion chamber using Medtherm coaxial thermocouples and Gardon heat flux gauges. Injectors were tested with shear coaxial and swirl coaxial elements, including recessed, flush and scarfed oxidizer post configurations, and concentric and non-concentric fuel annuli. This paper includes general descriptions of the experimental hardware, instrumentation, and results of the hot-fire testing for three of the single element injectors - recessed-post shear coaxial with concentric fuel, flush-post swirl coaxial with concentric fuel, and scarfed-post swirl coaxial with concentric fuel. Detailed geometry and test results will be published elsewhere to provide well-defined data sets for injector development and model validatation.

  7. Gaseous fuel reactors for power systems

    NASA Technical Reports Server (NTRS)

    Kendall, J. S.; Rodgers, R. J.

    1977-01-01

    Gaseous-fuel nuclear reactors have significant advantages as energy sources for closed-cycle power systems. The advantages arise from the removal of temperature limits associated with conventional reactor fuel elements, the wide variety of methods of extracting energy from fissioning gases, and inherent low fissile and fission product in-core inventory due to continuous fuel reprocessing. Example power cycles and their general performance characteristics are discussed. Efficiencies of gaseous fuel reactor systems are shown to be high with resulting minimal environmental effects. A technical overview of the NASA-funded research program in gaseous fuel reactors is described and results of recent tests of uranium hexafluoride (UF6)-fueled critical assemblies are presented.

  8. Two-Dimensional Diffusion Theory Analysis of Reactivity Effects of a Fuel-Plate-Removal Experiment

    NASA Technical Reports Server (NTRS)

    Gotsky, Edward R.; Cusick, James P.; Bogart, Donald

    1959-01-01

    Two-dimensional two-group diffusion calculations were performed on the NASA reactor simulator in order to evaluate the reactivity effects of fuel plates removed successively from the center experimental fuel element of a seven- by three-element core loading at the Oak Ridge Bulk Shielding Facility. The reactivity calculations were performed by two methods: In the first, the slowing-down properties of the experimental fuel element were represented by its infinite media parameters; and, in the second, the finite size of the experimental fuel element was recognized, and the slowing-down properties of the surrounding core were attributed to this small region. The latter calculation method agreed very well with the experimented reactivity effects; the former method underestimated the experimental reactivity effects.

  9. An Optimization-Based Approach to Injector Element Design

    NASA Technical Reports Server (NTRS)

    Tucker, P. Kevin; Shyy, Wei; Vaidyanathan, Rajkumar; Turner, Jim (Technical Monitor)

    2000-01-01

    An injector optimization methodology, method i, is used to investigate optimal design points for gaseous oxygen/gaseous hydrogen (GO2/GH2) injector elements. A swirl coaxial element and an unlike impinging element (a fuel-oxidizer-fuel triplet) are used to facilitate the study. The elements are optimized in terms of design variables such as fuel pressure drop, APf, oxidizer pressure drop, deltaP(sub f), combustor length, L(sub comb), and full cone swirl angle, theta, (for the swirl element) or impingement half-angle, alpha, (for the impinging element) at a given mixture ratio and chamber pressure. Dependent variables such as energy release efficiency, ERE, wall heat flux, Q(sub w), injector heat flux, Q(sub inj), relative combustor weight, W(sub rel), and relative injector cost, C(sub rel), are calculated and then correlated with the design variables. An empirical design methodology is used to generate these responses for both element types. Method i is then used to generate response surfaces for each dependent variable for both types of elements. Desirability functions based on dependent variable constraints are created and used to facilitate development of composite response surfaces representing the five dependent variables in terms of the input variables. Three examples illustrating the utility and flexibility of method i are discussed in detail for each element type. First, joint response surfaces are constructed by sequentially adding dependent variables. Optimum designs are identified after addition of each variable and the effect each variable has on the element design is illustrated. This stepwise demonstration also highlights the importance of including variables such as weight and cost early in the design process. Secondly, using the composite response surface that includes all five dependent variables, unequal weights are assigned to emphasize certain variables relative to others. Here, method i is used to enable objective trade studies on design issues

  10. Magnesium transport extraction of transuranium elements from LWR fuel

    DOEpatents

    Ackerman, John P.; Battles, James E.; Johnson, Terry R.; Miller, William E.; Pierce, R. Dean

    1992-01-01

    A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels which contain rare earth and noble metal fission products. The oxide fuel is reduced with Ca metal in the presence of CaCl.sub.2 and a U-Fe alloy containing not less than about 84% by weight uranium at a temperature in the range of from about 800.degree. C. to about 850.degree. C. to produce additional uranium metal which dissolves in the U-Fe alloy raising the uranium concentration and having transuranium actinide metals and rare earth fission product metals and the noble metal fission products dissolved therein. The CaCl.sub.2 having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein is separated and electrolytically treated with a carbon electrode to reduce the CaO to Ca metal while converting the carbon electrode to CO and CO.sub.2. The Ca metal and CaCl.sub.2 is recycled to reduce additional oxide fuel. The U-Fe alloy having transuranium actinide metals and rare earth fission product metals and the noble metal fission products dissolved therein is contacted with Mg metal which takes up the actinide and rare earth fission product metals. The U-Fe alloy retains the noble metal fission products and is stored while the Mg is distilled and recycled leaving the transuranium actinide and rare earth fission products isolated.

  11. Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Upgrade Activities

    NASA Technical Reports Server (NTRS)

    Emrich, William J. Jr.; Moran, Robert P.; Pearson, J. Boise

    2012-01-01

    To support the on-going nuclear thermal propulsion effort, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The facility to perform this testing is referred to as the Nuclear Thermal Rocket Element Environment Simulator (NTREES). This device can simulate the environmental conditions (minus the radiation) to which nuclear rocket fuel components will be subjected during reactor operation. Test articles mounted in the simulator are inductively heated in such a manner so as to accurately reproduce the temperatures and heat fluxes which would normally occur as a result of nuclear fission and would be exposed to flowing hydrogen. Initial testing of a somewhat prototypical fuel element has been successfully performed in NTREES and the facility has now been shutdown to allow for an extensive reconfiguration of the facility which will result in a significant upgrade in its capabilities

  12. Thermo-Physics Technical Note No. 60: thermal analysis of SNAP 10A reactor core during atmospheric reentry and resulting core disintegration and fuel element separation

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Mouradian, E.M.

    1966-02-16

    A thermal analysis is carried out to determine the temperature distribution throughout a SNAP 10A reactor core, particularly in the vicinity of the grid plates, during atmospheric reentry. The transient temperatue distribution of the grid plate indicates when sufficient melting occurs so that fuel elements are free to be released and continue their descent individually.

  13. Multidimensional Fuel Performance Code: BISON

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    BISON is a finite element based nuclear fuel performance code applicable to a variety of fuel forms including light water reactor fuel rods, TRISO fuel particles, and metallic rod and plate fuel (Refs. [a, b, c]). It solves the fully-coupled equations of thermomechanics and species diffusion and includes important fuel physics such as fission gas release and material property degradation with burnup. BISON is based on the MOOSE framework (Ref. [d]) and can therefore efficiently solve problems on 1-, 2- or 3-D meshes using standard workstations or large high performance computers. BISON is also coupled to a MOOSE-based mesoscale phasemore » field material property simulation capability (Refs. [e, f]). As described here, BISON includes the code library named FOX, which was developed concurrent with BISON. FOX contains material and behavioral models that are specific to oxide fuels.« less

  14. Process to separate transuranic elements from nuclear waste

    DOEpatents

    Johnson, T.R.; Ackerman, J.P.; Tomczuk, Z.; Fischer, D.F.

    1988-07-12

    A process for removing transuranic elements from a waste chloride electrolytic salt containing transuranic elements in addition to rare earth and other fission product elements so the salt waste may be disposed of more easily and the valuable transuranic elements may be recovered for reuse. The salt is contacted with a cadmium-uranium alloy which selectively extracts the transuranic elements from the salt. The waste salt is generated during the reprocessing of nuclear fuel associated with the Integral Fast Reactor (IFR). 2 figs.

  15. Process to separate transuranic elements from nuclear waste

    DOEpatents

    Johnson, Terry R.; Ackerman, John P.; Tomczuk, Zygmunt; Fischer, Donald F.

    1989-01-01

    A process for removing transuranic elements from a waste chloride electrolytic salt containing transuranic elements in addition to rare earth and other fission product elements so the salt waste may be disposed of more easily and the valuable transuranic elements may be recovered for reuse. The salt is contacted with a cadmium-uranium alloy which selectively extracts the transuranic elements from the salt. The waste salt is generated during the reprocessing of nuclear fuel associated with the Integral Fast Reactor (IFR).

  16. Film bonded fuel cell interface configuration

    DOEpatents

    Kaufman, Arthur; Terry, Peter L.

    1985-01-01

    An improved interface configuration for use between adjacent elements of a fuel cell stack. The interface is impervious to gas and liquid and provides resistance to corrosion by the electrolyte of the fuel cell. A multi-layer arrangement for the interface provides bridging electrical contact with a hot-pressed resin filling the void space.

  17. Spacecraft attitude impacts on COLD-SAT non-vacuum jacketed LH2 supply tank thermal performance

    NASA Technical Reports Server (NTRS)

    Arif, Hugh

    1990-01-01

    The Cryogenic On-Orbit Liquid Depot - Storage, Acquisition and Transfer (COLD-SAT) spacecraft will be launched into low earth orbit to perform fluid management experiments on the behavior of subcritical liquid hydrogen (LH2). For determining the optimum on-orbit attitude for the COLD-SAT satellite, a comparative analytical study was performed to determine the thermal impacts of spacecraft attitude on the performance of the COLD-SAT non-vacuum jacketed LH2 supply tank. Tank thermal performance was quantitied by total conductive and radiative heat leakage into the pressure vessel due to the absorbed solar, earth albedo and infra-red on-orbit fluxes, and also by the uniformity of the variation of this leakage on the vessel surface area. Geometric and thermal analysis math models were developed for the spacecraft and the tank as part of this analysis, based on their individual thermal/structural designs. Two quasi-inertial spacecraft attitudes were investigated and their effects on the tank performance compared. The results are one of the criteria by which the spacecraft orientation in orbit was selected for the in-house NASA Lewis Research Center design.

  18. Spacecraft attitude impacts on COLD-SAT non-vacuum jacketed LH2 supply tank thermal performance

    NASA Technical Reports Server (NTRS)

    Arif, Hugh

    1990-01-01

    The Cryogenic On-Orbit Liquid Depot - Storage, Acquisition and Transfer (COLD-SAT) spacecraft will be launched into low earth orbit to perform fluid management experiments on the behavior of subcritical liquid hydrogen (LH2). For determining the optimum on-orbit attitude for the COLD-SAT satellite, a comparative analytical study was performed to determine the thermal impacts of spacecraft attitude on the performance of the COLD-SAT non-vacuum jacketed LH2 supply tank. Tank thermal performance was quantified by total conductive and radiative heat leakage into the pressure vessel due to the absorbed solar, earth albedo and infra-red on-orbit fluxes, and also by the uniformity of the variation of this leakage on the vessel surface area. Geometric and thermal analysis math models were developed for the spacecraft and the tank as part of this analysis, based on their individual thermal/structural designs. Two quasi-inertial spacecraft attitudes were investigated and their effects on the tank performance compared. The results are one of the criteria by which the spacecraft orientation in orbit was selected for the in-house NASA Lewis Research Center design.

  19. Finite element simulation of gap opening between cladding tube and spacer grid in a fuel rod assembly using crystallographic models of irradiation growth and creep

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Patra, Anirban; Tomé, Carlos N.

    A physically-based crystal plasticity framework for modeling irradiation growth and creep is interfaced with the finite element code ABAQUS in order to study the contact forces and the gap evolution between the spacer grid and the cladding tube as a function of irradiation in a representative section of a fuel rod assembly. Deformation mechanisms governing the gap opening are identified and correlated to the texture-dependent material response. Thus, in the absence of coolant flow-induced vibrations, these simulations predict the contribution of irradiation growth and creep to the gap opening between the cladding tube and the springs and dimples on themore » spacer grid. The simulated contact forces on the springs and dimples are compared to available experimental and modeling data. Various combinations of external loads are applied on the springs and dimples to simulate fuel rods in the interior and at the periphery of the fuel rod assembly. Furthermore, we found that loading conditions representative (to a first order approximation) of fuel rods at the periphery show higher gap opening. This is in agreement with in-reactor data, where rod leakages due to the synergistic effects of gap opening and coolant flow-induced vibrations were generally found to occur at the periphery of the fuel rod assembly.« less

  20. Finite element simulation of gap opening between cladding tube and spacer grid in a fuel rod assembly using crystallographic models of irradiation growth and creep

    DOE PAGES

    Patra, Anirban; Tomé, Carlos N.

    2017-03-06

    A physically-based crystal plasticity framework for modeling irradiation growth and creep is interfaced with the finite element code ABAQUS in order to study the contact forces and the gap evolution between the spacer grid and the cladding tube as a function of irradiation in a representative section of a fuel rod assembly. Deformation mechanisms governing the gap opening are identified and correlated to the texture-dependent material response. Thus, in the absence of coolant flow-induced vibrations, these simulations predict the contribution of irradiation growth and creep to the gap opening between the cladding tube and the springs and dimples on themore » spacer grid. The simulated contact forces on the springs and dimples are compared to available experimental and modeling data. Various combinations of external loads are applied on the springs and dimples to simulate fuel rods in the interior and at the periphery of the fuel rod assembly. Furthermore, we found that loading conditions representative (to a first order approximation) of fuel rods at the periphery show higher gap opening. This is in agreement with in-reactor data, where rod leakages due to the synergistic effects of gap opening and coolant flow-induced vibrations were generally found to occur at the periphery of the fuel rod assembly.« less

  1. Pellet Cladding Mechanical Interaction Modeling Using the Extended Finite Element Method

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Spencer, Benjamin W.; Jiang, Wen; Dolbow, John E.

    As a brittle material, the ceramic UO2 used as light water reactor fuel experiences significant fracturing throughout its life, beginning with the first rise to power of fresh fuel. This has multiple effects on the thermal and mechanical response of the fuel/cladding system. One such effect that is particularly important is that when there is mechanical contact between the fuel and cladding, cracks that extending from the outer surface of the fuel into the volume of the fuel cause elevated stresses in the adjacent cladding, which can potentially lead to cladding failure. Modeling the thermal and mechanical response of themore » cladding in the vicinity of these surface-breaking cracks in the fuel can provide important insights into this behavior to help avoid operating conditions that could lead to cladding failure. Such modeling has traditionally been done in the context of finite-element-based fuel performance analysis by modifying the fuel mesh to introduce discrete cracks. While this approach is effective in capturing the important behavior at the fuel/cladding interface, there are multiple drawbacks to explicitly incorporating the cracks in the finite element mesh. Because the cracks are incorporated in the original mesh, the mesh must be modified for cracks of specified location and depth, so it is difficult to account for crack propagation and the formation of new cracks at other locations. The extended finite element method (XFEM) has emerged in recent years as a powerful method to represent arbitrary, evolving, discrete discontinuities within the context of the finite element method. Development work is underway by the authors to implement XFEM in the BISON fuel performance code, and this capability has previously been demonstrated in simulations of fracture propagation in ceramic nuclear fuel. These preliminary demonstrations have included only the fuel, and excluded the cladding for simplicity. This paper presents initial results of efforts to apply

  2. Durability test on irradiated rock-like oxide fuels

    NASA Astrophysics Data System (ADS)

    Kuramoto, K.; Nitani, N.; Yamashita, T.

    2003-06-01

    For a profitable use of Pu, Japan Atomic Energy Research Institute has been promoting researches for once-through type fuels. The strategy consists of stable rock-like oxide fuel fabrication in conventional fuel facilities followed by almost complete Pu burning in LWR and disposal of chemically stable spent fuel without further processing. Because leach rates of hazardous nuclides, such as TRU and β-emitters, that have long half-lives, are very important for the evaluation of geological safety, leaching tests in deionized water at 363 K were performed with reference to the MCC-1 method. Five irradiated fuel pellets, a single phase fuel of a yttria-stabilized zirconia (YSZ) containing UO 2 (U-YSZ), two fuels of U-YSZ particle dispersed in MgAl 2O 4 (SPI) or Al 2O 3 (COR) matrix, two homogeneous-blended fuels of U-YSZ and SPI or COR powders, were submitted to the tests. Stainless steel containers with Au coating and ethylene propylene diene monomer were used as leaching vessels and packing, respectively. The evaluated normalized leach rates of Zr, U and Pu were obviously lower than those of the other important elements and nuclides. Americium, Np and especially Y showed unexpectedly high evaluated normalized leach rates. The volatile elements, Cs and I, showed enhanced leaching within particle-dispersed type fuels because of crack formation around the particle.

  3. Molten carbonate fuel cell separator

    DOEpatents

    Nickols, Richard C.

    1986-09-02

    In a stacked array of molten carbonate fuel cells, a fuel cell separator is positioned between adjacent fuel cells to provide isolation as well as a conductive path therebetween. The center portion of the fuel cell separator includes a generally rectangular, flat, electrical conductor. Around the periphery of the flat portion of the separator are positioned a plurality of elongated resilient flanges which form a gas-tight seal around the edges of the fuel cell. With one elongated flange resiliently engaging a respective edge of the center portion of the separator, the sealing flanges, which are preferably comprised of a noncorrosive material such as an alloy of yttrium, iron, aluminum or chromium, form a tight-fitting wet seal for confining the corrosive elements of the fuel cell therein. This arrangement permits a good conductive material which may be highly subject to corrosion and dissolution to be used in combination with a corrosion-resistant material in the fuel cell separator of a molten carbonate fuel cell for improved fuel cell conductivity and a gas-tight wet seal.

  4. Molten carbonate fuel cell separator

    DOEpatents

    Nickols, R.C.

    1984-10-17

    In a stacked array of molten carbonate fuel cells, a fuel cell separator is positioned between adjacent fuel cells to provide isolation as well as a conductive path therebetween. The center portion of the fuel cell separator includes a generally rectangular, flat, electrical conductor. Around the periphery of the flat portion of the separator are positioned a plurality of elongated resilient flanges which form a gas-tight seal around the edges of the fuel cell. With one elongated flange resiliently engaging a respective edge of the center portion of the separator, the sealing flanges, which are preferably comprised of a noncorrosive material such as an alloy of yttrium, iron, aluminum or chromium, form a tight-fitting wet seal for confining the corrosive elements of the fuel cell therein. This arrangement permits a good conductive material which may be highly subject to corrosion and dissolution to be used in combination with a corrosion-resistant material in the fuel cell separator of a molten carbonate fuel cell for improved fuel cell conductivity and a gas-tight wet seal.

  5. Process to separate transuranic elements from nuclear waste

    DOEpatents

    Johnson, T.R.; Ackerman, J.P.; Tomczuk, Z.; Fischer, D.F.

    1989-03-21

    A process is described for removing transuranic elements from a waste chloride electrolytic salt containing transuranic elements in addition to rare earth and other fission product elements so the salt waste may be disposed of more easily and the valuable transuranic elements may be recovered for reuse. The salt is contacted with a cadmium-uranium alloy which selectively extracts the transuranic elements from the salt. The waste salt is generated during the reprocessing of nuclear fuel associated with the Integral Fast Reactor (IFR). 2 figs.

  6. Aerosol emissions of a ship diesel engine operated with diesel fuel or heavy fuel oil.

    PubMed

    Streibel, Thorsten; Schnelle-Kreis, Jürgen; Czech, Hendryk; Harndorf, Horst; Jakobi, Gert; Jokiniemi, Jorma; Karg, Erwin; Lintelmann, Jutta; Matuschek, Georg; Michalke, Bernhard; Müller, Laarnie; Orasche, Jürgen; Passig, Johannes; Radischat, Christian; Rabe, Rom; Reda, Ahmed; Rüger, Christopher; Schwemer, Theo; Sippula, Olli; Stengel, Benjamin; Sklorz, Martin; Torvela, Tiina; Weggler, Benedikt; Zimmermann, Ralf

    2017-04-01

    Gaseous and particulate emissions from a ship diesel research engine were elaborately analysed by a large assembly of measurement techniques. Applied methods comprised of offline and online approaches, yielding averaged chemical and physical data as well as time-resolved trends of combustion by-products. The engine was driven by two different fuels, a commonly used heavy fuel oil (HFO) and a standardised diesel fuel (DF). It was operated in a standardised cycle with a duration of 2 h. Chemical characterisation of organic species and elements revealed higher concentrations as well as a larger number of detected compounds for HFO operation for both gas phase and particulate matter. A noteworthy exception was the concentration of elemental carbon, which was higher in DF exhaust aerosol. This may prove crucial for the assessment and interpretation of biological response and impact via the exposure of human lung cell cultures, which was carried out in parallel to this study. Offline and online data hinted at the fact that most organic species in the aerosol are transferred from the fuel as unburned material. This is especially distinctive at low power operation of HFO, where low volatility structures are converted to the particulate phase. The results of this study give rise to the conclusion that a mere switching to sulphur-free fuel is not sufficient as remediation measure to reduce health and environmental effects of ship emissions.

  7. Finite element analysis for the evaluation of protective functions of helmets against ballistic impact.

    PubMed

    Lee, H P; Gong, S W

    2010-10-01

    The ballistic impact of a human head model protected by a Personnel Armor System Ground Troops Kevlar® helmet is analysed using the finite element method. The emphasis is to examine the effect of the interior cushioning system as a shock absorber in mitigating ballistic impact to the head. The simulations of the frontal and side impacts of the full metal jacket (FMJ) and fragment-simulating projectile (FSP) were carried out using LS-DYNA. It was found that the Kevlar® helmet with its interior nylon and leather strap was able to defeat both the FMJ and FSP without the projectiles penetrating the helmet. However, the head injuries caused by the FMJ impact can be fatal due to the high stiffness of the interior strap. The bulge section at the side of the Kevlar® helmet had more room for deformation that resulted in less serious head injuries.

  8. Gaseous fuel nuclear reactor research

    NASA Technical Reports Server (NTRS)

    Schwenk, F. C.; Thom, K.

    1975-01-01

    Gaseous-fuel nuclear reactors are described; their distinguishing feature is the use of fissile fuels in a gaseous or plasma state, thereby breaking the barrier of temperature imposed by solid-fuel elements. This property creates a reactor heat source that may be able to heat the propellant of a rocket engine to 10,000 or 20,000 K. At this temperature level, gas-core reactors would provide the breakthrough in propulsion needed to open the entire solar system to manned and unmanned spacecraft. The possibility of fuel recycling makes possible efficiencies of up to 65% and nuclear safety at reduced cost, as well as high-thrust propulsion capabilities with specific impulse up to 5000 sec.

  9. A finite element analysis modeling tool for solid oxide fuel cell development: coupled electrochemistry, thermal and flow analysis in MARC®

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Khaleel, Mohammad A.; Lin, Zijing; Singh, Prabhakar

    2004-05-03

    A 3D simulation tool for modeling solid oxide fuel cells is described. The tool combines the versatility and efficiency of a commercial finite element analysis code, MARC{reg_sign}, with an in-house developed robust and flexible electrochemical (EC) module. Based upon characteristic parameters obtained experimentally and assigned by the user, the EC module calculates the current density distribution, heat generation, and fuel and oxidant species concentration, taking the temperature profile provided by MARC{reg_sign} and operating conditions such as the fuel and oxidant flow rate and the total stack output voltage or current as the input. MARC{reg_sign} performs flow and thermal analyses basedmore » on the initial and boundary thermal and flow conditions and the heat generation calculated by the EC module. The main coupling between MARC{reg_sign} and EC is for MARC{reg_sign} to supply the temperature field to EC and for EC to give the heat generation profile to MARC{reg_sign}. The loosely coupled, iterative scheme is advantageous in terms of memory requirement, numerical stability and computational efficiency. The coupling is iterated to self-consistency for a steady-state solution. Sample results for steady states as well as the startup process for stacks with different flow designs are presented to illustrate the modeling capability and numerical performance characteristic of the simulation tool.« less

  10. Micro-structural study and Rietveld analysis of fast reactor fuels: U-Mo fuels

    NASA Astrophysics Data System (ADS)

    Chakraborty, S.; Choudhuri, G.; Banerjee, J.; Agarwal, Renu; Khan, K. B.; Kumar, Arun

    2015-12-01

    U-Mo alloys are the candidate fuels for both research reactors and fast breeder reactors. In-reactor performance of the fuel depends on the microstructural stability and thermal properties of the fuel. To improve the fuel performance, alloying elements viz. Zr, Mo, Nb, Ti and fissium are added in the fuel. The first reactor fuels are normally prepared by injection casting. The objective of this work is to compare microstructure, phase-fields and hardness of as-cast four different U-Mo alloy (2, 5, 10 and 33 at.% Mo) fuels with the equilibrium microstructure of the alloys. Scanning electron microscope with energy dispersive spectrometer and optical microscope have been used to characterize the morphology of the as-cast and annealed alloys. The monoclinic α'' phase in as-cast U-10 at.% Mo alloy has been characterized through Rietveld analysis. A comparison of metallographic and Rietveld analysis of as-cast (dendritic microstructure) and annealed U-33 at.% Mo alloy, corresponding to intermetallic compound, has been reported here for the first time. This study will provide in depth understanding of microstructural and phase evolution of U-Mo alloys as fast reactor fuel.

  11. Analysis of Accidents at the Pakistan Research Reactor-1 Using Proposed Mixed-Fuel (HEU and LEU) Core

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Bokhari, Ishtiaq H.

    2004-12-15

    The Pakistan Research Reactor-1 (PARR-1) was converted from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel in 1991. The reactor is running successfully, with an upgraded power level of 10 MW. To save money on the purchase of costly fresh LEU fuel elements, the use of less burnt HEU spent fuel elements along with the present LEU fuel elements is being considered. The proposal calls for the HEU fuel elements to be placed near the thermal column to gain the required excess reactivity. In the present study the safety analysis of a proposed mixed-fuel core has been carried outmore » at a calculated steady-state power level of 9.8 MW. Standard computer codes and correlations were employed to compute various parameters. Initiating events in reactivity-induced accidents involve various modes of reactivity insertion, namely, start-up accident, accidental drop of a fuel element on the core, flooding of a beam tube with water, and removal of an in-pile experiment during reactor operation. For each of these transients, time histories of reactor power, energy released, temperature, and reactivity were determined.« less

  12. Comprehensive Fuel Spray Modeling and Impacts on Chamber Acoustics in Combustion Dynamics Simulations

    DTIC Science & Technology

    2013-05-01

    multiple swirler configurations and fuel injector locations at atmospheric pressure con- ditions. Both single-element and multiple-element LDI...the swirl number, Reynolds’ number and injector location in the LDI element. Besides the multi-phase flow characteristics, several experimen- tal...region downstream of the fuel injector on account of a sta- ble and compact precessing vortex core. Recent ex- periments conducted by the Purdue group have

  13. Issues and Potential Program on Denatured Fuel Utilization.

    DTIC Science & Technology

    1978-12-01

    HTGR fuel develop - ment program ; 4. coated particles of (U,Th)02 have been extensively tested as potential HTGR fuels . A detailed summary of the...current scrap and waste treatment requirements. dBase case for all HTGR (Prismatic Fuel Element) cases based on data in "Summary Program Plan...Alternate Program for HTGR Fuel Recycle," April 11, 1975, Draft. 19 a --- AC8NCi09 The principal factors that result in a nominally-higher cost for

  14. High density dispersion fuel

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Hofman, G.L.

    1996-09-01

    A fuel development campaign that results in an aluminum plate-type fuel of unlimited LEU burnup capability with an uranium loading of 9 grams per cm{sup 3} of meat should be considered an unqualified success. The current worldwide approved and accepted highest loading is 4.8 g cm{sup {minus}3} with U{sub 3}Si{sub 2} as fuel. High-density uranium compounds offer no real density advantage over U{sub 3}Si{sub 2} and have less desirable fabrication and performance characteristics as well. Of the higher-density compounds, U{sub 3}Si has approximately a 30% higher uranium density but the density of the U{sub 6}X compounds would yield the factormore » 1.5 needed to achieve 9 g cm{sup {minus}3} uranium loading. Unfortunately, irradiation tests proved these peritectic compounds have poor swelling behavior. It is for this reason that the authors are turning to uranium alloys. The reason pure uranium was not seriously considered as a dispersion fuel is mainly due to its high rate of growth and swelling at low temperatures. This problem was solved at least for relatively low burnup application in non-dispersion fuel elements with small additions of Si, Fe, and Al. This so called adjusted uranium has nearly the same density as pure {alpha}-uranium and it seems prudent to reconsider this alloy as a dispersant. Further modifications of uranium metal to achieve higher burnup swelling stability involve stabilization of the cubic {gamma} phase at low temperatures where normally {alpha} phase exists. Several low neutron capture cross section elements such as Zr, Nb, Ti and Mo accomplish this in various degrees. The challenge is to produce a suitable form of fuel powder and develop a plate fabrication procedure, as well as obtain high burnup capability through irradiation testing.« less

  15. Requirements to the procedure and stages of innovative fuel development

    NASA Astrophysics Data System (ADS)

    Troyanov, V.; Zabudko, L.; Grachyov, A.; Zhdanova, O.

    2016-04-01

    According to the accepted current understanding under the nuclear fuel we will consider the assembled active zone unit (Fuel assembly) with its structural elements, fuel rods, pellet column, structural materials of fuel rods and fuel assemblies. The licensing process includes justification of safe application of the proposed modifications, including design-basis and experimental justification of the modified items under normal operating conditions and in violation of normal conditions, including accidents as well. Besides the justification of modified units itself, it is required to show the influence of modifications on the performance and safety of the other Reactor Unit’ and Nuclear Plant’ elements (e.g. burst can detection system, transportation and processing operations during fuel handling), as well as to justify the new standards of fuel storage etc. Finally, the modified fuel should comply with the applicable regulations, which often becomes a very difficult task, if only because those regulations, such as the NP-082-07, are not covered modification issues. Making amendments into regulations can be considered as the only solution, but the process is complicated and requires deep grounds for amendments. Some aspects of licensing new nuclear fuel are considered the example of mixed nitride uranium -plutonium fuel application for the BREST reactor unit.

  16. Trace elements by instrumental neutron activation analysis for pollution monitoring

    NASA Technical Reports Server (NTRS)

    Sheibley, D. W.

    1975-01-01

    Methods and technology were developed to analyze 1000 samples/yr of coal and other pollution-related samples. The complete trace element analysis of 20-24 samples/wk averaged 3-3.5 man-hours/sample. The computerized data reduction scheme could identify and report data on as many as 56 elements. In addition to coal, samples of fly ash, bottom ash, crude oil, fuel oil, residual oil, gasoline, jet fuel, kerosene, filtered air particulates, ore, stack scrubber water, clam tissue, crab shells, river sediment and water, and corn were analyzed. Precision of the method was plus or minus 25% based on all elements reported in coal and other sample matrices. Overall accuracy was estimated at 50%.

  17. Discrete Element Model for Simulations of Early-Life Thermal Fracturing Behaviors in Ceramic Nuclear Fuel Pellets

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Hai Huang; Ben Spencer; Jason Hales

    2014-10-01

    A discrete element Model (DEM) representation of coupled solid mechanics/fracturing and heat conduction processes has been developed and applied to explicitly simulate the random initiations and subsequent propagations of interacting thermal cracks in a ceramic nuclear fuel pellet during initial rise to power and during power cycles. The DEM model clearly predicts realistic early-life crack patterns including both radial cracks and circumferential cracks. Simulation results clearly demonstrate the formation of radial cracks during the initial power rise, and formation of circumferential cracks as the power is ramped down. In these simulations, additional early-life power cycles do not lead to themore » formation of new thermal cracks. They do, however clearly indicate changes in the apertures of thermal cracks during later power cycles due to thermal expansion and shrinkage. The number of radial cracks increases with increasing power, which is consistent with the experimental observations.« less

  18. FACILITY LAYOUT OF FUEL STORAGE BUILDING (CPP603) SHOWING STORAGE BASINS, ...

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

    FACILITY LAYOUT OF FUEL STORAGE BUILDING (CPP-603) SHOWING STORAGE BASINS, FUEL ELEMENT CUTTING FACILITY, AND DRY GRAPHITE STORAGE FACILITY. INL DRAWING NUMBER 200-0603-00-030-056329. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  19. Salt transport extraction of transuranium elements from LWR fuel

    DOEpatents

    Pierce, R.D.; Ackerman, J.P.; Battles, J.E.; Johnson, T.R.; Miller, W.E.

    1992-11-03

    A process is described for separating transuranium actinide values from uranium values present in spent nuclear oxide fuels which contain rare earth and noble metal fission products. The oxide fuel is reduced with Ca metal in the presence of CaCl[sub 2] and a Cu--Mg alloy containing not less than about 25% by weight Mg at a temperature in the range of from about 750 C to about 850 C to precipitate uranium metal and some of the noble metal fission products leaving the Cu--Mg alloy having transuranium actinide metals and rare earth fission product metals and some of the noble metal fission products dissolved therein. The CaCl[sub 2] having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein is separated and electrolytically treated with a carbon electrode to reduce the CaO to Ca metal while converting the carbon electrode to CO and CO[sub 2]. The Ca metal and CaCl[sub 2] is recycled to reduce additional oxide fuel. The Cu--Mg alloy having transuranium metals and rare earth fission product metals and the noble metal fission products dissolved therein is contacted with a transport salt including MgCl[sub 2] to transfer Mg values from the transport salt to the Cu--Mg alloy while transuranium actinide and rare earth fission product metals transfer from the Cu--Mg alloy to the transport salt. Then the transport salt is mixed with a Mg--Zn alloy to transfer Mg values from the alloy to the transport salt while the transuranium actinide and rare earth fission product values dissolved in the salt are reduced and transferred to the Mg--Zn alloy. 2 figs.

  20. Salt transport extraction of transuranium elements from lwr fuel

    DOEpatents

    Pierce, R. Dean; Ackerman, John P.; Battles, James E.; Johnson, Terry R.; Miller, William E.

    1992-01-01

    A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels which contain rare earth and noble metal fission products. The oxide fuel is reduced with Ca metal in the presence of CaCl.sub.2 and a Cu--Mg alloy containing not less than about 25% by weight Mg at a temperature in the range of from about 750.degree. C. to about 850.degree. C. to precipitate uranium metal and some of the noble metal fission products leaving the Cu--Mg alloy having transuranium actinide metals and rare earth fission product metals and some of the noble metal fission products dissolved therein. The CaCl.sub.2 having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein is separated and electrolytically treated with a carbon electrode to reduce the CaO to Ca metal while converting the carbon electrode to CO and CO.sub.2. The Ca metal and CaCl.sub.2 is recycled to reduce additional oxide fuel. The Cu--Mg alloy having transuranium metals and rare earth fission product metals and the noble metal fission products dissolved therein is contacted with a transport salt including Mg Cl.sub.2 to transfer Mg values from the transport salt to the Cu--Mg alloy while transuranium actinide and rare earth fission product metals transfer from the Cu--Mg alloy to the transport salt. Then the transport salt is mixed with a Mg--Zn alloy to transfer Mg values from the alloy to the transport salt while the transuranium actinide and rare earth fission product values dissolved in the salt are reduced and transferred to the Mg--Zn alloy.

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