National Library of Energy BETA

Sample records for reactor components shield

  1. MEANS FOR SHIELDING REACTORS

    DOE Patents [OSTI]

    Garrison, W.M.; McClinton, L.T.; Burton, M.

    1959-03-10

    A reactor of the heterageneous, heavy water moderated type is described. The reactor is comprised of a plurality of vertically disposed fuel element tubes extending through a tank of heavy water moderator and adapted to accommodate a flow of coolant water in contact with the fuel elements. A tank containing outgoing coolant water is disposed above the core to function is a radiation shield. Unsaturated liquid hydrocarbon is floated on top of the water in the shield tank to reduce to a minimum the possibility of the occurrence of explosive gaseous mixtures resulting from the neutron bombardment of the water in the shield tank.

  2. Nuclear reactor shield including magnesium oxide

    DOE Patents [OSTI]

    Rouse, Carl A. (Del Mar, CA); Simnad, Massoud T. (La Jolla, CA)

    1981-01-01

    An improvement in nuclear reactor shielding of a type used in reactor applications involving significant amounts of fast neutron flux, the reactor shielding including means providing structural support, neutron moderator material, neutron absorber material and other components as described below, wherein at least a portion of the neutron moderator material is magnesium in the form of magnesium oxide either alone or in combination with other moderator materials such as graphite and iron.

  3. NEUTRONIC REACTOR SHIELD AND SPACER CONSTRUCTION

    DOE Patents [OSTI]

    Wigner, E.P.; Ohlinger, L.A.

    1958-11-18

    Reactors of the heterogeneous, graphite moderated, fluid cooled type and shielding and spacing plugs for the coolant channels thereof are reported. In this design, the coolant passages extend horizontally through the moderator structure, accommodating the fuel elements in abutting end-to-end relationship, and have access openings through the outer shield at one face of the reactor to facilitate loading of the fuel elements. In the outer ends of the channels which extend through the shields are provided spacers and shielding plugs designed to offer minimal reslstance to coolant fluid flow while preventing emanation of harmful radiation through the access openings when closed between loadings.

  4. MEANS FOR SHIELDING AND COOLING REACTORS

    DOE Patents [OSTI]

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

    1959-02-10

    Reactors of the water-cooled type and a means for shielding such a rcactor to protect operating personnel from harmful radiation are discussed. In this reactor coolant tubes which contain the fissionable material extend vertically through a mass of moderator. Liquid coolant enters through the bottom of the coolant tubes and passes upwardly over the fissionable material. A shield tank is disposed over the top of the reactor and communicates through its bottom with the upper end of the coolant tubes. A hydrocarbon shielding fluid floats on the coolant within the shield tank. With this arrangements the upper face of the reactor can be opened to the atmosphere through the two superimposed liquid layers. A principal feature of the invention is that in the event radioactive fission products enter thc coolant stream. imposed layer of hydrocarbon reduces the intense radioactivity introduced into the layer over the reactors and permits removal of the offending fuel material by personnel shielded by the uncontaminated hydrocarbon layer.

  5. Space Reactor Radiation Shield Design Summary, for Information

    SciTech Connect (OSTI)

    EC Pheil

    2006-02-17

    The purpose of this letter is to provide a summary of the Prometheus space reactor radiation shield design status at the time of program restructuring.

  6. Recovery Act Workers Clear Reactor Shields from Brookhaven Lab

    Broader source: Energy.gov [DOE]

    American Recovery and Reinvestment Act workers are in the final stage of decommissioning a nuclear reactor after they recently removed thick steel shields once used to absorb neutrons produced for...

  7. Shielded fluid stream injector for particle bed reactor

    DOE Patents [OSTI]

    Notestein, John E. (Morgantown, WV)

    1993-01-01

    A shielded fluid-stream injector assembly is provided for particle bed reactors. The assembly includes a perforated pipe injector disposed across the particle bed region of the reactor and an inverted V-shaped shield placed over the pipe, overlapping it to prevent descending particles from coming into direct contact with the pipe. The pipe and shield are fixedly secured at one end to the reactor wall and slidably secured at the other end to compensate for thermal expansion. An axially extending housing aligned with the pipe and outside the reactor and an in-line reamer are provided for removing deposits from the inside of the pipe. The assembly enables fluid streams to be injected and distributed uniformly into the particle bed with minimized clogging of injector ports. The same design may also be used for extraction of fluid streams from particle bed reactors.

  8. Recovery Act Workers Clear Reactor Shields from Brookhaven Lab

    Office of Environmental Management (EM)

    UPTON, N.Y. - American Recovery and Reinvestment Act workers are in the final stage of decommissioning a nuclear reactor after they recently removed thick steel shields once used to absorb neutrons produced for research. The Brookhaven National Laboratory is using $39 million from the Recovery Act to decommission the Brookhaven Graphite Research Reactor, the world's first reactor built solely for peaceful research purposes. The decommissioning is slated for completion later this year and will

  9. Neutron shielding panels for reactor pressure vessels

    DOE Patents [OSTI]

    Singleton, Norman R. (Murrysville, PA)

    2011-11-22

    In a nuclear reactor neutron panels varying in thickness in the circumferential direction are disposed at spaced circumferential locations around the reactor core so that the greatest radial thickness is at the point of highest fluence with lesser thicknesses at adjacent locations where the fluence level is lower. The neutron panels are disposed between the core barrel and the interior of the reactor vessel to maintain radiation exposure to the vessel within acceptable limits.

  10. Superconducting shielded core reactor with reduced AC losses

    DOE Patents [OSTI]

    Cha, Yung S.; Hull, John R.

    2006-04-04

    A superconducting shielded core reactor (SSCR) operates as a passive device for limiting excessive AC current in a circuit operating at a high power level under a fault condition such as shorting. The SSCR includes a ferromagnetic core which may be either closed or open (with an air gap) and extends into and through a superconducting tube or superconducting rings arranged in a stacked array. First and second series connected copper coils each disposed about a portion of the iron core are connected to the circuit to be protected and are respectively wound inside and outside of the superconducting tube or rings. A large impedance is inserted into the circuit by the core when the shielding capability of the superconducting arrangement is exceeded by the applied magnetic field generated by the two coils under a fault condition to limit the AC current in the circuit. The proposed SSCR also affords reduced AC loss compared to conventional SSCRs under continuous normal operation.

  11. Solid tags for identifying failed reactor components

    DOE Patents [OSTI]

    Bunch, Wilbur L. (Richland, WA); Schenter, Robert E. (Richland, WA)

    1987-01-01

    A solid tag material which generates stable detectable, identifiable, and measurable isotopic gases on exposure to a neutron flux to be placed in a nuclear reactor component, particularly a fuel element, in order to identify the reactor component in event of its failure. Several tag materials consisting of salts which generate a multiplicity of gaseous isotopes in predetermined ratios are used to identify different reactor components.

  12. Nuclear reactor having a polyhedral primary shield and removable vessel insulation

    DOE Patents [OSTI]

    Ekeroth, Douglas E. (Delmont, PA); Orr, Richard (Pittsburgh, PA)

    1993-01-01

    A nuclear reactor is provided having a generally cylindrical reactor vessel disposed within an opening in a primary shield. The opening in the primary shield is defined by a plurality of generally planar side walls forming a generally polyhedral-shaped opening. The reactor vessel is supported within the opening in the primary shield by reactor vessel supports which are in communication and aligned with central portions of some of the side walls. The reactor vessel is connected to the central portions of the reactor vessel supports. A thermal insulation polyhedron formed from a plurality of slidably insertable and removable generally planar insulation panels substantially surrounds at least a portion of the reactor vessel and is disposed between the reactor vessel and the side walls of the primary shield. The shape of the insulation polyhedron generally corresponds to the shape of the opening in the primary shield. Reactor monitoring instrumentation may be mounted in the corners of the opening in the primary shield between the side walls and the reactor vessel such that insulation is not disposed between the instrumentation and the reactor vessel.

  13. Nuclear reactor having a polyhedral primary shield and removable vessel insulation

    DOE Patents [OSTI]

    Ekeroth, D.E.; Orr, R.

    1993-12-07

    A nuclear reactor is provided having a generally cylindrical reactor vessel disposed within an opening in a primary shield. The opening in the primary shield is defined by a plurality of generally planar side walls forming a generally polyhedral-shaped opening. The reactor vessel is supported within the opening in the primary shield by reactor vessel supports which are in communication and aligned with central portions of some of the side walls. The reactor vessel is connected to the central portions of the reactor vessel supports. A thermal insulation polyhedron formed from a plurality of slidably insertable and removable generally planar insulation panels substantially surrounds at least a portion of the reactor vessel and is disposed between the reactor vessel and the side walls of the primary shield. The shape of the insulation polyhedron generally corresponds to the shape of the opening in the primary shield. Reactor monitoring instrumentation may be mounted in the corners of the opening in the primary shield between the side walls and the reactor vessel such that insulation is not disposed between the instrumentation and the reactor vessel. 5 figures.

  14. Energy deposition in STARFIRE reactor components

    SciTech Connect (OSTI)

    Gohar, Y.; Brooks, J.N.

    1985-04-01

    The energy deposition in the STARFIRE commercial tokamak reactor was calculated based on detailed models for the different reactor components. The heat deposition and the 14 MeV neutron flux poloidal distributions in the first wall were obtained. The poloidal surface heat load distribution in the first wall was calculated from the plasma radiation. The Monte Carlo method was used for the calculation to allow an accurate modeling for the reactor geometry.

  15. CHARACTERIZATION OF RADIOACTIVITY IN THE REACTOR VESSEL OF THE HEAVY WATER COMPONENT TEST REACTOR

    SciTech Connect (OSTI)

    Vinson, Dennis

    2010-06-01

    The Heavy Water Component Test Reactor (HWCTR) facility is a pressurized heavy water reactor that was used to test candidate fuel designs for heavy water power reactors. The reactor operated at nominal power of 50 MW{sub th}. The reactor coolant loop operated at 1200 psig and 250 C. Two isolated test loop were designed into the reactor to provide special test conditions. Fig. 1 shows a cut-away view of the reactor. The two loops are contained in four inch diameter stainless steel piping. The HWCTR was operated for only a short duration, from March 1962 to December 1964 in order to test the viability of test fuel elements and other reactor components for use in a heavy water power reactor. The reactor achieved 13,882 MWd of total power while testing 36 different fuel assemblies. In the course of operation, HWCTR experienced the cladding failures of 10 separate test fuel assemblies. In each case, the cladding was breached with some release of fuel core material into the isolated test loop, causing fission product and actinide contamination in the main coolant loop and the liquid and boiling test loops. Despite the contribution of the contamination from the failed fuel, the primary source of radioactivity in the HWCTR vessel and internals is the activation products in the thermal shields, and to a lesser degree, activation products in the reactor vessel walls and liner. A detailed facility characterization report of the HWCTR facility was completed in 1996. Many of the inputs and assumptions in the 1996 characterization report were derived from the HWCTR decommissioning plan published in 1975. The current paper provides an updated assessment of the radioisotopic characteristics of the HWCTR vessel and internals to support decommissioning activities on the facility.

  16. ITER (International Thermonuclear Experimental Reactor) shield and blanket work package report

    SciTech Connect (OSTI)

    Not Available

    1988-06-01

    This report summarizes nuclear-related work in support of the US effort for the International Thermonuclear Experimental Reactor (ITER) Study. The purpose of this work was to prepare for the first international ITER workshop devoted to defining a basic ITER concept that will serve as a basis for an indepth conceptual design activity over the next 2-1/2 years. Primary tasks carried out during the past year included: design improvements of the inboard shield developed for the TIBER concept, scoping studies of a variety of tritium breeding blanket options, development of necessary design guidelines and evaluation criteria for the blanket options, further safety considerations related to nuclear components and issues regarding structural materials for an ITER device. 44 refs., 31 figs., 29 tabs.

  17. US ITER (International Thermonuclear Experimental Reactor) shield and blanket design activities

    SciTech Connect (OSTI)

    Baker, C.C.

    1988-08-01

    This paper summarizes nuclear-related work in support of the US effort for the International Thermonuclear Experimental Reactor (ITER) Study. Primary tasks carried out during the past year include design improvements of the inboard shield developed for the TIBER concept, scoping studies of a variety of tritium breeding blanket options, development of necessary design guidelines and evaluation criteria for the blanket options, further safety considerations related to nuclear components, and issues regarding structural materials for an ITER device. The blanket concepts considered are the aqueous/Li salt solution, a water-cooled, solid breeder blanket, a helium-cooled, solid-breeder blanket, a blanket cooled by helium containing lithium-bearing particulates, and a blanket concept based on breeding tritium from He/sup 3/. 1 ref., 2 tabs.

  18. Heavy Water Components Test Reactor Decommissioning - Major Component Removal

    SciTech Connect (OSTI)

    Austin, W.; Brinkley, D.

    2010-05-05

    The Heavy Water Components Test Reactor (HWCTR) facility (Figure 1) was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR facility is on high, well-drained ground, about 30 meters above the water table. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. It was not a defense-related facility like the materials production reactors at SRS. The reactor was moderated with heavy water and was rated at 50 megawatts thermal power. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In 1965, fuel assemblies were removed, systems that contained heavy water were drained, fluid piping systems were drained, deenergized and disconnected and the spent fuel basin was drained and dried. The doors of the reactor facility were shut and it wasn't until 10 years later that decommissioning plans were considered and ultimately postponed due to budget constraints. In the early 1990s, DOE began planning to decommission HWCTR again. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. The $1.6 billion allocation from the American Recovery and Reinvestment Act to SRS for site clean up at SRS has opened the doors to the HWCTR again - this time for final decommissioning. During the lifetime of HWCTR, 36 different fuel assemblies were tested in the facility. Ten of these experienced cladding failures as operational capabilities of the different designs were being established. In addition, numerous spills of heavy water occurred within the facility. Currently, radiation and radioactive contamination levels are low within HWCTR with most of the radioactivity contained within the reactor vessel. There are no known insults to the environment, however with the increasing deterioration of the facility, the possibility exists that contamination could spread outside the facility if it is not decommissioned. An interior panoramic view of the ground floor elevation taken in August 2009 is shown in Figure 2. The foreground shows the transfer coffin followed by the reactor vessel and control rod drive platform in the center. Behind the reactor vessel is the fuel pool. Above the ground level are the polar crane and the emergency deluge tank at the top of the dome. Note the considerable rust and degradation of the components and the interior of the containment building. Alternative studies have concluded that the most environmentally safe, cost effective option for final decommissioning is to remove the reactor vessel, steam generators, and all equipment above grade including the dome. Characterization studies along with transport models have concluded that the remaining below grade equipment that is left in place including the transfer coffin will not contribute any significant contamination to the environment in the future. The below grade space will be grouted in place. A concrete cover will be placed over the remaining footprint and the groundwater will be monitored for an indefinite period to ensure compliance with environmental regulations. The schedule for completion of decommissioning is late FY2011. This paper describes the concepts planned in order to remove the major components including the dome, the reactor vessel (RV), the two steam generators (SG), and relocating the transfer coffin (TC).

  19. Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

    DOE Patents [OSTI]

    Schreiber, Roger B. (Penn Twp., PA); Fero, Arnold H. (New Kensington, PA); Sejvar, James (Murrysville, PA)

    1997-01-01

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel to form a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor.

  20. Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

    DOE Patents [OSTI]

    Schreiber, R.B.; Fero, A.H.; Sejvar, J.

    1997-12-16

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel to form a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor. 8 figs.

  1. HEAVY WATER COMPONENTS TEST REACTOR DECOMMISSIONING

    SciTech Connect (OSTI)

    Austin, W.; Brinkley, D.

    2011-10-13

    The Heavy Water Components Test Reactor (HWCTR) Decommissioning Project was initiated in 2009 as a Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) Removal Action with funding from the American Recovery and Reinvestment Act (ARRA). This paper summarizes the history prior to 2009, the major D&D activities, and final end state of the facility at completion of decommissioning in June 2011. The HWCTR facility was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In the early 1990s, DOE began planning to decommission HWCTR. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. In 2009 the $1.6 billion allocation from the ARRA to SRS for site footprint reduction at SRS reopened the doors to HWCTR - this time for final decommissioning. Alternative studies concluded that the most environmentally safe, cost effective option for final decommissioning was to remove the reactor vessel, both steam generators, and all equipment above grade including the dome. The transfer coffin, originally above grade, was to be placed in the cavity vacated by the reactor vessel and the remaining below grade spaces would be grouted. Once all above equipment including the dome was removed, a concrete cover was to be placed over the remaining footprint and the groundwater monitored for an indefinite period to ensure compliance with environmental regulations.

  2. Heat insulating system for a fast reactor shield slab

    DOE Patents [OSTI]

    Kotora, Jr., James (LaGrange Park, IL); Groh, Edward F. (Naperville, IL); Kann, William J. (Park Ridge, IL); Burelbach, James P. (Glen Ellyn, IL)

    1986-01-01

    Improved thermal insulation for a nuclear reactor deck comprising many helical coil springs disposed in generally parallel, side-by-side laterally overlapping or interfitted relationship to one another so as to define a three-dimensional composite having both metal and voids between the metal, and enclosure means for holding the composite to the underside of the deck.

  3. Heat insulating system for a fast reactor shield slab

    DOE Patents [OSTI]

    Kotora, J. Jr.; Groh, E.F.; Kann, W.J.; Burelbach, J.P.

    1984-04-10

    Improved thermal insulation for a nuclear reactor deck comprises many helical coil springs disposed in generally parallel, side-by-side laterally overlapping or interfitted relationship to one another so as to define a three-dimensional composite having both metal and voids between the metal, and enclosure means for holding the composite to the underside of the deck.

  4. Nuclear reactor removable radial shielding assembly having a self-bowing feature

    DOE Patents [OSTI]

    Pennell, William E. (Greensburg, PA); Kalinowski, Joseph E. (Smithton, PA); Waldby, Robert N. (New Stanton, PA); Rylatt, John A. (Monroeville, PA); Swenson, Daniel V. (Greensburg, PA)

    1978-01-01

    A removable radial shielding assembly for use in the periphery of the core of a liquid-metal-cooled fast-breeder reactor, for closing interassembly gaps in the reactor core assembly load plane prior to reactor criticality and power operation to prevent positive reactivity insertion. The assembly has a lower nozzle portion for inserting into the core support and a flexible heat-sensitive bimetallic central spine surrounded by blocks of shielding material. At refueling temperature and below the spine is relaxed and in a vertical position so that the tolerances permitted by the interassembly gaps allow removal and replacement of the various reactor core assemblies. During an increase in reactor temperature from refueling to hot standby, the bimetallic spine expands, bowing the assembly toward the core center line, exerting a radially inward gap-closing-force on the above core load plane of the reactor core assembly, closing load plane interassembly gaps throughout the core prior to startup and preventing positive reactivity insertion.

  5. Technological Assessment of Plasma Facing Components for DEMO Reactors |

    Office of Environmental Management (EM)

    Department of Energy Technological Assessment of Plasma Facing Components for DEMO Reactors Technological Assessment of Plasma Facing Components for DEMO Reactors Presentation from the 34th Tritium Focus Group Meeting held in Idaho Falls, Idaho on September 23-25, 2014. PDF icon Technological Assessment of Plasma Facing Components for DEMO Reactors More Documents & Publications Tritium Plasma Experiment and Its Role in PHENIX Program Tritium research activities in Safety and Tritium

  6. Component failures that lead to reactor scrams. [PWR; BWR

    SciTech Connect (OSTI)

    Burns, E. T.; Wilson, R. J.; Lim, E. Y.

    1980-04-01

    This report summarizes the operating experience scram data compiled from 35 operating US light water reactors (LWRs) to identify the principal components/systems related to reactor scrams. The data base utilized to identify the scram causes is developed from a EPRI-utility sponsored survey conducted by SAI coupled with recent data from the USNRC Gray Books. The reactor population considered in this evaluation is limited to 23 PWRs and 12 BWRs because of the limited scope of the program. The population includes all the US NSSS vendors. It is judged that this population accurately characterizes the component-related scrams in LWRs over the first 10 years of plant operation.

  7. Austenitic alloy and reactor components made thereof

    DOE Patents [OSTI]

    Bates, John F. (Ogden, UT); Brager, Howard R. (Richland, WA); Korenko, Michael K. (Wexford, PA)

    1986-01-01

    An austenitic stainless steel alloy is disclosed, having excellent fast neutron irradiation swelling resistance and good post irradiation ductility, making it especially useful for liquid metal fast breeder reactor applications. The alloy contains: about 0.04 to 0.09 wt. % carbon; about 1.5 to 2.5 wt. % manganese; about 0.5 to 1.6 wt. % silicon; about 0.030 to 0.08 wt. % phosphorus; about 13.3 to 16.5 wt. % chromium; about 13.7 to 16.0 wt. % nickel; about 1.0 to 3.0 wt. % molybdenum; and about 0.10 to 0.35 wt. % titanium.

  8. Nuclear reactor spacer grid and ductless core component

    DOE Patents [OSTI]

    Christiansen, David W.; Karnesky, Richard A.

    1989-01-01

    The invention relates to a nuclear reactor spacer grid member for use in a liquid cooled nuclear reactor and to a ductless core component employing a plurality of these spacer grid members. The spacer grid member is of the egg-shell type and is constructed so that the walls of the cell members of the grid member are formed of a single thickness of metal to avoid tolerance problems. Within each cell member is a hydraulic spring which laterally constrains the nuclear material bearing rod which passes through each cell member against a hardstop in response to coolant flow through the cell member. This hydraulic spring is also suitable for use in a water cooled nuclear reactor. A core component constructed of, among other components, a plurality of these spacer grid members, avoids the use of a full length duct by providing spacer sleeves about the sodium tubes passing through the spacer grid members at locations between the grid members, thereby maintaining a predetermined space between adjacent grid members.

  9. GFR Sub-Assembly Shielding Design Studies

    SciTech Connect (OSTI)

    J. R. Parry

    2006-01-01

    This report presents the methodology and results for a preliminary study for Gas-Cooled Fast Reactor (GFR) subassembly fast neutron shielding configurations. The purpose of the shielding in the subassembly is to protect reactor components from fast (E>0.1 MeV) neutrons. The subassembly is modeled in MCNP version 5 release 1.30. Parametric studies were performed varying the thickness of the shielding and calculating the fast neutron flux at the vessel head and the core grid plate. This data was used to determine the minimum thickness needed to protect the vessel head and the core grid plate. These thicknesses were used to analyze different shielding configurations incorporating coolant passages and also to estimate the neutron and photon energy deposition in the shielding material.

  10. Nuclear reactor heat transport system component low friction support system

    DOE Patents [OSTI]

    Wade, Elman E. (Ruffs Dale, PA)

    1980-01-01

    A support column for a heavy component of a liquid metal fast breeder reactor heat transport system which will deflect when the pipes leading coolant to and from the heavy component expand or contract due to temperature changes includes a vertically disposed pipe, the pipe being connected to the heavy component by two longitudinally spaced cycloidal dovetail joints wherein the distal end of each of the dovetails constitutes a part of the surface of a large diameter cylinder and the centerlines of these large diameter cylinders intersect at right angles and the pipe being supported through two longitudinally spaced cycloidal dovetail joints wherein the distal end of each of the dovetails constitutes a part of the surface of a large diameter cylinder and the centerlines of these large diameter cylinders intersect at right angles, each of the cylindrical surfaces bearing on a flat and horizontal surface.

  11. Component failures at pressurized water reactors. Final report

    SciTech Connect (OSTI)

    Reisinger, M.F.

    1980-10-01

    Objectives of this study were to identify those systems having major impact on safety and availability (i.e. to identify those systems and components whose failures have historically caused the greatest number of challenges to the reactor protective systems and which have resulted in greatest loss of electric generation time). These problems were identified for engineering solutions and recommendations made for areas and programs where research and development should be concentrated. The program was conducted in three major phases: Data Analysis, Engineering Evaluation, Cost Benefit Analysis.

  12. EM Removes Radioactive Components from Former Reactor at Oak Ridge National

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

    Laboratory | Department of Energy EM Removes Radioactive Components from Former Reactor at Oak Ridge National Laboratory EM Removes Radioactive Components from Former Reactor at Oak Ridge National Laboratory September 24, 2015 - 12:10pm Addthis Federal and contractor employees who worked on the project to remove irradiated components from a reactor pool gather to watch the transport of the shipment offsite for disposition. Federal and contractor employees who worked on the project to remove

  13. Concepts and Tests for the Remote-Controlled Dismantling of the Biological Shield and Form work of the KNK Reactor - 13425

    SciTech Connect (OSTI)

    Neff, Sylvia; Graf, Anja; Petrick, Holger; Rothschmitt, Stefan; Klute, Stefan

    2013-07-01

    The compact sodium-cooled nuclear reactor facility Karlsruhe (KNK), a prototype Fast Breeder, is currently in an advanced stage of dismantling. Complete dismantling is based on 10 partial licensing steps. In the frame of the 9. decommissioning permit, which is currently ongoing, the dismantling of the biological shield is foreseen. The biological shield consists of heavy reinforced concrete with built-in steel fitments, such as form-work of the reactor tank, pipe sleeves, ventilation channels, and measuring devices. Due to the activation of the inner part of the biological shield, dismantling has to be done remote-controlled. During a comprehensive basic design phase a practical dismantling strategy was developed. Necessary equipment and tools were defined. Preliminary tests revealed that hot wire plasma cutting is the most favorable cutting technology due to the geometrical boundary conditions, the varying distance between cutter and material, and the heavy concrete behind the steel form-work. The cutting devices will be operated remotely via a carrier system with an industrial manipulator. The carrier system has expandable claws to adjust to the varying diameter of the reactor shaft during dismantling progress. For design approval of this prototype development, interaction between manipulator and hot wire plasma cutting was tested in a real configuration. For the demolition of the concrete structure, an excavator with appropriate tools, such as a hydraulic hammer, was selected. Other mechanical cutting devices, such as a grinder or rope saw, were eliminated because of concrete containing steel spheres added to increase the shielding factor of the heavy concrete. Dismantling of the biological shield will be done in a ring-wise manner due to static reasons. During the demolition process, the excavator is positioned on its tripod in three concrete recesses made prior to the dismantling of the separate concrete rings. The excavator and the manipulator carrier system will be operated alternately. Main boundary condition for all the newly designed equipment is the decommissioning housing of limited space within the reactor building containment. To allow for a continuous removal of the concrete rubble, an additional opening on the lowest level of the reactor shaft will be made. All equipment and the interaction of the tools have to be tested before use in the controlled area. Therefore a full-scale model of the biological shield will be provided in a mock-up. The tests will be performed in early 2014. The dismantling of the biological shield is scheduled for 2015. (authors)

  14. DISMANTLING OF THE UPPER RPV COMPONENTS OF THE KARLSRUHE MULTI-PURPOSE RESEARCH REACTOR (MZFR), GERMANY

    SciTech Connect (OSTI)

    Prechtl, E.; Suessdorf, W.

    2003-02-27

    The Multi-purpose Research Reactor was a pressurized-water reactor cooled and moderated with heavy water. It was built from 1961 to 1966 and went critical for the first time on 29 September 1965. After nineteen years of successful operation, the reactor was de-activated on 3 May 1984. The reactor had a thermal output of 200 MW and an electrical output of 50 MW. The MZFR not only served to supply electrical power, but also as a test bed for: - research into various materials for reactor building (e. g. zirkaloy), - the manufacturing and operating industry to gain experience in erection and operation, - training scientific and technical reactor staff, and - power supply (first nuclear combined-heat-and-power system, 1979-1984). The experience gained in operating the MZFR was very helpful for the development and operation of power reactors. At first, safe containment and enclosure of the plant was planned, but then it was decided to dismantle the plant completely, step by step, in view o f the clear advantages of this approach. The decommissioning concept for the complete elimination of the plant down to a green-field site provides for eight steps. A separate decommissioning license is required for each step. As part of the dismantling, about 72,000 Mg [metric tons] of concrete and 7,200 Mg of metal (400 Mg RPV) must be removed. About 700 Mg of concrete (500 Mg biological shield) and 1300 Mg of metal must be classified as radioactive waste.

  15. EM Removes Radioactive Components from Former Reactor at Oak...

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

    Employees use long-handled tools to move irradiated components into a container for retrieval. Employees use long-handled tools to move irradiated components into a container for ...

  16. Radiation Shielding Design and Orientation Considerations for a 1 kWe Heat Pipe Cooled Reactor Utilized to Bore Through the Ice Caps of Mars

    SciTech Connect (OSTI)

    Fensin, Michael L.; Elliott, John O.; Lipinski, Ronald J.; Poston, David I.

    2006-01-20

    The goal in designing any space power system is to develop a system able to meet the mission requirements for success while minimizing the overall costs. The mission requirements for the this study was to develop a reactor (with Stirling engine power conversion) and shielding configuration able to fit, along with all the other necessary science equipment, in a Cryobot 3 m high with {approx}0.5 m diameter hull, produce 1 kWe for 5yrs, and not adversely affect the mission science by keeping the total integrated dose to the science equipment below 150 krad. Since in most space power missions the overall system mass dictates the mission cost, the shielding designs in this study incorporated Martian water extracted at the startup site in order to minimize the tungsten and LiH mass loading at launch. Different reliability and mass minimization concerns led to three design configuration evolutions. With the help of implementing Martian water and configuring the reactor as far from the science equipment as possible, the needed tungsten and LiH shield mass was minimized. This study further characterizes the startup dose and the necessary mission requirements in order to ensure integrity of the surface equipment during reactor startup phase.

  17. An Assessment of Remote Visual Methods to Detect Cracking in Reactor Components

    SciTech Connect (OSTI)

    Cumblidge, Stephen E.; Anderson, Michael T.; Doctor, Steven R.; Simonen, Fredric A.; Elliot, Anthony J.

    2008-01-01

    Recently, the U.S. nuclear industry has proposed replacing current volumetric and/or surface examinations of certain components in commercial nuclear power plants, as required by the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, “Inservice Inspection of Nuclear Power Plant Components,” with a simpler visual testing (VT) method. The advantages of VT are that these tests generally involve much less radiation exposure and time to perform the examination than do volumetric examinations such as ultrasonic testing. The issues relative to the reliability of VT in determining the structural integrity of reactor components were examined. Some piping and pressure vessel components in a nuclear power station are examined using VT as they are either in high radiation fields or component geometry precludes the use of ultrasonic testing (UT) methodology. Remote VT with radiation-hardened video systems has been used by nuclear utilities to find cracks in pressure vessel cladding in pressurized water reactors, core shrouds in boiling water reactors, and to investigate leaks in piping and reactor components. These visual tests are performed using a wide variety of procedures and equipment. The techniques for remote VT use submersible closed-circuit video cameras to examine reactor components and welds. PNNL conducted a parametric study that examined the important variables influencing the effectiveness of a remote visual test. Tested variables included lighting techniques, camera resolution, camera movement, and magnification. PNNL also conducted a limited laboratory test using a commercial visual testing camera system to experimentally determine the ability of the camera system to detect cracks of various widths under ideal conditions. The results of these studies and their implications are presented in this paper.

  18. Nuclear heat source component design considerations for HTGR process heat reactor plant concept

    SciTech Connect (OSTI)

    McDonald, C.F.; Kapich, D.; King, J.H.; Venkatesh, M.C.

    1982-05-01

    The coupling of a high-temperature gas-cooled reactor (HTGR) and a chemical process facility has the potential for long-term synthetic fuel production (i.e., oil, gasoline, aviation fuel, hydrogen, etc) using coal as the carbon source. Studies are in progress to exploit the high-temperature capability of an advanced HTGR variant for nuclear process heat. The process heat plant discussed in this paper has a 1170-MW(t) reactor as the heat source and the concept is based on indirect reforming, i.e., the high-temperature nuclear thermal energy is transported (via an intermediate heat exchanger (IHX)) to the externally located process plant by a secondary helium transport loop. Emphasis is placed on design considerations for the major nuclear heat source (NHS) components, and discussions are presented for the reactor core, prestressed concrete reactor vessel (PCRV), rotating machinery, and heat exchangers.

  19. Progress Towards Prognostic Health Management of Passive Components in Advanced Small Modular Reactors

    SciTech Connect (OSTI)

    Meyer, Ryan M.; Ramuhalli, Pradeep; Hirt, Evelyn H.; Pardini, Allan F.; Suter, Jonathan D.; Prowant, Matthew S.

    2014-08-01

    Sustainable nuclear power to promote energy security and to reduce greenhouse gas emissions are two key national energy priorities. The development of deployable small modular reactors (SMRs) is expected to support these objectives by developing technologies that improve the reliability, sustain safety, and improve affordability of new reactors. Advanced SMRs (AdvSMRs) refer to a specific class of SMRs and are based on modularization of advanced reactor concepts. Prognostic health management (PHM) systems can benefit both the safety and economics of deploying AdvSMRs and can play an essential role in managing the inspection and maintenance of passive components in AdvSMR systems. This paper describes progress on development of a prototypic PHM system for AdvSMR passive components, with thermal creep chosen as the target degradation mechanism.

  20. The Evaluation of Lithium Hydride for Use in a Space Nuclear Reactor Shield, Including a Historical Perspective

    SciTech Connect (OSTI)

    D. Poeth

    2005-12-09

    LiH was one of the five primary shield materials the NRPCT intended to develop (along with beryllium, boron carbide, tungsten, and water) for potential Prometheus application. It was also anticipated that {sup 10}B metal would be investigated for feasibility at a low level of effort. LiH historically has been selected as a low mass, neutron absorption material for space shields (Systems for Nuclear Auxiliary Power (SNAP), Topaz, SP-100). Initial NRPCT investigations did not produce convincing evidence that LiH was desirable or feasible for a Prometheus mission due to material property issues (primarily swelling and hydrogen cover gas containment), and related thermal design complexity. Furthermore, if mass limits allowed, an option to avoid use of LiH was being contemplated to lower development costs and associated risks. However, LiH remains theoretically the most efficient neutron shield material per unit mass, and, with sufficient testing and development, could be an optimal material choice for future flights.

  1. Reactor Safety Gap Evaluation of Accident Tolerant Components and Severe Accident Analysis

    SciTech Connect (OSTI)

    Farmer, Mitchell T.; Bunt, R.; Corradini, M.; Ellison, Paul B.; Francis, M.; Gabor, John D.; Gauntt, R.; Henry, C.; Linthicum, R.; Luangdilok, W.; Lutz, R.; Paik, C.; Plys, M.; Rabiti, Cristian; Rempe, J.; Robb, K.; Wachowiak, R.

    2015-01-31

    The overall objective of this study was to conduct a technology gap evaluation on accident tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist, given the current state of light water reactor (LWR) severe accident research, and additionally augmented by insights obtained from the Fukushima accident. The ultimate benefit of this activity is that the results can be used to refine the Department of Energy’s (DOE) Reactor Safety Technology (RST) research and development (R&D) program plan to address key knowledge gaps in severe accident phenomena and analyses that affect reactor safety and that are not currently being addressed by the industry or the Nuclear Regulatory Commission (NRC).

  2. Swelling in light water reactor internal components: Insights from computational modeling

    SciTech Connect (OSTI)

    Stoller, Roger E.; Barashev, Alexander V.; Golubov, Stanislav I.

    2015-08-01

    A modern cluster dynamics model has been used to investigate the materials and irradiation parameters that control microstructural evolution under the relatively low-temperature exposure conditions that are representative of the operating environment for in-core light water reactor components. The focus is on components fabricated from austenitic stainless steel. The model accounts for the synergistic interaction between radiation-produced vacancies and the helium that is produced by nuclear transmutation reactions. Cavity nucleation rates are shown to be relatively high in this temperature regime (275 to 325C), but are sensitive to assumptions about the fine scale microstructure produced under low-temperature irradiation. The cavity nucleation rates observed run counter to the expectation that void swelling would not occur under these conditions. This expectation was based on previous research on void swelling in austenitic steels in fast reactors. This misleading impression arose primarily from an absence of relevant data. The results of the computational modeling are generally consistent with recent data obtained by examining ex-service components. However, it has been shown that the sensitivity of the model s predictions of low-temperature swelling behavior to assumptions about the primary damage source term and specification of the mean-field sink strengths is somewhat greater that that observed at higher temperatures. Further assessment of the mathematical model is underway to meet the long-term objective of this research, which is to provide a predictive model of void swelling at relevant lifetime exposures to support extended reactor operations.

  3. Adhesive particle shielding

    DOE Patents [OSTI]

    Klebanoff, Leonard Elliott (Dublin, CA); Rader, Daniel John (Albuquerque, NM); Walton, Christopher (Berkeley, CA); Folta, James (Livermore, CA)

    2009-01-06

    An efficient device for capturing fast moving particles has an adhesive particle shield that includes (i) a mounting panel and (ii) a film that is attached to the mounting panel wherein the outer surface of the film has an adhesive coating disposed thereon to capture particles contacting the outer surface. The shield can be employed to maintain a substantially particle free environment such as in photolithographic systems having critical surfaces, such as wafers, masks, and optics and in the tools used to make these components, that are sensitive to particle contamination. The shield can be portable to be positioned in hard-to-reach areas of a photolithography machine. The adhesive particle shield can incorporate cooling means to attract particles via the thermophoresis effect.

  4. Modular shield

    DOE Patents [OSTI]

    Snyder, Keith W. (Sandia Park, NM)

    2002-01-01

    A modular system for containing projectiles has a sheet of material including at least a polycarbonate layer held by a metal frame having a straight frame member corresponding to each straight edge of the sheet. Each frame member has a U-shaped shield channel covering and holding a straight edge of the sheet and an adjacent U-shaped clamp channel rigidly held against the shield channel. A flexible gasket separates each sheet edge from its respective shield channel; and each frame member is fastened to each adjacent frame member only by clamps extending between adjacent clamp channels.

  5. Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility

    SciTech Connect (OSTI)

    S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

    2008-04-01

    A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: Identifies pre-conceptual design requirements Develops test loop equipment schematics and layout Identifies space allocations for each of the facility functions, as required Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems Identifies pre-conceptual utility and support system needs Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs.

  6. Radiation shielding composition

    DOE Patents [OSTI]

    Quapp, W.J.; Lessing, P.A.

    1998-07-28

    A composition is disclosed for use as a radiation shield. The shield is a concrete product containing a stable uranium aggregate for attenuating gamma rays and a neutron absorbing component, the uranium aggregate and neutron absorbing component being present in the concrete product in sufficient amounts to provide a concrete having a density between about 4 and about 15 grams/cm{sup 3} and which will at a predetermined thickness, attenuate gamma rays and absorb neutrons from a radioactive material of projected gamma ray and neutron emissions over a determined time period. The composition is preferably in the form of a container for storing radioactive materials that emit gamma rays and neutrons. The concrete container preferably comprises a metal liner and/or a metal outer shell. The resulting radiation shielding container has the potential of being structurally sound, stable over a long period of time, and, if desired, readily mobile. 5 figs.

  7. Radiation shielding composition

    DOE Patents [OSTI]

    Quapp, William J. (Idaho Falls, ID); Lessing, Paul A. (Idaho Falls, ID)

    2000-12-26

    A composition for use as a radiation shield. The shield is a concrete product containing a stable uranium aggregate for attenuating gamma rays and a neutron absorbing component, the uranium aggregate and neutron absorbing component being present in the concrete product in sufficient amounts to provide a concrete having a density between about 4 and about 15 grams/cm.sup.3 and which will at a predetermined thickness, attenuate gamma rays and absorb neutrons from a radioactive material of projected gamma ray and neutron emissions over a determined time period. The composition is preferably in the form of a container for storing radioactive materials that emit gamma rays and neutrons. The concrete container preferably comprises a metal liner and/or a metal outer shell. The resulting radiation shielding container has the potential of being structurally sound, stable over a long period of time, and, if desired, readily mobile.

  8. Radiation shielding composition

    DOE Patents [OSTI]

    Quapp, William J. (Idaho Falls, ID); Lessing, Paul A. (Idaho Falls, ID)

    1998-01-01

    A composition for use as a radiation shield. The shield is a concrete product containing a stable uranium aggregate for attenuating gamma rays and a neutron absorbing component, the uranium aggregate and neutron absorbing component being present in the concrete product in sufficient amounts to provide a concrete having a density between about 4 and about 15 grams/cm.sup.3 and which will at a predetermined thickness, attenuate gamma rays and absorb neutrons from a radioactive material of projected gamma ray and neutron emissions over a determined time period. The composition is preferably in the form of a container for storing radioactive materials that emit gamma rays and neutrons. The concrete container preferably comprises a metal liner and/or a metal outer shell. The resulting radiation shielding container has the potential of being structurally sound, stable over a long period of time, and, if desired, readily mobile.

  9. A Review of Stress Corrosion Cracking/Fatigue Modeling for Light Water Reactor Cooling System Components

    Broader source: Energy.gov [DOE]

    In the United States currently there are approximately 104 operating light water reactors. Of these, 69 are pressurized water reactors (PWRs) and 35 are boiling water reactors (BWRs). In 2007, the...

  10. Technical Needs for Prototypic Prognostic Technique Demonstration for Advanced Small Modular Reactor Passive Components

    SciTech Connect (OSTI)

    Meyer, Ryan M.; Coble, Jamie B.; Hirt, Evelyn H.; Ramuhalli, Pradeep; Mitchell, Mark R.; Wootan, David W.; Berglin, Eric J.; Bond, Leonard J.; Henager, Charles H.

    2013-05-17

    This report identifies a number of requirements for prognostics health management of passive systems in AdvSMRs, documents technical gaps in establishing a prototypical prognostic methodology for this purpose, and describes a preliminary research plan for addressing these technical gaps. AdvSMRs span multiple concepts; therefore a technology- and design-neutral approach is taken, with the focus being on characteristics that are likely to be common to all or several AdvSMR concepts. An evaluation of available literature is used to identify proposed concepts for AdvSMRs along with likely operational characteristics. Available operating experience of advanced reactors is used in identifying passive components that may be subject to degradation, materials likely to be used for these components, and potential modes of degradation of these components. This information helps in assessing measurement needs for PHM systems, as well as defining functional requirements of PHM systems. An assessment of current state-of-the-art approaches to measurements, sensors and instrumentation, diagnostics and prognostics is also documented. This state-of-the-art evaluation, combined with the requirements, may be used to identify technical gaps and research needs in the development, evaluation, and deployment of PHM systems for AdvSMRs. A preliminary research plan to address high-priority research needs for the deployment of PHM systems to AdvSMRs is described, with the objective being the demonstration of prototypic prognostics technology for passive components in AdvSMRs. Greater efficiency in achieving this objective can be gained through judicious selection of materials and degradation modes that are relevant to proposed AdvSMR concepts, and for which significant knowledge already exists. These selections were made based on multiple constraints including the analysis performed in this document, ready access to laboratory-scale facilities for materials testing and measurement, and potential synergies with other national laboratory and university partners.

  11. Thermocouple shield

    DOE Patents [OSTI]

    Ripley, Edward B. (Knoxville, TN)

    2009-11-24

    A thermocouple shield for use in radio frequency fields. In some embodiments the shield includes an electrically conductive tube that houses a standard thermocouple having a thermocouple junction. The electrically conductive tube protects the thermocouple from damage by an RF (including microwave) field and mitigates erroneous temperature readings due to the microwave or RF field. The thermocouple may be surrounded by a ceramic sheath to further protect the thermocouple. The ceramic sheath is generally formed from a material that is transparent to the wavelength of the microwave or RF energy. The microwave transparency property precludes heating of the ceramic sheath due to microwave coupling, which could affect the accuracy of temperature measurements. The ceramic sheath material is typically an electrically insulating material. The electrically insulative properties of the ceramic sheath help avert electrical arcing, which could damage the thermocouple junction. The electrically conductive tube is generally disposed around the thermocouple junction and disposed around at least a portion of the ceramic sheath. The concepts of the thermocouple shield may be incorporated into an integrated shielded thermocouple assembly.

  12. Physics-Based Multi-State Models of Passive Component Degradation for the R7 Reactor Simulation Environment

    SciTech Connect (OSTI)

    Unwin, Stephen D.; Layton, Robert F.; Johnson, Kenneth I.; Lowry, Peter P.

    2012-06-25

    Abstract: The Next Generation Systems Analysis Code - referred to as R7 - is reactor systems simulation software being developed to support the Risk-Informed Safety Margin Characterization Pathway of the U.S. Department of Energy's Light Water Reactor Sustainability Program. It will provide an integrated multi-physics environment, implemented in an uncertainty quantification (UQ) framework that can produce risk and other performance insights on long-term reactor operations. An element of this simulation environment will be the performance of passive components and materials. Conventional models of component reliability are largely parametric, relying on plant service data to estimate component lifetimes and failure rates. This type of model has limited usefulness in the R7 environment where the intent is to explicitly determine the influence of physical stressors on component degradation. In this paper, we describe a new class of multi-state physics-based component models designed to be R7-compatible. These models capture the physics of materials degradation while also incorporating the effects of interventions and component rejuvenation. The models are implemented in a cumulative damage framework that allows the impact of an evolving physical environment to be addressed without recourse to resampling within the Monte Carlo-based UQ framework. The paper describes an application to stress corrosion cracking in dissimilar metal welds - a principal contributor to potential loss of coolant accidents. So while R7 will have the more conventional capability of reactor simulation codes to model the impact of degraded components and systems on plant performance, the methodology described here allows R7 to model the inverse effect; the impact of the physical environment on component degradation and performance.

  13. System Engineering Program Applicability for the High Temperature Gas-Cooled Reactor (HTGR) Component Test Capability (CTC)

    SciTech Connect (OSTI)

    Jeffrey Bryan

    2009-06-01

    This white paper identifies where the technical management and systems engineering processes and activities to be used in establishing the High Temperature Gas-cooled Reactor (HTGR) Component Test Capability (CTC) should be addressed and presents specific considerations for these activities under each CTC alternative

  14. Fundamental Understanding of Crack Growth in Structural Components of Generation IV Supercritical Light Water Reactors

    SciTech Connect (OSTI)

    Iouri I. Balachov; Takao Kobayashi; Francis Tanzella; Indira Jayaweera; Palitha Jayaweera; Petri Kinnunen; Martin Bojinov; Timo Saario

    2004-11-17

    This work contributes to the design of safe and economical Generation-IV Super-Critical Water Reactors (SCWRs) by providing a basis for selecting structural materials to ensure the functionality of in-vessel components during the entire service life. During the second year of the project, we completed electrochemical characterization of the oxide film properties and investigation of crack initiation and propagation for candidate structural materials steels under supercritical conditions. We ranked candidate alloys against their susceptibility to environmentally assisted degradation based on the in situ data measure with an SRI-designed controlled distance electrochemistry (CDE) arrangement. A correlation between measurable oxide film properties and susceptibility of austenitic steels to environmentally assisted degradation was observed experimentally. One of the major practical results of the present work is the experimentally proven ability of the economical CDE technique to supply in situ data for ranking candidate structural materials for Generation-IV SCRs. A potential use of the CDE arrangement developed ar SRI for building in situ sensors monitoring water chemistry in the heat transport circuit of Generation-IV SCWRs was evaluated and proved to be feasible.

  15. Drip Shield Emplacement Gantry Concept

    SciTech Connect (OSTI)

    Silva, R.A.; Cron, J.

    2000-03-29

    This design analysis has shown that, on a conceptual level, the emplacement of drip shields is feasible with current technology and equipment. A plan for drip shield emplacement was presented using a Drip Shield Transporter, a Drip Shield Emplacement Gantry, a locomotive, and a Drip Shield Gantry Carrier. The use of a Drip Shield Emplacement Gantry as an emplacement concept results in a system that is simple, reliable, and interfaces with the numerous other exising repository systems. Using the Waste Emplacement/Retrieval System design as a basis for the drip shield emplacement concept proved to simplify the system by using existing equipment, such as the gantry carrier, locomotive, Electrical and Control systems, and many other systems, structures, and components. Restricted working envelopes for the Drip Shield Emplacement System require further consideration and must be addressed to show that the emplacement operations can be performed as the repository design evolves. Section 6.1 describes how the Drip Shield Emplacement System may use existing equipment. Depending on the length of time between the conclusion of waste emplacement and the commencement of drip shield emplacement, this equipment could include the locomotives, the gantry carrier, and the electrical, control, and rail systems. If the exisiting equipment is selected for use in the Drip Shield Emplacement System, then the length of time after the final stages of waste emplacement and start of drip shield emplacement may pose a concern for the life cycle of the system (e.g., reliability, maintainability, availability, etc.). Further investigation should be performed to consider the use of existing equipment for drip shield emplacement operations. Further investigation will also be needed regarding the interfaces and heat transfer and thermal effects aspects. The conceptual design also requires further design development. Although the findings of this analysis are accurate for the assumptions made, further refinements of this analysis are needed as the project parameters change. The designs of the drip shield, the Emplacement Drift, and the other drip shield emplacement equipment all have a direct effect on the overall design feasibility.

  16. The Capabilities and Limitation of Remote Visual Methods to Detect Service-Induced Cracks in Reactor Components

    SciTech Connect (OSTI)

    Cumblidge, Stephen E.; Doctor, Steven R.; Anderson, Michael T.

    2006-11-01

    Since 1977, the U.S. Nuclear Regulatory Commission (NRC) Office of Nuclear Regulatory Research has funded a multiyear program at the Pacific Northwest National Laboratory (PNNL) to evaluate the reliability and accuracy of nondestructive evaluation (NDE) techniques employed for inservice inspection (ISI). Recently, the U.S. nuclear industry proposed replacing current volumetric and/or surface examinations of certain components in commercial nuclear power plants, as required by ASME Boiler and Pressure Vessel Code Section XI, with a simpler visual testing (VT) method. The advantages of VT are that these tests generally involve much less radiation exposure and examination times than do volumetric examinations such as ultrasonic testing (UT). However, for industry to justify supplamenting volumetric metods with VT, and analysis of pertinent issues is needed to support the reliability of VT in determining the structural intefrity of reactor components. As piping and pressure vessel compoents in a nuclear power station are generally underwater and in high radiation field, they need to be examined by VT from a distance with radiation-hardened video systems. Remote visual testing has been used by nuclear utilities to find cracks in pressure vessel cladding in pressurized water reactors, for shrouds in boiling water reactors, and to investigate leaks in piping and reactor components. These visual tests are performed using a wide variety of procedures and equipment. The techniques for remote visual testing use submersible closed-circuit video cameras to examine reactor components and welds. PNNL has conducted a parametric study that examines the important variables that affect the effectiveness of a remote visual test. Tested variables include lighting techniques, camera resolution, camera movement, and magnification. PNNL has also conductrd a laboratory test using a commercial visual testing camera system to experimentally determine the ability of the camera system to detect cracks of various widths under ideal conditions.

  17. Magnetic shielding

    DOE Patents [OSTI]

    Kerns, J.A.; Stone, R.R.; Fabyan, J.

    1985-02-12

    A magnetically-conductive filler material bridges the gap between a multi-part magnetic shield structure which substantially encloses a predetermined volume so as to minimize the ingress or egress of magnetic fields with respect to that volume. The filler material includes a heavy concentration of single-magnetic-domain-sized particles of a magnetically conductive material (e.g. soft iron, carbon steel or the like) dispersed throughout a carrier material which is generally a non-magnetic material that is at least sometimes in a plastic or liquid state. The maximum cross-sectional particle dimension is substantially less than the nominal dimension of the gap to be filled. An epoxy base material (i.e. without any hardening additive) low volatility vacuum greases or the like may be used for the carrier material. The structure is preferably exposed to the expected ambient field while the carrier is in a plastic or liquid state so as to facilitate alignment of the single-magnetic-domain-sized particles with the expected magnetic field lines.

  18. Magnetic shielding

    DOE Patents [OSTI]

    Kerns, J.A.; Stone, R.R.; Fabyan, J.

    1987-10-06

    A magnetically-conductive filler material bridges the gap between a multi-part magnetic shield structure which substantially encloses a predetermined volume so as to minimize the ingress or egress of magnetic fields with respect to that volume. The filler material includes a heavy concentration of single-magnetic-domain-sized particles of a magnetically conductive material (e.g. soft iron, carbon steel or the like) dispersed throughout a carrier material which is generally a non-magnetic material that is at least sometimes in a plastic or liquid state. The maximum cross-sectional particle dimension is substantially less than the nominal dimension of the gap to be filled. An epoxy base material (i.e. without any hardening additive) low volatility vacuum greases or the like may be used for the carrier material. The structure is preferably exposed to the expected ambient magnetic field while the carrier is in a plastic or liquid state so as to facilitate alignment of the single-magnetic-domain-sized particles with the expected magnetic field lines. 3 figs.

  19. Magnetic shielding

    DOE Patents [OSTI]

    Kerns, John A. (Livermore, CA); Stone, Roger R. (Walnut Creek, CA); Fabyan, Joseph (Livermore, CA)

    1987-01-01

    A magnetically-conductive filler material bridges the gap between a multi-part magnetic shield structure which substantially encloses a predetermined volume so as to minimize the ingress or egress of magnetic fields with respect to that volume. The filler material includes a heavy concentration of single-magnetic-domain-sized particles of a magnetically conductive material (e.g. soft iron, carbon steel or the like) dispersed throughout a carrier material which is generally a non-magnetic material that is at least sometimes in a plastic or liquid state. The maximum cross-sectional particle dimension is substantially less than the nominal dimension of the gap to be filled. An epoxy base material (i.e. without any hardening additive) low volatility vacuum greases or the like may be used for the carrier material. The structure is preferably exposed to the expected ambient magnetic field while the carrier is in a plastic or liquid state so as to facilitate alignment of the single-magnetic-domain-sized particles with the expected magnetic field lines.

  20. Analytical Study of High Concentration PCB Paint at the Heavy Water Components Test Reactor

    SciTech Connect (OSTI)

    Lowry, N.J.

    1998-10-21

    This report provides results of an analytical study of high concentration PCB paint in a shutdown nuclear test reactor located at the US Department of Energy's Savannah River Site (SRS). The study was designed to obtain data relevant for an evaluation of potential hazards associated with the use of and exposure to such paints.

  1. Mechanical design of core components for a high performance light water reactor with a three pass core

    SciTech Connect (OSTI)

    Fischer, Kai; Schneider, Tobias; Redon, Thomas; Schulenberg, Thomas; Starflinger, Joerg

    2007-07-01

    Nuclear reactors using supercritical water as coolant can achieve more than 500 deg. C core outlet temperature, if the coolant is heated up in three steps with intermediate mixing to avoid hot streaks. This method reduces the peak cladding temperatures significantly compared with a single heat up. The paper presents an innovative mechanical design which has been developed recently for such a High Performance Light Water Reactor. The core is built with square assemblies of 40 fuel pins each, using wire wraps as grid spacers. Nine of these assemblies are combined to a cluster having a common head piece and a common foot piece. A downward flow of additional moderator water, separated from the coolant, is provided in gaps between the assemblies and in a water box inside each assembly. The cluster head and foot pieces and mixing chambers, which are key components for this design, are explained in detail. (authors)

  2. Early test facilities and analytic methods for radiation shielding: Proceedings

    SciTech Connect (OSTI)

    Ingersoll, D.T. ); Ingersoll, J.K. )

    1992-11-01

    This report represents a compilation of eight papers presented at the 1992 American Nuclear Society/European Nuclear Society International Meeting. The meeting is of special significance since it commemorates the fiftieth anniversary of the first controlled nuclear chain reaction. The papers contained in this report were presented in a special session organized by the Radiation Protection and Shielding Division in keeping with the historical theme of the meeting. The paper titles are good indicators of their content and are: (1) The origin of radiation shielding research: The Oak Ridge experience, (2) Shielding research at the hanford site, (3) Aircraft shielding experiments at General Dynamics Fort Worth, 1950-1962, (4) Where have the neutrons gone , a history of the tower shielding facility, (5) History and evolution of buildup factors, (6) Early shielding research at Bettis atomic power laboratory, (7) UK reactor shielding: then and now, (8) A very personal view of the development of radiation shielding theory.

  3. Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.

    SciTech Connect (OSTI)

    Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division

    2007-06-30

    This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition from normal high-flow operation to natural circulation. Low-flow coolant events are the most difficult to design for because they involve the most complex thermal-hydraulic behavior induced by the dominance of thermal-buoyancy forces acting on the coolants. Such behavior can cause multiple-component flow interaction phenomena, which are not adequately understood or appreciated by reactor designers as to their impact on reactor performance and safety. Since the early 1990s, when DOE canceled the U.S. Liquid Metal Fast Breeder Reactor (LMFBR) program, little has been done experimentally to further understand the importance of the complex thermal-buoyancy phenomena and their impact on reactor design or to improve the ability of three-dimensional (3-D) transient computational fluid dynamics (CFD) and structures codes to model the phenomena. An improved experimental data base and the associated improved validated codes would provide needed design tools to the reactor community. The improved codes would also facilitate scale-up from small-scale testing to prototype size and would facilitate comparing performance of one reactor/component design with another. The codes would also have relevance to the design and safety of water-cooled reactors. To accomplish the preceding, it is proposed to establish a national GNEP-LMR research and development center at Argonne having as its foundation state-of-art science-based infrastructure consisting of: (a) thermal-hydraulic experimental capabilities for conducting both water and sodium testing of individual reactor components and complete reactor in-vessel models and (b) a computational modeling development and validation capability that is strongly interfaced with the experimental facilities. The proposed center would greatly advance capabilities for reactor development by establishing the validity of high-fidelity (i.e., close to first principles) models and tools. Such tools could be used directly for reactor design or for qualifying/tuning of lower-fidelity models, which now require costly experimental qualification for each different type of design

  4. Apparatus, components and operating methods for circulating fluidized bed transport gasifiers and reactors

    DOE Patents [OSTI]

    Vimalchand, Pannalal; Liu, Guohai; Peng, Wan Wang

    2015-02-24

    The improvements proposed in this invention provide a reliable apparatus and method to gasify low rank coals in a class of pressurized circulating fluidized bed reactors termed "transport gasifier." The embodiments overcome a number of operability and reliability problems with existing gasifiers. The systems and methods address issues related to distribution of gasification agent without the use of internals, management of heat release to avoid any agglomeration and clinker formation, specific design of bends to withstand the highly erosive environment due to high solid particles circulation rates, design of a standpipe cyclone to withstand high temperature gasification environment, compact design of seal-leg that can handle high mass solids flux, design of nozzles that eliminate plugging, uniform aeration of large diameter Standpipe, oxidant injection at the cyclone exits to effectively modulate gasifier exit temperature and reduction in overall height of the gasifier with a modified non-mechanical valve.

  5. The Tower Shielding Facility: Its glorious past

    SciTech Connect (OSTI)

    Muckenthaler, F.J.

    1997-05-07

    The Tower Shielding Facility (TSF) is the only reactor facility in the US that was designed and built for radiation-shielding studies in which both the reactor source and shield samples could be raised into the air to allow measurements to be made without interference from ground scattering or other spurious effects. The TSF proved its usefulness as many different programs were successfully completed. It became active in work for the Defense Atomic Support Agency (DASA) Space Nuclear Auxiliary Power, Defense Nuclear Agency, Liquid Metal Fast Breeder Reactor Program, the Gas-Cooled and High-Temperature Gas-Cooled Reactor programs, and the Japanese-American Shielding Program of Experimental Research, just to mention a few of the more extensive ones. The history of the TSF as presented in this report describes the various experiments that were performed using the different reactors. The experiments are categorized as to the programs which they supported and placed in corresponding chapters. The experiments are described in modest detail, along with their purpose when appropriate. Discussion of the results is minimal, but references are given to more extensive topical reports.

  6. REACTORS

    DOE Patents [OSTI]

    Spitzer, L. Jr.

    1961-10-01

    Thermonuclear reactors, methods, and apparatus are described for controlling and confining high temperature plasma. Main axial confining coils in combination with helical windings provide a rotational transform that avoids the necessity of a figure-eight shaped reactor tube. The helical windings provide a multipolar helical magnetic field transverse to the axis of the main axial confining coils so as to improve the effectiveness of the confining field by counteracting the tendency of the more central lines of force in the stellarator tube to exchange positions with the magnetic lines of force nearer the walls of the tube. (AEC)

  7. Balance of Plant System Analysis and Component Design of Turbo-Machinery for High Temperature Gas Reactor Systems

    SciTech Connect (OSTI)

    Ballinger, Ronald G.; Wang, Chun Yun; Kadak, Andrew; Todreas, Neil; Mirick, Bradley; Demetri, Eli; Koronowski, Martin

    2004-08-30

    The Modular Pebble Bed Reactor system (MPBR) requires a gas turbine cycle (Brayton cycle) as the power conversion system for it to achieve economic competitiveness as a Generation IV nuclear system. The availability of controllable helium turbomachinery and compact heat exchangers are thus the critical enabling technology for the gas turbine cycle. The development of an initial reference design for an indirect helium cycle has been accomplished with the overriding constraint that this design could be built with existing technology and complies with all current codes and standards. Using the initial reference design, limiting features were identified. Finally, an optimized reference design was developed by identifying key advances in the technology that could reasonably be expected to be achieved with limited R&D. This final reference design is an indirect, intercooled and recuperated cycle consisting of a three-shaft arrangement for the turbomachinery system. A critical part of the design process involved the interaction between individual component design and overall plant performance. The helium cycle overall efficiency is significantly influenced by performance of individual components. Changes in the design of one component, a turbine for example, often required changes in other components. To allow for the optimization of the overall design with these interdependencies, a detailed steady state and transient control model was developed. The use of the steady state and transient models as a part of an iterative design process represents a key contribution of this work. A dynamic model, MPBRSim, has been developed. The model integrates the reactor core and the power conversion system simultaneously. Physical parameters such as the heat exchangers; weights and practical performance maps such as the turbine characteristics and compressor characteristics are incorporated into the model. The individual component models as well as the fully integrated model of the power conversion system have been verified with an industry-standard general thermal-fluid code Flownet. With respect to the dynamic model, bypass valve control and inventory control have been used as the primary control methods for the power conversion system. By performing simulation using the dynamic model with the designed control scheme, the combination of bypass and inventory control was optimized to assure system stability within design temperature and pressure limits. Bypass control allows for rapid control system response while inventory control allows for ultimate steady state operation at part power very near the optimum operating point for the system. Load transients simulations show that the indirect, three-shaft arrangement gas turbine power conversion system is stable and controllable. For the indirect cycle the intermediate heat exchanger (IHX) is the interface between the reactor and the turbomachinery systems. As a part of the design effort the IHX was identified as the key component in the system. Two technologies, printed circuit and compact plate-fin, were investigated that have the promise of meeting the design requirements for the system. The reference design incorporates the possibility of using either technology although the compact plate-fin design was chosen for subsequent analysis. The thermal design and parametric analysis with an IHX and recuperator using the plate-fin configuration have been performed. As a three-shaft arrangement, the turbo-shaft sets consist of a pair of turbine/compressor sets (high pressure and low pressure turbines with same-shaft compressor) and a power turbine coupled with a synchronous generator. The turbines and compressors are all axial type and the shaft configuration is horizontal. The core outlet/inlet temperatures are 900/520 C, and the optimum pressure ratio in the power conversion cycle is 2.9. The design achieves a plant net efficiency of approximately 48%.

  8. Method for producing components with internal architectures, such as micro-channel reactors, via diffusion bonding sheets

    DOE Patents [OSTI]

    Alman, David E.; Wilson, Rick D.; Davis, Daniel L.

    2011-03-08

    This invention relates to a method for producing components with internal architectures, and more particularly, this invention relates to a method for producing structures with microchannels via the use of diffusion bonding of stacked laminates. Specifically, the method involves weakly bonding a stack of laminates forming internal voids and channels with a first generally low uniaxial pressure and first temperature such that bonding at least between the asperites of opposing laminates occurs and pores are isolated in interfacial contact areas, followed by a second generally higher isostatic pressure and second temperature for final bonding. The method thereby allows fabrication of micro-channel devices such as heat exchangers, recuperators, heat-pumps, chemical separators, chemical reactors, fuel processing units, and combustors without limitation on the fin aspect ratio.

  9. Rotating shielded crane system

    DOE Patents [OSTI]

    Commander, John C. (Idaho Falls, ID)

    1988-01-01

    A rotating, radiation shielded crane system for use in a high radiation test cell, comprises a radiation shielding wall, a cylindrical ceiling made of radiation shielding material and a rotatable crane disposed above the ceiling. The ceiling rests on an annular ledge intergrally attached to the inner surface of the shielding wall. Removable plugs in the ceiling provide access for the crane from the top of the ceiling into the test cell. A seal is provided at the interface between the inner surface of the shielding wall and the ceiling.

  10. The role of trace impurities in classification of in-core reactor components. Final report

    SciTech Connect (OSTI)

    Luksic, A.T.; Pauley, K.A.; Reid, B.d.

    1993-08-01

    Low-level radioactive waste is classified into three distinct categories under Title 10 of the Code of Federal Regulations Part 61 (10CFR61). These categories, Class A, B, and C, are determined by the concentrations of specified radionuclides, with Class C being the most restrictive. Waste which is determined to exceed the Class C limits is referred to as Greater Than Class C (GTCC) and is generally deemed to be unacceptable for disposal in a shallow land burial site. Under the Low-Level Waste Policy Act Amendments of 1985 (Public Law 99-240), the responsibility for disposal of waste classified as GTCC rests with the US government. To provide a better understanding of the elemental constituents that are present in nuclear fuel related materials, 41 samples of unirradiated materials used in the manufacture of fuel assembly hardware and nonfuel assembly components were obtained and analyzed. Particular emphasis was placed on evaluating the concentration of niobium in Zircaloy. The results of this work indicate that insignificant amounts of niobium are found in the current generation of Zircaloy. However, the reason for the inconsistency between the trace amounts of elemental niobium (<10 ppM) observed in the current analysis and the significant amounts (>100 ppM) that were observed in the previous analysis has not been identified.

  11. Large Component Removal/Disposal

    SciTech Connect (OSTI)

    Wheeler, D. M.

    2002-02-27

    This paper describes the removal and disposal of the large components from Maine Yankee Atomic Power Plant. The large components discussed include the three steam generators, pressurizer, and reactor pressure vessel. Two separate Exemption Requests, which included radiological characterizations, shielding evaluations, structural evaluations and transportation plans, were prepared and issued to the DOT for approval to ship these components; the first was for the three steam generators and one pressurizer, the second was for the reactor pressure vessel. Both Exemption Requests were submitted to the DOT in November 1999. The DOT approved the Exemption Requests in May and July of 2000, respectively. The steam generators and pressurizer have been removed from Maine Yankee and shipped to the processing facility. They were removed from Maine Yankee's Containment Building, loaded onto specially designed skid assemblies, transported onto two separate barges, tied down to the barges, th en shipped 2750 miles to Memphis, Tennessee for processing. The Reactor Pressure Vessel Removal Project is currently under way and scheduled to be completed by Fall of 2002. The planning, preparation and removal of these large components has required extensive efforts in planning and implementation on the part of all parties involved.

  12. Gamma ray detector shield

    DOE Patents [OSTI]

    Ohlinger, R.D.; Humphrey, H.W.

    1985-08-26

    A gamma ray detector shield comprised of a rigid, lead, cylindrical-shaped vessel having upper and lower portions with an pneumatically driven, sliding top assembly. Disposed inside the lead shield is a gamma ray scintillation crystal detector. Access to the gamma detector is through the sliding top assembly.

  13. Cosmic Ray Interactions in Shielding Materials

    SciTech Connect (OSTI)

    Aguayo Navarrete, Estanislao; Kouzes, Richard T.; Ankney, Austin S.; Orrell, John L.; Berguson, Timothy J.; Troy, Meredith D.

    2011-09-08

    This document provides a detailed study of materials used to shield against the hadronic particles from cosmic ray showers at Earths surface. This work was motivated by the need for a shield that minimizes activation of the enriched germanium during transport for the MAJORANA collaboration. The materials suitable for cosmic-ray shield design are materials such as lead and iron that will stop the primary protons, and materials like polyethylene, borated polyethylene, concrete and water that will stop the induced neutrons. The interaction of the different cosmic-ray components at ground level (protons, neutrons, muons) with their wide energy range (from kilo-electron volts to giga-electron volts) is a complex calculation. Monte Carlo calculations have proven to be a suitable tool for the simulation of nucleon transport, including hadron interactions and radioactive isotope production. The industry standard Monte Carlo simulation tool, Geant4, was used for this study. The result of this study is the assertion that activation at Earths surface is a result of the neutronic and protonic components of the cosmic-ray shower. The best material to shield against these cosmic-ray components is iron, which has the best combination of primary shielding and minimal secondary neutron production.

  14. HWMA/RCRA CLOSURE PLAN FOR THE MATERIALS TEST REACTOR WING (TRA-604) LABORATORY COMPONENTS VOLUNTARY CONSENT ORDER ACTION PLAN VCO-5.8 D REVISION2

    SciTech Connect (OSTI)

    KIRK WINTERHOLLER

    2008-02-25

    This Hazardous Waste Management Act/Resource Conservation and Recovery Act closure plan was developed for the laboratory components of the Test Reactor Area Catch Tank System (TRA-630) that are located in the Materials Test Reactor Wing (TRA-604) at the Reactor Technology Complex, Idaho National Laboratory Site, to meet a further milestone established under Voluntary Consent Order Action Plan VCO-5.8.d. The TRA-604 laboratory components addressed in this closure plan were deferred from the TRA-630 Catch Tank System closure plan due to ongoing laboratory operations in the areas requiring closure actions. The TRA-604 laboratory components include the TRA-604 laboratory warm wastewater drain piping, undersink drains, subheaders, and the east TRA-604 laboratory drain header. Potentially contaminated surfaces located beneath the TRA-604 laboratory warm wastewater drain piping and beneath the island sinks located in Laboratories 126 and 128 (located in TRA-661) are also addressed in this closure plan. The TRA-604 laboratory components will be closed in accordance with the interim status requirements of the Hazardous Waste Management Act/Resource Conservation and Recovery Act as implemented by the Idaho Administrative Procedures Act 58.01.05.009 and 40 Code of Federal Regulations 265, Subparts G and J. This closure plan presents the closure performance standards and the methods for achieving those standards.

  15. Light Water Reactor Sustainability Technical Documents | Department...

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

    in Light Water Reactors: Life After 60 Nuclear reactors present a very harsh environment for components service. Components within a reactor core must tolerate high...

  16. Technical Letter Report, An Evaluation of Ultrasonic Phased Array Testing for Reactor Piping System Components Containing Dissimilar Metal Welds, JCN N6398, Task 2A

    SciTech Connect (OSTI)

    Diaz, Aaron A.; Cinson, Anthony D.; Crawford, Susan L.; Anderson, Michael T.

    2009-11-30

    Research is being conducted for the U.S. Nuclear Regulatory Commission at the Pacific Northwest National Laboratory to assess the effectiveness and reliability of advanced nondestructive examination (NDE) methods for the inspection of light-water reactor components. The scope of this research encom¬passes primary system pressure boundary materials including dissimilar metal welds (DMWs), cast austenitic stainless steels (CASS), piping with corrosion-resistant cladding, weld overlays, inlays and onlays, and far-side examinations of austenitic piping welds. A primary objective of this work is to evaluate various NDE methods to assess their ability to detect, localize, and size cracks in steel components that challenge standard and/or conventional inspection methodologies. This interim technical letter report provides a summary of a technical evaluation aimed at assessing the capabilities of phased-array (PA) ultrasonic testing (UT) methods as applied to the inspection of small-bore DMW components that exist in the reactor coolant systems (RCS) of pressurized water reactors (PWRs). Operating experience and events such as the circumferential cracking in the reactor vessel nozzle-to-RCS hot leg pipe at V.C. Summer nuclear power station, identified in 2000, show that in PWRs where primary coolant water (or steam) are present under normal operation, Alloy 82/182 materials are susceptible to pressurized water stress corrosion cracking. The extent and number of occurrences of DMW cracking in nuclear power plants (domestically and internationally) indicate the necessity for reliable and effective inspection techniques. The work described herein was performed to provide insights for evaluating the utility of advanced NDE approaches for the inspection of DMW components such as a pressurizer surge nozzle DMW, a shutdown cooling pipe DMW, and a ferritic (low-alloy carbon steel)-to-CASS pipe DMW configuration.

  17. Composition for radiation shielding

    DOE Patents [OSTI]

    Kronberg, James W. (Aiken, SC)

    1994-01-01

    A composition for use as a radiation shield. The shield has a depleted urum core for absorbing gamma rays and a bismuth coating for preventing chemical corrosion and absorbing gamma rays. Alternatively, a sheet of gadolinium may be positioned between the uranium core and the bismuth coating for absorbing neutrons. The composition is preferably in the form of a container for storing materials that emit radiation such as gamma rays and neutrons. The container is preferably formed by casting bismuth around a pre-formed uranium container having a gadolinium sheeting, and allowing the bismuth to cool. The resulting container is a structurally sound, corrosion-resistant, radiation-absorbing container.

  18. Glove box shield

    DOE Patents [OSTI]

    Brackenbush, Larry W. (Richland, WA); Hoenes, Glenn R. (Richland, WA)

    1981-01-01

    According to the present invention, a shield for a glove box housing radioactive material is comprised of spaced apart clamping members which maintain three overlapping flaps in place therebetween. There is a central flap and two side flaps, the side flaps overlapping at the interior edges thereof and the central flap extending past the intersection of the side flaps in order to insure that the shield is always closed when the user withdraws his hand from the glove box. Lead loaded neoprene rubber is the preferred material for the three flaps, the extent of lead loading depending upon the radiation levels within the glove box.

  19. Lightweight blast shield

    DOE Patents [OSTI]

    Mixon, Larry C. (Madison, AL); Snyder, George W. (Huntsville, AL); Hill, Scott D. (Toney, AL); Johnson, Gregory L. (Decatur, AL); Wlodarski, J. Frank (Huntsville, AL); von Spakovsky, Alexis P. (Huntsville, AL); Emerson, John D. (Arab, AL); Cole, James M. (Huntsville, AL); Tipton, John P. (Huntsville, AL)

    1991-01-01

    A tandem warhead missile arrangement that has a composite material housing structure with a first warhead mounted at one end and a second warhead mounted near another end of the composite structure with a dome shaped composite material blast shield mounted between the warheads to protect the second warhead from the blast of the first warhead.

  20. Steam Generator Component Model in a Combined Cycle of Power Conversion Unit for Very High Temperature Gas-Cooled Reactor

    SciTech Connect (OSTI)

    Oh, Chang H; Han, James; Barner, Robert; Sherman, Steven R

    2007-06-01

    The Department of Energy and the Idaho National Laboratory are developing a Next Generation Nuclear Plant (NGNP), Very High Temperature Gas-Cooled Reactor (VHTR) to serve as a demonstration of state-of-the-art nuclear technology. The purpose of the demonstration is two fold 1) efficient low cost energy generation and 2) hydrogen production. Although a next generation plant could be developed as a single-purpose facility, early designs are expected to be dual-purpose. While hydrogen production and advanced energy cycles are still in its early stages of development, research towards coupling a high temperature reactor, electrical generation and hydrogen production is under way. A combined cycle is considered as one of the power conversion units to be coupled to the very high-temperature gas-cooled reactor (VHTR). The combined cycle configuration consists of a Brayton top cycle coupled to a Rankine bottoming cycle by means of a steam generator. A detailed sizing and pressure drop model of a steam generator is not available in the HYSYS processes code. Therefore a four region model was developed for implementation into HYSYS. The focus of this study was the validation of a HYSYS steam generator model of two phase flow correlations. The correlations calculated the size and heat exchange of the steam generator. To assess the model, those calculations were input into a RELAP5 model and its results were compared with HYSYS results. The comparison showed many differences in parameters such as the heat transfer coefficients and revealed the different methods used by the codes. Despite differences in approach, the overall results of heat transfer were in good agreement.

  1. Gas shielding apparatus

    DOE Patents [OSTI]

    Brandt, D.

    1985-12-31

    An apparatus is disclosed for preventing oxidation by uniformly distributing inert shielding gas over the weld area of workpieces such as pipes being welded together. The apparatus comprises a chamber and a gas introduction element. The chamber has an annular top wall, an annular bottom wall, an inner side wall and an outer side wall connecting the top and bottom walls. One side wall is a screen and the other has a portion defining an orifice. The gas introduction element has a portion which encloses the orifice and can be one or more pipes. The gas introduction element is in fluid communication with the chamber and introduces inert shielding gas into the chamber. The inert gas leaves the chamber through the screen side wall and is dispersed evenly over the weld area. 3 figs.

  2. Composition for radiation shielding

    DOE Patents [OSTI]

    Kronberg, J.W.

    1994-08-02

    A composition for use as a radiation shield is disclosed. The shield has a depleted uranium core for absorbing gamma rays and a bismuth coating for preventing chemical corrosion and absorbing gamma rays. Alternatively, a sheet of gadolinium may be positioned between the uranium core and the bismuth coating for absorbing neutrons. The composition is preferably in the form of a container for storing materials that emit radiation such as gamma rays and neutrons. The container is preferably formed by casting bismuth around a pre-formed uranium container having a gadolinium sheeting, and allowing the bismuth to cool. The resulting container is a structurally sound, corrosion-resistant, radiation-absorbing container. 2 figs.

  3. Gas shielding apparatus

    DOE Patents [OSTI]

    Brandt, Daniel (Los Alamos, NM)

    1985-01-01

    An apparatus for preventing oxidation by uniformly distributing inert shielding gas over the weld area of workpieces such as pipes being welded together. The apparatus comprises a chamber and a gas introduction element. The chamber has an annular top wall, an annular bottom wall, an inner side wall and an outer side wall connecting the top and bottom walls. One side wall is a screen and the other has a portion defining an orifice. The gas introduction element has a portion which encloses the orifice and can be one or more pipes. The gas introduction element is in fluid communication with the chamber and introduces inert shielding gas into the chamber. The inert gas leaves the chamber through the screen side wall and is dispersed evenly over the weld area.

  4. Gas shielding apparatus

    DOE Patents [OSTI]

    Brandt, D.

    1984-06-05

    An apparatus for preventing oxidation by uniformly distributing inert shielding gas over the weld area of workpieces such as pipes being welded together. The apparatus comprises a chamber and a gas introduction element. The chamber has an annular top wall, an annular bottom wall, an inner side wall and an outer side wall connecting the top and bottom walls. One side wall is a screen and the other has a portion defining an orifice. The gas introduction element has a portion which encloses the orifice and can be one or more pipes. The gas introduction element is in fluid communication with the chamber and introduces inert shielding gas into the chamber. The inert gas leaves the chamber through the screen side wall and is dispersed evenly over the weld area.

  5. Multilayer radiation shield

    DOE Patents [OSTI]

    Urbahn, John Arthur (Saratoga Springs, NY); Laskaris, Evangelos Trifon (Niskayuna, NY)

    2009-06-16

    A power generation system including: a generator including a rotor including a superconductive rotor coil coupled to a rotatable shaft; a first prime mover drivingly coupled to the rotatable shaft; and a thermal radiation shield, partially surrounding the rotor coil, including at least a first sheet and a second sheet spaced apart from the first sheet by centripetal force produced by the rotatable shaft. A thermal radiation shield for a generator including a rotor including a super-conductive rotor coil including: a first sheet having at least one surface formed from a low emissivity material; and at least one additional sheet having at least one surface formed from a low emissivity material spaced apart from the first sheet by centripetal force produced by the rotatable shaft, wherein each successive sheet is an incrementally greater circumferential arc length and wherein the centripetal force shapes the sheets into a substantially catenary shape.

  6. shields-98.pdf

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    7 The Whole Sky Imager - A Year of Progress J. E. Shields and M. E. Karr Marine Physical Laboratory Scripps Institution of Oceanography University of California, San Diego San Diego, California T. P. Tooman Sandia National Laboratories Livermore, California D. H. Sowle and S. T. Moore Mission Research Corporation Santa Barbara, California Abstract Much progress has been made this last year in realizing the potential of the whole sky imager (WSI). Two imagers are deployed [at the Southern Great

  7. Spallation Neutron Source Radiation Shielding Issues

    SciTech Connect (OSTI)

    Azmy, Y.Y.; Barnes, J.M.; Drischler, J.D.; Johnston, J.O.; Lillie, R.A.; McNeilly, G.S.; Santoro, R.T.

    1999-11-14

    This paper summarizes results of Spallation Neutron Source calculations to estimate radiation hazards and shielding requirements for activated Mercury, target components, target cooling water, and {sup 7}Be plateout. Dose rates in the accelerator tunnel from activation of magnets and concrete were investigated. The impact of gaps and other streaming paths on the radiation environment inside the test cell during operation and after shutdown were also assessed.

  8. 2016 SHIELDS Workshop: Shielding Society from Space Weather

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    April » 2016 SHIELDS Workshop 2016 SHIELDS Workshop: Shielding Society from Space Weather WHEN: Apr 04, 2016 8:00 AM - Apr 08, 2016 5:00 PM WHERE: Drury Plaza Hotel, Santa Fe, NM CONTACT: Gian Luca Delzanno (505) 667-2604 CATEGORY: Science TYPE: Workshop INTERNAL: Calendar Login Event Description Space weather is an emerging research area within space science that is rapidly gaining importance and public recognition because of its technological and societal impact. A recent report from the

  9. Evaluation of aging degradation of structural components

    SciTech Connect (OSTI)

    Chopra, O.K.; Shack, W.J.

    1992-03-01

    Irradiation embrittlement of the neutron shield tank (NST) A212 Grade B steel from the Shippingport reactor, as well as thermal embrittlement of CF-8 cast stainless steel components from the Shippingport and KRB reactors, has been characterized. Increases in Charpy transition temperature (CTT), yield stress, and hardness of the NST material in the low-temperature low-flux environment are consistent with the test reactor data for irradiations at < 232{degrees}C. The shift in CTT is not as severe as that observed in surveillance samples from the High Flux Isotope Reactor (HFIR): however, it shows very good agreement with the results for HFIR A212-B steel irradiated in the Oak Ridge Research Reactor. The results indicate that fluence rate has not effect on radiation embrittlement at rates as low as 2 {times} 10{sup 8} n/cm{sup 2}{center dot}s at the low operating temperature of the Shippingport NST, i.e., 55{degrees}C. This suggest that radiation damage in Shippingport NST and HFIR surveillance samples may be different because of the neutron spectra and/or Cu and Ni content of the two materials. Cast stainless steel components show relatively modest decreases in fracture toughness and Charpy-impact properties and a small increase in tensile strength. Correlations for estimating mechanical properties of cast stainless steels predict accurate or slightly conservative values for Charpy-impact energy, tensile flow stress, fracture toughness J-R curve, and J{sub IC} of the materials. The kinetics of thermal embrittlement and degree of embrittlement at saturation, i.e., the minimum impact energy achieved after long-term aging, were established from materials that were aged further in the laboratory. The results were consistent with the estimates. The correlations successfully predict the mechanical properties of the Ringhals 2 reactor hot- and crossover-leg elbows (CF-8M steel) after service of {approx}15 y.

  10. Evaluation of aging degradation of structural components

    SciTech Connect (OSTI)

    Chopra, O.K.; Shack, W.J.

    1992-03-01

    Irradiation embrittlement of the neutron shield tank (NST) A212 Grade B steel from the Shippingport reactor, as well as thermal embrittlement of CF-8 cast stainless steel components from the Shippingport and KRB reactors, has been characterized. Increases in Charpy transition temperature (CTT), yield stress, and hardness of the NST material in the low-temperature low-flux environment are consistent with the test reactor data for irradiations at < 232{degrees}C. The shift in CTT is not as severe as that observed in surveillance samples from the High Flux Isotope Reactor (HFIR): however, it shows very good agreement with the results for HFIR A212-B steel irradiated in the Oak Ridge Research Reactor. The results indicate that fluence rate has not effect on radiation embrittlement at rates as low as 2 {times} 10{sup 8} n/cm{sup 2}{center_dot}s at the low operating temperature of the Shippingport NST, i.e., 55{degrees}C. This suggest that radiation damage in Shippingport NST and HFIR surveillance samples may be different because of the neutron spectra and/or Cu and Ni content of the two materials. Cast stainless steel components show relatively modest decreases in fracture toughness and Charpy-impact properties and a small increase in tensile strength. Correlations for estimating mechanical properties of cast stainless steels predict accurate or slightly conservative values for Charpy-impact energy, tensile flow stress, fracture toughness J-R curve, and J{sub IC} of the materials. The kinetics of thermal embrittlement and degree of embrittlement at saturation, i.e., the minimum impact energy achieved after long-term aging, were established from materials that were aged further in the laboratory. The results were consistent with the estimates. The correlations successfully predict the mechanical properties of the Ringhals 2 reactor hot- and crossover-leg elbows (CF-8M steel) after service of {approx}15 y.

  11. Actively driven thermal radiation shield

    DOE Patents [OSTI]

    Madden, Norman W. (Livermore, CA); Cork, Christopher P. (Pleasant Hill, CA); Becker, John A. (Alameda, CA); Knapp, David A. (Livermore, CA)

    2002-01-01

    A thermal radiation shield for cooled portable gamma-ray spectrometers. The thermal radiation shield is located intermediate the vacuum enclosure and detector enclosure, is actively driven, and is useful in reducing the heat load to mechanical cooler and additionally extends the lifetime of the mechanical cooler. The thermal shield is electrically-powered and is particularly useful for portable solid-state gamma-ray detectors or spectrometers that dramatically reduces the cooling power requirements. For example, the operating shield at 260K (40K below room temperature) will decrease the thermal radiation load to the detector by 50%, which makes possible portable battery operation for a mechanically cooled Ge spectrometer.

  12. Sterically shielded diboron-containing metallocene olefin polymerization catalysts

    DOE Patents [OSTI]

    Marks, Tobin J. (Evanston, IL); Ja, Li (Chicago, IL); Yang, Xinmin (Evanston, IL)

    1995-09-05

    A non-coordinating anion, preferably containing a sterically shielded diboron hydride, if combined with a cyclopenta-dienyl-substituted metallocene cation component, such as a zirconocene metallocene, is a useful olefin polymerization catalyst component. The anion preferably has the formula ##STR1## where R is branched lower alkyl, such as t-butyl.

  13. Portable convertible blast effects shield

    DOE Patents [OSTI]

    Pastrnak, John W. (Livermore, CA); Hollaway, Rocky (Modesto, CA); Henning, Carl D. (Livermore, CA); Deteresa, Steve (Livermore, CA); Grundler, Walter (Hayward, CA); Hagler,; Lisle B. (Berkeley, CA); Kokko, Edwin (Dublin, CA); Switzer, Vernon A (Livermore, CA)

    2010-10-26

    A rapidly deployable portable convertible blast effects shield/ballistic shield includes a set two or more telescoping cylindrical rings operably connected to each other to convert between a telescopically-collapsed configuration for storage and transport, and a telescopically-extended upright configuration forming an expanded inner volume. In a first embodiment, the upright configuration provides blast effects shielding, such as against blast pressures, shrapnel, and/or fire balls. And in a second embodiment, the upright configuration provides ballistic shielding, such as against incoming weapons fire, shrapnel, etc. Each ring has a high-strength material construction, such as a composite fiber and matrix material, capable of substantially inhibiting blast effects and impinging projectiles from passing through the shield. And the set of rings are releasably securable to each other in the telescopically-extended upright configuration, such as by click locks.

  14. Portable convertible blast effects shield

    DOE Patents [OSTI]

    Pastrnak, John W. (Livermore, CA); Hollaway, Rocky (Modesto, CA); Henning, Carl D. (Livermore, CA); Deteresa, Steve (Livermore, CA); Grundler, Walter (Hayward, CA); Hagler, Lisle B. (Berkeley, CA); Kokko, Edwin (Dublin, CA); Switzer, Vernon A. (Livermore, CA)

    2011-03-15

    A rapidly deployable portable convertible blast effects shield/ballistic shield includes a set two or more frusto-conically-tapered telescoping rings operably connected to each other to convert between a telescopically-collapsed configuration for storage and transport, and a telescopically-extended upright configuration forming an expanded inner volume. In a first embodiment, the upright configuration provides blast effects shielding, such as against blast pressures, shrapnel, and/or fire balls. And in a second embodiment, the upright configuration provides ballistic shielding, such as against incoming weapons fire, shrapnel, etc. Each ring has a high-strength material construction, such as a composite fiber and matrix material, capable of substantially inhibiting blast effects and impinging projectiles from passing through the shield. And the set of rings are releasably securable to each other in the telescopically-extended upright configuration by the friction fit of adjacent pairs of frusto-conically-tapered rings to each other.

  15. Portable convertible blast effects shield

    DOE Patents [OSTI]

    Pastrnak, John W. (Livermore, CA); Hollaway, Rocky (Modesto, CA); Henning, Carl D. (Livermore, CA); Deteresa, Steve (Livermore, CA); Grundler, Walter (Hayward, CA); Hagler, Lisle B. (Berkeley, CA); Kokko, Edwin (Dublin, CA); Switzer, Vernon A (Livermore, CA)

    2007-05-22

    A rapidly deployable portable convertible blast effects shield/ballistic shield includes a set two or more telescoping cylindrical rings operably connected to each other to convert between a telescopically-collapsed configuration for storage and transport, and a telescopically-extended upright configuration forming an expanded inner volume. In a first embodiment, the upright configuration provides blast effects shielding, such as against blast pressures, shrapnel, and/or fire balls. And in a second embodiment, the upright configuration provides ballistic shielding, such as against incoming weapons fire, shrapnel, etc. Each ring has a high-strength material construction, such as a composite fiber and matrix material, capable of substantially inhibiting blast effects and impinging projectiles from passing through the shield. And the set of rings are releasably securable to each other in the telescopically-extended upright configuration, such as by click locks.

  16. Welding shield for coupling heaters

    DOE Patents [OSTI]

    Menotti, James Louis (Dickinson, TX)

    2010-03-09

    Systems for coupling end portions of two elongated heater portions and methods of using such systems to treat a subsurface formation are described herein. A system may include a holding system configured to hold end portions of the two elongated heater portions so that the end portions are abutted together or located near each other; a shield for enclosing the end portions, and one or more inert gas inlets configured to provide at least one inert gas to flush the system with inert gas during welding of the end portions. The shield may be configured to inhibit oxidation during welding that joins the end portions together. The shield may include a hinged door that, when closed, is configured to at least partially isolate the interior of the shield from the atmosphere. The hinged door, when open, is configured to allow access to the interior of the shield.

  17. SHIELDING ANALYSIS FOR PORTABLE GAUGING COMBINATION SOURCES

    SciTech Connect (OSTI)

    J. TOMPKINS; L. LEONARD; ET AL

    2000-08-01

    Radioisotopic decay has been used as a source of photons and neutrons for industrial gauging operations since the late 1950s. Early portable moisture/density gauging equipment used Americium (Am)-241/Beryllium (Be)/Cesium (Cs)-137 combination sources to supply the required nuclear energy for gauging. Combination sources typically contained 0.040 Ci of Am-241 and 0.010 Ci of CS-137 in the same source capsule. Most of these sources were manufactured approximately 30 years ago. Collection, transportation, and storage of these sources once removed from their original device represent a shielding problem with distinct gamma and neutron components. The Off-Site Source Recovery (OSR) Project is planning to use a multi-function drum (MFD) for the collection, shipping, and storage of AmBe sources, as well as the eventual waste package for disposal. The MFD is an approved TRU waste container design for DOE TRU waste known as the 12 inch Pipe Component Overpack. As the name indicates, this drum is based on a 12 inch ID stainless steel weldment approximately 25 inch in internal length. The existing drum design allows for addition of shielding within the pipe component up to the 110 kg maximum pay load weight. The 12 inch pipe component is packaged inside a 55-gallon drum, with the balance of the interior space filled with fiberboard dunnage. This packaging geometry is similar to the design of a DOT 6M, Type B shipping container.

  18. SNF shipping cask shielding analysis

    SciTech Connect (OSTI)

    Johnson, J.O.; Pace, J.V. III

    1996-01-01

    The Waste Management and Remedial Action Division has planned a modification sequence for storage facility 7827 in the Solid Waste Storage Area (SWSA). The modification cycle is: (1) modify an empty caisson, (2) transfer the spent nuclear fuel (SNF) of an occupied caisson to a hot cell in building 3525 for inspection and possible repackaging, and (3) return the package to the modified caisson in the SWSA. Although the SNF to be moved is in the solid form, it has different levels of activity. Thus, the following 5 shipping casks will be available for the task: the Loop Transport Carrier, the In- Pile Loop LITR HB-2 Carrier, the 6.5-inch HRLEL Carrier, the HFIR Hot Scrap Carrier, and the 10-inch ORR Experiment Removal Shield Cask. This report describes the shielding tasks for the 5 casks: determination of shielding characteristics, any streaming avenues, estimation of thermal limits, and shielding calculational uncertainty for use in the transportation plan.

  19. Reflector and Shield Material Properties for Project Prometheus

    SciTech Connect (OSTI)

    J. Nash

    2005-11-02

    This letter provides updated reflector and shield preliminary material property information to support reactor design efforts. The information provided herein supersedes the applicable portions of Revision 1 to the Space Power Program Preliminary Reactor Design Basis (Reference (a)). This letter partially answers the request in Reference (b) to provide unirradiated and irradiated material properties for beryllium, beryllium oxide, isotopically enriched boron carbide ({sup 11}B{sub 4}C) and lithium hydride. With the exception of {sup 11}B{sub 4}C, the information is provided in Attachments 1 and 2. At the time of issuance of this document, {sup 11}B{sub 4}C had not been studied.

  20. Thermal neutron shield and method of manufacture

    DOE Patents [OSTI]

    Metzger, Bert Clayton; Brindza, Paul Daniel

    2014-03-04

    A thermal neutron shield comprising boron shielding panels with a high percentage of the element Boron. The panel is least 46% Boron by weight which maximizes the effectiveness of the shielding against thermal neutrons. The accompanying method discloses the manufacture of boron shielding panels which includes enriching the pre-cursor mixture with varying grit sizes of Boron Carbide.

  1. Radiation shielding materials and containers incorporating same

    DOE Patents [OSTI]

    Mirsky, Steven M. (Greenbelt, MD); Krill, Stephen J. (Arlington, VA); Murray, Alexander P. (Gaithersburg, MD)

    2005-11-01

    An improved radiation shielding material and storage systems for radioactive materials incorporating the same. The PYRolytic Uranium Compound ("PYRUC") shielding material is preferably formed by heat and/or pressure treatment of a precursor material comprising microspheres of a uranium compound, such as uranium dioxide or uranium carbide, and a suitable binder. The PYRUC shielding material provides improved radiation shielding, thermal characteristic, cost and ease of use in comparison with other shielding materials. The shielding material can be used to form containment systems, container vessels, shielding structures, and containment storage areas, all of which can be used to house radioactive waste. The preferred shielding system is in the form of a container for storage, transportation, and disposal of radioactive waste. In addition, improved methods for preparing uranium dioxide and uranium carbide microspheres for use in the radiation shielding materials are also provided.

  2. Radiation Shielding Materials and Containers Incorporating Same

    DOE Patents [OSTI]

    Mirsky, Steven M.; Krill, Stephen J.; and Murray, Alexander P.

    2005-11-01

    An improved radiation shielding material and storage systems for radioactive materials incorporating the same. The PYRolytic Uranium Compound (''PYRUC'') shielding material is preferably formed by heat and/or pressure treatment of a precursor material comprising microspheres of a uranium compound, such as uranium dioxide or uranium carbide, and a suitable binder. The PYRUC shielding material provides improved radiation shielding, thermal characteristic, cost and ease of use in comparison with other shielding materials. The shielding material can be used to form containment systems, container vessels, shielding structures, and containment storage areas, all of which can be used to house radioactive waste. The preferred shielding system is in the form of a container for storage, transportation, and disposal of radioactive waste. In addition, improved methods for preparing uranium dioxide and uranium carbide microspheres for use in the radiation shielding materials are also provided.

  3. Catalytic reactor

    DOE Patents [OSTI]

    Aaron, Timothy Mark; Shah, Minish Mahendra; Jibb, Richard John

    2009-03-10

    A catalytic reactor is provided with one or more reaction zones each formed of set(s) of reaction tubes containing a catalyst to promote chemical reaction within a feed stream. The reaction tubes are of helical configuration and are arranged in a substantially coaxial relationship to form a coil-like structure. Heat exchangers and steam generators can be formed by similar tube arrangements. In such manner, the reaction zone(s) and hence, the reactor is compact and the pressure drop through components is minimized. The resultant compact form has improved heat transfer characteristics and is far easier to thermally insulate than prior art compact reactor designs. Various chemical reactions are contemplated within such coil-like structures such that as steam methane reforming followed by water-gas shift. The coil-like structures can be housed within annular chambers of a cylindrical housing that also provide flow paths for various heat exchange fluids to heat and cool components.

  4. Small Modular Nuclear Reactors: Parametric Modeling of Integrated Reactor

    Energy Savers [EERE]

    Vessel Manufacturing Within a Factory Environment - Volume 1 | Department of Energy 1 Small Modular Nuclear Reactors: Parametric Modeling of Integrated Reactor Vessel Manufacturing Within a Factory Environment - Volume 1 This study focused on the learning process for the factory built components of the Integrated Reactor Vessel of a generic 100MWe SMR using Pressurized Water Reactor Technology. PDF icon Small Modular Nuclear Reactors: Parametric Modeling of Integrated Reactor Vessel

  5. ENGINEERING TEST REACTOR

    DOE Patents [OSTI]

    De Boisblanc, D.R.; Thomas, M.E.; Jones, R.M.; Hanson, G.H.

    1958-10-21

    Heterogeneous reactors of the type which is both cooled and moderated by the same fluid, preferably water, and employs highly enriched fuel are reported. In this design, an inner pressure vessel is located within a main outer pressure vessel. The reactor core and its surrounding reflector are disposed in the inner pressure vessel which in turn is surrounded by a thermal shield, Coolant fluid enters the main pressure vessel, fiows downward into the inner vessel where it passes through the core containing tbe fissionable fuel assemblies and control rods, through the reflector, thence out through the bottom of the inner vessel and up past the thermal shield to the discharge port in the main vessel. The fuel assemblles are arranged in the core in the form of a cross having an opening extending therethrough to serve as a high fast flux test facility.

  6. Materials Degradation in Light Water Reactors: Life After 60

    Broader source: Energy.gov [DOE]

    Nuclear reactors present a very harsh environment for components service. Components within a reactor core must tolerate high temperature water, stress, vibration, and an intense neutron field....

  7. LOADING MACHINE FOR REACTORS

    DOE Patents [OSTI]

    Simon, S.L.

    1959-07-01

    An apparatus is described for loading or charging slugs of fissionable material into a nuclear reactor. The apparatus of the invention is a "muzzle loading" type comprising a delivery tube or muzzle designed to be brought into alignment with any one of a plurality of fuel channels. The delivery tube is located within the pressure shell and it is also disposed within shielding barriers while the fuel cantridges or slugs are forced through the delivery tube by an externally driven flexible ram.

  8. Shielding calculations and verifications for the new Radiation Instrument Calibration Facility at Los Alamos National Laboratory

    SciTech Connect (OSTI)

    George, G. L.; Olsher, R. H.; Seagraves, D. T.

    2002-01-01

    MCNP-4C1 was used to perform the shielding design for the new Central Health Physics Calibration Facility (CHPCF) at Los Alamos National Laboratory (LANL). The problem of shielding the facility was subdivided into three separate components: (1) Transmission; (2) Skyshine; and (3) Maze Streaming/ Transmission. When possible, actual measurements were taken to verify calculation results. The comparison of calculation versus measurement results shows excellent agreement for neutron calculations. For photon comparisons, calculations resulted in conservative estimates of the Effective Dose Equivalent (EDE) compared to measured results. This disagreement in the photon measurements versus calculations is most likely due to several conservative assumptions regarding shield density and composition. For example, reinforcing steel bars (Rebar) in the concrete shield walls were not included in the shield model.

  9. Facility target insert shielding assessment

    SciTech Connect (OSTI)

    Mocko, Michal

    2015-10-06

    Main objective of this report is to assess the basic shielding requirements for the vertical target insert and retrieval port. We used the baseline design for the vertical target insert in our calculations. The insert sits in the 12-diameter cylindrical shaft extending from the service alley in the top floor of the facility all the way down to the target location. The target retrieval mechanism is a long rod with the target assembly attached and running the entire length of the vertical shaft. The insert also houses the helium cooling supply and return lines each with 2 diameter. In the present study we focused on calculating the neutron and photon dose rate fields on top of the target insert/retrieval mechanism in the service alley. Additionally, we studied a few prototypical configurations of the shielding layers in the vertical insert as well as on the top.

  10. Light shield for solar concentrators

    DOE Patents [OSTI]

    Plesniak, Adam P.; Martins, Guy L.

    2014-08-26

    A solar receiver unit including a housing defining a recess, a cell assembly received in the recess, the cell assembly including a solar cell, and a light shield received in the recess and including a body and at least two tabs, the body defining a window therein, the tabs extending outward from the body and being engaged with the recess, wherein the window is aligned with the solar cell.

  11. Nuclear component horizontal seismic restraint

    DOE Patents [OSTI]

    Snyder, Glenn J. (Lynchburg, VA)

    1988-01-01

    A nuclear component horizontal seismic restraint. Small gaps limit horizontal displacement of components during a seismic occurrence and therefore reduce dynamic loadings on the free lower end. The reactor vessel and reactor guard vessel use thicker section roll-forged rings welded between the vessel straight shell sections and the bottom hemispherical head sections. The inside of the reactor guard vessel ring forging contains local vertical dovetail slots and upper ledge pockets to mount and retain field fitted and installed blocks. As an option, the horizontal displacement of the reactor vessel core support cone can be limited by including shop fitted/installed local blocks in opposing alignment with the reactor vessel forged ring. Beams embedded in the wall of the reactor building protrude into apertures in the thermal insulation shell adjacent the reactor guard vessel ring and have motion limit blocks attached thereto to provide to a predetermined clearance between the blocks and reactor guard vessel ring.

  12. Small Modular Nuclear Reactors: Parametric Modeling of Integrated Reactor

    Energy Savers [EERE]

    Vessel Manufacturing Within a Factory Environment - Volume 2 | Department of Energy 2 Small Modular Nuclear Reactors: Parametric Modeling of Integrated Reactor Vessel Manufacturing Within a Factory Environment - Volume 2 This study presents a detailed analysis of the economics of Small Modular Reactors (SMRs), specifically a generic 100MWe conceptual design at the component level. PDF icon Small Modular Nuclear Reactors: Parametric Modeling of Integrated Reactor Vessel Manufacturing Within a

  13. Advanced Nuclear Reactors | Department of Energy

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

    Advanced Nuclear Reactors Advanced Nuclear Reactors Turbulent Flow of Coolant in an Advanced Nuclear Reactor Visualizing Coolant Flow in Sodium Reactor Subassemblies Sodium-cooled Fast Reactor (SFR) Coolant Flow At the heart of a nuclear power plant is the reactor. The fuel assembly is placed inside a reactor vessel where all the nuclear reactions occur to produce the heat and steam used for power generation. Nonetheless, an entire power plant consists of many other support components and key

  14. Waste Package Component Design Methodology Report

    SciTech Connect (OSTI)

    D.C. Mecham

    2004-07-12

    This Executive Summary provides an overview of the methodology being used by the Yucca Mountain Project (YMP) to design waste packages and ancillary components. This summary information is intended for readers with general interest, but also provides technical readers a general framework surrounding a variety of technical details provided in the main body of the report. The purpose of this report is to document and ensure appropriate design methods are used in the design of waste packages and ancillary components (the drip shields and emplacement pallets). The methodology includes identification of necessary design inputs, justification of design assumptions, and use of appropriate analysis methods, and computational tools. This design work is subject to ''Quality Assurance Requirements and Description''. The document is primarily intended for internal use and technical guidance for a variety of design activities. It is recognized that a wide audience including project management, the U.S. Department of Energy (DOE), the U.S. Nuclear Regulatory Commission, and others are interested to various levels of detail in the design methods and therefore covers a wide range of topics at varying levels of detail. Due to the preliminary nature of the design, readers can expect to encounter varied levels of detail in the body of the report. It is expected that technical information used as input to design documents will be verified and taken from the latest versions of reference sources given herein. This revision of the methodology report has evolved with changes in the waste package, drip shield, and emplacement pallet designs over many years and may be further revised as the design is finalized. Different components and analyses are at different stages of development. Some parts of the report are detailed, while other less detailed parts are likely to undergo further refinement. The design methodology is intended to provide designs that satisfy the safety and operational requirements of the YMP. Four waste package configurations have been selected to illustrate the application of the methodology during the licensing process. These four configurations are the 21-pressurized water reactor absorber plate waste package (21-PWRAP), the 44-boiling water reactor waste package (44-BWR), the 5 defense high-level radioactive waste (HLW) DOE spent nuclear fuel (SNF) codisposal short waste package (5-DHLWDOE SNF Short), and the naval canistered SNF long waste package (Naval SNF Long). Design work for the other six waste packages will be completed at a later date using the same design methodology. These include the 24-boiling water reactor waste package (24-BWR), the 21-pressurized water reactor control rod waste package (21-PWRCR), the 12-pressurized water reactor waste package (12-PWR), the 5 defense HLW DOE SNF codisposal long waste package (5-DHLWDOE SNF Long), the 2 defense HLW DOE SNF codisposal waste package (2-MC012-DHLW), and the naval canistered SNF short waste package (Naval SNF Short). This report is only part of the complete design description. Other reports related to the design include the design reports, the waste package system description documents, manufacturing specifications, and numerous documents for the many detailed calculations. The relationships between this report and other design documents are shown in Figure 1.

  15. Hot Cell Window Shielding Analysis Using MCNP

    SciTech Connect (OSTI)

    Chad L. Pope; Wade W. Scates; J. Todd Taylor

    2009-05-01

    The Idaho National Laboratory Materials and Fuels Complex nuclear facilities are undergoing a documented safety analysis upgrade. In conjunction with the upgrade effort, shielding analysis of the Fuel Conditioning Facility (FCF) hot cell windows has been conducted. This paper describes the shielding analysis methodology. Each 4-ft thick window uses nine glass slabs, an oil film between the slabs, numerous steel plates, and packed lead wool. Operations in the hot cell center on used nuclear fuel (UNF) processing. Prior to the shielding analysis, shield testing with a gamma ray source was conducted, and the windows were found to be very effective gamma shields. Despite these results, because the glass contained significant amounts of lead and little neutron absorbing material, some doubt lingered regarding the effectiveness of the windows in neutron shielding situations, such as during an accidental criticality. MCNP was selected as an analysis tool because it could model complicated geometry, and it could track gamma and neutron radiation. A bounding criticality source was developed based on the composition of the UNF. Additionally, a bounding gamma source was developed based on the fission product content of the UNF. Modeling the windows required field inspections and detailed examination of drawings and material specifications. Consistent with the shield testing results, MCNP results demonstrated that the shielding was very effective with respect to gamma radiation, and in addition, the analysis demonstrated that the shielding was also very effective during an accidental criticality.

  16. RESONANCE SELF-SHIELDING METHODOLOGY IN MPACT

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    RESONANCE SELF-SHIELDING METHODOLOGY IN MPACT Yuxuan Liu, Benjamin Collins, Brendan Kochunas and William Martin Department of Nuclear Engineering and Radiological Sciences...

  17. Thermal neutron shield and method of manufacture

    DOE Patents [OSTI]

    Brindza, Paul Daniel; Metzger, Bert Clayton

    2013-05-28

    A thermal neutron shield comprising concrete with a high percentage of the element Boron. The concrete is least 54% Boron by weight which maximizes the effectiveness of the shielding against thermal neutrons. The accompanying method discloses the manufacture of Boron loaded concrete which includes enriching the concrete mixture with varying grit sizes of Boron Carbide.

  18. EA-415 Lion Shield Energy, LLC | Department of Energy

    Energy Savers [EERE]

    5 Lion Shield Energy, LLC EA-415 Lion Shield Energy, LLC Order authorizing Lion Shield to export electric energy to Mexico. PDF icon EA-415 Lion Shield Energy (MX).pdf More Documents & Publications Application to Export Electric Energy OE Docket No. EA-415 Lion Shield Energy, LLC Application to Export Electric Energy OE Docket No. EA-415 Lion Shield Energy, LLC: Federal Register Notice, Volume 80, No. 164 - Aug. 25, 2015 EA-387 Energia Renovable S.C., LLC

  19. Specifications for the development of BUGLE-93: An ENDF/B-VI multigroup cross section library for LWR shielding and pressure vessel dosimetry

    SciTech Connect (OSTI)

    White, J.E.; Wright, R.Q.; Roussin, R.W.; Ingersoll, D.T.

    1992-11-01

    This report discusses specifications which have been developed for a new multigroup cross section library based on ENDF/B-VI data for light water reactor shielding and reactor pressure vessel dosimetry applications. The resulting broad-group library and an intermediate fine-group library are defined by the specifications provided in this report. Processing ENDF/B-VI into multigroup format for use in radiation transport codes will provide radiation shielding analysts with the most currently available nuclear data. it is expected that the general nature of the specifications will be useful in other applications such as reactor physics.

  20. Impact of External Heat-shielding Techniques on Shell Surface...

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

    External Heat-shielding Techniques on Shell Surface Temperatures and Dynamic Shell Thermal Deformation of Diesel Engine Emission Control Systems Impact of External Heat-shielding...

  1. Hot cell shield plug extraction apparatus

    DOE Patents [OSTI]

    Knapp, Philip A. (Moore, ID); Manhart, Larry K. (Pingree, ID)

    1995-01-01

    An apparatus is provided for moving shielding plugs into and out of holes in concrete shielding walls in hot cells for handling radioactive materials without the use of external moving equipment. The apparatus provides a means whereby a shield plug is extracted from its hole and then swung approximately 90 degrees out of the way so that the hole may be accessed. The apparatus uses hinges to slide the plug in and out and to rotate it out of the way, the hinge apparatus also supporting the weight of the plug in all positions, with the load of the plug being transferred to a vertical wall by means of a bolting arrangement.

  2. Tank evaluation system shielded annular tank application

    SciTech Connect (OSTI)

    Freier, D.A.

    1988-10-04

    TEST (Tank Evaluation SysTem) is a research project utilizing neutron interrogation techniques to analyze the content of nuclear poisons and moderators in tank shielding. TEST experiments were performed on an experimental SAT (Shielded Annular Tank) at the Rocky Flats Plant. The purpose of these experiments was threefold: (1) to assess TEST application to SATs, (2) to determine if Nuclear Safety inspection criteria could be met, and (3) to perform a preliminary calibration of TEST for SATs. Several experiments were performed, including measurements of 11 tank shielding configurations, source-simulated holdup experiments, analysis of three detector modes, resolution studies, and TEST scanner geometry experiments. 1 ref., 21 figs., 4 tabs.

  3. Thermomagnetic burn control for magnetic fusion reactor

    DOE Patents [OSTI]

    Rawls, John M. (Del Mar, CA); Peuron, Unto A. (Solana Beach, CA)

    1982-01-01

    Apparatus is provided for controlling the plasma energy production rate of a magnetic-confinement fusion reactor, by controlling the magnetic field ripple. The apparatus includes a group of shield sectors (30a, 30b, etc.) formed of ferromagnetic material which has a temperature-dependent saturation magnetization, with each shield lying between the plasma (12) and a toroidal field coil (18). A mechanism (60) for controlling the temperature of the magnetic shields, as by controlling the flow of cooling water therethrough, thereby controls the saturation magnetization of the shields and therefore the amount of ripple in the magnetic field that confines the plasma, to thereby control the amount of heat loss from the plasma. This heat loss in turn determines the plasma state and thus the rate of energy production.

  4. Thermomagnetic burn control for magnetic fusion reactor

    DOE Patents [OSTI]

    Rawls, J.M.; Peuron, A.U.

    1980-07-01

    Apparatus is provided for controlling the plasma energy production rate of a magnetic-confinement fusion reactor, by controlling the magnetic field ripple. The apparatus includes a group of shield sectors formed of ferromagnetic material which has a temperature-dependent saturation magnetization, with each shield lying between the plasma and a toroidal field coil. A mechanism for controlling the temperature of the magnetic shields, as by controlling the flow of cooling water therethrough, thereby controls the saturation magnetization of the shields and therefore the amount of ripple in the magnetic field that confines the plasma, to thereby control the amount of heat loss from the plasma. This heat loss in turn determines the plasma state and thus the rate of energy production.

  5. Shielded beam delivery apparatus and method

    DOE Patents [OSTI]

    Hershcovitch, Ady; Montano, Rory Dominick

    2006-07-11

    An apparatus includes a plasma generator aligned with a beam generator for producing a plasma to shield an energized beam. An electrode is coaxially aligned with the plasma generator and followed in turn by a vortex generator coaxially aligned with the electrode. A target is spaced from the vortex generator inside a fluid environment. The electrode is electrically biased relative to the electrically grounded target for driving the plasma toward the target inside a vortex shield.

  6. X-ray transmissive debris shield

    DOE Patents [OSTI]

    Spielman, Rick B. (Albuquerque, NM)

    1994-01-01

    A composite window structure is described for transmitting x-ray radiation and for shielding radiation generated debris. In particular, separate layers of different x-ray transmissive materials are laminated together to form a high strength, x-ray transmissive debris shield which is particularly suited for use in high energy fluences. In one embodiment, the composite window comprises alternating layers of beryllium and a thermoset polymer.

  7. Advances toward a transportable antineutrino detector system for reactor monitoring and safeguards

    SciTech Connect (OSTI)

    Reyna, D.; Bernstein, A.; Lund, J.; Kiff, S.; Cabrera-Palmer, B.; Bowden, N. S.; Dazeley, S.; Keefer, G.

    2011-07-01

    Nuclear reactors have served as the neutrino source for many fundamental physics experiments. The techniques developed by these experiments make it possible to use these very weakly interacting particles for a practical purpose. The large flux of antineutrinos that leaves a reactor carries information about two quantities of interest for safeguards: the reactor power and fissile inventory. Our SNL/LLNL collaboration has demonstrated that such antineutrino based monitoring is feasible using a relatively small cubic meter scale liquid scintillator detector at tens of meters standoff from a commercial Pressurized Water Reactor (PWR). With little or no burden on the plant operator we have been able to remotely and automatically monitor the reactor operational status (on/off), power level, and fuel burnup. The initial detector was deployed in an underground gallery that lies directly under the containment dome of an operating PWR. The gallery is 25 meters from the reactor core center, is rarely accessed by plant personnel, and provides a muon-screening effect of some 20-30 meters of water equivalent earth and concrete overburden. Unfortunately, many reactor facilities do not contain an equivalent underground location. We have therefore attempted to construct a complete detector system which would be capable of operating in an aboveground location and could be transported to a reactor facility with relative ease. A standard 6-meter shipping container was used as our transportable laboratory - containing active and passive shielding components, the antineutrino detector and all electronics, as well as climate control systems. This aboveground system was deployed and tested at the San Onofre Nuclear Generating Station (SONGS) in southern California in 2010 and early 2011. We will first present an overview of the initial demonstrations of our below ground detector. Then we will describe the aboveground system and the technological developments of the two antineutrino detectors that were deployed. Finally, some preliminary results of our aboveground test will be shown. (authors)

  8. Reactor Vessel and Reactor Vessel Internals Segmentation at Zion Nuclear Power Station - 13230

    SciTech Connect (OSTI)

    Cooke, Conrad; Spann, Holger

    2013-07-01

    Zion Nuclear Power Station (ZNPS) is a dual-unit Pressurized Water Reactor (PWR) nuclear power plant located on the Lake Michigan shoreline, in the city of Zion, Illinois approximately 64 km (40 miles) north of Chicago, Illinois and 67 km (42 miles) south of Milwaukee, Wisconsin. Each PWR is of the Westinghouse design and had a generation capacity of 1040 MW. Exelon Corporation operated both reactors with the first unit starting production of power in 1973 and the second unit coming on line in 1974. The operation of both reactors ceased in 1996/1997. In 2010 the Nuclear Regulatory Commission approved the transfer of Exelon Corporation's license to ZionSolutions, the Long Term Stewardship subsidiary of EnergySolutions responsible for the decommissioning of ZNPS. In October 2010, ZionSolutions awarded Siempelkamp Nuclear Services, Inc. (SNS) the contract to plan, segment, remove, and package both reactor vessels and their respective internals. This presentation discusses the tools employed by SNS to remove and segment the Reactor Vessel Internals (RVI) and Reactor Vessels (RV) and conveys the recent progress. SNS's mechanical segmentation tooling includes the C-HORCE (Circumferential Hydraulically Operated Cutting Equipment), BMT (Bolt Milling Tool), FaST (Former Attachment Severing Tool) and the VRS (Volume Reduction Station). Thermal segmentation of the reactor vessels will be accomplished using an Oxygen- Propane cutting system. The tools for internals segmentation were designed by SNS using their experience from other successful reactor and large component decommissioning and demolition (D and D) projects in the US. All of the designs allow for the mechanical segmentation of the internals remotely in the water-filled reactor cavities. The C-HORCE is designed to saw seven circumferential cuts through the Core Barrel and Thermal Shield walls with individual thicknesses up to 100 mm (4 inches). The BMT is designed to remove the bolts that fasten the Baffle Plates to the Baffle Former Plates. The FaST is designed to remove the Baffle Former Plates from the Core Barrel. The VRS further volume reduces segmented components using multiple configurations of the 38i and horizontal reciprocating saws. After the successful removal and volume reduction of the Internals, the RV will be segmented using a 'First in the US' thermal cutting process through a co-operative effort with Siempelkamp NIS Ingenieurgesellschaft mbH using their experience at the Stade NPP and Karlsruhe in Germany. SNS mobilized in the fall of 2011 to commence execution of the project in order to complete the RVI segmentation, removal and packaging activities for the first unit (Unit 2) by end of the 2012/beginning 2013 and then mobilize to the second unit, Unit 1. Parallel to the completion of the segmentation of the reactor vessel internals at Unit 1, SNS will segment the Unit 2 pressure vessel and at completion move to Unit 1. (authors)

  9. Microscreen radiation shield for thermoelectric generator

    DOE Patents [OSTI]

    Hunt, Thomas K. (Ann Arbor, MI); Novak, Robert F. (Farmington Hills, MI); McBride, James R. (Ypsilanti, MI)

    1990-01-01

    The present invention provides a microscreen radiation shield which reduces radiative heat losses in thermoelectric generators such as sodium heat engines without reducing the efficiency of operation of such devices. The radiation shield is adapted to be interposed between a reaction zone and a means for condensing an alkali metal vapor in a thermoelectric generator for converting heat energy directly to electrical energy. The radiation shield acts to reflect infrared radiation emanating from the reaction zone back toward the reaction zone while permitting the passage of the alkali metal vapor to the condensing means. The radiation shield includes a woven wire mesh screen or a metal foil having a plurality of orifices formed therein. The orifices in the foil and the spacing between the wires in the mesh is such that radiant heat is reflected back toward the reaction zone in the interior of the generator, while the much smaller diameter alkali metal atoms such as sodium pass directly through the orifices or along the metal surfaces of the shield and through the orifices with little or no impedance.

  10. Spherical torus fusion reactor

    DOE Patents [OSTI]

    Martin Peng, Y.K.M.

    1985-10-03

    The object of this invention is to provide a compact torus fusion reactor with dramatic simplification of plasma confinement design. Another object of this invention is to provide a compact torus fusion reactor with low magnetic field and small aspect ratio stable plasma confinement. In accordance with the principles of this invention there is provided a compact toroidal-type plasma confinement fusion reactor in which only the indispensable components inboard of a tokamak type of plasma confinement region, mainly a current conducting medium which carries electrical current for producing a toroidal magnet confinement field about the toroidal plasma region, are retained.

  11. Power converter having improved EMI shielding

    DOE Patents [OSTI]

    Beihoff, Bruce C.; Kehl, Dennis L.; Gettelfinger, Lee A.; Kaishian, Steven C.; Phillips, Mark G.; Radosevich, Lawrence D.

    2006-06-13

    EMI shielding is provided for power electronics circuits and the like via a direct-mount reference plane support and shielding structure. The thermal support may receive one or more power electronic circuits. The support may aid in removing heat from the circuits through fluid circulating through the support. The support forms a shield from both external EMI/RFI and from interference generated by operation of the power electronic circuits. Features may be provided to permit and enhance connection of the circuitry to external circuitry, such as improved terminal configurations. Modular units may be assembled that may be coupled to electronic circuitry via plug-in arrangements or through interface with a backplane or similar mounting and interconnecting structures.

  12. Vehicle drive module having improved EMI shielding

    DOE Patents [OSTI]

    Beihoff, Bruce C.; Kehl, Dennis L.; Gettelfinger, Lee A.; Kaishian, Steven C.; Phillips, Mark G.; Radosevich, Lawrence D.

    2006-11-28

    EMI shielding in an electric vehicle drive is provided for power electronics circuits and the like via a direct-mount reference plane support and shielding structure. The thermal support may receive one or more power electronic circuits. The support may aid in removing heat from the circuits through fluid circulating through the support. The support forms a shield from both external EMI/RFI and from interference generated by operation of the power electronic circuits. Features may be provided to permit and enhance connection of the circuitry to external circuitry, such as improved terminal configurations. Modular units may be assembled that may be coupled to electronic circuitry via plug-in arrangements or through interface with a backplane or similar mounting and interconnecting structures.

  13. Repair welding of fusion reactor components

    SciTech Connect (OSTI)

    Chin, B.A.

    1993-05-15

    Experiments have shown that irradiated Type 316 stainless steel is susceptible to heat-affected-zone (HAZ) cracking upon cooling when welded using the gas tungsten arc (GTA) process under lateral constraint. The cracking has been hypothesized to be caused by stress-assisted helium bubble growth and rupture at grain boundaries. This study utilized an experimental welding setup which enabled different compressive stresses to be applied to the plates during welding. Autogenous GTA welds were produced in Type 316 stainless steel doped with 256 appm helium. The application of a compressive stress, 55 MPa, during welding suppressed the previously observed catastrophic cracking. Detailed examinations conducted after welding showed a dramatic change in helium bubble morphology. Grain boundary bubble growth along directions parallel to the weld was suppressed. Results suggest that stress-modified welding techniques may be used to suppress or eliminate helium-induced cracking during joining of irradiated materials.

  14. Hewlett and Duncan - Atomic Shield | Department of Energy

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

    Duncan - Atomic Shield Hewlett and Duncan - Atomic Shield Hewlett, Richard G. and Francis Duncan. Atomic Shield, 1947-1952. U.S. Atomic Energy Comission, 1972. The second volume of the three volume A History of the United States Atomic Energy Commission. Text in each PDF is fully searchable. "Hewlett and Duncan - Atomic Shield (complete).pdf" contains the complete text and images from Atomic Shield. 12mb "Hewlett and Duncan - Atomic Shield (figures only).pdf" contains hi-res

  15. Shielding analysis and design of the KIPT experimental neutron source facility of Ukraine.

    SciTech Connect (OSTI)

    Zhong, Z.; Gohar, M. Y. A.; Naberezhnev, D.; Duo, J.; Nuclear Engineering Division

    2008-10-31

    Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an experimental neutron source facility based on the use of an electron accelerator driven subcritical (ADS) facility [1]. The facility uses the existing electron accelerators of KIPT in Ukraine. The neutron source of the sub-critical assembly is generated from the interaction of 100 KW electron beam with a natural uranium target. The electron beam has a uniform spatial distribution and the electron energy in the range of 100 to 200 MeV, [2]. The main functions of the facility are the production of medical isotopes and the support of the Ukraine nuclear power industry. Reactor physics experiments and material performance characterization will also be carried out. The subcritical assembly is driven by neutrons generated by the electron beam interactions with the target material. A fraction of these neutrons has an energy above 50 MeV generated through the photo nuclear interactions. This neutron fraction is very small and it has an insignificant contribution to the subcritical assembly performance. However, these high energy neutrons are difficult to shield and they can be slowed down only through the inelastic scattering with heavy isotopes. Therefore the shielding design of this facility is more challenging relative to fission reactors. To attenuate these high energy neutrons, heavy metals (tungsten, iron, etc.) should be used. To reduce the construction cost, heavy concrete with 4.8 g/cm{sup 3} density is selected as a shielding material. The iron weight fraction in this concrete is about 0.6. The shape and thickness of the heavy concrete shield are defined to reduce the biological dose equivalent outside the shield to an acceptable level during operation. At the same time, special attention was give to reduce the total shield mass to reduce the construction cost. The shield design is configured to maintain the biological dose equivalent during operation {le} 0.5 mrem/h inside the subcritical hall, which is five times less than the allowable dose for working forty hours per week for 50 weeks per year. This study analyzed and designed the thickness and the shape of the radial and top shields of the neutron source based on the biological dose equivalent requirements inside the subcritical hall during operation. The Monte Carlo code MCNPX is selected because of its capabilities for transporting electrons, photons, and neutrons. Mesh based weight windows variance reduction technique is utilized to estimate the biological dose outside the shield with good statistics. A significant effort dedicated to the accurate prediction of the biological dose equivalent outside the shield boundary as a function of the shield thickness without geometrical approximations or material homogenization. The building wall was designed with ordinary concrete to reduce the biological dose equivalent to the public with a safety factor in the range of 5 to 20.

  16. Summary of Prometheus Radiation Shielding Nuclear Design Analysis

    SciTech Connect (OSTI)

    J. Stephens

    2006-01-13

    This report transmits a summary of radiation shielding nuclear design studies performed to support the Prometheus project. Together, the enclosures and references associated with this document describe NRPCT (KAPL & Bettis) shielding nuclear design analyses done for the project.

  17. A Neutral Beam Pole Shield with Copper Plates and Serviceable...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    A Neutral Beam Pole Shield with Copper Plates and Serviceable Molybdenum Inserts The copper pole shields for the neutral beam lines that have been in service at DIII-D have...

  18. Integrity evaluation of lower thermal shield under exposure to HFBR environment

    SciTech Connect (OSTI)

    Kassir, M.; Weeks, J.; Bandyopadhyay, K.; Shewmon, P.

    1998-01-01

    The effects of exposure to the HFBR environment on the carbon steel in the HFBR lower thermal shield were evaluated. Corrosion was found to be a non-significant degradation process. Radiation embrittlement has occurred; portions of the plate closest to the reactor are currently operating in the lower-shelf region of the Charpy impact curve (i.e., below the fracture toughness transition temperature). In this region, the effects of radiation on the mechanical properties of carbon steel are believed to have been saturated, so that no further deterioration is anticipated. A fracture toughness analysis shows that a large factor of safety (> 1.5) exists against propagation of credible hypothetical flaws. Therefore, the existing lower thermal shield structure is suitable for continued operation of the HFBR.

  19. Construction of a Post-Irradiated Fuel Examination Shielded Enclosure Facility

    SciTech Connect (OSTI)

    Michael A. Lehto, Ph.D.; Boyd D. Christensen

    2008-05-01

    The U.S. Department of Energy (DOE) has committed to provide funding to the Idaho National Laboratory (INL) for new post-irradiation examination (PIE) equipment in support of advanced fuels development. This equipment will allow researchers at the INL to accurately characterize the behavior of experimental test fuels after they are removed from an experimental reactor also located at the INL. The accurate and detailed characterization of the fuel from the reactor, when used in conjunction with computer modeling, will allow DOE to more quickly understand the behavior of the fuel and to guide further development activities consistent with the missions of the INL and DOE. Due to the highly radioactive nature of the specimen samples that will be prepared and analyzed by the PIE equipment, shielded enclosures are required. The shielded cells will be located in the existing Analytical Laboratory (AL) basement (Rooms B-50 and B-51) at the INL Material and Fuels Complex (MFC). AL Rooms B-50 and B-51 will be modified to establish an area where sample containment and shielding will be provided for the analysis of radioactive fuels and materials while providing adequate protection for personnel and the environment. The area is comprised of three separate shielded cells for PIE instrumentation. Each cell contains an atmosphere interface enclosure (AIE) for contamination containment. The shielding will provide a work area consistent with the as-low-as-reasonably-achievable (ALARA) concept, assuming a source term of 10 samples in each of the three shielded areas. Source strength is assumed to be a maximum of 3 Ci at 0.75 MeV gamma for each sample. Each instrument listed below will be installed in an individual shielded enclosure: Shielded electron probe micro-analyzer (EPMA) Focused ion beam instrument (FIB) Micro-scale x-ray diffractometer (MXRD). The project is designed and expected to be built incrementally as funds are allocated. The initial phase will be to fund the construction activities, which will include facility modifications and construction of one shielded enclosure. Follow-up activities will be to construct two additional shielded enclosures to complete the suite of three separate but connected remote operated examination areas. Equipment purchases are to be capital procurement spread out over several years on a funded schedule. This paper discusses safety and operational considerations given during the conceptual design phase of the project. The paper considers such things as project material at risk (MAR), new processes and equipment, potential hazards, and the major modification evaluation process to determine if a preliminary Documented Safety Analysis (PDSA) is required. As part of that process, an evaluation was made of the potential hazards with the new project compared to the existing and historical work and associated hazards in the affected facility.

  20. CONVECTION REACTOR

    DOE Patents [OSTI]

    Hammond, R.P.; King, L.D.P.

    1960-03-22

    An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

  1. Electronically shielded solid state charged particle detector

    DOE Patents [OSTI]

    Balmer, D.K.; Haverty, T.W.; Nordin, C.W.; Tyree, W.H.

    1996-08-20

    An electronically shielded solid state charged particle detector system having enhanced radio frequency interference immunity includes a detector housing with a detector entrance opening for receiving the charged particles. A charged particle detector having an active surface is disposed within the housing. The active surface faces toward the detector entrance opening for providing electrical signals representative of the received charged particles when the received charged particles are applied to the active surface. A conductive layer is disposed upon the active surface. In a preferred embodiment, a nonconductive layer is disposed between the conductive layer and the active surface. The conductive layer is electrically coupled to the detector housing to provide a substantially continuous conductive electrical shield surrounding the active surface. The inner surface of the detector housing is supplemented with a radio frequency absorbing material such as ferrite. 1 fig.

  2. Electronically shielded solid state charged particle detector

    DOE Patents [OSTI]

    Balmer, David K. (155 Coral Way, Broomfield, CO 80020); Haverty, Thomas W. (1173 Logan, Northglenn, CO 80233); Nordin, Carl W. (7203 W. 32nd Ave., Wheatridge, CO 80033); Tyree, William H. (1977 Senda Rocosa, Boulder, CO 80303)

    1996-08-20

    An electronically shielded solid state charged particle detector system having enhanced radio frequency interference immunity includes a detector housing with a detector entrance opening for receiving the charged particles. A charged particle detector having an active surface is disposed within the housing. The active surface faces toward the detector entrance opening for providing electrical signals representative of the received charged particles when the received charged particles are applied to the active surface. A conductive layer is disposed upon the active surface. In a preferred embodiment, a nonconductive layer is disposed between the conductive layer and the active surface. The conductive layer is electrically coupled to the detector housing to provide a substantially continuous conductive electrical shield surrounding the active surface. The inner surface of the detector housing is supplemented with a radio frequency absorbing material such as ferrite.

  3. Supplemental heating of deposition tooling shields

    DOE Patents [OSTI]

    Ohlhausen, James A. (Albuquerque, NM); Peebles, Diane E. (Albuquerque, NM); Hunter, John A. (Albuquerque, NM); Eckelmeyer, Kenneth H. (Albuquerque, NM)

    2000-01-01

    A method of reducing particle generation from the thin coating deposited on the internal surfaces of a deposition chamber which undergoes temperature variation greater than 100.degree. C. comprising maintaining the temperature variation of the internal surfaces low enough during the process cycle to keep thermal expansion stresses between the coating and the surfaces under 500 MPa. For titanium nitride deposited on stainless steel, this means keeping temperature variations under approximately 70.degree. C. in a chamber that may be heated to over 350.degree. C. during a typical processing operation. Preferably, a supplemental heater is mounted behind the upper shield and controlled by a temperature sensitive element which provides feedback control based on the temperature of the upper shield.

  4. In-situ Condition Monitoring of Components in Small Modular Reactors Using Process and Electrical Signature Analysis. Final report, volume 1. Development of experimental flow control loop, data analysis and plant monitoring

    SciTech Connect (OSTI)

    Upadhyaya, Belle; Hines, J. Wesley; Damiano, Brian; Mehta, Chaitanya; Collins, Price; Lish, Matthew; Cady, Brian; Lollar, Victor; de Wet, Dane; Bayram, Duygu

    2015-12-15

    The research and development under this project was focused on the following three major objectives: Objective 1: Identification of critical in-vessel SMR components for remote monitoring and development of their low-order dynamic models, along with a simulation model of an integral pressurized water reactor (iPWR). Objective 2: Development of an experimental flow control loop with motor-driven valves and pumps, incorporating data acquisition and on-line monitoring interface. Objective 3: Development of stationary and transient signal processing methods for electrical signatures, machinery vibration, and for characterizing process variables for equipment monitoring. This objective includes the development of a data analysis toolbox. The following is a summary of the technical accomplishments under this project: - A detailed literature review of various SMR types and electrical signature analysis of motor-driven systems was completed. A bibliography of literature is provided at the end of this report. Assistance was provided by ORNL in identifying some key references. - A review of literature on pump-motor modeling and digital signal processing methods was performed. - An existing flow control loop was upgraded with new instrumentation, data acquisition hardware and software. The upgrading of the experimental loop included the installation of a new submersible pump driven by a three-phase induction motor. All the sensors were calibrated before full-scale experimental runs were performed. - MATLAB-Simulink model of a three-phase induction motor and pump system was completed. The model was used to simulate normal operation and fault conditions in the motor-pump system, and to identify changes in the electrical signatures. - A simulation model of an integral PWR (iPWR) was updated and the MATLAB-Simulink model was validated for known transients. The pump-motor model was interfaced with the iPWR model for testing the impact of primary flow perturbations (upsets) on plant parameters and the pump electrical signatures. Additionally, the reactor simulation is being used to generate normal operation data and data with instrumentation faults and process anomalies. A frequency controller was interfaced with the motor power supply in order to vary the electrical supply frequency. The experimental flow control loop was used to generate operational data under varying motor performance characteristics. Coolant leakage events were simulated by varying the bypass loop flow rate. The accuracy of motor power calculation was improved by incorporating the power factor, computed from motor current and voltage in each phase of the induction motor.- A variety of experimental runs were made for steady-state and transient pump operating conditions. Process, vibration, and electrical signatures were measured using a submersible pump with variable supply frequency. High correlation was seen between motor current and pump discharge pressure signal; similar high correlation was exhibited between pump motor power and flow rate. Wide-band analysis indicated high coherence (in the frequency domain) between motor current and vibration signals. - Wide-band operational data from a PWR were acquired from AMS Corporation and used to develop time-series models, and to estimate signal spectrum and sensor time constant. All the data were from different pressure transmitters in the system, including primary and secondary loops. These signals were pre-processed using the wavelet transform for filtering both low-frequency and high-frequency bands. This technique of signal pre-processing provides minimum distortion of the data, and results in a more optimal estimation of time constants of plant sensors using time-series modeling techniques.

  5. Grounding and shielding in the accelerator environment

    SciTech Connect (OSTI)

    Kerns, Q.

    1991-12-31

    Everyday features of the accelerator environment include long cable runs, high power and low level equipment sharing building space, stray electromagnetic fields and ground voltage differences between the sending and receiving ends of an installation. This paper pictures some Fermilab installations chosen to highlight significant features and presents practices, test methods and equipment that have been helpful in achieving successful shielding. Throughout the report are numbered statements aimed at summarizing good practices and avoiding pitfalls.

  6. Shielded serpentine traveling wave tube deflection structure

    DOE Patents [OSTI]

    Hudson, C.L.; Spector, J.

    1994-12-27

    A shielded serpentine slow wave deflection structure is disclosed having a serpentine signal conductor within a channel groove. The channel groove is formed by a serpentine channel in a trough plate and a ground plane. The serpentine signal conductor is supported at its ends by coaxial feed through connectors. A beam interaction trough intersects the channel groove to form a plurality of beam interaction regions wherein an electron beam may be deflected relative to the serpentine signal conductor. 4 figures.

  7. Transparent self-cleaning dust shield

    DOE Patents [OSTI]

    Mazumder, Malay K.; Sims, Robert A.; Wilson, James D.

    2005-06-28

    A transparent electromagnetic shield to protect solar panels and the like from dust deposition. The shield is a panel of clear non-conducting (dielectric) material with embedded parallel electrodes. The panel is coated with a semiconducting film. Desirably the electrodes are transparent. The electrodes are connected to a single-phase AC signal or to a multi-phase AC signal that produces a travelling electromagnetic wave. The electromagnetic field produced by the electrodes lifts dust particles away from the shield and repels charged particles. Deposited dust particles are removed when the electrodes are activated, regardless of the resistivity of the dust. Electrostatic charges on the panel are discharged by the semiconducting film. When used in conjunction with photovoltaic cells, the power for the device may be obtained from the cells themselves. For other surfaces, such as windshields, optical windows and the like, the power must be derived from an external source. One embodiment of the invention employs monitoring and detection devices to determine when the level of obscuration of the screen by dust has reached a threshold level requiring activation of the dust removal feature.

  8. Distributed resonance self-shielding using the equivalence principle

    SciTech Connect (OSTI)

    Altiparmakov, D.

    2012-07-01

    This paper presents an extension of the equivalence principle to allow distributed resonance self-shielding in a multi-region fuel configuration. Rational expansion of fuel-to-fuel collision probability is applied in order to establish equivalence between the actual fuel configuration and a homogeneous mixture of hydrogen and resonant absorber, which is a commonly used model to calculate library tables of resonance integrals. The main steps in derivation are given along with the basic physics assumptions on which the presented approach relies. The method has been implemented in the lattice code WIMS-AECL and routinely used for calculation of CANDU-type reactor lattices. Its capabilities are illustrated by comparison of WIMS-AECL and MCNP results of {sup 238}U resonance capture in a CANDU lattice cell. In order to determine optimal rational expansion of fuel-to-fuel collision probability, the calculations were carried out by varying the number of rational terms from 1 to 6. The results show that 4 terms are sufficient. The further increase of the number of terms affects the computing time, while the impact on accuracy is negligible. To illustrate the convergence of the results, the fuel subdivision is gradually refined varying the number of fuel pin subdivisions from 1 to 32 equal-area annuli. The results show very good agreement with the reference MCNP calculation. (authors)

  9. Including shielding effects in application of the TPCA method for detection of embedded radiation sources.

    SciTech Connect (OSTI)

    Johnson, William C.; Shokair, Isaac R.

    2011-12-01

    Conventional full spectrum gamma spectroscopic analysis has the objective of quantitative identification of all the radionuclides present in a measurement. For low-energy resolution detectors such as NaI, when photopeaks alone are not sufficient for complete isotopic identification, such analysis requires template spectra for all the radionuclides present in the measurement. When many radionuclides are present it is difficult to make the correct identification and this process often requires many attempts to obtain a statistically valid solution by highly skilled spectroscopists. A previous report investigated using the targeted principal component analysis method (TPCA) for detection of embedded sources for RPM applications. This method uses spatial/temporal information from multiple spectral measurements to test the hypothesis of the presence of a target spectrum of interest in these measurements without the need to identify all the other radionuclides present. The previous analysis showed that the TPCA method has significant potential for automated detection of target radionuclides of interest, but did not include the effects of shielding. This report complements the previous analysis by including the effects of spectral distortion due to shielding effects for the same problem of detection of embedded sources. Two examples, one with one target radionuclide and the other with two, show that the TPCA method can successfully detect shielded targets in the presence of many other radionuclides. The shielding parameters are determined as part of the optimization process using interpolation of library spectra that are defined on a 2D grid of atomic numbers and areal densities.

  10. Nuclear reactor sealing system

    DOE Patents [OSTI]

    McEdwards, James A. (Calabasas, CA)

    1983-01-01

    A liquid metal-cooled nuclear reactor sealing system. The nuclear reactor includes a vessel sealed at its upper end by a closure head. The closure head comprises at least two components, one of which is rotatable; and the two components define an annulus therebetween. The sealing system includes at least a first and second inflatable seal disposed in series in an upper portion of the annulus. The system further includes a dip seal extending into a body of insulation located adjacent a bottom portion of the closure head. The dip seal comprises a trough formed by a lower portion of one of the components, and a seal blade pendently supported from the other component and extending downwardly into the trough. A body of liquid metal is contained in the trough which submerges a portion of the seal blade. The seal blade is provided with at least one aperture located above the body of liquid metal for providing fluid communication between the annulus intermediate the dip seal and the inflatable seals, and a body of cover gas located inside the vessel. There also is provided means for introducing a purge gas into the annulus intermediate the inflatable seals and the seal blade. The purge gas is introduced in an amount sufficient to substantially reduce diffusion of radioactive cover gas or sodium vapor up to the inflatable seals. The purge gas mixes with the cover gas in the reactor vessel where it can be withdrawn from the vessel for treatment and recycle to the vessel.

  11. X-ray transmissive debris shield

    DOE Patents [OSTI]

    Spielman, R.B.

    1996-05-21

    An X-ray debris shield for use in X-ray lithography that is comprised of an X-ray window having a layer of low density foam exhibits increased longevity without a substantial increase in exposure time. The low density foam layer serves to absorb the debris emitted from the X-ray source and attenuate the shock to the window so as to reduce the chance of breakage. Because the foam is low density, the X-rays are hardly attenuated by the foam and thus the exposure time is not substantially increased.

  12. X-ray transmissive debris shield

    DOE Patents [OSTI]

    Spielman, Rick B. (Albuquerque, NM)

    1996-01-01

    An X-ray debris shield for use in X-ray lithography that is comprised of an X-ray window having a layer of low density foam exhibits increased longevity without a substantial increase in exposure time. The low density foam layer serves to absorb the debris emitted from the X-ray source and attenuate the shock to the window so as to reduce the chance of breakage. Because the foam is low density, the X-rays are hardly attenuated by the foam and thus the exposure time is not substantially increased.

  13. Shielded serpentine traveling wave tube deflection structure

    DOE Patents [OSTI]

    Hudson, Charles L. (Santa Barbara, CA); Spector, Jerome (Berkeley, CA)

    1994-01-01

    A shielded serpentine slow wave deflection structure (10) having a serpene signal conductor (12) within a channel groove (46). The channel groove (46) is formed by a serpentine channel (20) in a trough plate (18) and a ground plane (14). The serpentine signal conductor (12) is supported at its ends by coaxial feed through connectors 28. A beam interaction trough (22) intersects the channel groove (46) to form a plurality of beam interaction regions (56) wherein an electron beam (54) may be deflected relative to the serpentine signal conductor (12).

  14. Radiation shielding properties of barite coated fabric by computer programme

    SciTech Connect (OSTI)

    Akarslan, F.; Molla, T.; nc, I. S.; K?l?ncarslan, S.; Akkurt, I.

    2015-03-30

    With the development of technology radiation started to be used in variety of different fields. As the radiation is hazardous for human health, it is important to keep radiation dose as low as possible. This is done mainly using shielding materials. Barite is one of the important materials in this purpose. As the barite is not used directly it can be used in some other materials such as fabric. For this purposes barite has been coated on fabric in order to improve radiation shielding properties of fabric. Determination of radiation shielding properties of coated fabric has been done by using computer program written C# language. With this program the images obtained from digital Rontgen films is used to determine radiation shielding properties in terms of image processing numerical values. Those values define radiation shielding and in this way the coated barite effect on radiation shielding properties of fabric has been obtained.

  15. Optimal Shielding for Minimum Materials Cost of Mass

    SciTech Connect (OSTI)

    Woolley, Robert D.

    2014-08-01

    Material costs dominate some shielding design problems. This is certainly the case for manned nuclear power space applications for which shielding is essential and the cost of launching by rocket from earth is high. In such situations or in those where shielding volume or mass is constrained, it is important to optimize the design. Although trial and error synthesis methods may succeed a more systematic approach is warranted. Design automation may also potentially reduce engineering costs.

  16. Polyethylene/Boron Composites for Radiation Shielding Applications

    SciTech Connect (OSTI)

    Harrison, Courtney; Grulke, Eric; Burgett, Eric; Hertel, Nolan

    2008-01-21

    Multifunctional composites made with boron are absorbers of low energy nuetrons, and could be used for structural shielding materials. Polyethylene/boron carbide composites were fabricated using conventional polymer processing techniques, and were evaluated for mechanical and radiation shielding properties. Addition of neat boron carbide (powder and nanoparticles) to an injection molding grade HPDE showed superior mechanical properties compared to neat HDPE. Radiation shielding measurements of a 2 wt% boron carbide composite were improved over those of the neat polyethylene.

  17. Nuclear Rocket Facility Decommissioning Project: Controlled Explosive Demolition of Neutron-Activated Shield Wall

    SciTech Connect (OSTI)

    Michael R. Kruzic

    2008-06-01

    Located in Area 25 of the Nevada Test Site (NTS), the Test Cell A (TCA) Facility (Figure 1) was used in the early to mid-1960s for testing of nuclear rocket engines, as part of the Nuclear Rocket Development Program, to further space travel. Nuclear rocket testing resulted in the activation of materials around the reactors and the release of fission products and fuel particles. The TCA facility, known as Corrective Action Unit 115, was decontaminated and decommissioned (D&D) from December 2004 to July 2005 using the Streamlined Approach for Environmental Restoration (SAFER) process, under the Federal Facility Agreement and Consent Order. The SAFER process allows environmental remediation and facility closure activities (i.e., decommissioning) to occur simultaneously, provided technical decisions are made by an experienced decision maker within the site conceptual site model. Facility closure involved a seven-step decommissioning strategy. First, preliminary investigation activities were performed, including review of process knowledge documentation, targeted facility radiological and hazardous material surveys, concrete core drilling and analysis, shield wall radiological characterization, and discrete sampling, which proved to be very useful and cost-effective in subsequent decommissioning planning and execution and worker safety. Second, site setup and mobilization of equipment and personnel were completed. Third, early removal of hazardous materials, including asbestos, lead, cadmium, and oil, was performed ensuring worker safety during more invasive demolition activities. Process piping was to be verified void of contents. Electrical systems were de-energized and other systems were rendered free of residual energy. Fourth, areas of high radiological contamination were decontaminated using multiple methods. Contamination levels varied across the facility. Fixed beta/gamma contamination levels ranged up to 2 million disintegrations per minute (dpm)/100 centimeters squared (cm2) beta/gamma. Removable beta/gamma contamination levels seldom exceeded 1,000 dpm/100 cm2, but, in railroad trenches on the reactor pad containing soil on the concrete pad in front of the shield wall, the beta dose rates ranged up to 120 milli-roentgens per hour from radioactivity entrained in the soil. General area dose rates were less than 100 micro-roentgens per hour. Prior to demolition of the reactor shield wall, removable and fixed contaminated surfaces were decontaminated to the best extent possible, using traditional decontamination methods. Fifth, large sections of the remaining structures were demolished by mechanical and open-air controlled explosive demolition (CED). Mechanical demolition methods included the use of conventional demolition equipment for removal of three main buildings, an exhaust stack, and a mobile shed. The 5-foot (ft), 5-inch (in.) thick, neutron-activated reinforced concrete shield was demolished by CED, which had never been performed at the NTS.

  18. Revised RTG source term and dose rate shielding evaluation

    SciTech Connect (OSTI)

    Goldberg, H.J., Westinghouse Hanford

    1996-07-31

    A source term was developed for the RTG and the adequacy with regards to shielding of the shipping container in which it will be transported was evaluated.

  19. Graphene shield enhanced photocathodes and methods for making the same

    DOE Patents [OSTI]

    Moody, Nathan Andrew

    2014-09-02

    Disclosed are graphene shield enhanced photocathodes, such as high QE photocathodes. In certain embodiments, a monolayer graphene shield membrane ruggedizes a high quantum efficiency photoemission electron source by protecting a photosensitive film of the photocathode, extending operational lifetime and simplifying its integration in practical electron sources. In certain embodiments of the disclosed graphene shield enhanced photocathodes, the graphene serves as a transparent shield that does not inhibit photon or electron transmission but isolates the photosensitive film of the photocathode from reactive gas species, preventing contamination and yielding longer lifetime.

  20. Neutronics at Wisconsin, ORNL advances ITER shielding and internationa...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Neutronics at Wisconsin, ORNL advances ITER shielding and international collaboration American Fusion News Category: U.S. ITER Link: Neutronics at Wisconsin, ORNL advances ITER...

  1. MicroShield/ISOCS gamma modeling comparison.

    SciTech Connect (OSTI)

    Sansone, Kenneth R

    2013-08-01

    Quantitative radiological analysis attempts to determine the quantity of activity or concentration of specific radionuclide(s) in a sample. Based upon the certified standards that are used to calibrate gamma spectral detectors, geometric similarities between sample shape and the calibration standards determine if the analysis results developed are qualitative or quantitative. A sample analyzed that does not mimic a calibrated sample geometry must be reported as a non-standard geometry and thus the results are considered qualitative and not quantitative. MicroShieldR or ISOCSR calibration software can be used to model non-standard geometric sample shapes in an effort to obtain a quantitative analytical result. MicroShieldR and Canberra's ISOCSR software contain several geometry templates that can provide accurate quantitative modeling for a variety of sample configurations. Included in the software are computational algorithms that are used to develop and calculate energy efficiency values for the modeled sample geometry which can then be used with conventional analysis methodology to calculate the result. The response of the analytical method and the sensitivity of the mechanical and electronic equipment to the radionuclide of interest must be calibrated, or standardized, using a calibrated radiological source that contains a known and certified amount of activity.

  2. An Account of Oak Ridge National Laboratory's Thirteen Research Reactors

    SciTech Connect (OSTI)

    Rosenthal, Murray Wilford

    2009-08-01

    The Oak Ridge National Laboratory has built and operated 13 nuclear reactors in its 66-year history. The first was the graphite reactor, the world's first operational nuclear reactor, which served as a plutonium production pilot plant during World War II. It was followed by two aqueous-homogeneous reactors and two red-hot molten-salt reactors that were parts of power-reactor development programs and by eight others designed for research and radioisotope production. One of the eight was an all-metal fast burst reactor used for health physics studies. All of the others were light-water cooled and moderated, including the famous swimming-pool reactor that was copied dozens of times around the world. Two of the reactors were hoisted 200 feet into the air to study the shielding needs of proposed nuclear-powered aircraft. The final reactor, and the only one still operating today, is the High Flux Isotope Reactor (HFIR) that was built particularly for the production of californium and other heavy elements. With the world's highest flux and recent upgrades that include the addition of a cold neutron source, the 44-year-old HFIR continues to be a valuable tool for research and isotope production, attracting some 500 scientific visitors and guests to Oak Ridge each year. This report describes all of the reactors and their histories.

  3. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Moore, R.V.; Bowen, J.H.; Dent, K.H.

    1958-12-01

    A heterogeneous, natural uranium fueled, solid moderated, gas cooled reactor is described, in which the fuel elements are in the form of elongated rods and are dlsposed within vertical coolant channels ln the moderator symmetrically arranged as a regular lattice in groups. This reactor employs control rods which operate in vertical channels in the moderator so that each control rod is centered in one of the fuel element groups. The reactor is enclosed in a pressure vessel which ls provided with access holes at the top to facilitate loading and unloadlng of the fuel elements, control rods and control rod driving devices.

  4. Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges

    Broader source: Energy.gov [DOE]

    The most life-limiting structural component in light-water reactors (LWR) is the reactor pressure vessel (RPV) because replacement of the RPV is not considered a viable option at this time. LWR...

  5. Three-dimensional analysis of AP600 standard plant shield building roof

    SciTech Connect (OSTI)

    Greimann, L.; Fanous, F.; Safar, S.; Khalil, A.; Bluhm, D.

    1999-06-01

    The AP600 passive containment vessel is surrounded by a concrete cylindrical shell covered with a truncated conical roof. This roof supports the passive containment cooling system (PCS) annular tank, shield plate and other nonstructural attachments. When the shield building is subjected to different loading combinations as defined in the Standard Review Plan (SRP), some of the sections in the shield building could experience forces in excess of their design values. This report summarized the three-dimensional finite element analysis that was conducted to review the adequacy of the proposed Westinghouse shield building design. The ANSYS finite element software was utilized to analyze the Shield Building Roof (SBR) under dead, snow, wind, thermal and seismic loadings. A three-dimensional model that included a portion of the shield building cylindrical shell, the conical roof and its attachments, the eccentricities at the cone-cylinder connection and at the compression ring and the PCS tank was developed. Mesh sensitivity studies were conducted to select appropriate element size in the cylinder, cone, near air intakes and in the vicinity of the eccentricities. Also, a study was carried out to correctly idealize the water-structure interaction in the PCS tank. Response spectrum analysis was used to calculate the internal forces at different sections in the SBR under Safe Shutdown Earthquake (SSE). Forty-nine structural modes and twenty sloshing modes were used. Two horizontal components of the SSE together with a vertical component were used. Modal stress resultants were combined taking into account the effects of closely spaced modes. The three earthquake directions were combined by the Square Root of the Sum Squares method. Two load combinations were studied. The load combination that included dead, snow, fluid, thermal and seismic loads was selected to be the most critical. Interaction diagrams for critical sections were developed and used to check the design adequacy. The results demonstrated that provided area of steal on each face of several sections of the AP600 SBR was inadequate. This was also noticed when comparing the total provided area of steel per section, i.e., the area of steel on both faces. The discrepancy between Westinghouse results and these reported herein could have resulted from the different finite element mesh sizes and the assumption used in Westinghouse design.

  6. Gas Reactor Plant Analyzer and Simulator for Hydrogen Production

    Energy Science and Technology Software Center (OSTI)

    2004-01-01

    This software is used to study and analyze various configurations of plant equipment for gas cooled nuclear reactor applications. The user of this software would likely be interested in optimizing the economic, safety, and operating performance of this type of reactor. The code provides the capability for the user through his input to configure networks of nuclear reactor components. The components available include turbine, compressor, heat exchanger, reactor core, coolers, bypass valves, and control systems.

  7. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Wigner, E.P.

    1960-11-22

    A nuclear reactor is described wherein horizontal rods of thermal- neutron-fissionable material are disposed in a body of heavy water and extend through and are supported by spaced parallel walls of graphite.

  8. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  9. Reactor apparatus

    DOE Patents [OSTI]

    Echtler, J. Paul (Pittsburgh, PA)

    1981-01-01

    A reactor apparatus for hydrocracking a polynuclear aromatic hydrocarbonaceous feedstock to produce lighter hydrocarbon fuels by contacting the hydrocarbonaceous feedstock with hydrogen in the presence of a molten metal halide catalyst.

  10. General Corrosion and Localized Corrosion of the Drip Shield

    SciTech Connect (OSTI)

    F. Hua

    2004-09-16

    The repository design includes a drip shield (BSC 2004 [DIRS 168489]) that provides protection for the waste package both as a barrier to seepage water contact and a physical barrier to potential rockfall. The purpose of the process-level models developed in this report is to model dry oxidation, general corrosion, and localized corrosion of the drip shield plate material, which is made of Ti Grade 7. This document is prepared according to ''Technical Work Plan For: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]). The models developed in this report are used by the waste package degradation analyses for TSPA-LA and serve as a basis to determine the performance of the drip shield. The drip shield may suffer from other forms of failure such as the hydrogen induced cracking (HIC) or stress corrosion cracking (SCC), or both. Stress corrosion cracking of the drip shield material is discussed in ''Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural Material'' (BSC 2004 [DIRS 169985]). Hydrogen induced cracking of the drip shield material is discussed in ''Hydrogen Induced Cracking of Drip Shield'' (BSC 2004 [DIRS 169847]).

  11. Plasma shield for in-air beam processes

    SciTech Connect (OSTI)

    Hershcovitch, Ady

    2008-05-15

    A novel concept/apparatus, the Plasma Shield, is introduced in this paper. The purpose of the Plasma Shield is designed to shield a target object chemically and thermally by engulfing an area subjected to beam treatment with inert plasma. The shield consists of a vortex-stabilized arc that is employed to shield beams and workpiece area of interaction from an atmospheric or liquid environment. A vortex-stabilized arc is established between a beam generating device (laser, ion or electron gun) and a target object. The arc, which is composed of a pure noble gas, engulfs the interaction region and shields it from any surrounding liquids like water or reactive gases. The vortex is composed of a sacrificial gas or liquid that swirls around and stabilizes the arc. The successful Plasma Shield was experimentally established and very high-quality electron beam welding with partial plasma shielding was performed. The principle of the operation and experimental results are discussed in the paper.

  12. Superconducting magnetic shielding apparatus and method

    DOE Patents [OSTI]

    Clem, J.R.

    1982-07-09

    Disclosed is a method and apparatus for providing magnetic shielding around a working volume. The apparatus includes a hollow elongated superconducting shell or cylinder having an elongated low magnetic pinning central portion, and two high magnetic pinning end regions. Transition portions of varying magnetic pinning properties are interposed between the central and end portions. The apparatus further includes a solenoid substantially coextensive with and overlying the superconducting cylinder, so as to be magnetically coupled therewith. The method includes the steps passing a longitudinally directed current through the superconducting cylinder so as to depin magnetic reservoirs trapped in the cylinder. Next, a circumferentially directed current is passed through the cylinder, while a longitudinally directed current is maintained. Depinned magnetic reservoirs are moved to the end portions of the cylinder, where they are trapped.

  13. Fan-fold shielded electrical leads

    DOE Patents [OSTI]

    Rohatgi, R.R.; Cowan, T.E.

    1996-06-11

    Disclosed are fan-folded electrical leads made from copper cladded Kapton, for example, with the copper cladding on one side serving as a ground plane and the copper cladding on the other side being etched to form the leads. The Kapton is fan folded with the leads located at the bottom of the fan-folds. Electrical connections are made by partially opening the folds of the fan and soldering, for example, the connections directly to the ground plane and/or the lead. The fan folded arrangement produces a number of advantages, such as electrically shielding the leads from the environment, is totally non-magnetic, and has a very low thermal conductivity, while being easy to fabricate. 3 figs.

  14. Fan-fold shielded electrical leads

    DOE Patents [OSTI]

    Rohatgi, Rajeev R. (Mountain View, CA); Cowan, Thomas E. (Livermore, CA)

    1996-01-01

    Fan-folded electrical leads made from copper cladded Kapton, for example, with the copper cladding on one side serving as a ground plane and the copper cladding on the other side being etched to form the leads. The Kapton is fan folded with the leads located at the bottom of the fan-folds. Electrical connections are made by partially opening the folds of the fan and soldering, for example, the connections directly to the ground plane and/or the lead. The fan folded arrangement produces a number of advantages, such as electrically shielding the leads from the environment, is totally non-magnetic, and has a very low thermal conductivity, while being easy to fabricate.

  15. Superconducting magnetic shielding apparatus and method

    DOE Patents [OSTI]

    Clem, John R. (Ames, IA); Clem, John R. (Ames, IA)

    1983-01-01

    Disclosed is a method and apparatus for providing magnetic shielding around a working volume. The apparatus includes a hollow elongated superconducting shell or cylinder having an elongated low magnetic pinning central portion, and two high magnetic pinning end regions. Transition portions of varying magnetic pinning properties are interposed between the central and end portions. The apparatus further includes a solenoid substantially coextensive with and overlying the superconducting cylinder, so as to be magnetically coupled therewith. The method includes the steps passing a longitudinally directed current through the superconducting cylinder so as to depin magnetic reservoirs trapped in the cylinder. Next, a circumferentially directed current is passed through the cylinder, while a longitudinally directed current is maintained. Depinned magnetic reservoirs are moved to the end portions of the cylinder, where they are trapped.

  16. Superconducting magnetic shielding apparatus and method

    DOE Patents [OSTI]

    Clem, J.R.; Clem, J.R.

    1983-10-11

    Disclosed are a method and apparatus for providing magnetic shielding around a working volume. The apparatus includes a hollow elongated superconducting shell or cylinder having an elongated low magnetic pinning central portion, and two high magnetic pinning end regions. Transition portions of varying magnetic pinning properties are interposed between the central and end portions. The apparatus further includes a solenoid substantially coextensive with and overlying the superconducting cylinder, so as to be magnetically coupled therewith. The method includes the steps passing a longitudinally directed current through the superconducting cylinder so as to depin magnetic reservoirs trapped in the cylinder. Next, a circumferentially directed current is passed through the cylinder, while a longitudinally directed current is maintained. Depinned magnetic reservoirs are moved to the end portions of the cylinder, where they are trapped. 5 figs.

  17. Protective shield for an instrument probe

    DOE Patents [OSTI]

    Johnsen, Howard A.; Ross, James R.; Birtola, Sal R.

    2004-10-26

    A shield is disclosed that is particularly useful for protecting exposed optical elements at the end of optical probes used in the analysis of hazardous emissions in and around an industrial environment from the contaminating effects of those emissions. The instant invention provides a hood or cowl in the shape of a right circular cylinder that can be fitted over the end of such optical probes. The hood provides a clear aperture through which the probe can perform unobstructed analysis. The probe optical elements are protected from the external environment by passing a dry gas through the interior of the hood and out through the hood aperture in sufficient quantity and velocity to prevent any significant mixing between the internal and external environments. Additionally, the hood is provided with a cooling jacket to lessen the potential for damaging the probe due to temperature excursions.

  18. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Vernon, H.C.

    1959-01-13

    A neutronic reactor of the heterogeneous, fluid cooled tvpe is described. The reactor is comprised of a pressure vessel containing the moderator and a plurality of vertically disposed channels extending in spaced relationship through the moderator. Fissionable fuel material is placed within the channels in spaced relationship thereto to permit circulation of the coolant fluid. Separate means are provided for cooling the moderator and for circulating a fluid coolant thru the channel elements to cool the fuel material.

  19. Comments on shielding for dual energy accelerators

    SciTech Connect (OSTI)

    Rossi, M. C.; Lincoln, H. M.; Quarin, D. J.; Zwicker, R. D.

    2008-06-15

    Determination of shielding requirements for medical linear accelerators has been greatly facilitated by the publication of the National Council on Radiation Protection and Measurements (NCRP) latest guidelines on this subject in NCRP Report No. 151. In the present report the authors review their own recent experience with patient treatments on conventional dual energy linear accelerators to examine the various input parameters needed to follow the NCRP guidelines. Some discussion is included of workloads, occupancy, use factors, and field size, with the effects of intensity modulated radiotherapy (IMRT) treatments included. Studies of collimator settings showed average values of 13.1x16.2 cm{sup 2} for 6 MV and 14.1x16.8 cm{sup 2} for 18 MV conventional ports, and corresponding average unblocked areas of 228 and 254 cm{sup 2}, respectively. With an average of 77% of the field area unblocked, this gives a mean irradiated area of 196 cm{sup 2} for the 18 MV beam, which dominates shielding considerations for most dual energy machines. Assuming conservatively small room dimensions, a gantry bin angle of 18 deg. was found to represent a reasonable unit for tabulation of use factors. For conventional 18 MV treatments it was found that the usual treatment angles of 0, 90, 180, and 270 deg. were still favored, and use factors of 0.25 represent reasonable estimates for these beams. As expected, the IMRT fields (all at 6 MV) showed a high degree of gantry angle randomization, with no bin having a use factor in excess of 0.10. It is concluded that unless a significant number of patients are treated with high energy IMRT, the traditional use factors of 0.25 are appropriate for the dominant high energy beam.

  20. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Grebe, J.J.

    1959-07-14

    High temperature reactors which are uniquely adapted to serve as the heat source for nuclear pcwered rockets are described. The reactor is comprised essentially of an outer tubular heat resistant casing which provides the main coolant passageway to and away from the reactor core within the casing and in which the working fluid is preferably hydrogen or helium gas which is permitted to vaporize from a liquid storage tank. The reactor core has a generally spherical shape formed entirely of an active material comprised of fissile material and a moderator material which serves as a diluent. The active material is fabricated as a gas permeable porous material and is interlaced in a random manner with very small inter-connecting bores or capillary tubes through which the coolant gas may flow. The entire reactor is divided into successive sections along the direction of the temperature gradient or coolant flow, each section utilizing materials of construction which are most advantageous from a nuclear standpoint and which at the same time can withstand the operating temperature of that particular zone. This design results in a nuclear reactor characterized simultaneously by a minimum critiral size and mass and by the ability to heat a working fluid to an extremely high temperature.

  1. Passive heat transfer means for nuclear reactors

    DOE Patents [OSTI]

    Burelbach, James P. (Glen Ellyn, IL)

    1984-01-01

    An improved passive cooling arrangement is disclosed for maintaining adjacent or related components of a nuclear reactor within specified temperature differences. Specifically, heat pipes are operatively interposed between the components, with the vaporizing section of the heat pipe proximate the hot component operable to cool it and the primary condensing section of the heat pipe proximate the other and cooler component operable to heat it. Each heat pipe further has a secondary condensing section that is located outwardly beyond the reactor confinement and in a secondary heat sink, such as air ambient the containment, that is cooler than the other reactor component. Means such as shrouding normally isolated the secondary condensing section from effective heat transfer with the heat sink, but a sensor responds to overheat conditions of the reactor to open the shrouding, which thereby increases the cooling capacity of the heat pipe. By having many such heat pipes, an emergency passive cooling system is defined that is operative without electrical power.

  2. Energetic component treatability study

    SciTech Connect (OSTI)

    Gildea, P.D.; Brandon, S.L.; Brown, B.G. [and others

    1997-11-01

    The effectiveness of three environmentally sound processes for small energetic component disposal was examined experimentally in this study. The three destruction methods, batch reactor supercritical water oxidation, sodium hydroxide base hydrolysis and calcium carbonate cookoff were selected based on their potential for producing a clean solid residue and minimum release of toxic gases after component detonation. The explosive hazard was destroyed by all three processes. Batch supercritical water oxidation destroyed both the energetics and organics. Further development is desired to optimize process parameters. Sodium hydroxide base hydrolysis and calcium carbonate cookoff results indicated the potential for scrubbing gaseous detonation products. Further study and testing are needed to quantify the effectiveness of these later two processes for full-scale munition destruction. The preliminary experiments completed in this study have demonstrated the promise of these three processes as environmentally sound technologies for energetic component destruction. Continuation of these experimental programs is strongly recommended to optimize batch supercritical water oxidation processing, and to fully develop the sodium hydroxide base hydrolysis and calcium carbonate cookoff technologies.

  3. Expandable Metal Liner For Downhole Components

    DOE Patents [OSTI]

    Hall, David R. (Provo, UT); Fox, Joe R. (Provo, UT)

    2004-10-05

    A liner for an annular downhole component is comprised of an expandable metal tube having indentations along its surface. The indentations are formed in the wall of the tube either by drawing the tube through a die, by hydroforming, by stamping, or roll forming and may extend axially, radially, or spirally along its wall. The indentations accommodate radial and axial expansion of the tube within the downhole component. The tube is inserted into the annular component and deformed to match an inside surface of the component. The tube may be expanded using a hydroforming process or by drawing a mandrel through the tube. The tube may be expanded in such a manner so as to place it in compression against the inside wall of the component. The tube is useful for improving component hydraulics, shielding components from contamination, inhibiting corrosion, and preventing wear to the downhole component during use. It may also be useful for positioning conduit and insulated conductors within the component. An insulating material may be disposed between the tube and the component in order to prevent galvanic corrosion of the downhole component.

  4. Nuclear Reactors and Technology

    SciTech Connect (OSTI)

    Cason, D.L.; Hicks, S.C.

    1992-01-01

    This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.

  5. MODELING HEAT TRANSFER IN SPENT FUEL TRANSFER CASK NEUTRON SHIELDS A CHALLENGING PROBLEM IN NATURAL CONVECTION

    SciTech Connect (OSTI)

    Fort, James A.; Cuta, Judith M.; Bajwa, C.; Baglietto, E.

    2010-07-18

    In the United States, commercial spent nuclear fuel is typically moved from spent fuel pools to outdoor dry storage pads within a transfer cask system that provides radiation shielding to protect personnel and the surrounding environment. The transfer casks are cylindrical steel enclosures with integral gamma and neutron radiation shields. Since the transfer cask system must be passively cooled, decay heat removal from spent nuclear fuel canister is limited by the rate of heat transfer through the cask components, and natural convection from the transfer cask surface. The primary mode of heat transfer within the transfer cask system is conduction, but some cask designs incorporate a liquid neutron shield tank surrounding the transfer cask structural shell. In these systems, accurate prediction of natural convection within the neutron shield tank is an important part of assessing the overall thermal performance of the transfer cask system. The large-scale geometry of the neutron shield tank, which is typically an annulus approximately 2 meters in diameter but only 10-15 cm in thickness, and the relatively small scale velocities (typically less than 5 cm/s) represent a wide range of spatial and temporal scales that contribute to making this a challenging problem for computational fluid dynamics (CFD) modeling. Relevant experimental data at these scales are not available in the literature, but some recent modeling studies offer insights into numerical issues and solutions; however, the geometries in these studies, and for the experimental data in the literature at smaller scales, all have large annular gaps that are not prototypic of the transfer cask neutron shield. This paper proposes that there may be reliable CFD approaches to the transfer cask problem, specifically coupled steady-state solvers or unsteady simulations; however, both of these solutions take significant computational effort. Segregated (uncoupled) steady state solvers that were tested did not accurately capture the flow field and heat transfer distribution in this application. Mesh resolution, turbulence modeling, and the tradeoff between steady state and transient solutions are addressed. Because of the critical nature of this application, the need for new experiments at representative scales is clearly demonstrated.

  6. Bioconversion reactor

    DOE Patents [OSTI]

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

    A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

  7. Cold worked ferritic alloys and components

    DOE Patents [OSTI]

    Korenko, Michael K. (Wexford, PA)

    1984-01-01

    This invention relates to liquid metal fast breeder reactor and steam generator precipitation hardening fully ferritic alloy components which have a microstructure substantially free of the primary precipitation hardening phase while having cells or arrays of dislocations of varying population densities. It also relates to the process by which these components are produced, which entails solution treating the alloy followed by a final cold working step. In this condition, the first significant precipitation hardening of the component occurs during high temperature use.

  8. Radiolysis Concerns for Water Shielding in Fission Surface Power Applications

    SciTech Connect (OSTI)

    Schoenfeld, Michael P.; Anghaie, Samim

    2008-01-21

    This paper presents an overview of radiolysis concerns with regard to water shields for fission surface power. A review of the radiolysis process is presented and key parameters and trends are identified. From this understanding of the radiolytic decomposition of water, shield pressurization and corrosion are identified as the primary concerns. Existing experimental and modeling data addressing concerns are summarized. It was found that radiolysis of pure water in a closed volume results in minimal, if any net decomposition, and therefore reduces the potential for shield pressurization and corrosion.

  9. Spherical torus fusion reactor

    DOE Patents [OSTI]

    Peng, Yueng-Kay M.

    1989-01-01

    A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

  10. Nuclear Reactors and Technology; (USA)

    SciTech Connect (OSTI)

    Cason, D.L.; Hicks, S.C.

    1991-01-01

    Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database (EDB) during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency's Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on EDB and Nuclear Science Abstracts (NSA) database. Current information, added daily to EDB, is available to DOE and its contractors through the DOE integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user's needs.

  11. A robust and well shielded thermal conductivity device for low temperature measurements

    SciTech Connect (OSTI)

    Toews, W. H.; Hill, R. W.

    2014-04-15

    We present a compact mechanically robust thermal conductivity measurement apparatus for measurements at low temperatures (<1 K) and high magnetic fields on small high-purity single crystal samples. A high-conductivity copper box is used to enclose the sample and all the components. The box provides protection for the thermometers, heater, and most importantly the sample increasing the portability of the mount. In addition to physical protection, the copper box is also effective at shielding radio frequency electromagnetic interference and thermal radiation, which is essential for low temperature measurements. A printed circuit board in conjunction with a braided ribbon cable is used to organize the delicate wiring and provide mechanical robustness.

  12. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Anderson, H.L.

    1958-10-01

    The design of control rods for nuclear reactors are described. In this design the control rod consists essentially of an elongated member constructed in part of a neutron absorbing material and having tube means extending therethrough for conducting a liquid to cool the rod when in use.

  13. Neutronic reactor

    DOE Patents [OSTI]

    Wende, Charles W. J. (Augusta, GA); Babcock, Dale F. (Wilmington, DE); Menegus, Robert L. (Wilmington, DE)

    1983-01-01

    A nuclear reactor includes an active portion with fissionable fuel and neutron moderating material surrounded by neutron reflecting material. A control element in the active portion includes a group of movable rods constructed of neutron-absorbing material. Each rod is movable with respect to the other rods to vary the absorption of neutrons and effect control over neutron flux.

  14. Technology Development Program for an Advanced Potassium Rankine Power Conversion System Compatible with Several Space Reactor Designs

    SciTech Connect (OSTI)

    Yoder, G.L.

    2005-10-03

    This report documents the work performed during the first phase of the National Aeronautics and Space Administration (NASA), National Research Announcement (NRA) Technology Development Program for an Advanced Potassium Rankine Power Conversion System Compatible with Several Space Reactor Designs. The document includes an optimization of both 100-kW{sub e} and 250-kW{sub e} (at the propulsion unit) Rankine cycle power conversion systems. In order to perform the mass optimization of these systems, several parametric evaluations of different design options were investigated. These options included feed and reheat, vapor superheat levels entering the turbine, three different material types, and multiple heat rejection system designs. The overall masses of these Nb-1%Zr systems are approximately 3100 kg and 6300 kg for the 100- kW{sub e} and 250-kW{sub e} systems, respectively, each with two totally redundant power conversion units, including the mass of the single reactor and shield. Initial conceptual designs for each of the components were developed in order to estimate component masses. In addition, an overall system concept was presented that was designed to fit within the launch envelope of a heavy lift vehicle. A technology development plan is presented in the report that describes the major efforts that are required to reach a technology readiness level of 6. A 10-year development plan was proposed.

  15. Experimental Test of Self-Shielding in VUV Photodissociation...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Experimental Test of Self-Shielding in VUV Photodissociation of CO Print One way to test models of the solar system's formation is to compare the isotopic abundances of the ...

  16. Noise modeling from high-permeability shields using Kirchhoff equations

    SciTech Connect (OSTI)

    Sandin, Henrik J; Volegov, Petr L; Espy, Michelle A; Matlashov, Andrei N; Savukov, Igor M; Schultz, Larry J

    2010-01-01

    Progress in the development of high-sensitivity magnetic-field measurements has stimulated interest in understanding magnetic noise of conductive materials, especially of magnetic shields (DC or rf) based on high-permeability materials and/or high-conductivity materials. For example, SQUIDs and atomic magnetometers have been used in many experiments with mu-metal shields, and additionally SQUID systems frequently have rf shielding based on thin conductive materials. Typical existing approaches to modeling noise only work with simple shield and sensor geometries while common experimental setups today consist of multiple sensor systems arbitrary shapes and complex shield geometries. With complex sensor arrays used in, for example, MEG and Ultra Low Field MRI studies the knowledge of the noise correlation between sensors is as important as the knowledge of the noise itself. This is crucial for incorporating efficient noise cancelation schemes for the system. We developed an approach that allows us to calculate the Johnson noise for any geometrically shaped shield and multiple sensor systems. The approach uses a fraction of the processing power of other approaches and with a multiple sensor system our approach not only calculates the noise for each sensor but it also calculates the noise correlation matrix between sensors. Here we will show the algorithm and examples where it can be implemented.

  17. A 2-D Test Problem for CFD Modeling Heat Transfer in Spent Fuel Transfer Cask Neutron Shields

    SciTech Connect (OSTI)

    Zigh, Ghani; Solis, Jorge; Fort, James A.

    2011-01-14

    In the United States, commercial spent nuclear fuel is typically moved from spent fuel pools to outdoor dry storage pads within a transfer cask system that provides radiation shielding to protect personnel and the surrounding environment. The transfer casks are cylindrical steel enclosures with integral gamma and neutron radiation shields. Since the transfer cask system must be passively cooled, decay heat removal from spent nuclear fuel canister is limited by the rate of heat transfer through the cask components, and natural convection from the transfer cask surface. The primary mode of heat transfer within the transfer cask system is conduction, but some cask designs incorporate a liquid neutron shield tank surrounding the transfer cask structural shell. In these systems, accurate prediction of natural convection within the neutron shield tank is an important part of assessing the overall thermal performance of the transfer cask system. The large-scale geometry of the neutron shield tank, which is typically an annulus approximately 2 meters in diameter but only 5-10 cm in thickness, and the relatively small scale velocities (typically less than 5 cm/s) represent a wide range of spatial and temporal scales that contribute to making this a challenging problem for computational fluid dynamics (CFD) modeling. Relevant experimental data at these scales are not available in the literature, but some recent modeling studies offer insights into numerical issues and solutions; however, the geometries in these studies, and for the experimental data in the literature at smaller scales, all have large annular gaps that are not prototypic of the transfer cask neutron shield. This paper presents results for a simple 2-D problem that is an effective numerical analog for the neutron shield application. Because it is 2-D, solutions can be obtained relatively quickly allowing a comparison and assessment of sensitivity to model parameter changes. Turbulence models are considered as well as the tradeoff between steady state and transient solutions. Solutions are compared for two commercial CFD codes, FLUENT and STAR-CCM+. The results can be used to provide input to the CFD Best Practices for this application. Following study results for the 2-D test problem, a comparison of simulation results is provided for a high Rayleigh number experiment with large annular gap. Because the geometry of this validation is significantly different from the neutron shield, and due to the critical nature of this application, the argument is made for new experiments at representative scales

  18. Materials and Components Technology Division research summary, 1991

    SciTech Connect (OSTI)

    Not Available

    1991-04-01

    This division has the purpose of providing a R and D capability for design, fabrication, and testing of high-reliability materials, components, and instrumentation. Current divisional programs are in support of the Integral Fast Reactor, life extension for light water reactors, fuels development for the new production reactor and research and test reactors, fusion reactor first-wall and blanket technology, safe shipment of hazardous materials, fluid mechanics/materials/instrumentation for fossile energy systems, and energy conservation and renewables (including tribology, high- temperature superconductivity). Separate abstracts have been prepared for the data base.

  19. BlueCross BlueShield International Claim Form | Argonne National Laboratory

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    BlueCross BlueShield International Claim Form PDF icon BCBSIL International Claim Form

  20. REACTOR MONITORING

    DOE Patents [OSTI]

    Bugbee, S.J.; Hanson, V.F.; Babcock, D.F.

    1959-02-01

    A neutron density inonitoring means for reactors is described. According to this invention a tunnel is provided beneath and spaced from the active portion of the reactor and extends beyond the opposite faces of the activc portion. Neutron beam holes are provided between the active portion and the tunnel and open into the tunnel near the middle thereof. A carriage operates back and forth in the tunnel and is adapted to convey a neutron detector, such as an ion chamber, and position it beneath one of the neutron beam holes. This arrangement affords convenient access of neutron density measuring instruments to a location wherein direct measurement of neutron density within the piles can be made and at the same time affords ample protection to operating personnel.

  1. Synfuel production in nuclear reactors

    DOE Patents [OSTI]

    Henning, C.D.

    Apparatus and method for producing synthetic fuels and synthetic fuel components by using a neutron source as the energy source, such as a fusion reactor. Neutron absorbers are disposed inside a reaction pipe and are heated by capturing neutrons from the neutron source. Synthetic fuel feedstock is then placed into contact with the heated neutron absorbers. The feedstock is heated and dissociates into its constituent synfuel components, or alternatively is at least preheated sufficiently to use in a subsequent electrolysis process to produce synthetic fuels and synthetic fuel components.

  2. USE OF LEAD SHIELDING Nevada National Security Site Waste Acceptance Criteria (NNSSWAC)

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    USE OF LEAD SHIELDING Nevada National Security Site Waste Acceptance Criteria (NNSSWAC) Container ID: Waste profile: NOTE: If dose rate is < 5 mrem/hr at 30 cm from the standard waste package, lead shielding is unacceptable. Dose rate of loaded waste package without lead shielding: mrem/hr @ 30 cm (attach calculations) Describe lead shielding: Basis for demonstrating lead is not excessive (attached calculations, e.g. Microshield analysis, or attach dose rate survey): Maximum lead shielded

  3. Benefits of Small Modular Reactors (SMRs) | Department of Energy

    Energy Savers [EERE]

    Benefits of Small Modular Reactors (SMRs) Benefits of Small Modular Reactors (SMRs) Small modular reactors offer a lower initial capital investment, greater scalability, and siting flexibility for locations unable to accommodate more traditional larger reactors. They also have the potential for enhanced safety and security compared to earlier designs. Modularity: The term "modular" in the context of SMRs refers to the ability to fabricate major components of the nuclear steam supply

  4. State space modeling of reactor core in a pressurized water reactor

    SciTech Connect (OSTI)

    Ashaari, A.; Ahmad, T.; M, Wan Munirah W.; Shamsuddin, Mustaffa; Abdullah, M. Adib

    2014-07-10

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  5. Space-reactor electric systems: subsystem technology assessment

    SciTech Connect (OSTI)

    Anderson, R.V.; Bost, D.; Determan, W.R.

    1983-03-29

    This report documents the subsystem technology assessment. For the purpose of this report, five subsystems were defined for a space reactor electric system, and the report is organized around these subsystems: reactor; shielding; primary heat transport; power conversion and processing; and heat rejection. The purpose of the assessment was to determine the current technology status and the technology potentials for different types of the five subsystems. The cost and schedule needed to develop these potentials were estimated, and sets of development-compatible subsystems were identified.

  6. DETECTION OF COATING FAILURES IN A NEUTRONIC REACTOR

    DOE Patents [OSTI]

    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.

  7. TEST-HOLE CONSTRUCTION FOR A NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Ohlinger, L.A.; Seitz, F.; Young, G.J.

    1959-02-17

    Test-hole construction is described for a reactor which provides safe and ready access to the neutron flux region for specimen materials which are to be irradiated therein. An elongated tubular thimble adapted to be inserted in the access hole through the wall of the reactor is constructed of aluminum and is provided with a plurality of holes parallel to the axis of the thimble for conveying the test specimens into position for irradiation, and a conduit for the circulation of coolant. A laminated shield formed of alternate layers of steel and pressed wood fiber is disposed lengthwise of the thimble near the outer end thereof.

  8. Reference reactor module for NASA's lunar surface fission power system

    SciTech Connect (OSTI)

    Poston, David I; Kapernick, Richard J; Dixon, David D; Werner, James; Qualls, Louis; Radel, Ross

    2009-01-01

    Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. The AFSPS concept is now being further developed within the Fission Surface Power (FSP) Project, which is a near-term technology program to demonstrate system-level TRL-6 by 2013. This paper describes the reference FSP reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based. UO{sub 2}-fueled, pumped-NaK fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. The ultimate goal of this work is to provide a 'workhorse' power system that NASA can utilize in near-term and future Lunar and Martian mission architectures, with the eventual capability to evolve to very high power, low mass systems, for either surface, deep space, and/or orbital missions.

  9. Space reactor electric systems: system integration studies, Phase 1 report

    SciTech Connect (OSTI)

    Anderson, R.V.; Bost, D.; Determan, W.R.; Harty, R.B.; Katz, B.; Keshishian, V.; Lillie, A.F.; Thomson, W.B.

    1983-03-29

    This report presents the results of preliminary space reactor electric system integration studies performed by Rockwell International's Energy Systems Group (ESG). The preliminary studies investigated a broad range of reactor electric system concepts for powers of 25 and 100 KWe. The purpose of the studies was to provide timely system information of suitable accuracy to support ongoing mission planning activities. The preliminary system studies were performed by assembling the five different subsystems that are used in a system: the reactor, the shielding, the primary heat transport, the power conversion-processing, and the heat rejection subsystems. The subsystem data in this report were largely based on Rockwell's recently prepared Subsystem Technology Assessment Report. Nine generic types of reactor subsystems were used in these system studies. Several levels of technology were used for each type of reactor subsystem. Seven generic types of power conversion-processing subsystems were used, and several levels of technology were again used for each type. In addition, various types and levels of technology were used for the shielding, primary heat transport, and heat rejection subsystems. A total of 60 systems were studied.

  10. Dismantlement of the TSF-SNAP Reactor Assembly

    SciTech Connect (OSTI)

    Peretz, Fred J

    2009-01-01

    This paper describes the dismantlement of the Tower Shielding Facility (TSF)?Systems for Nuclear Auxiliary Power (SNAP) reactor, a SNAP-10A reactor used to validate radiation source terms and shield performance models at Oak Ridge National Laboratory (ORNL) from 1967 through 1973. After shutdown, it was placed in storage at the Y-12 National Security Complex (Y-12), eventually falling under the auspices of the Highly Enriched Uranium (HEU) Disposition Program. To facilitate downblending of the HEU present in the fuel elements, the TSF-SNAP was moved to ORNL on June 24, 2006. The reactor assembly was removed from its packaging, inspected, and the sodium-potassium (NaK) coolant was drained. A superheated steam process was used to chemically react the residual NaK inside the reactor assembly. The heat exchanger assembly was removed from the top of the reactor vessel, and the criticality safety sleeve was exchanged for a new safety sleeve that allowed for the removal of the vessel lid. A chain-mounted tubing cutter was used to separate the lid from the vessel, and the 36 fuel elements were removed and packaged in four U.S. Department of Transportation 2R/6M containers. The fuel elements were returned to Y-12 on July 13, 2006. The return of the fuel elements and disposal of all other reactor materials accomplished the formal objectives of the dismantlement project. In addition, a project model was established for the handling of a fully fueled liquid-metal?cooled reactor assembly. Current criticality safety codes have been benchmarked against experiments performed by Atomics International in the 1950s and 1960s. Execution of this project provides valuable experience applicable to future projects addressing space and liquid-metal-cooled reactors.

  11. Photocatalytic reactor

    DOE Patents [OSTI]

    Bischoff, B.L.; Fain, D.E.; Stockdale, J.A.D.

    1999-01-19

    A photocatalytic reactor is described for processing selected reactants from a fluid medium comprising at least one permeable photocatalytic membrane having a photocatalytic material. The material forms an area of chemically active sites when illuminated by light at selected wavelengths. When the fluid medium is passed through the illuminated membrane, the reactants are processed at these sites separating the processed fluid from the unprocessed fluid. A light source is provided and a light transmitting means, including an optical fiber, for transmitting light from the light source to the membrane. 4 figs.

  12. Evaluation Of Shielding Efficacy Of A Ferrite Containing Ceramic Material

    SciTech Connect (OSTI)

    Verst, C.

    2015-10-12

    The shielding evaluation of the ferrite based Mitsuishi ceramic material has produced for several radiation sources and possible shielding sizes comparative dose attenuation measurements and simulated projections. High resolution gamma spectroscopy provided uncollided and scattered photon spectra at three energies, confirming theoretical estimates of the ceramic’s mass attenuation coefficient, μ/ρ. High level irradiation experiments were performed using Co-60, Cs-137, and Cf-252 sources to measure penetrating dose rates through steel, lead, concrete, and the provided ceramic slabs. The results were used to validate the radiation transport code MCNP6 which was then used to generate dose rate attenuation curves as a function of shielding material, thickness, and mass for photons and neutrons ranging in energy from 200 keV to 2 MeV.

  13. Radiation Shielding Properties of Some Marbles in Turkey

    SciTech Connect (OSTI)

    Guenoglu, K.; Akkurt, I.

    2011-12-26

    Especially after development of technology, radiation started to be used in a large fields such as medicine, industry and energy. Using radiation in those fields bring hazardous effect of radiation into humancell. Thus radiation protection becomes important in physics. Although there are three ways for radiation protection, shielding of the radiation is the most commonly used method. Natural Stones such as marble is used as construction material especially in critical building and thus its radiation shielding capability should be determined.In this study, gamma ray shielding properties of some different types of marble mined in Turkey, have been measured using a NaI(Tl) scintillator detector. The measured results were also compared with the theoretical calculations XCOM.

  14. Investigating Radiation Shielding Properties of Different Mineral Origin Heavyweight Concretes

    SciTech Connect (OSTI)

    Basyigit, Celalettin; Uysal, Volkan; Kilincarslan, Semsettin; Akkas, Ayse; Mavi, Betuel; Guenoglu, Kadir; Akkurt, Iskender

    2011-12-26

    The radiation although has hazardous effects for human health, developing technologies bring lots of usage fields to radiation like in medicine and nuclear power station buildings. In this case protecting from undesirable radiation is a necessity for human health. Heavyweight concrete is one of the most important materials used in where radiation should be shielded, like those areas. In this study, used heavyweight aggregates of different mineral origin (Limonite, Siderite), in order to prepare different series in concrete mixtures and investigated radiation shielding properties. The experimental results on measuring the radiation shielding, the heavyweight concrete prepared with heavyweight aggregates of different mineral origin show that, are useful radiation absorbents when they used in concrete mixtures.

  15. Shielded helix traveling wave cathode ray tube deflection structure

    DOE Patents [OSTI]

    Norris, N.J.; Hudson, C.L.

    1992-12-15

    Various embodiments of a helical coil deflection structure of a CRT are described and illustrated which provide shielding between adjacent turns of the coil on either three or four sides of each turn in the coil. Threaded members formed with either male or female threads and having the same pitch as the deflection coil are utilized for shielding the deflection coil with each turn of the helical coil placed between adjacent threads which act to shield each coil turn from adjacent turns and to confine the field generated by the coil to prevent or inhibit cross-coupling between adjacent turns of the coil to thereby prevent generation of fast fields which might otherwise deflect the beam out of time synchronization with the electron beam pulse. 13 figs.

  16. Shielded helix traveling wave cathode ray tube deflection structure

    DOE Patents [OSTI]

    Norris, Neil J. (Santa Barbara, CA); Hudson, Charles L. (Santa Barbara, CA)

    1992-01-01

    Various embodiments of a helical coil deflection structure of a CRT are described and illustrated which provide shielding between adjacent turns of the coil on either three or four sides of each turn in the coil. Threaded members formed with either male or female threads and having the same pitch as the deflection coil are utilized for shielding the deflection coil with each turn of the helical coil placed between adjacent threads which act to shield each coil turn from adjacent turns and to confine the field generated by the coil to prevent or inhibit cross-coupling between adjacent turns of the coil to thereby prevent generation of fast fields which might otherwise deflect the beam out of time synchronization with the electron beam pulse.

  17. A Basic LEGO Reactor Design for the Provision of Lunar Surface Power

    SciTech Connect (OSTI)

    John Darrell Bess

    2008-06-01

    A final design has been established for a basic Lunar Evolutionary Growth-Optimized (LEGO) Reactor using current and near-term technologies. The LEGO Reactor is a modular, fast-fission, heatpipe-cooled, clustered-reactor system for lunar-surface power generation. The reactor is divided into subcritical units that can be safely launched with lunar shipments from Earth, and then emplaced directly into holes drilled into the lunar regolith to form a critical reactor assembly. The regolith would not just provide radiation shielding, but serve as neutron-reflector material as well. The reactor subunits are to be manufactured using proven and tested materials for use in radiation environments, such as uranium-dioxide fuel, stainless-steel cladding and structural support, and liquid-sodium heatpipes. The LEGO Reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of an increase in launch mass per overall rated power level and a reduction in neutron economy when compared to a single-reactor system. A single unshielded LEGO Reactor subunit has an estimated mass of approximately 448 kg and provides approximately 5 kWe. The overall envelope for a single subunit with fully extended radiator panels has a height of 8.77 m and a diameter of 0.50 m. Six subunits could provide sufficient power generation throughout the initial stages of establishing a lunar outpost. Portions of the reactor may be neutronically decoupled to allow for reduced power production during unmanned periods of base operations. During later stages of lunar-base development, additional subunits may be emplaced and coupled into the existing LEGO Reactor network, subject to lunar base power demand. Improvements in reactor control methods, fuel form and matrix, shielding, as well as power conversion and heat rejection techniques can help generate an even more competitive LEGO Reactor design. Further modifications in the design could provide power generative opportunities for use on other extraterrestrial surfaces.

  18. Hybrid adsorptive membrane reactor

    DOE Patents [OSTI]

    Tsotsis, Theodore T. (Huntington Beach, CA); Sahimi, Muhammad (Altadena, CA); Fayyaz-Najafi, Babak (Richmond, CA); Harale, Aadesh (Los Angeles, CA); Park, Byoung-Gi (Yeosu, KR); Liu, Paul K. T. (Lafayette Hill, PA)

    2011-03-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  19. Hanford Waste Treatment Plant Sets Massive Protective Shield door in

    Office of Environmental Management (EM)

    Pretreatment Facility | Department of Energy Waste Treatment Plant Sets Massive Protective Shield door in Pretreatment Facility Hanford Waste Treatment Plant Sets Massive Protective Shield door in Pretreatment Facility January 12, 2011 - 12:00pm Addthis The carbon steel doors come together to form an upside-down L-shape. The 102-ton door was set on top of the 85-ton door that was installed at the end of December. The carbon steel doors come together to form an upside-down L-shape. The

  20. SU-D-BRE-07: Neutron Shielding Assessment for a Compact Proton Therapy Vault

    SciTech Connect (OSTI)

    Prusator, M; Ahmad, S; Chen, Y [University of Oklahoma Health Sciences Center, Oklahoma City, OK (United States)

    2014-06-01

    Purpose: To perform a neutron shielding assessment of a commercially available compact proton therapy system. Methods: TOPAS (TOol for PArticle Simulation) beta release was used to model beam line components for Mevion S250 proton treatment system the design of which is that the cyclotron is present in the treatment room. Three neutron production sources were taken into account in the simulation. These are the cyclotron, the treatment nozzle and the patient itself, respectively. The cyclotron was modeled as a cylindrical iron target (r =5 cm, length = 8 cm). A water phantom (10 cm 10 cm 60 cm) was used to model the patient and various structures (scattering foils, range modulator wheel, applicator and compensator) defaulted in TOPAS were used to model the passive scattering treatment nozzle. Neutron fluences and energy spectra were counted in a spherical scoring geometry per incident proton in 18 angular bins (10 degree each). Fluence to dose conversion factors from ICRU publication 74 were used to acquire neutron ambient dose equivalent H*(10). A point source line of sight model was then used to calculate neutron dose at eight locations beyond shielding barriers. Results: The neutron ambient dose equivalent was calculated at the 8 points of interest around the proton treatment vault. The highest dose was found to be less than 0.781 mSv/year outside south barrier wall. However, the dose is less than 0.05 mSv/year at the control room area of the proton vault. Conclusion: All Points of interest were well under annual dose limits. This suggests that the shielding design of this compact proton therapy system is sufficient for radiation protection purpose. However, it is important to note that the workload and the occupancy factors are direct multipliers for dose calculations beyond the barrier and must be accurately estimated for validation of our results.

  1. Application to Export Electric Energy OE Docket No. EA-415 Lion Shield

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

    Energy, LLC | Department of Energy 415 Lion Shield Energy, LLC Application to Export Electric Energy OE Docket No. EA-415 Lion Shield Energy, LLC Application from Lion Shield Energy to export electric energy to Mexico. PDF icon EA-415 Lion Shield Energy (MX).pdf More Documents & Publications EA-415 Lion Shield Energy, LLC Application to Export Electric Energy OE Docket No. EA-415 Lion Shield Energy, LLC: Federal Register Notice, Volume 80, No. 164 - Aug. 25, 2015 Application to Export

  2. Small Modular Reactors - SRSCRO

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    River National Laboratory (SRNL) has announced several partnerships to bring refrigerator-sized modular nuclear reactors, known as Small Modular Reactors or SMRs, to the...

  3. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Stewart, H.B.

    1958-12-23

    A nuclear reactor of the type speclfically designed for the irradiation of materials is discussed. In this design a central cyllndrical core of moderating material ls surrounded by an active portlon comprlsed of an annular tank contalning fissionable material immersed ln a liquid moderator. The active portion ls ln turn surrounded by a reflector, and a well ls provided in the center of the core to accommodate the materlals to be irradiated. The over-all dimensions of the core ln at least one plane are equal to or greater than twice the effective slowing down length and equal to or less than twlce the effective diffuslon length for neutrons in the core materials.

  4. Vanadium recycling for fusion reactors

    SciTech Connect (OSTI)

    Dolan, T.J.; Butterworth, G.J.

    1994-04-01

    Very stringent purity specifications must be applied to low activation vanadium alloys, in order to meet recycling goals requiring low residual dose rates after 50--100 years. Methods of vanadium production and purification which might meet these limits are described. Following a suitable cooling period after their use, the vanadium alloy components can be melted in a controlled atmosphere to remove volatile radioisotopes. The aim of the melting and decontamination process will be the achievement of dose rates low enough for ``hands-on`` refabrication of new reactor components from the reclaimed metal. The processes required to permit hands-on recycling appear to be technically feasible, and demonstration experiments are recommended. Background information relevant to the use of vanadium alloys in fusion reactors, including health hazards, resources, and economics, is provided.

  5. Testing and Commissioning of a Multifunctional Tool for the Dismantling of the Activated Internals of the KNK Reactor Shaft - 13524

    SciTech Connect (OSTI)

    Rothschmitt, Stefan; Graf, Anja; Bauer, Stefan; Klute, Stefan; Koselowski, Eiko

    2013-07-01

    The Compact Sodium Cooled Reactor Facility Karlsruhe (KNK), a prototype reactor to demonstrate the Fast Breeder Reactor Technology in Germany, was in operation from 1971 to 1991. The dismantling activities started in 1991. The project aim is the green field in 2020. Most of the reactor internals as well as the primary and secondary cooling loops are already dismantled. The total contaminated sodium inventory has already been disposed of. Only the high activated reactor vessel shielding structures are remaining. Due to the high dose rates these structures must be dismantled remotely. For the dismantling of the primary shielding of the reactor vessel, 12 stacked cast iron blocks with a total mass of 90 Mg and single masses up to 15.5 Mg, a remote-controlled multifunctional dismantling device (HWZ) was designed, manufactured and tested in a mock-up. After successful approval of the test sequences by the authorities, the HWZ was implemented into the reactor building containment for final assembling of the auxiliary equipment and subsequent hot commissioning in 2012. Dismantling of the primary shielding blocks is scheduled for early 2013. (authors)

  6. Nuclear Reactor Safety Design Criteria

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1993-01-19

    The order establishes nuclear safety criteria applicable to the design, fabrication, construction, testing, and performance requirements of nuclear reactor facilities and safety class structures, systems, and components (SSCs) within these facilities. Cancels paragraphs 8a and 8b of DOE 5480.6. Cancels DOE O 5480.6 in part. Supersedes DOE 5480.1, dated 1-19-93. Certified 11-18-10.

  7. Fault current limiter with shield and adjacent cores

    DOE Patents [OSTI]

    Darmann, Francis Anthony; Moriconi, Franco; Hodge, Eoin Patrick

    2013-10-22

    In a fault current limiter (FCL) of a saturated core type having at least one coil wound around a high permeability material, a method of suppressing the time derivative of the fault current at the zero current point includes the following step: utilizing an electromagnetic screen or shield around the AC coil to suppress the time derivative current levels during zero current conditions.

  8. 2008-01 "Expedite Planned Change Request for Shielded Containers"

    Broader source: Energy.gov [DOE]

    Approved July 30, 2008 The intent of this recommendation is to encourage DOE to provide strong support for using shielded containers for removal of RH TRU waste, so that the TRU wastes from the LANL Area G can be removed on schedule.

  9. RZ calculations for self shielded multigroup cross sections

    SciTech Connect (OSTI)

    Li, M.; Sanchez, R.; Zmijarevic, I.; Stankovski, Z.

    2006-07-01

    A collision probability method has been implemented for RZ geometries. The method accounts for white albedo, specular and translation boundary condition on the top and bottom surfaces of the geometry and for a white albedo condition on the outer radial surface. We have applied the RZ CP method to the calculation of multigroup self shielded cross sections for Gadolinia absorbers in BWRs. (authors)

  10. Gamma Shield Thunder Exercise Concludes | National Nuclear Security

    National Nuclear Security Administration (NNSA)

    Administration Gamma Shield Thunder Exercise Concludes | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Countering Nuclear Terrorism About Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Library Bios Congressional Testimony Fact Sheets Newsletters Press Releases Photo Gallery Jobs

  11. SRS Small Modular Reactors

    ScienceCinema (OSTI)

    None

    2014-05-21

    The small modular reactor program at the Savannah River Site and the Savannah River National Laboratory.

  12. Reactor safety method

    DOE Patents [OSTI]

    Vachon, Lawrence J. (Clairton, PA)

    1980-03-11

    This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

  13. Blue Cross Blue Shield of Illinois PPO Claim Form | Argonne National

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Laboratory Blue Cross Blue Shield of Illinois PPO Claim Form PDF icon BCBSIL PPO Claim Form

  14. Application to Export Electric Energy OE Docket No. EA-415 Lion Shield

    Energy Savers [EERE]

    Energy, LLC: Federal Register Notice, Volume 80, No. 164 - Aug. 25, 2015 | Department of Energy 5 Lion Shield Energy, LLC: Federal Register Notice, Volume 80, No. 164 - Aug. 25, 2015 Application to Export Electric Energy OE Docket No. EA-415 Lion Shield Energy, LLC: Federal Register Notice, Volume 80, No. 164 - Aug. 25, 2015 Application from Lion Shield Energy to export electric energy to Mexico. Federal Register Notice. PDF icon EA-415 Lion Shield Energy (MX).pdf More Documents &

  15. Nuclear reactor

    DOE Patents [OSTI]

    Thomson, Wallace B. (Severna Park, MD)

    2004-03-16

    A nuclear reactor comprising a cylindrical pressure vessel, an elongated annular core centrally disposed within and spaced from the pressure vessel, and a plurality of ducts disposed longitudinally of the pressure vessel about the periphery thereof, said core comprising an annular active portion, an annular reflector just inside the active portion, and an annular reflector just outside the active a portion, said annular active portion comprising rectangular slab, porous fuel elements radially disposed around the inner reflector and extending the length of the active portion, wedge-shaped, porous moderator elements disposed adjacent one face of each fuel element and extending the length of the fuel element, the fuel and moderator elements being oriented so that the fuel elements face each other and the moderator elements do likewise, adjacent moderator elements being spaced to provide air inlet channels, and adjacent fuel elements being spaced to provide air outlet channels which communicate with the interior of the peripheral ducts, and means for introducing air into the air inlet channels which passes through the porous moderator elements and porous fuel elements to the outlet channel.

  16. Density functional investigation of intermolecular effects on {sup 13}C NMR chemical-shielding tensors modeled with molecular clusters

    SciTech Connect (OSTI)

    Holmes, Sean T.; Dybowski, Cecil; Iuliucci, Robbie J.; Mueller, Karl T.

    2014-10-28

    A quantum-chemical method for modeling solid-state nuclear magnetic resonance chemical-shift tensors by calculations on large symmetry-adapted clusters of molecules is demonstrated. Four hundred sixty five principal components of the {sup 13}C chemical-shielding tensors of 24 organic materials are analyzed. The comparison of calculations on isolated molecules with molecules in clusters demonstrates that intermolecular effects can be successfully modeled using a cluster that represents a local portion of the lattice structure, without the need to use periodic-boundary conditions (PBCs). The accuracy of calculations which model the solid state using a cluster rivals the accuracy of calculations which model the solid state using PBCs, provided the cluster preserves the symmetry properties of the crystalline space group. The size and symmetry conditions that the model cluster must satisfy to obtain significant agreement with experimental chemical-shift values are discussed. The symmetry constraints described in the paper provide a systematic approach for incorporating intermolecular effects into chemical-shielding calculations performed at a level of theory that is more advanced than the generalized gradient approximation. Specifically, NMR parameters are calculated using the hybrid exchange-correlation functional B3PW91, which is not available in periodic codes. Calculations on structures of four molecules refined with density plane waves yield chemical-shielding values that are essentially in agreement with calculations on clusters where only the hydrogen sites are optimized and are used to provide insight into the inherent sensitivity of chemical shielding to lattice structure, including the role of rovibrational effects.

  17. Repair welding of fusion reactor components. Second year technical report

    SciTech Connect (OSTI)

    Chin, B.A.

    1993-05-15

    Experiments have shown that irradiated Type 316 stainless steel is susceptible to heat-affected-zone (HAZ) cracking upon cooling when welded using the gas tungsten arc (GTA) process under lateral constraint. The cracking has been hypothesized to be caused by stress-assisted helium bubble growth and rupture at grain boundaries. This study utilized an experimental welding setup which enabled different compressive stresses to be applied to the plates during welding. Autogenous GTA welds were produced in Type 316 stainless steel doped with 256 appm helium. The application of a compressive stress, 55 MPa, during welding suppressed the previously observed catastrophic cracking. Detailed examinations conducted after welding showed a dramatic change in helium bubble morphology. Grain boundary bubble growth along directions parallel to the weld was suppressed. Results suggest that stress-modified welding techniques may be used to suppress or eliminate helium-induced cracking during joining of irradiated materials.

  18. Validation of nuclear models used in space radiation shielding applications

    SciTech Connect (OSTI)

    Norman, Ryan B.; Blattnig, Steve R.

    2013-01-15

    A program of verification and validation has been undertaken to assess the applicability of models to space radiation shielding applications and to track progress as these models are developed over time. In this work, simple validation metrics applicable to testing both model accuracy and consistency with experimental data are developed. The developed metrics treat experimental measurement uncertainty as an interval and are therefore applicable to cases in which epistemic uncertainty dominates the experimental data. To demonstrate the applicability of the metrics, nuclear physics models used by NASA for space radiation shielding applications are compared to an experimental database consisting of over 3600 experimental cross sections. A cumulative uncertainty metric is applied to the question of overall model accuracy, while a metric based on the median uncertainty is used to analyze the models from the perspective of model development by examining subsets of the model parameter space.

  19. Accordian-folded boot shield for flexible swivel connection

    DOE Patents [OSTI]

    Hoh, Joseph C. (Naperville, IL)

    1986-01-01

    A flexible swivel boot connector for connecting a first boot shield section to a second boot shield section, both first and second boot sections having openings therethrough, the second boot section having at least two adjacent accordian folds at the end having the opening, the second boot section being positioned through the opening of the first boot section such that a first of the accordian folds is within the first boot section and a second of the accordian folds is outside of the first boot, includes first and second annular discs, the first disc being positioned within and across the first accordian fold, the second disc being positioned within and across the second accordian fold, such that the first boot section is moveably and rigidly connected between the first and second accordian folds of the second boot section.

  20. Spent nuclear fuel discharges from U.S. reactors 1994

    SciTech Connect (OSTI)

    1996-02-01

    Spent Nuclear Fuel Discharges from US Reactors 1994 provides current statistical data on fuel assemblies irradiated at commercial nuclear reactors operating in the US. This year`s report provides data on the current inventories and storage capacities at these reactors. Detailed statistics on the data are presented in four chapters that highlight 1994 spent fuel discharges, storage capacities and inventories, canister and nonfuel component data, and assembly characteristics. Five appendices, a glossary, and bibliography are also included. 10 figs., 34 tabs.

  1. Physical analysis of the radiation shielding for the medical accelerators

    SciTech Connect (OSTI)

    Li, Q. F.; Xing, Q. Z.; Kong, C. C.

    2009-02-01

    Radiation safety standards today require comprehensive shielding protection schemes for all particle accelerators. The original shielding system of BJ-20 (BeiJing-20 MeV), the high-energy medical electron linac, was designed only for the 18 MeV level. And the dose caused by the lost electrons in the 270 deg. bending magnet system was neglected. In this paper, the leakage dose of BJ-20 is carefully analyzed. The radiation leakage dose distribution of the photons coming from the accelerator head is obtained for energy levels of 6, 12, 14, and 18 MeV. The dose of the photoneutrons is especially analyzed for the 18 MeV level. The result gives that even neglecting the dose from the 270 deg. bending magnet system, the shielding system is still not enough for the energy levels lower than 18 MeV. The radiation leakage produced by electrons that are lost in the 270 deg. bending magnet system has been particularly studied. Using beam transport theory and Monte Carlo sampling methods, which have been combined in calculations, we have obtained the distribution of the energy, position, and direction of the lost electrons. These data were then further processed by the Monte Carlo N-particle (MCNP) code as input data. The results show that when the electron loss rate in the 270 deg. bending magnet system is 13.5%, the radiation leakage dose of the photons generated by the lost electrons is 0.1% higher than that at the isocenter, and the corresponding relative leakage dose of the photoneutrons reaches 0.045% around an angle of 170 deg. at 18 MeV level. Both of these parameters exceed radioprotection safety standards for medical accelerators. The original shielding design is therefore not suitable and is also incomplete since the radiation produced by the electrons being lost in the 270 deg. bending magnet system was neglected and the leakage dose for the low-energy levels was not considered in the original design. Our calculations provide a very useful tool for further optimization and design improvement that will enable this radiation shielding to conform to present day safety standards.

  2. BWR spent fuel storage cask performance test. Volume 1. Cask handling experience and decay heat, heat transfer, and shielding data

    SciTech Connect (OSTI)

    McKinnon, M.A.; Doman, J.W.; Tanner, J.E.; Guenther, R.J.; Creer, J.M.; King, C.E.

    1986-02-01

    This report documents a heat transfer and shielding performance test conducted on a Ridihalgh, Eggers and Associates REA 2023 boiling water reactor (BWR) spent fuel storage cask. The testing effort consisted of three parts: pretest preparations, performance testing, and post-test activities. Pretest preparations included conducting cask handling dry runs and characterizing BWR spent fuel assemblies from Nebraska Public Power District's Cooper Nuclear Station. The performance test matrix included 14 runs consisting of two loadings, two cask orientations, and three backfill environments. Post-test activities included calorimetry and axial radiation scans of selected fuel assemblies, in-basin sipping of each assembly, crud collection, video and photographic scans, and decontamination of the cask interior and exterior.

  3. Fluid sampling system for a nuclear reactor

    DOE Patents [OSTI]

    Lau, L.K.; Alper, N.I.

    1994-11-22

    A system of extracting fluid samples, either liquid or gas, from the interior of a nuclear reactor containment utilizes a jet pump. To extract the sample fluid, a nonradioactive motive fluid is forced through the inlet and discharge ports of a jet pump located outside the containment, creating a suction that draws the sample fluid from the containment through a sample conduit connected to the pump suction port. The mixture of motive fluid and sample fluid is discharged through a return conduit to the interior of the containment. The jet pump and means for removing a portion of the sample fluid from the sample conduit can be located in a shielded sample grab station located next to the containment. A non-nuclear grade active pump can be located outside the grab sampling station and the containment to pump the nonradioactive motive fluid through the jet pump. 1 fig.

  4. Fluid sampling system for a nuclear reactor

    DOE Patents [OSTI]

    Lau, Louis K. (Monroeville, PA); Alper, Naum I. (Monroeville, PA)

    1994-01-01

    A system of extracting fluid samples, either liquid or gas, from the interior of a nuclear reactor containment utilizes a jet pump. To extract the sample fluid, a nonradioactive motive fluid is forced through the inlet and discharge ports of a jet pump located outside the containment, creating a suction that draws the sample fluid from the containment through a sample conduit connected to the pump suction port. The mixture of motive fluid and sample fluid is discharged through a return conduit to the interior of the containment. The jet pump and means for removing a portion of the sample fluid from the sample conduit can be located in a shielded sample grab station located next to the containment. A non-nuclear grade active pump can be located outside the grab sampling station and the containment to pump the nonradioactive motive fluid through the jet pump.

  5. Attrition reactor system

    DOE Patents [OSTI]

    Scott, Charles D. (Oak Ridge, TN); Davison, Brian H. (Knoxvile, TN)

    1993-01-01

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur.

  6. Attrition reactor system

    DOE Patents [OSTI]

    Scott, C.D.; Davison, B.H.

    1993-09-28

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur. 2 figures.

  7. B Reactor - Hanford Site

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    War II, B Reactor produced plutonium used in the Trinity Test, as well as for the atomic bomb dropped on Nagasaki, Japan, to end World War II. The reactor was designed and built...

  8. Period meter for reactors

    DOE Patents [OSTI]

    Rusch, Gordon K.

    1976-01-06

    An improved log N amplifier type nuclear reactor period meter with reduced probability for noise-induced scrams is provided. With the reactor at low power levels a sampling circuit is provided to determine the reactor period by measuring the finite change in the amplitude of the log N amplifier output signal for a predetermined time period, while at high power levels, differentiation of the log N amplifier output signal provides an additional measure of the reactor period.

  9. Improved vortex reactor system

    DOE Patents [OSTI]

    Diebold, James P. (Lakewood, CO); Scahill, John W. (Evergreen, CO)

    1995-01-01

    An improved vortex reactor system for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor.

  10. Advanced Test Reactor Tour

    ScienceCinema (OSTI)

    Miley, Don

    2013-05-28

    The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

  11. High solids fermentation reactor

    DOE Patents [OSTI]

    Wyman, Charles E.; Grohmann, Karel; Himmel, Michael E.; Richard, Christopher J.

    1993-03-02

    A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

  12. High solids fermentation reactor

    DOE Patents [OSTI]

    Wyman, Charles E.; Grohmann, Karel; Himmel, Michael E.; Richard, Christopher J.

    1993-01-01

    A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

  13. NUCLEAR REACTOR CONTROL SYSTEM

    DOE Patents [OSTI]

    Epler, E.P.; Hanauer, S.H.; Oakes, L.C.

    1959-11-01

    A control system is described for a nuclear reactor using enriched uranium fuel of the type of the swimming pool and other heterogeneous nuclear reactors. Circuits are included for automatically removing and inserting the control rods during the course of normal operation. Appropriate safety circuits close down the nuclear reactor in the event of emergency.

  14. NGNP Reactor Coolant Chemistry Control Study

    SciTech Connect (OSTI)

    Brian Castle

    2010-11-01

    The main focus of this paper is to identify the most desirable ranges of impurity levels in the primary coolant to optimize component life in the primary circuit of the Next Generation Nuclear Plant (NGNP), which will either be a prismatic block or pebble bed reactor.

  15. A NEW ALGORITHM FOR RADIOISOTOPE IDENTIFICATION OF SHIELDED AND MASKED SNM/RDD MATERIALS

    SciTech Connect (OSTI)

    Jeffcoat, R.

    2012-06-05

    Detection and identification of shielded and masked nuclear materials is crucial to national security, but vast borders and high volumes of traffic impose stringent requirements for practical detection systems. Such tools must be be mobile, and hence low power, provide a low false alarm rate, and be sufficiently robust to be operable by non-technical personnel. Currently fielded systems have not achieved all of these requirements simultaneously. Transport modeling such as that done in GADRAS is able to predict observed spectra to a high degree of fidelity; our research is focusing on a radionuclide identification algorithm that inverts this modeling within the constraints imposed by a handheld device. Key components of this work include incorporation of uncertainty as a function of both the background radiation estimate and the hypothesized sources, dimensionality reduction, and nonnegative matrix factorization. We have partially evaluated performance of our algorithm on a third-party data collection made with two different sodium iodide detection devices. Initial results indicate, with caveats, that our algorithm performs as good as or better than the on-board identification algorithms. The system developed was based on a probabilistic approach with an improved approach to variance modeling relative to past work. This system was chosen based on technical innovation and system performance over algorithms developed at two competing research institutions. One key outcome of this probabilistic approach was the development of an intuitive measure of confidence which was indeed useful enough that a classification algorithm was developed based around alarming on high confidence targets. This paper will present and discuss results of this novel approach to accurately identifying shielded or masked radioisotopes with radiation detection systems.

  16. 2013 R&D 100 Award: 'SHIELD' protects NIF optics from harmful pulses

    ScienceCinema (OSTI)

    Chou, Jason

    2014-07-22

    In the past, it took as long as 12 hours to manually screen 48 critical checkpoints at the National Ignition Facility (NIF) for harmful laser pulses. The screening equipment had to be moved from point to point throughout a facility the size of three football fields. Now with a new technology, called Laser SHIELD (Screening at High-throughput to Identify Energetic Laser Distortion), and with the push of a button, the screening can be done in less than one second. Proper screening of pulses is critical for the operation of high-energy lasers to ensure that the laser does not exceed safe operating conditions for optics. The energetic beams of light are so powerful that, when left uncontrolled, they can shatter the extremely valuable glass inside the laser. If a harmful pulse is found, immediate adjustments can be made in order to protect the optics for the facility. Laser SHIELD is a custom-designed high-throughput screening system built from low-cost and commercially available components found in the telecommunications industry. Its all-fiber design makes it amenable to the unique needs of high-energy laser facilities, including routing to intricate pick-off locations, immunity to electromagnetic interference and low-loss transport (up to several kilometers). The technology offers several important benefits for NIF. First, the facility is able to fire more shots in less time-an efficiency that saves the facility millions of dollars each year. Second, high-energy lasers are more flexible to wavelength changes requested by target physicists. Third, by identifying harmful pulses before they damage the laser's optics, the facility potentially saves hundreds of thousands of dollars in maintenance costs each year.

  17. The Muon System of the Daya Bay Reactor Antineutrino Experiment

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    An, F. P.; Hackenburg, R. W.; Brown, R. E.; Chasman, C.; Dale, E.; Diwan, M. V.; Gill, R.; Hans, S.; Isvan, Z.; Jaffe, D. E.; et al

    2014-10-05

    The Daya Bay experiment consists of functionally identical antineutrino detectors immersed in pools of ultrapure water in three well-separated underground experimental halls near two nuclear reactor complexes. These pools serve both as shields against natural, low-energy radiation, and as water Cherenkov detectors that efficiently detect cosmic muons using arrays of photomultiplier tubes. Each pool is covered by a plane of resistive plate chambers as an additional means of detecting muons. Design, construction, operation, and performance of these muon detectors are described. (auth)

  18. Reactor vessel support system

    DOE Patents [OSTI]

    Golden, Martin P. (Trafford, PA); Holley, John C. (McKeesport, PA)

    1982-01-01

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  19. Spinning fluids reactor

    DOE Patents [OSTI]

    Miller, Jan D; Hupka, Jan; Aranowski, Robert

    2012-11-20

    A spinning fluids reactor, includes a reactor body (24) having a circular cross-section and a fluid contactor screen (26) within the reactor body (24). The fluid contactor screen (26) having a plurality of apertures and a circular cross-section concentric with the reactor body (24) for a length thus forming an inner volume (28) bound by the fluid contactor screen (26) and an outer volume (30) bound by the reactor body (24) and the fluid contactor screen (26). A primary inlet (20) can be operatively connected to the reactor body (24) and can be configured to produce flow-through first spinning flow of a first fluid within the inner volume (28). A secondary inlet (22) can similarly be operatively connected to the reactor body (24) and can be configured to produce a second flow of a second fluid within the outer volume (30) which is optionally spinning.

  20. Reactor water cleanup system

    DOE Patents [OSTI]

    Gluntz, D.M.; Taft, W.E.

    1994-12-20

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling. 1 figure.

  1. Reactor water cleanup system

    DOE Patents [OSTI]

    Gluntz, Douglas M.; Taft, William E.

    1994-01-01

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling.

  2. Brachytherapy structural shielding calculations using Monte Carlo generated, monoenergetic data

    SciTech Connect (OSTI)

    Zourari, K.; Peppa, V.; Papagiannis, P.; Ballester, Facundo; Siebert, Frank-Andr

    2014-04-15

    Purpose: To provide a method for calculating the transmission of any broad photon beam with a known energy spectrum in the range of 201090 keV, through concrete and lead, based on the superposition of corresponding monoenergetic data obtained from Monte Carlo simulation. Methods: MCNP5 was used to calculate broad photon beam transmission data through varying thickness of lead and concrete, for monoenergetic point sources of energy in the range pertinent to brachytherapy (201090 keV, in 10 keV intervals). The three parameter empirical model introduced byArcher et al. [Diagnostic x-ray shielding design based on an empirical model of photon attenuation, Health Phys. 44, 507517 (1983)] was used to describe the transmission curve for each of the 216 energy-material combinations. These three parameters, and hence the transmission curve, for any polyenergetic spectrum can then be obtained by superposition along the lines of Kharrati et al. [Monte Carlo simulation of x-ray buildup factors of lead and its applications in shielding of diagnostic x-ray facilities, Med. Phys. 34, 13981404 (2007)]. A simple program, incorporating a graphical user interface, was developed to facilitate the superposition of monoenergetic data, the graphical and tabular display of broad photon beam transmission curves, and the calculation of material thickness required for a given transmission from these curves. Results: Polyenergetic broad photon beam transmission curves of this work, calculated from the superposition of monoenergetic data, are compared to corresponding results in the literature. A good agreement is observed with results in the literature obtained from Monte Carlo simulations for the photon spectra emitted from bare point sources of various radionuclides. Differences are observed with corresponding results in the literature for x-ray spectra at various tube potentials, mainly due to the different broad beam conditions or x-ray spectra assumed. Conclusions: The data of this work allow for the accurate calculation of structural shielding thickness, taking into account the spectral variation with shield thickness, and broad beam conditions, in a realistic geometry. The simplicity of calculations also obviates the need for the use of crude transmission data estimates such as the half and tenth value layer indices. Although this study was primarily designed for brachytherapy, results might also be useful for radiology and nuclear medicine facility design, provided broad beam conditions apply.

  3. Early Test Facilities and Analytic Methods for Radiation Shielding

    SciTech Connect (OSTI)

    Ingersoll, D.T.

    1992-01-01

    This report represents a compilation of eight papers presented at the 1992 American Nuclear Society/European Nuclear Society International Meeting held in Chicago, Illinois on November 15 20,1992. The meeting is of special significance since it commemorates the 50th anniversary of the first controlled nuclear chain reaction, which occurred, not coincidentally, in Chicago. The papers contained in this report were presented in a special session organized by the Radiation Protection and Shielding Division in keeping with the historical theme of the meeting.

  4. Passive heat-transfer means for nuclear reactors. [LMFBR

    DOE Patents [OSTI]

    Burelbach, J.P.

    1982-06-10

    An improved passive cooling arrangement is disclosed for maintaining adjacent or related components of a nuclear reactor within specified temperature differences. Specifically, heat pipes are operatively interposed between the components, with the vaporizing section of the heat pipe proximate the hot component operable to cool it and the primary condensing section of the heat pipe proximate the other and cooler component operable to heat it. Each heat pipe further has a secondary condensing section that is located outwardly beyond the reactor confinement and in a secondary heat sink, such as air ambient the containment, that is cooler than the other reactor component. By having many such heat pipes, an emergency passive cooling system is defined that is operative without electrical power.

  5. RSE Pulp & Chemical, LLC (Subsidiary of Red Shield Environmental, LLC) |

    Energy Savers [EERE]

    Department of Energy RSE Pulp & Chemical, LLC (Subsidiary of Red Shield Environmental, LLC) RSE Pulp & Chemical, LLC (Subsidiary of Red Shield Environmental, LLC) A fact sheet detailling a proposal of a biorefinery facility in an existing pulp mill to demonstrate the production of cellulosic ethanol from lignocellulosic (wood) extract. PDF icon RSE Pulp & Chemical, LLC (Subsidiary of Red Shield Environmental, LLC) More Documents & Publications Pacific Ethanol, Inc EA-1888:

  6. Source Terms for HFIR Beam Tube Shielding Analyses, and a Complete Shielding Analysis of the HB-3 Tube

    SciTech Connect (OSTI)

    Bucholz, J.A.

    2000-07-01

    The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory is in the midst of a massive upgrade program to enhance experimental facilities. The reactor presently has four horizontal experimental beam tubes, all of which will be replaced or redesigned. The HB-2 beam tube will be enlarged to support more guide tubes, while the HB-4 beam tube will soon include a cold neutron source.

  7. Conceptual design of the Clinch River Breeder Reactor spent-fuel shipping cask

    SciTech Connect (OSTI)

    Pope, R B; Diggs, J M [eds.

    1982-04-01

    Details of a baseline conceptual design of a spent fuel shipping cask for the Clinch River Breeder Reactor (CRBR) are presented including an assessment of shielding, structural, thermal, fabrication and cask/plant interfacing problems. A basis for continued cask development and for new technological development is established. Alternates to the baseline design are briefly presented. Estimates of development schedules, cask utilization and cost schedules, and of personnel dose commitments during CRBR in-plant handling of the cask are also presented.

  8. Numerical Simulation of Earth Pressure on Head Chamber of Shield Machine with FEM

    SciTech Connect (OSTI)

    Li Shouju; Kang Chengang [State Key Laboratory of structural analysis for industrial equipment, Dalian University of Technology, Dalian 116023 (China); Sun, Wei [School of Mechanical Engineering, Dalian University of Technology, Dalian 116023 (China); Shangguan Zichang [School of Civil and Hydraulic Engineering, Dalian University of Technology, Dalian 116023 (China); Institute of Civil Engineering, Dalian Fishery University, Dalian 116023 (China)

    2010-05-21

    Model parameters of conditioned soils in head chamber of shield machine are determined based on tree-axial compression tests in laboratory. The loads acting on tunneling face are estimated according to static earth pressure principle. Based on Duncan-Chang nonlinear elastic constitutive model, the earth pressures on head chamber of shield machine are simulated in different aperture ratio cases for rotating cutterhead of shield machine. Relationship between pressure transportation factor and aperture ratio of shield machine is proposed by using aggression analysis.

  9. Structural Design and Thermal Analysis for Thermal Shields of the MICE Coupling Magnets

    SciTech Connect (OSTI)

    Green, Michael A.; Pan, Heng; Liu, X. K.; Wang, Li; Wu, Hong; Chen, A. B.; Guo, X.L.

    2009-07-01

    A superconducting coupling magnet made from copper matrix NbTi conductors operating at 4 K will be used in the Muon Ionization Cooling Experiment (MICE) to produce up to 2.6 T on the magnet centerline to keep the muon beam within the thin RF cavity indows. The coupling magnet is to be cooled by two cryocoolers with a total cooling capacity of 3 W at 4.2 K. In order to keep a certain operating temperature margin, the most important is to reduce the heat leakage imposed on cold surfaces of coil cold mass assembly. An ntermediate temperature shield system placed between the coupling coil and warm vacuum chamber is adopted. The shield system consists of upper neck shield, main shields, flexible connections and eight supports, which is to be cooled by the first stage cold heads of two ryocoolers with cooling capacity of 55 W at 60 K each. The maximum temperature difference on the shields should be less than 20 K, so the thermal analyses for the shields with different thicknesses, materials, flexible connections for shields' cooling and structure design for heir supports were carried out. 1100 Al is finally adopted and the maximum temperature difference is around 15 K with 4 mm shield thickness. The paper is to present detailed analyses on the shield system design.

  10. Method to Reduce Neutron Production in Small Clean Fusion Reactors Inventor

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    --- Samuel A. Cohen | Princeton Plasma Physics Lab Method to Reduce Neutron Production in Small Clean Fusion Reactors Inventor --- Samuel A. Cohen This invention describes a method to reduce neutron production of D-3He-fueled, steady state, small FRC fusion reactors using periodic, co-streaming, energetic ion beams generated by RF. Use of this method will lessen damage to and activation of reactor components and, in doing so, can advance the development of fusion reactors for electrical

  11. Two component-three dimensional catalysis

    DOE Patents [OSTI]

    Schwartz, Michael (Boulder, CO); White, James H. (Boulder, CO); Sammells, Anthony F. (Boulder, CO)

    2002-01-01

    This invention relates to catalytic reactor membranes having a gas-impermeable membrane for transport of oxygen anions. The membrane has an oxidation surface and a reduction surface. The membrane is coated on its oxidation surface with an adherent catalyst layer and is optionally coated on its reduction surface with a catalyst that promotes reduction of an oxygen-containing species (e.g., O.sub.2, NO.sub.2, SO.sub.2, etc.) to generate oxygen anions on the membrane. The reactor has an oxidation zone and a reduction zone separated by the membrane. A component of an oxygen containing gas in the reduction zone is reduced at the membrane and a reduced species in a reactant gas in the oxidation zone of the reactor is oxidized. The reactor optionally contains a three-dimensional catalyst in the oxidation zone. The adherent catalyst layer and the three-dimensional catalyst are selected to promote a desired oxidation reaction, particularly a partial oxidation of a hydrocarbon.

  12. Pulsed deuterium lithium nuclear reactor

    SciTech Connect (OSTI)

    Fischer, A.G.

    1980-01-08

    A nuclear reactor that burns hydrogen bomb material 6-lithium deuterotritide to helium in successive microexplosions which are ignited electrically and enclosed by this same molten material, and that permits the conversion of the reaction heat into useful electrical power. A specially-constructed high-current pulse machine is discharged via a thermally-preformed highly conducting path through a mass of the molten salt 6lid1-xtx (0shield. The shock wave is attenuated in it by a curtain of rising deuterium bubbles. The heat shock is buffered by partial melting of the external solid crust. The reaction heat is carried by the liquid metal of the external cooling jacket to the heat exchanger of the associated turbo-generator. Every few seconds, a new pulse can take place.

  13. CORAL: a stepping stone for establishing the Indian fast reactor fuel reprocessing technology

    SciTech Connect (OSTI)

    Venkataraman, M.; Natarajan, R.; Raj, Baldev

    2007-07-01

    The reprocessing of spent fuel from Fast Breeder Test Reactor (FBTR) has been successfully demonstrated in the pilot plant, CORAL (COmpact Reprocessing facility for Advanced fuels in Lead shielded cell). Since commissioning in 2003, spent mixed carbide fuel from FBTR of different burnups and varying cooling period, have been reprocessed in this facility. Reprocessing of the spent fuel with a maximum burnup of 100 GWd/t has been successfully carried out so far. The feed backs from these campaigns with progressively increasing specific activities, have been useful in establishing a viable process flowsheet for reprocessing the Prototype Fast Breeder Reactor (PFBR) spent fuel. Also, the design of various equipments and processes for the future plants, which are either under design for construction, namely, the Demonstration Fast Reactor Fuel Reprocessing Plant (DFRP) and the Fast reactor fuel Reprocessing Plant (FRP) could be finalized. (authors)

  14. BDDR, a new CEA technological and operating reactor database

    SciTech Connect (OSTI)

    Soldevilla, M.; Salmons, S.; Espinosa, B.

    2013-07-01

    The new application BDDR (Reactor database) has been developed at CEA in order to manage nuclear reactors technological and operating data. This application is a knowledge management tool which meets several internal needs: -) to facilitate scenario studies for any set of reactors, e.g. non-proliferation assessments; -) to make core physics studies easier, whatever the reactor design (PWR-Pressurized Water Reactor-, BWR-Boiling Water Reactor-, MAGNOX- Magnesium Oxide reactor-, CANDU - CANada Deuterium Uranium-, FBR - Fast Breeder Reactor -, etc.); -) to preserve the technological data of all reactors (past and present, power generating or experimental, naval propulsion,...) in a unique repository. Within the application database are enclosed location data and operating history data as well as a tree-like structure containing numerous technological data. These data address all kinds of reactors features and components. A few neutronics data are also included (neutrons fluxes). The BDDR application is based on open-source technologies and thin client/server architecture. The software architecture has been made flexible enough to allow for any change. (authors)

  15. Improved vortex reactor system

    DOE Patents [OSTI]

    Diebold, J.P.; Scahill, J.W.

    1995-05-09

    An improved vortex reactor system is described for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor. 12 figs.

  16. On Perturbation Components Correspondence between Diffusion and Transport

    SciTech Connect (OSTI)

    G. Palmiotti

    2012-11-01

    We have established a correspondence between perturbation components in diffusion and transport theory. In particular we have established the correspondence between the leakage perturbation component of the diffusion theory to that of the group self scattering in transport theory. This has been confirmed by practical applications on sodium void reactivity calculations of fast reactors. Why this is important for current investigations? Recently, there has been a renewed interest in designing fast reactors where the sodium void reactivity coefficient is minimized. In particular the ASTRID8,9 reactor concept has been optimized with this goal in mind. The correspondence on the leakage term that has been established here has a twofold implication for the design of this kind of reactors. First, this type of reactor has a radial reflector; therefore, as shown before, the sodium void reactivity coefficient calculation requires the use of transport theory. The minimization of the sodium reactivity coefficient is normally done by increasing the leakage component that has a negative sign. The correspondence established in this paper allows to directly look at this component in transport theory. The second implication is related to the uncertainty evaluation on sodium void reactivity. As it has shown before, the total sodium void reactivity effect is the result of a large compensation (opposite sign) between the scattering (called often spectral) component and the leakage one. Consequently, one has to evaluate separately the uncertainty on each separate component and then combine them statistically. If one wants to compute the cross section sensitivity coefficients of the two different components, the formulation established in this paper allows to achieve this goal by playing on the contribution to the sodium void reactivity coming from the group self scattering of the sodium cross section.

  17. Pressurized fluidized bed reactor

    DOE Patents [OSTI]

    Isaksson, Juhani (Karhula, FI)

    1996-01-01

    A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine.

  18. Pressurized fluidized bed reactor

    DOE Patents [OSTI]

    Isaksson, J.

    1996-03-19

    A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine. 1 fig.

  19. H Reactor - Hanford Site

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    About Us Projects & Facilities H Reactor About Us About Hanford Cleanup Hanford History Hanford Site Wide Programs Contact Us 100 Area 118-K-1 Burial Ground 200 Area 222-S Laboratory 242-A Evaporator 300 Area 324 Building 325 Building 400 Area/Fast Flux Test Facility 618-10 and 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and Interim Storage Area Canyon Facilities Cold Test Facility D and DR Reactors Effluent Treatment Facility Environmental

  20. Tokamak reactor first wall

    DOE Patents [OSTI]

    Creedon, R.L.; Levine, H.E.; Wong, C.; Battaglia, J.

    1984-11-20

    This invention relates to an improved first wall construction for a tokamak fusion reactor vessel, or other vessels subjected to similar pressure and thermal stresses.

  1. Experimental Study on Flow Optimization in Upper Plenum of Reactor Vessel for a Compact Sodium-Cooled Fast Reactor

    SciTech Connect (OSTI)

    Kimura, Nobuyuki; Hayashi, Kenji; Kamide, Hideki; Itoh, Masami; Sekine, Tadashi

    2005-11-15

    An innovative sodium-cooled fast reactor has been investigated in a feasibility study of fast breeder reactor cycle systems in Japan. A compact reactor vessel and a column-type upper inner structure with a radial slit for an arm of a fuel-handling machine (FHM) are adopted. Dipped plates are set in the reactor vessel below the free surface to prevent gas entrainment. We performed a one-tenth-scaled model water experiment for the upper plenum of the reactor vessel. Gas entrainment was not observed in the experiment under the same velocity condition as the reactor. Three vortex cavitations were observed near the hot-leg inlet. A vertical rib on the reactor vessel wall was set to restrict the rotating flow near the hot leg. The vortex cavitation between the reactor vessel wall and the hot leg was suppressed by the rib under the same cavitation factor condition as in the reactor. The cylindrical plug was installed through the hole in the dipped plates for the FHM to reduce the flow toward the free surface. It was effective when the plug was submerged into the middle height in the upper plenum. This combination of two components had a possibility to optimize the flow in the compact reactor vessel.

  2. Method for fabricating fan-fold shielded electrical leads

    DOE Patents [OSTI]

    Rohatgi, R.R.; Cowan, T.E.

    1994-12-27

    Fan-folded electrical leads made from copper cladded Kapton, for example, with the copper cladding on one side serving as a ground plane and the copper cladding on the other side being etched to form the leads. The Kapton is fan folded with the leads located at the bottom of the fan-folds. Electrical connections are made by partially opening the folds of the fan and soldering, for example, the connections directly to the ground plane and/or the lead. The fan folded arrangement produces a number of advantages, such as electrically shielding the leads from the environment, is totally non-magnetic, and has a very low thermal conductivity, while being easy to fabricate. 3 figures.

  3. Underground barrier construction apparatus with soil-retaining shield

    DOE Patents [OSTI]

    Gardner, B.M.; Smith, A.M.; Hanson, R.W.; Hodges, R.T.

    1998-08-04

    An apparatus is described for building a horizontal underground barrier by cutting through soil and depositing a slurry, preferably one which cures into a hardened material. The apparatus includes a digging means for cutting and removing soil to create a void under the surface of the ground, a shield means for maintaining the void, and injection means for inserting barrier-forming material into the void. In one embodiment, the digging means is a continuous cutting chain. Mounted on the continuous cutting chain are cutter teeth for cutting through soil and discharge paddles for removing the loosened soil. This invention includes a barrier placement machine, a method for building an underground horizontal containment barrier using the barrier placement machine, and the underground containment system. Preferably the underground containment barrier goes underneath and around the site to be contained in a bathtub-type containment. 17 figs.

  4. Underground barrier construction apparatus with soil-retaining shield

    DOE Patents [OSTI]

    Gardner, Bradley M. (Idaho Falls, ID); Smith, Ann Marie (Pocatello, ID); Hanson, Richard W. (Spokane, WA); Hodges, Richard T. (Deer Park, WA)

    1998-01-01

    An apparatus for building a horizontal underground barrier by cutting through soil and depositing a slurry, preferably one which cures into a hardened material. The apparatus includes a digging means for cutting and removing soil to create a void under the surface of the ground, a shield means for maintaining the void, and injection means for inserting barrier-forming material into the void. In one embodiment, the digging means is a continuous cutting chain. Mounted on the continuous cutting chain are cutter teeth for cutting through soil and discharge paddles for removing the loosened soil. This invention includes a barrier placement machine, a method for building an underground horizontal containment barrier using the barrier placement machine, and the underground containment system. Preferably the underground containment barrier goes underneath and around the site to be contained in a bathtub-type containment.

  5. Method for fabricating fan-fold shielded electrical leads

    DOE Patents [OSTI]

    Rohatgi, Rajeev R. (Mountain View, CA); Cowan, Thomas E. (Livermore, CA)

    1994-01-01

    Fan-folded electrical leads made from copper cladded Kapton, for example, with the copper cladding on one side serving as a ground plane and the copper cladding on the other side being etched to form the leads. The Kapton is fan folded with the leads located at the bottom of the fan-folds. Electrical connections are made by partially opening the folds of the fan and soldering, for example, the connections directly to the ground plane and/or the lead. The fan folded arrangement produces a number of advantages, such as electrically shielding the leads from the environment, is totally non-magnetic, and has a very low thermal conductivity, while being easy to fabricate.

  6. Probabilistic Safety Assessment of Tehran Research Reactor

    SciTech Connect (OSTI)

    Hosseini, Seyed Mohammad Hadi; Nematollahi, Mohammad Reza; Sepanloo, Kamran

    2004-07-01

    Probabilistic Safety Assessment (PSA) application is found to be a practical tool for research reactor safety due to intense involvement of human interactions in an experimental facility. In this paper the application of the Probabilistic Safety Assessment to the Tehran Research Reactor (TRR) is presented. The level 1 PSA application involved: Familiarization with the plant, selection of accident initiators, mitigating functions and system definitions, event tree constructions and quantification, fault tree constructions and quantification, human reliability, component failure data base development and dependent failure analysis. Each of the steps of the analysis given above is discussed with highlights from the selected results. Quantification of the constructed models is done using SAPHIRE software. This Study shows that the obtained core damage frequency for Tehran Research Reactor (8.368 E-6 per year) well meets the IAEA criterion for existing nuclear power plants (1E-4). But safety improvement suggestions are offered to decrease the most probable accidents. (authors)

  7. Estimation of the Performance of Multiple Active Neutron Interrogation Signatures for Detecting Shielded HEU

    SciTech Connect (OSTI)

    David L. Chichester; Scott J. Thompson; Scott M. Watson; James T. Johnson; Edward H. Seabury

    2012-10-01

    A comprehensive modeling study has been carried out to evaluate the utility of multiple active neutron interrogation signatures for detecting shielded highly enriched uranium (HEU). The modeling effort focused on varying HEU masses from 1 kg to 20 kg; varying types of shields including wood, steel, cement, polyethylene, and borated polyethylene; varying depths of the HEU in the shields, and varying engineered shields immediately surrounding the HEU including steel, tungsten, and cadmium. Neutron and gamma-ray signatures were the focus of the study and false negative detection probabilities versus measurement time were used as a performance metric. To facilitate comparisons among different approaches an automated method was developed to generate receiver operating characteristic (ROC) curves for different sets of model variables for multiple background count rate conditions. This paper summarizes results or the analysis, including laboratory benchmark comparisons between simulations and experiments. The important impact engineered shields can play towards degrading detectability and methods for mitigating this will be discussed.

  8. REFLECTOR FOR NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Fraas, A.P.

    1963-08-01

    A reflector for nuclear reactors that comprises an assembly of closely packed graphite rods disposed with their major axes substantially perpendicular to the interface between the reactor core and the reflector is described. Each graphite rod is round in transverse cross section at (at least) its interface end and is provided, at that end, with a coaxial, inwardly tapering hole. (AEC)

  9. Montena Components | Open Energy Information

    Open Energy Info (EERE)

    Components Jump to: navigation, search Name: Montena Components Place: Rossens, Switzerland Zip: CH 1728 Product: Montena Components is a manufacturer of ultracapacitors and...

  10. Integrating Program Component Executables

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Integrating Program Component Executables on Distributed Memory Architectures via MPH Chris Ding and Yun He Computational Research Division, Lawrence Berkeley National Laboratory University of California, Berkeley, CA 94720, USA chqding@lbl.gov, yhe@lbl.gov Abstract A growing trend in developing large and complex ap- plications on today's Teraflop computers is to integrate stand-alone and/or semi-independent program components into a comprehensive simulation package. One example is the climate

  11. Evaluation of Integrated High Temperature Component Testing Needs

    SciTech Connect (OSTI)

    Rafael Soto; David Duncan; Vincent Tonc

    2009-05-01

    This paper describes the requirements for a large-scale component test capability to support the development of advanced nuclear reactor technology and their adaptation to commercial applications that advance U.S. energy economy, reliability, and security and reduce carbon emissions.

  12. Final Technical Report [Cosmogenic background and shielding R&D for a Ge Neutrinoless Double Beta Decay Experiment

    SciTech Connect (OSTI)

    Guiseppe, Vince

    2013-10-01

    The USD Majorana group focused all of its effort in support of the MAJORANA DEMONSTRATOR (MJD) experiment. Final designs of the shielding subsystems are complete. Construction of the MJD shielding systems at SURF has begun and the proposed activities directly support the completion of the shield systems. The PI and the group contribute heavily to the onsite construction activities of the MJD experiment. The group led investigations into neutron and neutron-?induced backgrounds, shielding effectiveness and design, and radon backgrounds.

  13. Risk Management for Sodium Fast Reactors.

    SciTech Connect (OSTI)

    Denman, Matthew R; Groth, Katrina; Cardoni, Jeffrey N; Wheeler, Timothy A.

    2015-01-01

    Accident management is an important component to maintaining risk at acceptable levels for all complex systems, such as nuclear power plants. With the introduction of self - correcting, or inherently safe, reactor designs the focus has shifted from management by operators to allowing the syste m's design to manage the accident. While inherently and passively safe designs are laudable, extreme boundary conditions can interfere with the design attributes which facilitate inherent safety , thus resulting in unanticipated and undesirable end states. This report examines an inherently safe and small sodium fast reactor experiencing a beyond design basis seismic event with the intend of exploring two issues : (1) can human intervention either improve or worsen the potential end states and (2) can a Bayes ian Network be constructed to infer the state of the reactor to inform (1). ACKNOWLEDGEMENTS The author s would like to acknowledge the U.S. Department of E nergy's Office of Nuclear Energy for funding this research through Work Package SR - 14SN100303 under the Advanced Reactor Concepts program. The authors also acknowledge the PRA teams at A rgonne N ational L aborator y , O ak R idge N ational L aborator y , and I daho N ational L aborator y for their continue d contributions to the advanced reactor PRA mission area.

  14. Advanced Reactor Technologies | Department of Energy

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

    Nuclear Reactor Technologies » Advanced Reactor Technologies Advanced Reactor Technologies Advanced Reactor Technologies Advanced Reactor Technologies The Office of Advanced Reactor Technologies (ART) sponsors research, development and deployment (RD&D) activities through its Next Generation Nuclear Plant (NGNP), Advanced Reactor Concepts (ARC), and Advanced Small Modular Reactor (aSMR) programs to promote safety, technical, economical, and environmental advancements of innovative

  15. Slurry reactor design studies

    SciTech Connect (OSTI)

    Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. ); Akgerman, A. ); Smith, J.M. )

    1990-06-01

    The objective of these studies was to perform a realistic evaluation of the relative costs of tublar-fixed-bed and slurry reactors for methanol, mixed alcohols and Fischer-Tropsch syntheses under conditions where they would realistically be expected to operate. The slurry Fischer-Tropsch reactor was, therefore, operated at low H{sub 2}/CO ratio on gas directly from a Shell gasifier. The fixed-bed reactor was operated on 2.0 H{sub 2}/CO ratio gas after adjustment by shift and CO{sub 2} removal. Every attempt was made to give each reactor the benefit of its optimum design condition and correlations were developed to extend the models beyond the range of the experimental pilot plant data. For the methanol design, comparisons were made for a recycle plant with high methanol yield, this being the standard design condition. It is recognized that this is not necessarily the optimum application for the slurry reactor, which is being proposed for a once-through operation, coproducing methanol and power. Consideration is also given to the applicability of the slurry reactor to mixed alcohols, based on conditions provided by Lurgi for an Octamix{trademark} plant using their standard tubular-fixed reactor technology. 7 figs., 26 tabs.

  16. Reactor Safety Research Programs

    SciTech Connect (OSTI)

    Edler, S. K.

    1981-07-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipeto- pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-ofcoolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  17. Nuclear reactor control column

    DOE Patents [OSTI]

    Bachovchin, Dennis M.

    1982-01-01

    The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest cross-sectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor.

  18. 105-H Reactor Interim Safe Storage Project Final Report

    SciTech Connect (OSTI)

    E.G. Ison

    2008-11-08

    The following information documents the decontamination and decommissioning of the 105-H Reactor facility, and placement of the reactor core into interim safe storage. The D&D of the facility included characterization, engineering, removal of hazardous and radiologically contaminated materials, equipment removal, decontamination, demolition of the structure, and restoration of the site. The ISS work also included construction of the safe storage enclosure, which required the installation of a new roofing system, power and lighting, a remote monitoring system, and ventilation components.

  19. Light Water Reactor Sustainability Program: Materials Aging and Degradation

    Office of Environmental Management (EM)

    Technical Program Plan | Department of Energy Program: Materials Aging and Degradation Technical Program Plan Light Water Reactor Sustainability Program: Materials Aging and Degradation Technical Program Plan Components serving in a nuclear reactor plant must withstand a very harsh environment including extended time at temperature, neutron irradiation, stress, and/or corrosive media. The many modes of degradation are complex and vary depending on location and material. However,

  20. NUCLEAR REACTOR FUEL SYSTEMS

    DOE Patents [OSTI]

    Thamer, B.J.; Bidwell, R.M.; Hammond, R.P.

    1959-09-15

    Homogeneous reactor fuel solutions are reported which provide automatic recombination of radiolytic gases and exhibit large thermal expansion characteristics, thereby providing stability at high temperatures and enabling reactor operation without the necessity of apparatus to recombine gases formed by the radiolytic dissociation of water in the fuel and without the necessity of liquid fuel handling outside the reactor vessel except for recovery processes. The fuels consist of phosphoric acid and water solutions of enriched uranium, wherein the uranium is in either the hexavalent or tetravalent state.

  1. Nuclear reactor reflector

    DOE Patents [OSTI]

    Hopkins, R.J.; Land, J.T.; Misvel, M.C.

    1994-06-07

    A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled. 12 figs.

  2. Nuclear reactor reflector

    DOE Patents [OSTI]

    Hopkins, Ronald J. (Pensacola, FL); Land, John T. (Pensacola, FL); Misvel, Michael C. (Pensacola, FL)

    1994-01-01

    A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled.

  3. CONTROL FOR NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Lichtenberger, H.V.; Cameron, R.A.

    1959-03-31

    S>A control rod operating device in a nuclear reactor of the type in which the control rod is gradually withdrawn from the reactor to a position desired during stable operation is described. The apparatus is comprised essentially of a stop member movable in the direction of withdrawal of the control rod, a follower on the control rod engageable with the stop and means urging the follower against the stop in the direction of withdrawal. A means responsive to disengagement of the follower from the stop is provided for actuating the control rod to return to the reactor shut-down position.

  4. Microfluidic electrochemical reactors

    DOE Patents [OSTI]

    Nuzzo, Ralph G.; Mitrovski, Svetlana M.

    2011-03-22

    A microfluidic electrochemical reactor includes an electrode and one or more microfluidic channels on the electrode, where the microfluidic channels are covered with a membrane containing a gas permeable polymer. The distance between the electrode and the membrane is less than 500 micrometers. The microfluidic electrochemical reactor can provide for increased reaction rates in electrochemical reactions using a gaseous reactant, as compared to conventional electrochemical cells. Microfluidic electrochemical reactors can be incorporated into devices for applications such as fuel cells, electrochemical analysis, microfluidic actuation, pH gradient formation.

  5. Fast Breeder Reactor studies

    SciTech Connect (OSTI)

    Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

    1980-07-01

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

  6. Crack growth rates and metallographic examinations of Alloy 600 and Alloy 82/182 from field components and laboratory materials tested in PWR environments.

    SciTech Connect (OSTI)

    Alexandreanu, B.; Chopra, O. K.; Shack, W. J.

    2008-05-05

    In light water reactors, components made of nickel-base alloys are susceptible to environmentally assisted cracking. This report summarizes the crack growth rate results and related metallography for field and laboratory-procured Alloy 600 and its weld alloys tested in pressurized water reactor (PWR) environments. The report also presents crack growth rate (CGR) results for a shielded-metal-arc weld of Alloy 182 in a simulated PWR environment as a function of temperature between 290 C and 350 C. These data were used to determine the activation energy for crack growth in Alloy 182 welds. The tests were performed by measuring the changes in the stress corrosion CGR as the temperatures were varied during the test. The difference in electrochemical potential between the specimen and the Ni/NiO line was maintained constant at each temperature by adjusting the hydrogen overpressure on the water supply tank. The CGR data as a function of temperature yielded activation energies of 252 kJ/mol for a double-J weld and 189 kJ/mol for a deep-groove weld. These values are in good agreement with the data reported in the literature. The data reported here and those in the literature suggest that the average activation energy for Alloy 182 welds is on the order of 220-230 kJ/mol, higher than the 130 kJ/mol commonly used for Alloy 600. The consequences of using a larger value of activation energy for SCC CGR data analysis are discussed.

  7. Nuclear reactor with low-level core coolant intake

    DOE Patents [OSTI]

    Challberg, Roy C. (Livermore, CA); Townsend, Harold E. (Campbell, CA)

    1993-01-01

    A natural-circulation boiling-water reactor has skirts extending downward from control rod guide tubes to about 10 centimeters from the reactor vessel bottom. The skirts define annular channels about control rod drive housings that extend through the reactor vessel bottom. Recirculating water is forced in through the low-level entrances to these channels, sweeping bottom water into the channels in the process. The sweeping action prevents cooler water from accumulating at the bottom. This in turn minimizes thermal shock to bottom-dwelling components as would occur when accumulated cool water is swept away and suddenly replaced by warmer water.

  8. Reactor User Interface Technology Development Roadmaps for a High Temperature Gas-Cooled Reactor Outlet Temperature of 750 degrees C

    SciTech Connect (OSTI)

    Ian Mckirdy

    2010-12-01

    This report evaluates the technology readiness of the interface components that are required to transfer high-temperature heat from a High Temperature Gas-Cooled Reactor (HTGR) to selected industrial applications. This report assumes that the HTGR operates at a reactor outlet temperature of 750°C and provides electricity and/or process heat at 700°C to conventional process applications, including the production of hydrogen.

  9. Components in the Pipeline

    SciTech Connect (OSTI)

    Gorton, Ian; Wynne, Adam S.; Liu, Yan; Yin, Jian

    2011-02-24

    Scientists commonly describe their data processing systems metaphorically as software pipelines. These pipelines input one or more data sources and apply a sequence of processing steps to transform the data and create useful results. While conceptually simple, pipelines often adopt complex topologies and must meet stringent quality of service requirements that place stress on the software infrastructure used to construct the pipeline. In this paper we describe the MeDICi Integration Framework, which is a component-based framework for constructing complex software pipelines. The framework supports composing pipelines from distributed heterogeneous software components and provides mechanisms for controlling qualities of service to meet demanding performance, reliability and communication requirements.

  10. Ceramic component for electrodes

    DOE Patents [OSTI]

    Marchant, David D. (Richland, WA); Bates, J. Lambert (Richland, WA)

    1980-01-01

    A ceramic component suitable for preparing MHD generator electrodes having the compositional formula: Y.sub.x (Mg.sub.y Cr.sub.z).sub.w Al.sub.(1-w) O.sub.3 where x=0.9 to 1.05, y=0.02 to 0.2, z=0.8 to 1.05 and w=1.0 to 0.5. The component is resistant to the formation of hydration products in an MHD environment, has good electrical conductivity and exhibits a lower electrochemical corrosion rate than do comparable compositions of lanthanum chromite.

  11. N Reactor - Hanford Site

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Projects & Facilities N Reactor About Us About Hanford Cleanup Hanford History Hanford Site Wide Programs Contact Us 100 Area 118-K-1 Burial Ground 200 Area 222-S Laboratory 242-A...

  12. C Reactor - Hanford Site

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    C Reactor About Us About Hanford Cleanup Hanford History Hanford Site Wide Programs Contact Us 100 Area 118-K-1 Burial Ground 200 Area 222-S Laboratory 242-A Evaporator 300 Area...

  13. F Reactor - Hanford Site

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    About Us Projects & Facilities F Reactor About Us About Hanford Cleanup Hanford History Hanford Site Wide Programs Contact Us 100 Area 118-K-1 Burial Ground 200 Area 222-S...

  14. CHARACTERIZATION OF AN ACTIVELY COOLED METAL FOIL THERMAL RADIATION SHIELD

    SciTech Connect (OSTI)

    Feller, J. R.; Salerno, L. J.; Kashani, A.; Helvensteijn, B. P. M.

    2010-04-09

    Zero boil-off (ZBO) or reduced boil-off (RBO) systems that involve active cooling of large cryogenic propellant tanks will most likely be required for future space exploration missions. For liquid oxygen or methane, such systems could be implemented using existing high technology readiness level (TRL) cryocoolers. However, for liquid hydrogen temperatures (approx20 K) no such coolers exist. In order to partially circumvent this technology gap, the concept of broad area cooling (BAC) has been developed, whereby a low mass thermal radiation shield could be maintained at temperatures around 100 K by steady circulation of cold pressurized gas through a network of narrow tubes. By this method it is possible to dramatically reduce the radiative heat leak to the 20 K tank. A series of experiments, designed to investigate the heat transfer capabilities of BAC systems, have been conducted at NASA Ames Research Center (ARC). Results of the final experiment in this series, investigating heat transfer from a metal foil film to a distributed cooling line, are presented here.

  15. MAGNETIC SHIELDING OF EXOMOONS BEYOND THE CIRCUMPLANETARY HABITABLE EDGE

    SciTech Connect (OSTI)

    Heller, Ren; Zuluaga, Jorge I. E-mail: jzuluaga@fisica.udea.edu.co

    2013-10-20

    With most planets and planetary candidates detected in the stellar habitable zone (HZ) being super-Earths and gas giants rather than Earth-like planets, we naturally wonder if their moons could be habitable. The first detection of such an exomoon has now become feasible, and due to observational biases it will be at least twice as massive as Mars. However, formation models predict that moons can hardly be as massive as Earth. Hence, a giant planet's magnetosphere could be the only possibility for such a moon to be shielded from cosmic and stellar high-energy radiation. Yet, the planetary radiation belt could also have detrimental effects on exomoon habitability. Here we synthesize models for the evolution of the magnetic environment of giant planets with thresholds from the runaway greenhouse (RG) effect to assess the habitability of exomoons. For modest eccentricities, we find that satellites around Neptune-sized planets in the center of the HZ around K dwarf stars will either be in an RG state and not be habitable, or they will be in wide orbits where they will not be affected by the planetary magnetosphere. Saturn-like planets have stronger fields, and Jupiter-like planets could coat close-in habitable moons soon after formation. Moons at distances between about 5 and 20 planetary radii from a giant planet can be habitable from an illumination and tidal heating point of view, but still the planetary magnetosphere would critically influence their habitability.

  16. General Corrosion and Localized Corrosion of the Drip Shield

    SciTech Connect (OSTI)

    F. Hua; K. Mon

    2003-06-24

    The recommended waste package (WP) design is described in BSC (2001a). The design includes a double-wall WP underneath a protective drip shield (DS) (BSC 2003a). The purpose of the process-level models developed in this report is to model dry oxidation (DOX), general corrosion (GC) and localized corrosion (LC) of the DS plate material, which is made of Ti Grade 7. The DS design also includes structural supports fabricated from Ti Grade 24. Degradation of Ti Grade 24 is not considered in this report. The DS provides protection for the waste package outer barrier (WPOB) both as a barrier to seepage water contact and a physical barrier to potential rockfall. This Model Report (MR) serves as a feed to the Integrated Waste Package Degradation Model (IWPD) analyses, and was developed in accordance with the Technical Work Plan (TWP) (BSC 2002a). The models contained in this report serve as a basis to determine whether or not the performance requirements for the DS can be met.

  17. B Reactor | Department of Energy

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

    » Signature Facilities » B Reactor B Reactor B Reactor Completed in September 1944, the B Reactor was the world's first large-scale plutonium production reactor. As at Oak Ridge, the need for labor turned Hanford into an atomic boomtown, with the population reaching 50,000 by summer 1944. Similar to the X-10 Graphite Reactor at Oak Ridge in terms of loading and unloading fuel, the B Reactor was built on a much larger scale and used water rather than air as a coolant. Whereas the X-10 had an

  18. Compact power reactor

    DOE Patents [OSTI]

    Wetch, Joseph R.; Dieckamp, Herman M.; Wilson, Lewis A.

    1978-01-01

    There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector.

  19. Molten metal reactors

    DOE Patents [OSTI]

    Bingham, Dennis N; Klingler, Kerry M; Turner, Terry D; Wilding, Bruce M

    2013-11-05

    A molten metal reactor for converting a carbon material and steam into a gas comprising hydrogen, carbon monoxide, and carbon dioxide is disclosed. The reactor includes an interior crucible having a portion contained within an exterior crucible. The interior crucible includes an inlet and an outlet; the outlet leads to the exterior crucible and may comprise a diffuser. The exterior crucible may contain a molten alkaline metal compound. Contained between the exterior crucible and the interior crucible is at least one baffle.

  20. F Reactor Inspection

    ScienceCinema (OSTI)

    Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

    2014-11-24

    Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor before stepping inside the reactor to complete its periodic inspection. This is the first time the Department of Energy (DOE) has had the reactor open since 2008. The F Reactor is one of nine reactors along the Columbia River at the Department's Hanford Site in southeastern Washington State, where environmental cleanup has been ongoing since 1989. As part of the Tri-Party Agreement, the Department completes surveillance and maintenance activities of cocooned reactors periodically to evaluate the structural integrity of the safe storage enclosure and to ensure confinement of any remaining hazardous materials. "This entry marks a transition of sorts because the Hanford Long-Term Stewardship Program, for the first time, was responsible for conducting the entry and surveillance and maintenance activities," said Keith Grindstaff, Energy Department Long-Term Stewardship Program Manager. "As the River Corridor cleanup work is completed and transitioned to long-term stewardship, our program will manage any on-going requirements."

  1. Research Gaps and Technology Needs in Development of PHM for Passive AdvSMR Components

    SciTech Connect (OSTI)

    Meyer, Ryan M.; Ramuhalli, Pradeep; Coble, Jamie B.; Hirt, Evelyn H.; Mitchell, Mark R.; Wootan, David W.; Berglin, Eric J.; Bond, Leonard J.; Henager, Charles H.

    2014-01-01

    Advanced small modular reactors (AdvSMRs), which are based on modularization of advanced reactor concepts, may provide a longer-term alternative to traditional light-water reactors and near term small modular reactors (SMRs), which are based on integral pressurized water reactor (iPWR) concepts. SMRs are challenged economically due to losses in economy of scale, thus, there is increased motivation to reduce the controllable operations and maintenance (O&M) costs through automation technologies including prognostics health management (PHM) systems. In this regard, PHM systems have the potential to play a vital role in supporting the deployment of AdvSMRs and face several unique challenges with respect to implementation for passive AdvSMR components. This paper presents a summary of a research gaps and technical needs assessment performed for implementation of PHM for passive AdvSMR components. state-of-the-art in PHM.

  2. Research gaps and technology needs in development of PHM for passive AdvSMR components

    SciTech Connect (OSTI)

    Meyer, Ryan M.; Ramuhalli, Pradeep; Hirt, Evelyn H.; Mitchell, Mark R.; Wootan, David W.; Berglin, Eric J.; Henagar, Chuck H. Jr.; Coble, Jamie B.; Bond, Leonard J.

    2014-02-18

    Advanced small modular reactors (AdvSMRs), which are based on modularization of advanced reactor concepts, may provide a longer-term alternative to traditional light-water reactors and near-term small modular reactors (SMRs), which are based on integral pressurized water reactor (iPWR) concepts. SMRs are challenged economically because of losses in economy of scale; thus, there is increased motivation to reduce the controllable operations and maintenance costs through automation technologies including prognostics health management (PHM) systems. In this regard, PHM systems have the potential to play a vital role in supporting the deployment of AdvSMRs and face several unique challenges with respect to implementation for passive AdvSMR components. This paper presents a summary of a research gaps and technical needs assessment performed for implementation of PHM for passive AdvSMR components.

  3. Evaluate fundamental approaches to longwall dust control: Subprogram G, Reduction of shield generated dust

    SciTech Connect (OSTI)

    Mangolds, A.; Rajan, S.; Ruggieri, S.

    1990-05-01

    As longwall shield supports have become more popular and shearer-generated dust has been reduced, many mines have noticed that support movement can be a significant respirable dust contributor. The objective of this program was to isolate and document the mechanics of support dust generation and mobilization and to devise appropriate control techniques. This was accomplished through a literature review, laboratory testing and underground surveys. The literature review studied past research related to support dust generation and existing control measures. Laboratory testing investigated a variety of ventilation factors such as canopy cross flow, eddy effects and the effectiveness of canopy sprays. Airflow over and around advancing shields was shown to be a significant factor in dust dispersion, and spray systems indicated promise for control. The underground surveys provided additional insight into the patterns of shield dust generation and dispersion, and confirmed that shield dust can be a significant fraction of workers' dust exposures. 17 figs., 3 tabs.

  4. Array of virtual Frisch-grid detectors with common cathode and reduced length of shielding electrodes

    DOE Patents [OSTI]

    Bolotnikov, Aleksey; James, Ralph B.; De Geronimo, Gianluigi; Vernon, Emerson

    2015-09-15

    A radiation detector system that effectively solves the electron trapping problem by optimizing shielding of individual virtual Frisch-grid detectors in an array configuration with a common cathode.

  5. Array of virtual Frisch-grid detectors with common cathode and reduced length of shielding electrodes

    DOE Patents [OSTI]

    Bolotnikov, Aleksey E.; Yang, Ge; Camarda, Giuseppe; Cui, Yonggang; Hossain, Anwar; Kim, Ki Hyun; James, Ralph B.

    2015-09-01

    A radiation detector system that solves the electron trapping problem by optimizing shielding of the individual virtual Frisch-grid detectors in an array configuration with a common cathode.

  6. Graphit-ceramic RF Faraday-thermal shield and plasma limiter

    DOE Patents [OSTI]

    Hwang, David L. (Princeton Junction, NJ); Hosea, Joel C. (Princeton, NJ)

    1989-01-01

    The present invention is directed to a process of brazing a ceramic mater to graphite. In particular, the brazing procedure is directed to the production of a novel brazed ceramic graphite product useful as a Faraday shield.

  7. Recommendations for a Static Cosmic Ray Shield for Enriched Germanium Detectors

    SciTech Connect (OSTI)

    Aguayo Navarrete, Estanislao; Orrell, John L.; Ankney, Austin S.; Berguson, Timothy J.

    2011-09-21

    This document provides a detailed study of cost and materials that could be used to shield the detector material of the international Tonne-scale germanium neutrinoless double-beta decay experiment from hadronic particles from cosmic ray showers at the Earth's surface. This work was motivated by the need for a shield that minimizes activation of the enriched germanium during storage; in particular, when the detector material is being worked on at the detector manufacturer's facility. This work considers two options for shielding the detector material from cosmic ray particles. One option is to use a pre-existing structure already located near the detector manufacturer, such as Canberra Industries in Meriden, Connecticut. The other option is to build a shield onsite at a detector manufacturer's site. This paper presents a cost and efficiency analysis of such construction.

  8. Pollution prevention benefits of non-hazardous shielding glovebox gloves - 11000

    SciTech Connect (OSTI)

    Cournoyer, Michael E; Dodge, Robert L

    2011-01-11

    Radiation shielding is commonly used to protect the glovebox worker from unintentional direct and secondary radiation exposure, while working with plutonium-238 and plutonium-239. Shielding glovebox gloves are traditionally composed of lead-based materials, i.e., hazardous waste. This has prompted the development of new, non-hazardous shielding glovebox gloves. No studies, however, have investigated the pollution prevention benefits of these new glovebox gloves. We examined both leaded and non-hazardous shielding glovebox gloves. The nonhazardous substitutes are higher in cost, but this is offset by eliminating the costs associated with onsite waste handling of Resource Conservation and Recovery Act (RCRA) items. In the end, replacing lead with non-hazardous substitutes eliminates waste generation and future liability.

  9. Electromagnetic analysis of forces and torques on the ITER shield modules due to plasma disruption.

    SciTech Connect (OSTI)

    Kotulski, Joseph Daniel; Coats, Rebecca Sue; Pasik, Michael Francis; Ulrickson, Michael Andrew

    2009-06-01

    An electromagnetic analysis is performed on the ITER shield modules under different plasma disruption scenarios using the OPERA-3d software. The modeling procedure is explained, electromagnetic torques are presented, and results of the modeling are discussed.

  10. Renormalization plasma shielding effects on scattering entanglement fidelity in dense plasmas

    SciTech Connect (OSTI)

    Lee, Gyeong Won; Shim, Jaewon; Jung, Young-Dae

    2014-10-15

    The influence of renormalization plasma screening on the entanglement fidelity for the elastic electron-atom scattering is investigated in partially ionized dense hydrogen plasmas. The partial wave analysis and effective interaction potential are employed to obtain the scattering entanglement fidelity in dense hydrogen plasmas as functions of the collision energy, the Debye length, and the renormalization parameter. It is found that the renormalization plasma shielding enhances the scattering entanglement fidelity. Hence, we show that the transmission of the quantum information can be increased about 10% due to the renormalization shielding effect in dense hydrogen plasmas. It is also found that the renormalization shielding effect on the entanglement fidelity for the electron-atom collision increases with an increase of the collision energy. In addition, the renormalization shielding function increases with increasing collision energy and saturates to the unity with an increase of the Debye length.

  11. Effects of shielding gas compositions on arc plasma and metal transfer in gas metal arc welding

    SciTech Connect (OSTI)

    Rao, Z. H.; Liao, S. M.; Tsai, H. L.

    2010-02-15

    This article presents the effects of shielding gas compositions on the transient transport phenomena, including the distributions of temperature, flow velocity, current density, and electromagnetic force in the arc and the metal, and arc pressure in gas metal arc welding of mild steel at a constant current input. The shielding gas considered includes pure argon, 75% Ar, 50% Ar, and 25% Ar with the balance of helium. It is found that the shielding gas composition has significant influences on the arc characteristics; droplet formation, detachment, transfer, and impingement onto the workpiece; and weld pool dynamics and weld bead profile. As helium increases in the shielding gas, the droplet size increases but the droplet detachment frequency decreases. For helium-rich gases, the current converges at the workpiece with a 'ring' shape which produces non-Gaussian-like distributions of arc pressure and temperature along the workpiece surface. Detailed explanations to the physics of the very complex but interesting transport phenomena are given.

  12. On Enhancing Risk Monitors for Advanced Small Modular Reactors

    SciTech Connect (OSTI)

    Coble, Jamie B.; Coles, Garill A.; Meyer, Ryan M.; Ramuhalli, Pradeep

    2013-08-01

    Advanced small modular reactors (AdvSMRs) can contribute to safe, sustainable, and carbon-neutral energy production. However, the economics of AdvSMRs suffer from the loss of economy-of-scale for both construction and operation. The controllable day-to-day costs of AdvSMRs are expected to be dominated by operations and maintenance (O&M) costs. These expenses could potentially be managed through optimized scheduling of O&M activities for components, reactor modules, power blocks, and the full plant. Accurate, real-time risk assessment with integrated health monitoring of key active components can support scheduling of both online and offline inspection and maintenance activities.

  13. Impact of External Heat-shielding Techniques on Shell Surface Temperatures

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

    and Dynamic Shell Thermal Deformation of Diesel Engine Emission Control Systems | Department of Energy External Heat-shielding Techniques on Shell Surface Temperatures and Dynamic Shell Thermal Deformation of Diesel Engine Emission Control Systems Impact of External Heat-shielding Techniques on Shell Surface Temperatures and Dynamic Shell Thermal Deformation of Diesel Engine Emission Control Systems Presentation given at DEER 2006, August 20-24, 2006, Detroit, Michigan. Sponsored by the U.S.

  14. Progress on Establishing Guidelines for National Ignition Facility (NIF) Experiments to Extend Debris Shield Lifetime

    SciTech Connect (OSTI)

    Tobin, M; Eder, D; Braun, D; MacGowan, B

    2000-07-26

    The survivability and performance of the debris shields on the National Ignition Facility (NIF) are a key factor for the successful conduct and affordable operation of the facility. The improvements required over Nova debris shields are described. Estimates of debris shield lifetimes in the presence of target emissions with 4 - 5 J/cm{sup 2} laser fluences (and higher) indicate lifetimes that may contribute unacceptably to operations costs for NIF. We are developing detailed guidance for target and experiment designers for NIF to assist in minimizing the damage to, and therefore the cost of, maintaining NIF debris shields. The guidance limits the target mass that is allowed to become particulate on the debris shields (300 mg). It also limits the amount of material that can become shrapnel for any given shot (10 mg). Finally, it restricts the introduction of non-volatile residue (NVR) that is a threat to the sol-gel coatings on the debris shields to ensure that the chamber loading at any time is less than 1 pg/cm{sup 2}. We review the experimentation on the Nova chamber that included measuring quantities of particulate on debris shields by element and capturing shrapnel pieces in aerogel samples mounted in the chamber. We also describe computations of x-ray emissions from a likely NIF target and the associated ablation expected from this x-ray exposure on supporting target hardware. We describe progress in assessing the benefits of a pre-shield and the possible impact on the guidance for target experiments on NIF. Plans for possible experimentation on Omega and other facilities to improve our understanding of target emissions and their impacts are discussed. Our discussion of planned future work provides a forum to invite possible collaboration with the IFE community.

  15. Molecular based magnets comprising vanadium tetracyanoethylene complexes for shielding electromagnetic fields

    DOE Patents [OSTI]

    Epstein, A.J.; Morin, B.G.

    1998-10-13

    The invention presents a vanadium tetracyanoethylene solvent complex for electromagnetic field shielding, and a method for blocking low frequency and magnetic fields using these vanadium tetracyanoethylene compositions. The compositions of the invention can be produced at ambient temperature and are light weight, low density and flexible. The materials of the present invention are useful as magnetic shields to block low frequency fields and static fields, and for use in cores in transformers and motors. 21 figs.

  16. Boron cage compound materials and composites for shielding and absorbing neutrons

    DOE Patents [OSTI]

    Bowen, III, Daniel E; Eastwood, Eric A

    2014-03-04

    Boron cage compound-containing materials for shielding and absorbing neutrons. The materials include BCC-containing composites and compounds. BCC-containing compounds comprise a host polymer and a BCC attached thereto. BCC-containing composites comprise a mixture of a polymer matrix and a BCC filler. The BCC-containing materials can be used to form numerous articles of manufacture for shielding and absorbing neutrons.

  17. Shields Warren, 1971 | U.S. DOE Office of Science (SC)

    Office of Science (SC) Website

    Shields Warren, 1971 The Enrico Fermi Award Fermi Award Home Nomination & Selection Guidelines Award Laureates 2010's 2000's 1990's 1980's 1970's 1960's 1950's Ceremony The Life of Enrico Fermi Contact Information The Enrico Fermi Award U.S. Department of Energy SC-2/Germantown Building 1000 Independence Ave., SW Washington, DC 20585 P: (301) 903-2411 E: Email Us 1970's Shields

  18. FLEXIBLE NEUTRON SHIELDING FOR A GLOVEBOX WITHIN THE IDAHO NATIONAL LABORATORY RADIOISOTOPE POWER SYSTEM PROGRAM

    SciTech Connect (OSTI)

    Stephanie Walsh

    2007-07-01

    Neutron shielding was desired to reduce worker exposure during handling of plutonium-238 (Pu-238) in a glovebox at the Idaho National Laboratory. Due to the unusual shape of the glovebox, standard methods of neutron shielding were impractical and would have interfered with glovebox operations. A silicon-based, boron-impregnated material was chosen due to its flexibility. This paper discusses the material, the installation, and the results from neutron source testing.

  19. Study of a scattering shield in a high heat load monochromator (Journal

    Office of Scientific and Technical Information (OSTI)

    Article) | SciTech Connect Study of a scattering shield in a high heat load monochromator Citation Details In-Document Search Title: Study of a scattering shield in a high heat load monochromator Authors: Huang, Rong ; Meron, Mati [1] ; HWMRI) [2] + Show Author Affiliations (UC) [UC ( Publication Date: 2014-02-04 OSTI Identifier: 1091969 Resource Type: Journal Article Resource Relation: Journal Name: Nucl. Instrum. Methods A; Journal Volume: 716; Journal Issue: 07, 2013 Research Org:

  20. Molecular based magnets comprising vanadium tetracyanoethylene complexes for shielding electromagnetic fields

    DOE Patents [OSTI]

    Epstein, Arthur J. (Columbus, OH); Morin, Brian G. (Columbus, OH)

    1998-01-01

    The invention presents a vanadium tetracyanoethylene solvent complex for electromagnetic field shielding, and a method for blocking low frequency and magnetic fields using these vanadium tetracyanoethylene compositions. The compositions of the invention can be produced at ambient temperature and are light weight, low density and flexible. The materials of the present invention are useful as magnetic shields to block low frequency fields and static fields, and for use in cores in transformers and motors.

  1. Reactor Safety Planning for Prometheus Project, for Naval Reactors Information

    SciTech Connect (OSTI)

    P. Delmolino

    2005-05-06

    The purpose of this letter is to submit to Naval Reactors the initial plan for the Prometheus project Reactor Safety work. The Prometheus project is currently developing plans for cold physics experiments and reactor prototype tests. These tests and facilities may require safety analysis and siting support. In addition to the ground facilities, the flight reactor units will require unique analyses to evaluate the risk to the public from normal operations and credible accident conditions. This letter outlines major safety documents that will be submitted with estimated deliverable dates. Included in this planning is the reactor servicing documentation and shipping analysis that will be submitted to Naval Reactors.

  2. Injection molded component

    DOE Patents [OSTI]

    James, Allister W; Arrell, Douglas J

    2014-09-30

    An intermediate component includes a first wall member, a leachable material layer, and a precursor wall member. The first wall member has an outer surface and first connecting structure. The leachable material layer is provided on the first wall member outer surface. The precursor wall member is formed adjacent to the leachable material layer from a metal powder mixed with a binder material, and includes second connecting structure.

  3. Solid state lighting component

    DOE Patents [OSTI]

    Keller, Bernd; Ibbetson, James; Tarsa, Eric; Negley, Gerald; Yuan, Thomas

    2012-07-10

    An LED component comprising an array of LED chips mounted on a planar surface of a submount with the LED chips capable of emitting light in response to an electrical signal. The LED chips comprise respective groups emitting at different colors of light, with each of the groups interconnected in a series circuit. A lens is included over the LED chips. Other embodiments can comprise thermal spreading structures included integral to the submount and arranged to dissipate heat from the LED chips.

  4. Solid state lighting component

    DOE Patents [OSTI]

    Yuan, Thomas; Keller, Bernd; Ibbetson, James; Tarsa, Eric; Negley, Gerald

    2010-10-26

    An LED component comprising an array of LED chips mounted on a planar surface of a submount with the LED chips capable of emitting light in response to an electrical signal. The LED chips comprise respective groups emitting at different colors of light, with each of the groups interconnected in a series circuit. A lens is included over the LED chips. Other embodiments can comprise thermal spreading structures included integral to the submount and arranged to dissipate heat from the LED chips.

  5. Materials & Components Compatibility

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Components Compatibility - Sandia Energy Energy Search Icon Sandia Home Locations Contact Us Employee Locator Energy & Climate Secure & Sustainable Energy Future Stationary Power Energy Conversion Efficiency Solar Energy Wind Energy Water Power Supercritical CO2 Geothermal Natural Gas Safety, Security & Resilience of the Energy Infrastructure Energy Storage Nuclear Power & Engineering Grid Modernization Battery Testing Nuclear Fuel Cycle Defense Waste Management Programs Advanced

  6. Surface mount component jig

    DOE Patents [OSTI]

    Kronberg, James W. (Beech Island, SC)

    1990-08-07

    A device for bending and trimming the pins of a dual-inline-package component and the like for surface mounting rather than through mounting to a circuit board comprises, in a first part, in pin cutter astride a holder having a recess for holding the component, a first spring therebetween, and, in a second part, two flat members pivotally interconnected by a hinge and urged to an upward peaked position from a downward peaked position by a second spring. As a downward force is applied to the pin cutter it urges the holder downward, assisted by the first spring and a pair of ridges riding on shoulders of the holder, to carry the component against the upward peaked flat members which guide the pins outwardly. As the holder continues downwardly, the flat members pivot to the downward peaked position bending the pins upwardly against the sides of the holder. When the downward movement is met with sufficient resistance, the ridges of the pin cutter ride over the holder's shoulders to continue downward to cut any excess length of pin.

  7. Current Abstracts Nuclear Reactors and Technology

    SciTech Connect (OSTI)

    Bales, J.D.; Hicks, S.C.

    1993-01-01

    This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.

  8. Zero Power Reactor simulation | Argonne National Laboratory

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Zero Power Reactor simulation Share Description Ever wanted to see a nuclear reactor core in action? Here's a detailed simulation of the Zero Power Reactor experiment, run by...

  9. Chemical kinetic modeling of component mixtures relevant to gasoline

    SciTech Connect (OSTI)

    Mehl, M; Curran, H J; Pitz, W J; Westbrook, C K

    2009-02-13

    Real fuels are complex mixtures of thousands of hydrocarbon compounds including linear and branched paraffins, naphthenes, olefins and aromatics. It is generally agreed that their behavior can be effectively reproduced by simpler fuel surrogates containing a limited number of components. In this work, a recently revised version of the kinetic model by the authors is used to analyze the combustion behavior of several components relevant to gasoline surrogate formulation. Particular attention is devoted to linear and branched saturated hydrocarbons (PRF mixtures), olefins (1-hexene) and aromatics (toluene). Model predictions for pure components, binary mixtures and multi-component gasoline surrogates are compared with recent experimental information collected in rapid compression machine, shock tube and jet stirred reactors covering a wide range of conditions pertinent to internal combustion engines. Simulation results are discussed focusing attention on the mixing effects of the fuel components.

  10. Methanation assembly using multiple reactors

    DOE Patents [OSTI]

    Jahnke, Fred C.; Parab, Sanjay C.

    2007-07-24

    A methanation assembly for use with a water supply and a gas supply containing gas to be methanated in which a reactor assembly has a plurality of methanation reactors each for methanating gas input to the assembly and a gas delivery and cooling assembly adapted to deliver gas from the gas supply to each of said methanation reactors and to combine water from the water supply with the output of each methanation reactor being conveyed to a next methanation reactor and carry the mixture to such next methanation reactor.

  11. Nuclear reactor safety device

    DOE Patents [OSTI]

    Hutter, Ernest (Wilmette, IL)

    1986-01-01

    A safety device is disclosed for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of an upward thermal excursion. Such safety device comprises a laminated helical ribbon configured as a tube-like helical coil having contiguous helical turns with slidably abutting edges. The helical coil is disclosed as a portion of a drive member connected axially to the control rod. The laminated ribbon is formed of outer and inner laminae. The material of the outer lamina has a greater thermal coefficient of expansion than the material of the inner lamina. In the event of an upward thermal excursion, the laminated helical coil curls inwardly to a smaller diameter. Such inward curling causes the total length of the helical coil to increase by a substantial increment, so that the control rod is axially repositioned by a corresponding amount to reduce the power output of the reactor.

  12. Thermionic Reactor Design Studies

    SciTech Connect (OSTI)

    Schock, Alfred

    1994-08-01

    Paper presented at the 29th IECEC in Monterey, CA in August 1994. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their (thermionic reactor) performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling.

  13. Reactor for exothermic reactions

    DOE Patents [OSTI]

    Smith, Jr., Lawrence A. (Bellaire, TX); Hearn, Dennis (Houston, TX); Jones, Jr., Edward M. (Friendswood, TX)

    1993-01-01

    A liquid phase process for oligomerization of C.sub.4 and C.sub.5 isoolefins or the etherification thereof with C.sub.1 to C.sub.6 alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120.degree. to 300.degree. F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

  14. Dynamic bed reactor

    DOE Patents [OSTI]

    Stormo, Keith E. (Moscow, ID)

    1996-07-02

    A dynamic bed reactor is disclosed in which a compressible open cell foam matrix is periodically compressed and expanded to move a liquid or fluid through the matrix. In preferred embodiments, the matrix contains an active material such as an enzyme, biological cell, chelating agent, oligonucleotide, adsorbent or other material that acts upon the liquid or fluid passing through the matrix. The active material may be physically immobilized in the matrix, or attached by covalent or ionic bonds. Microbeads, substantially all of which have diameters less than 50 microns, can be used to immobilize the active material in the matrix and further improve reactor efficiency. A particularly preferred matrix is made of open cell polyurethane foam, which adsorbs pollutants such as polychlorophenol or o-nitrophenol. The reactors of the present invention allow unidirectional non-laminar flow through the matrix, and promote intimate exposure of liquid reactants to active agents such as microorganisms immobilized in the matrix.

  15. Reactor for exothermic reactions

    DOE Patents [OSTI]

    Smith, L.A. Jr.; Hearn, D.; Jones, E.M. Jr.

    1993-03-02

    A liquid phase process is described for oligomerization of C[sub 4] and C[sub 5] isoolefins or the etherification thereof with C[sub 1] to C[sub 6] alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120 to 300 F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

  16. Heat dissipating nuclear reactor

    DOE Patents [OSTI]

    Hunsbedt, A.; Lazarus, J.D.

    1985-11-21

    Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extend from the metal base plate downwardly and outwardly into the earth.

  17. Heat dissipating nuclear reactor

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA); Lazarus, Jonathan D. (Sunnyvale, CA)

    1987-01-01

    Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extends from the metal base plate downwardly and outwardly into the earth.

  18. Fast quench reactor method

    DOE Patents [OSTI]

    Detering, Brent A. (Idaho Falls, ID); Donaldson, Alan D. (Idaho Falls, ID); Fincke, James R. (Idaho Falls, ID); Kong, Peter C. (Idaho Falls, ID); Berry, Ray A. (Idaho Falls, ID)

    1999-01-01

    A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a means of rapidly expanding a reactant stream, such as a restrictive convergent-divergent nozzle at its outlet end. Metal halide reactants are injected into the reactor chamber. Reducing gas is added at different stages in the process to form a desired end product and prevent back reactions. The resulting heated gaseous stream is then rapidly cooled by expansion of the gaseous stream.

  19. Fusion reactor pumped laser

    DOE Patents [OSTI]

    Jassby, Daniel L. (Princeton, NJ)

    1988-01-01

    A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam.

  20. Plug Flow Reactor Simulator

    Energy Science and Technology Software Center (OSTI)

    1996-07-30

    PLUG is a computer program that solves the coupled steady state continuity, momentum, energy, and species balance equations for a plug flow reactor. Both homogeneous (gas-phase) and heterogenous (surface) reactions can be accommodated. The reactor may be either isothermal or adiabatic or may have a specified axial temperature or heat flux profile; alternatively, an ambient temperature and an overall heat-transfer coefficient can be specified. The crosssectional area and surface area may vary with axial position,more » and viscous drag is included. Ideal gas behavior and surface site conservation are assumed.« less

  1. Nuclear reactor apparatus

    DOE Patents [OSTI]

    Wade, Elman E.

    1978-01-01

    A lifting, rotating and sealing apparatus for nuclear reactors utilizing rotating plugs above the nuclear reactor core. This apparatus permits rotation of the plugs to provide under the plug refueling of a nuclear core. It also provides a means by which positive top core holddown can be utilized. Both of these operations are accomplished by means of the apparatus lifting the top core holddown structure off the nuclear core while stationary, and maintaining this structure in its elevated position during plug rotation. During both of these operations, the interface between the rotating member and its supporting member is sealingly maintained.

  2. Fast quench reactor method

    DOE Patents [OSTI]

    Detering, B.A.; Donaldson, A.D.; Fincke, J.R.; Kong, P.C.; Berry, R.A.

    1999-08-10

    A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a means of rapidly expanding a reactant stream, such as a restrictive convergent-divergent nozzle at its outlet end. Metal halide reactants are injected into the reactor chamber. Reducing gas is added at different stages in the process to form a desired end product and prevent back reactions. The resulting heated gaseous stream is then rapidly cooled by expansion of the gaseous stream. 8 figs.

  3. Perspectives on reactor safety

    SciTech Connect (OSTI)

    Haskin, F.E.; Camp, A.L.

    1994-03-01

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

  4. Instrumentation to Enhance Advanced Test Reactor Irradiations

    SciTech Connect (OSTI)

    J. L. Rempe; D. L. Knudson; K. G. Condie; J. E. Daw; S. C. Taylor

    2009-09-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

  5. Closure head for a nuclear reactor

    DOE Patents [OSTI]

    Wade, Elman E. (South Huntingdon, PA)

    1980-01-01

    A closure head for a nuclear reactor includes a stationary outer ring integral with the reactor vessel with a first rotatable plug disposed within the stationary outer ring and supported from the stationary outer ring by a bearing assembly. A sealing system is associated with the bearing assembly to seal the annulus defined between the first rotatable plug and the stationary outer ring. The sealing system comprises tubular seal elements disposed in the annulus with load springs contacting the tubular seal elements so as to force the tubular seal elements against the annulus in a manner to seal the annulus. The sealing system also comprises a sealing fluid which is pumped through the annulus and over the tubular seal elements causing the load springs to compress thereby reducing the friction between the tubular seal elements and the rotatable components while maintaining a gas-tight seal therebetween.

  6. Metal Cutting for Large Component Removal

    SciTech Connect (OSTI)

    Hulick, Robert M.

    2008-01-15

    Decommissioning of commercial nuclear power plants presents technological challenges. One major challenge is the removal of large components mainly consisting of the reactor vessel, steam generators and pressurizer. In order to remove and package these large components nozzles must be cut from the reactor vessel to precise tolerances. In some cases steam generators must be segmented for size and weight reduction. One innovative technology that has been used successfully at several commercial nuclear plant decommissioning is diamond wire sawing. Diamond wire sawing is performed by rotating a cable with diamond segments attached using a flywheel approximately 24 inches in diameter driven remotely by a hydraulic pump. Tension is provided using a gear rack drive which also takes up the slack in the wire. The wire is guided through the use of pulleys keeps the wire in a precise location. The diamond wire consists of 1/4 inch aircraft cable with diamond beads strung over the cable separated by springs and brass crimps. Standard wire contains 40 diamond beads per meter and can be made to any length. Cooling the wire and controlling the spread of contamination presents significant challenges. Under normal circumstances the wire is cooled and the cutting kerf cleaned by using water. In some cases of reactor nozzle cuts the use of water is prohibited because it cannot be controlled. This challenge was solved by using liquid Carbon Dioxide as the cooling agent. The liquid CO{sub 2} is passed through a special nozzle which atomizes the liquid into snowflakes which is introduced under pressure to the wire. The snowflakes attach to the wire keeping it cool and to the metal shavings. As the CO{sub 2} and metal shavings are released from the wire due to its fast rotation, the snowflakes evaporate leaving only the fine metal shavings as waste. Secondary waste produced is simply the small volume of fine metal shavings removed from the cut surface. Diamond wire sawing using CO{sub 2} cooling has been employed for cutting the reactor nozzles at San Onofre Unit 1 and at Connecticut Yankee. These carbon steel nozzles ranged up to 54 inch diameter with a 15 inch thick wall and an interior stainless cladding. Diamond wire sawing using traditional water cooling has been used to segment the reactor head at Rancho Seco and for cutting reactor nozzles and control rod drive tubes at Dairyland Power's Lacrosse BWR project. Advantages: - ALARA: All cutting is preformed remotely significantly reducing dose. Stringing of wires is accomplished using long handle tools. - Secondary waste is reduced to just the volume of material cut with the diamond wire. - The potential for airborne contamination is eliminated. Due to the flexibility of the wire, any access restrictions and interferences can be accommodated using pulleys and long handle tools. - The operation is quiet. Disadvantages: - With Liquid Carbon Dioxide cooling and cleaning, delivery of the material must be carefully planned. The longer the distance from the source to the cut area, the greater the chance for pressure drop and subsequent problems with line freezing. - Proper shrouding and ventilation are required for environmental reasons. In each case, the metal structures were cut at a precise location. Radiation dose was reduced significantly by operating the equipment from a remote location. The cuts were very smooth and completed on schedule. Each project must be analyzed individually and take into account many factors including access, radiological conditions, environmental conditions, schedule requirements, packaging requirements and size of cuts.

  7. B Reactor | Department of Energy

    Broader source: Energy.gov (indexed) [DOE]

    boomtown, with the population reaching 50,000 by summer 1944. Similar to the X-10 Graphite Reactor at Oak Ridge in terms of loading and unloading fuel, the B Reactor was built...

  8. Steam generator for liquid metal fast breeder reactor

    DOE Patents [OSTI]

    Gillett, James E. (Greensburg, PA); Garner, Daniel C. (Murrysville, PA); Wineman, Arthur L. (Greensburg, PA); Robey, Robert M. (North Huntingdon, PA)

    1985-01-01

    Improvements in the design of internal components of J-shaped steam generators for liquid metal fast breeder reactors. Complex design improvements have been made to the internals of J-shaped steam generators which improvements are intended to reduce tube vibration, tube jamming, flow problems in the upper portion of the steam generator, manufacturing complexities in tube spacer attachments, thermal stripping potentials and difficulties in the weld fabrication of certain components.

  9. Pressurized reactor system and a method of operating the same

    DOE Patents [OSTI]

    Isaksson, Juhani M. (Karhula, FI)

    1996-01-01

    A method and apparatus are provided for operating a pressurized reactor system in order to precisely control the temperature within a pressure vessel in order to minimize condensation of corrosive materials from gases on the surfaces of the pressure vessel or contained circulating fluidized bed reactor, and to prevent the temperature of the components from reaching a detrimentally high level, while at the same time allowing quick heating of the pressure vessel interior volume during start-up. Superatmospheric pressure gas is introduced from the first conduit into the fluidized bed reactor and heat derived reactions such as combustion and gassification are maintained in the reactor. Gas is exhausted from the reactor and pressure vessel through a second conduit. Gas is circulated from one part of the inside volume to another to control the temperature of the inside volume, such as by passing the gas through an exterior conduit which has a heat exchanger, control valve, blower and compressor associated therewith, or by causing natural convection flow of circulating gas within one or more generally vertically extending gas passages entirely within the pressure vessel (and containing heat exchangers, flow rate control valves, or the like therein). Preferably, inert gas is provided as a circulating gas, and the inert gas may also be used in emergency shut-down situations. In emergency shut-down reaction gas being supplied to the reactor is cut off, while inert gas from the interior gas volume of the pressure vessel is introduced into the reactor.

  10. Pressurized reactor system and a method of operating the same

    DOE Patents [OSTI]

    Isaksson, J.M.

    1996-06-18

    A method and apparatus are provided for operating a pressurized reactor system in order to precisely control the temperature within a pressure vessel in order to minimize condensation of corrosive materials from gases on the surfaces of the pressure vessel or contained circulating fluidized bed reactor, and to prevent the temperature of the components from reaching a detrimentally high level, while at the same time allowing quick heating of the pressure vessel interior volume during start-up. Super-atmospheric pressure gas is introduced from the first conduit into the fluidized bed reactor and heat derived reactions such as combustion and gasification are maintained in the reactor. Gas is exhausted from the reactor and pressure vessel through a second conduit. Gas is circulated from one part of the inside volume to another to control the temperature of the inside volume, such as by passing the gas through an exterior conduit which has a heat exchanger, control valve, blower and compressor associated therewith, or by causing natural convection flow of circulating gas within one or more generally vertically extending gas passages entirely within the pressure vessel (and containing heat exchangers, flow rate control valves, or the like therein). Preferably, inert gas is provided as a circulating gas, and the inert gas may also be used in emergency shut-down situations. In emergency shut-down reaction gas being supplied to the reactor is cut off, while inert gas from the interior gas volume of the pressure vessel is introduced into the reactor. 2 figs.

  11. Piqua, Ohio, Decommissioned Reactor Site

    Office of Legacy Management (LM)

    Piqua, Ohio, Decommissioned Reactor Site This fact sheet provides information about the Piqua, Ohio, Decommissioned Reactor. This site is managed by the U.S. Department of Energy Office of Legacy Management under the DOE Defense Decontamination and Decommissioning (D&D) Program. Location of the Piqua Decommissioned Reactor Site Description and History The Piqua, Ohio, Decommissioned Reactor site is located in southwestern Ohio in the city of Piqua on the east bank of the Great Miami River,

  12. Reactor operation safety information document

    SciTech Connect (OSTI)

    Not Available

    1990-01-01

    The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

  13. Reactor Materials | Department of Energy

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

    Reactor Materials Reactor Materials The reactor materials crosscut effort will enable the development of innovative and revolutionary materials and provide broad-based, modern materials science that will benefit all four DOE-NE objectives. This will be accomplished through innovative materials development, promoting the use of modern materials science and establishing new, shared research partnerships. Research into specific degradation modes or material needs unique to a particular reactor

  14. Hallam, Nebraska, Decommissioned Reactor Site

    Office of Legacy Management (LM)

    D&D D&D Hallam, Nebraska, Decommissioned Reactor Site This fact sheet provides information about the Hallam, Nebraska, Decommissioned Reactor Site. This site is managed by the U.S. Department of Energy Office of Legacy Management under the Defense Decontamination and Decommissioning (D&D) Program. Location of the Hallam Decommissioned Reactor Site Description and History The Hallam decommissioned reactor site is in southeastern Nebraska, approximately 19 miles south of Lincoln. The

  15. Thermal Reactor Safety

    SciTech Connect (OSTI)

    Not Available

    1980-06-01

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

  16. Cermet fuel reactors

    SciTech Connect (OSTI)

    Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.; Barner, J.O.

    1987-09-01

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs.

  17. Fossil fuel furnace reactor

    DOE Patents [OSTI]

    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.

  18. JACKETED REACTOR FUEL ELEMENT

    DOE Patents [OSTI]

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

    1958-12-01

    A fuel element is described for fast reactors comprised of a core of uranium metal containing material and a jacket around the core, the jacket consisting of from 2.5 to 15 percent of titanium, from 1 to 5 percent of niobium, and from 80 to 96.5 percent of vanadium.

  19. NETL - Chemical Looping Reactor

    SciTech Connect (OSTI)

    2013-07-24

    NETL's Chemical Looping Reactor unit is a high-temperature integrated CLC process with extensive instrumentation to improve computational simulations. A non-reacting test unit is also used to study solids flow at ambient temperature. The CLR unit circulates approximately 1,000 pounds per hour at temperatures around 1,800 degrees Fahrenheit.

  20. Nuclear reactor building

    DOE Patents [OSTI]

    Gou, Perng-Fei (Saratoga, CA); Townsend, Harold E. (Campbell, CA); Barbanti, Giancarlo (Sirtori, IT)

    1994-01-01

    A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed thereabove. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define therebetween an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin.

  1. NETL - Chemical Looping Reactor

    ScienceCinema (OSTI)

    None

    2014-06-26

    NETL's Chemical Looping Reactor unit is a high-temperature integrated CLC process with extensive instrumentation to improve computational simulations. A non-reacting test unit is also used to study solids flow at ambient temperature. The CLR unit circulates approximately 1,000 pounds per hour at temperatures around 1,800 degrees Fahrenheit.

  2. Nuclear reactor building

    DOE Patents [OSTI]

    Gou, P.F.; Townsend, H.E.; Barbanti, G.

    1994-04-05

    A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed there above. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define there between an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin. 4 figures.

  3. Sprayed skin turbine component

    DOE Patents [OSTI]

    Allen, David B

    2013-06-04

    Fabricating a turbine component (50) by casting a core structure (30), forming an array of pits (24) in an outer surface (32) of the core structure, depositing a transient liquid phase (TLP) material (40) on the outer surface of the core structure, the TLP containing a melting-point depressant, depositing a skin (42) on the outer surface of the core structure over the TLP material, and heating the assembly, thus forming both a diffusion bond and a mechanical interlock between the skin and the core structure. The heating diffuses the melting-point depressant away from the interface. Subsurface cooling channels (35) may be formed by forming grooves (34) in the outer surface of the core structure, filling the grooves with a fugitive filler (36), depositing and bonding the skin (42), then removing the fugitive material.

  4. REACTOR FUEL ELEMENTS TESTING CONTAINER

    DOE Patents [OSTI]

    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)

  5. EMERGENCY SHUTDOWN FOR NUCLEAR REACTORS

    DOE Patents [OSTI]

    Paget, J.A.; Koutz, S.L.; Stone, R.S.; Stewart, H.B.

    1963-12-24

    An emergency shutdown or scram apparatus for use in a nuclear reactor that includes a neutron absorber suspended from a temperature responsive substance that is selected to fail at a preselected temperature in excess of the normal reactor operating temperature, whereby the neutron absorber is released and allowed to fall under gravity to a preselected position within the reactor core is presented. (AEC)

  6. Passive cooling system for nuclear reactor containment structure

    DOE Patents [OSTI]

    Gou, Perng-Fei (Saratoga, CA); Wade, Gentry E. (Saratoga, CA)

    1989-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  7. Natural circulating passive cooling system for nuclear reactor containment structure

    DOE Patents [OSTI]

    Gou, Perng-Fei (Saratoga, CA); Wade, Gentry E. (Saratoga, CA)

    1990-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  8. Influence of nitrogen in the shielding gas on corrosion resistance of duplex stainless steel welds

    SciTech Connect (OSTI)

    Bhatt, R.B.; Kamat, H.S.; Ghosal, S.K.; De, P.K.

    1999-10-01

    The influence of nitrogen in shielding gas on the corrosion resistance of welds of a duplex stainless steel (grade U-50), obtained by gas tungsten arc (GTA) with filler wire, autogenous GTA (bead-on-plate), electron beam welding (EBW), and microplasma techniques, has been evaluated in chloride solutions at 30 C. Pitting attack has been observed in GTA, electron beam welding, and microplasma welds when welding has been carried out using pure argon as the shielding gas. Gas tungsten arc welding with 5 to 10% nitrogen and 90 to 95% argon, as the shielding gas, has been found to result in an improved pitting corrosion resistance of the weldments of this steel. However, the resistance of pitting of autogenous welds (bead-on-plate) obtained in pure argon as the shielding gas has been observed to remain unaffected. Microscopic examination, electron probe microanalysis (EPMA), and x-ray diffraction studies have revealed that the presence of nitrogen in the shielding gas in the GTA welds not only modifies the microstructure and the austenite to ferrite ratio but also results in a nearly uniform distribution of the various alloying elements, for example, chromium, nickel, and molybdenum among the constituent phases, which are responsible for improved resistance to pitting corrosion.

  9. Passive shielding effect on space profile of magnetic field emissions for wireless power transfer to vehicles

    SciTech Connect (OSTI)

    Batra, T. Schaltz, E.

    2015-05-07

    Magnetic fields emitted by wireless power transfer systems are of high importance with respect to human safety and health. Aluminum and ferrite are used in the system to reduce the fields and are termed as passive shielding. In this paper, the influence of these materials on the space profile has been investigated with the help of simulations on Comsol for the four possible geometriesno shielding, ferrite, aluminum, and full shielding. As the reflected impedance varies for the four geometries, the primary current is varied accordingly to maintain constant power transfer to the secondary side. Surrounding magnetic field plots in the vertical direction show that maxima's of the two coils for the no shielding geometry are centered at the respective coils and for the remaining three are displaced closer to each other. This closeness would lead to more effective addition of the two coil fields and an increase in the resultant field from space point of view. This closeness varies with distance in the horizontal direction and vertical gap between the coils and is explained in the paper. This paper provides a better understanding of effect of the passive shielding materials on the space nature of magnetic fields for wireless power transfer for vehicle applications.

  10. Advanced burner test reactor preconceptual design report.

    SciTech Connect (OSTI)

    Chang, Y. I.; Finck, P. J.; Grandy, C.; Cahalan, J.; Deitrich, L.; Dunn, F.; Fallin, D.; Farmer, M.; Fanning, T.; Kim, T.; Krajtl, L.; Lomperski, S.; Moisseytsev, A.; Momozaki, Y.; Sienicki, J.; Park, Y.; Tang, Y.; Reed, C.; Tzanos, C; Wiedmeyer, S.; Yang, W.; Chikazawa, Y.; JAEA

    2008-12-16

    The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. Implementation of the GNEP requires development and demonstration of three major technologies: (1) Light water reactor (LWR) spent fuel separations technologies that will recover transuranics to be recycled for fuel but not separate plutonium from other transuranics, thereby providing proliferation-resistance; (2) Advanced Burner Reactors (ABRs) based on a fast spectrum that transmute the recycled transuranics to produce energy while also reducing the long term radiotoxicity and decay heat loading in the repository; and (3) Fast reactor fuel recycling technologies to recover and refabricate the transuranics for repeated recycling in the fast reactor system. The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat. The Advanced Burner Test Reactor (ABTR) is the first step in demonstrating the transmutation technologies. It directly supports development of a prototype full-scale Advanced Burner Reactor, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR are: (1) To demonstrate reactor-based transmutation of transuranics as part of an advanced fuel cycle; (2) To qualify the transuranics-containing fuels and advanced structural materials needed for a full-scale ABR; and (3) To support the research, development and demonstration required for certification of an ABR standard design by the U.S. Nuclear Regulatory Commission. The ABTR should also address the following additional objectives: (1) To incorporate and demonstrate innovative design concepts and features that may lead to significant improvements in cost, safety, efficiency, reliability, or other favorable characteristics that could promote public acceptance and future private sector investment in ABRs; (2) To demonstrate improved technologies for safeguards and security; and (3) To support development of the U.S. infrastructure for design, fabrication and construction, testing and deployment of systems, structures and components for the ABRs. Based on these objectives, a pre-conceptual design of a 250 MWt ABTR has been developed; it is documented in this report. In addition to meeting the primary and additional objectives listed above, the lessons learned from fast reactor programs in the U.S. and worldwide and the operating experience of more than a dozen fast reactors around the world, in particular the Experimental Breeder Reactor-II have been incorporated into the design of the ABTR to the extent possible.

  11. Industrial Hygiene Concerns during the Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    SciTech Connect (OSTI)

    M.E. Lumia; C.A. Gentile

    2002-01-18

    A significant industrial hygiene concern during the Decontamination and Decommissioning (D and D) of the Tokamak Fusion Test Reactor (TFTR) was the oxidation of the lead bricks' surface, which were utilized for radiation shielding. This presented both airborne exposure and surface contamination issues for the workers in the field removing this material. This paper will detail the various protection and control methods tested and implemented to protect the workers, including those technologies deployed to decontaminate the work surfaces. In addition, those techniques employed to recycle the lead for additional use at the site will be discussed.

  12. Stationary Liquid Fuel Fast Reactor

    SciTech Connect (OSTI)

    Yang, Won Sik; Grandy, Andrew; Boroski, Andrew; Krajtl, Lubomir; Johnson, Terry

    2015-09-30

    For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named SLFFR (Stationary Liquid Fuel Fast Reactor) was proposed based on stationary molten metallic fuel. The fuel enters the reactor vessel in a solid form, and then it is heated to molten temperature in a small melting heater. The fuel is contained within a closed, thick container with penetrating coolant channels, and thus it is not mixed with coolant nor flow through the primary heat transfer circuit. The makeup fuel is semi- continuously added to the system, and thus a very small excess reactivity is required. Gaseous fission products are also removed continuously, and a fraction of the fuel is periodically drawn off from the fuel container to a processing facility where non-gaseous mixed fission products and other impurities are removed and then the cleaned fuel is recycled into the fuel container. A reference core design and a preliminary plant system design of a 1000 MWt TRU- burning SLFFR concept were developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches were adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses were performed to develop a reference core design. Region-dependent 33-group cross sections were generated based on the ENDF/B-VII.0 data using the MC2-3 code. Core and fuel cycle analyses were performed in theta-r-z geometries using the DIF3D and REBUS-3 codes. Reactivity coefficients and kinetics parameters were calculated using the VARI3D perturbation theory code. Thermo-fluidic analyses were performed using the ANSYS FLUENT computational fluid dynamics (CFD) code. Figure 0.1 shows a schematic radial layout of the reference 1000 MWt SLFFR core, and Table 0.1 summarizes the main design parameters of SLFFR-1000 loop plant. The fuel container is a 2.5 cm thick cylinder with an inner radius of 87.5 cm. The fuel container is penetrated by twelve hexagonal control assembly (CA) guide tubes, each of which has 3.0 mm thickness and 69.4 mm flat-to-flat outer distance. The distance between two neighboring CA guide tube is selected to be 26 cm to provide an adequate space for CA driving systems. The fuel container has 18181 penetrating coolant tubes of 6.0 mm inner diameter and 2.0 mm thickness. The coolant tubes are arranged in a triangular lattice with a lattice pitch of 1.21 cm. The fuel, structure, and coolant volume fractions inside the fuel container are 0.386, 0.383, and 0.231, respectively. Separate steel reflectors and B4C shields are used outside of the fuel container. Six gas expansion modules (GEMs) of 5.0 cm thickness are introduced in the radial reflector region. Between the radial reflector and the fuel container is a 2.5 cm sodium gap. The TRU inventory at the beginning of equilibrium cycle (BOEC) is 5081 kg, whereas the TRU inventory at the beginning of life (BOL) was 3541 kg. This is because the equilibrium cycle fuel contains a significantly smaller fissile fraction than the LWR TRU feed. The fuel inventory at BOEC is composed of 34.0 a/o TRU, 41.4 a/o Ce, 23.6 a/o Co, and 1.03 a/o solid fission products. Since uranium-free fuel is used, a theoretical maximum TRU consumption rate of 1.011 kg/day is achieved. The semi-continuous fuel cycle based on the 300-batch, 1- day cycle approximation yields a burnup reactivity loss of 26 pcm/day, and requires a daily reprocessing of 32.5 kg of SLFFR fuel. This yields a daily TRU charge rate of 17.45 kg, including a makeup TRU feed of 1.011 kg recovered from the LWR used fuel. The charged TRU-Ce-Co fuel is composed of 34.4 a/o TRU, 40.6 a/o Ce, and 25.0 a/o Co.

  13. Reactor vessel support system. [LMFBR

    DOE Patents [OSTI]

    Golden, M.P.; Holley, J.C.

    1980-05-09

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  14. Knowledge and abilities catalog for nuclear power plant operators: Pressurized water reactors. Revision 1

    SciTech Connect (OSTI)

    1995-08-01

    This document provides the basis for the development of content-valid licensing examinations for reactor operators and senior reactor operators. The examinations developed using the PWR catalog will cover those topics listed under Title 10, (ode of Federal Regulations Part 55. The PWR catalog contains approximately 5100 knowledge and ability (K/A) statements for reactor operators and senior reactor operators. The catalog is organized into six major sections: Catalog Organization; Generic Knowledge and Abilities; Plant Systems; Emergency and Abnormal Plant Evolutions; Components and Theory.

  15. Low-temperature Swelling in LWR Internal Components: Current Data and

    Office of Environmental Management (EM)

    Modeling Assessment | Department of Energy Low-temperature Swelling in LWR Internal Components: Current Data and Modeling Assessment Low-temperature Swelling in LWR Internal Components: Current Data and Modeling Assessment Recent experimental observations have made it clear that cavity formation can occur in light-water reactor internal components fabricated from austenitic stainless during the course of their service life. In order to assess the potential for cavity swelling in these

  16. Contoured inner after-heater shield for reducing stress in growing crystalline bodies

    DOE Patents [OSTI]

    Kalejs, Juris P. (Wellesley, MA)

    1996-09-24

    An apparatus for growing hollow crystalline bodies by the EFG process, comprising an EFG die having a top surface shaped for growing a hollow crystalline body having a cross-sectional configuration in the shape of a polygon having n faces, and a radiation shield adjacent to and surrounded by the top end surface of the die, characterized in that the shield has an inner edge defining a similar polygon with n sides, and the inner edge of the shield is notched so that the spacing between the n faces and the n sides is greatest between the central portions of the n faces and the n sides, whereby the greater spacing at the central portions helps to reduce lateral temperature gradients in the crystalline body that is grown by use of the die.

  17. Orthogonally interdigitated shielded serpentine travelling wave cathode ray tube deflection structure

    DOE Patents [OSTI]

    Hagen, E.C.; Hudson, C.L.

    1995-07-25

    A new deflection structure which deflects a beam of charged particles, such as an electron beam, includes a serpentine set for transmitting a deflection field, and a shielding frame for housing the serpentine set. The serpentine set includes a vertical serpentine deflection element and a horizontal serpentine deflection element. These deflection elements are identical, and are interdigitatedly and orthogonally disposed relative to each other, for forming a central transmission passage, through which the electron beam passes, and is deflected by the deflection field, so as to minimize drift space signal distortion. The shielding frame includes a plurality of ground blocks, and forms an internal serpentine trough within these ground blocks, for housing the serpentine set. The deflection structure further includes a plurality of feedthrough connectors which are inserted through the shielding frame, and which are electrically connected to the serpentine set. 10 figs.

  18. Nuclear reactor construction with bottom supported reactor vessel

    DOE Patents [OSTI]

    Sharbaugh, John E. (Bullskin Township, Fayette County, PA)

    1987-01-01

    An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core within the pool. The reactor vessel has an open top end, a closed flat bottom end wall and a continuous cylindrical closed side wall interconnecting the top end and bottom end wall. The reactor also has a generally cylindrical concrete containment structure surrounding the reactor vessel and being formed by a cylindrical side wall spaced outwardly from the reactor vessel side wall and a flat base mat spaced below the reactor vessel bottom end wall. A central support pedestal is anchored to the containment structure base mat and extends upwardly therefrom to the reactor vessel and upwardly therefrom to the reactor core so as to support the bottom end wall of the reactor vessel and the lower end of the reactor core in spaced apart relationship above the containment structure base mat. Also, an annular reinforced support structure is disposed in the reactor vessel on the bottom end wall thereof and extends about the lower end of the core so as to support the periphery thereof. In addition, an annular support ring having a plurality of inward radially extending linear members is disposed between the containment structure base mat and the bottom end of the reactor vessel wall and is connected to and supports the reactor vessel at its bottom end on the containment structure base mat so as to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event. The reactor construction also includes a bed of insulating material in sand-like granular form, preferably being high density magnesium oxide particles, disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall on the containment structure base mat so as to insulate the reactor vessel bottom end wall from the containment structure base mat and allow the reactor vessel bottom end wall to freely expand as it heats up while providing continuous support thereof. Further, a deck is supported upon the side wall of the containment structure above the top open end of the reactor vessel, and a plurality of serially connected extendible and retractable annular bellows extend between the deck and the top open end of the reactor vessel and flexibly and sealably interconnect the reactor vessel at its top end to the deck. An annular guide ring is disposed on the containment structure and extends between its side wall and the top open end of the reactor vessel for providing lateral support of the reactor vessel top open end by limiting imposition of lateral loads on the annular bellows by the occurrence of a lateral seismic event.

  19. TUNGSTEN SHIELDS FOR CS-137 INLINE MONITORS IN THE CAUSTIC SIDE SOLVENT EXTRACTION PROCESS

    SciTech Connect (OSTI)

    Casella, V; Mark Hogue, M; Javier Reyes-Jimenez, J; Paul Filpus-Luyckx, P; Timothy Riley, T; Fred Ogden, F; Donald Pak, D

    2007-05-10

    The Department of Energy (DOE) selected Caustic-Side Solvent Extraction (CSSX) as the preferred technology for the removal of radioactive cesium from High-Level Waste (HLW) at the Savannah River Site (SRS). The CSSX process is a continuous process that uses a novel solvent to extract cesium from highly radioactive waste and concentrate it in dilute nitric acid. In-line analyses are performed with gamma-ray monitors to measure the C-137 concentration in the decontaminated salt solution (DSS) and in the strip effluent (SE). Sodium iodide (NaI) monitors are used to measure the Cs-137 concentration before the DSS Hold Tank, while Geiger-Mueller (GM) monitors are used for Cs-137 measurements before the SE hold tank. Tungsten shields were designed using Monte Carlo calculations and fabricated to provide the needed reduction of the process background radiation at the detector positions. A one-inch tungsten cylindrical shield reduced the background radiation by a factor of fifty that was adequate for the GM detectors, while a three-and-one-half-inch tungsten cylindrical shield was required for the NaI detectors. Testing of the NaI shield was performed at the SRS Instrument Calibration Facility. Based on this testing, the as-built shield is predicted to be able to detect the MCU DSS stream at concentrations above 0.003 Ci/gal under the ''worst case'' field conditions with a MCU feed solution of 1.1 Ci/gal and all of the process tanks completely full. This paper discusses the design, fabrication, testing and implementation of the tungsten shields in the MCU facility.

  20. Optimizing moderation of He-3 neutron detectors for shielded fission sources

    SciTech Connect (OSTI)

    Rees, Lawrence B. [Brigham Young University, Provo, UT (United States); Czirr, J. Bart [Brigham Young University, Provo, UT (United States)

    2012-11-01

    Abstract: The response of 3-He neutron detectors is highly dependent on the amount of moderator incorporated into the detector system. If there is too little moderation, neutrons will not react with the 3-He. If there is too much moderation, neutrons will not reach the 3-He. In applications for portal or border monitors where 3He detectors are used to interdict illicit Importation of plutonium, the fission source is always shielded to some extent. Since the energy distribution of neutrons emitted from the source depends on the amount and type of shielding present, the optimum placement of moderating material around 3-He tubes is a function of shielding. In this paper, we use Monte Carlo techniques to model the response of 3-He tubes placed in polyethylene boxes for moderation. To model the shielded fission neutron source, we use a 252-Cf source placed in the center of spheres of water of varying radius. Detector efficiency as a function of box geometry and shielding are explored. We find that increasing the amount of moderator behind and to the sides of the detector generally improves the detector response, but that benefits are limited if the thickness of the polyethylene moderator is greater than about 5-7 cm. The thickness of the moderator in front of the 3He tubes, however, is very important. For bare sources, about 5-6 cm of moderator is optimum, but as the shielding increases, the optimum thickness of this moderator decreases to 0-1 cm. A two-tube box with a moderator thickness of 5 cm in front of the first tube and a thickness of 1 cm in front of the second tube is proposed to improve the detector's sensitivity to lower-energy neutrons.

  1. Optimizing moderation of He-3 neutron detectors for shielded fission sources

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Rees, Lawrence B.; Czirr, J. Bart

    2012-07-10

    Abstract: The response of 3-He neutron detectors is highly dependent on the amount of moderator incorporated into the detector system. If there is too little moderation, neutrons will not react with the 3-He. If there is too much moderation, neutrons will not reach the 3-He. In applications for portal or border monitors where 3He detectors are used to interdict illicit Importation of plutonium, the fission source is always shielded to some extent. Since the energy distribution of neutrons emitted from the source depends on the amount and type of shielding present, the optimum placement of moderating material around 3-He tubesmore » is a function of shielding. In this paper, we use Monte Carlo techniques to model the response of 3-He tubes placed in polyethylene boxes for moderation. To model the shielded fission neutron source, we use a 252-Cf source placed in the center of spheres of water of varying radius. Detector efficiency as a function of box geometry and shielding are explored. We find that increasing the amount of moderator behind and to the sides of the detector generally improves the detector response, but that benefits are limited if the thickness of the polyethylene moderator is greater than about 5-7 cm. The thickness of the moderator in front of the 3He tubes, however, is very important. For bare sources, about 5-6 cm of moderator is optimum, but as the shielding increases, the optimum thickness of this moderator decreases to 0-1 cm. A two-tube box with a moderator thickness of 5 cm in front of the first tube and a thickness of 1 cm in front of the second tube is proposed to improve the detector's sensitivity to lower-energy neutrons.« less

  2. HOMOGENEOUS NUCLEAR REACTOR

    DOE Patents [OSTI]

    Hammond, R.P.; Busey, H.M.

    1959-02-17

    Nuclear reactors of the homogeneous liquid fuel type are discussed. The reactor is comprised of an elongated closed vessel, vertically oriented, having a critical region at the bottom, a lower chimney structure extending from the critical region vertically upwardly and surrounded by heat exchanger coils, to a baffle region above which is located an upper chimney structure containing a catalyst functioning to recombine radiolyticallydissociated moderator gages. In operation the liquid fuel circulates solely by convection from the critical region upwardly through the lower chimney and then downwardly through the heat exchanger to return to the critical region. The gases formed by radiolytic- dissociation of the moderator are carried upwardly with the circulating liquid fuel and past the baffle into the region of the upper chimney where they are recombined by the catalyst and condensed, thence returning through the heat exchanger to the critical region.

  3. Nuclear reactor safety device

    DOE Patents [OSTI]

    Hutter, E.

    1983-08-15

    A safety device is described for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of a thermal excursion. It comprises a laminated strip helically configured to form a tube, said tube being in operative relation to said control rod. The laminated strip is formed of at least two materials having different thermal coefficients of expansion, and is helically configured such that the material forming the outer lamina of the tube has a greater thermal coefficient of expansion than the material forming the inner lamina of said tube. In the event of a thermal excursion the laminated strip will tend to curl inwardly so that said tube will increase in length, whereby as said tube increases in length it exerts a force on said control rod to axially reposition said control rod with respect to said core.

  4. Neutronic reactor construction

    DOE Patents [OSTI]

    Huston, Norman E.

    1976-07-06

    1. A neutronic reactor comprising a moderator including horizontal layers formed of horizontal rows of graphite blocks, alternate layers of blocks having the rows extending in one direction, the remaining alternate layers having the rows extending transversely to the said one direction, alternate rows of blocks in one set of alternate layers having longitudinal ducts, the moderator further including slotted graphite tubes positioned in the ducts, the reactor further comprising an aluminum coolant tube positioned within the slotted tube in spaced relation thereto, bodies of thermal-neutron-fissionable material, and jackets enclosing the bodies and being formed of a corrosion-resistant material having a low neutron-capture cross section, the bodies and jackets being positioned within the coolant tube so that the jackets are spaced from the coolant tube.

  5. Fusion reactor pumped laser

    DOE Patents [OSTI]

    Jassby, D.L.

    1987-09-04

    A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam. 10 figs.

  6. Nuclear reactor shutdown system

    DOE Patents [OSTI]

    Bhate, Suresh K. (Niskayuna, NY); Cooper, Martin H. (Monroeville, PA); Riffe, Delmar R. (Murrysville, PA); Kinney, Calvin L. (Penn Hills, PA)

    1981-01-01

    An inherent shutdown system for a nuclear reactor having neutron absorbing rods affixed to an armature which is held in an upper position by a magnetic flux flowing through a Curie temperature material. The Curie temperature material is fixedly positioned about the exterior of an inner duct in an annular region through which reactor coolant flows. Elongated fuel rods extending from within the core upwardly toward the Curie temperature material are preferably disposed within the annular region. Upon abnormal conditions which result in high neutron flux and coolant temperature, the Curie material loses its magnetic permeability, breaking the magnetic flux path and allowing the armature and absorber rods to drop into the core, thus shutting down the fissioning reaction. The armature and absorber rods are retrieved by lowering the housing for the electromagnet forming coils which create a magnetic flux path which includes the inner duct wall. The coil housing then is raised, resetting the armature.

  7. Apparatus for characterizing conductivity of materials by measuring the effect of induced shielding currents therein

    DOE Patents [OSTI]

    Doss, J.D.

    1991-05-14

    Apparatus and method for noncontact, radio-frequency shielding current characterization of materials is disclosed. Self- or mutual inductance changes in one or more inductive elements, respectively, occur when materials capable of supporting shielding currents are placed in proximity thereto, or undergo change in resistivity while in place. Such changes can be observed by incorporating the inductor(s) in a resonant circuit and determining the frequency of oscillation or by measuring the voltage induced on a coupled inductive element. The present invention is useful for determining the critical temperature and superconducting transition width for superconducting samples. 10 figures.

  8. Apparatus for characterizing conductivity of materials by measuring the effect of induced shielding currents therein

    DOE Patents [OSTI]

    Doss, James D. (Los Alamos, NM)

    1991-01-01

    Apparatus and method for noncontact, radio-frequency shielding current characterization of materials. Self- or mutual inductance changes in one or more inductive elements, respectively, occur when materials capable of supporting shielding currents are placed in proximity thereto, or undergo change in resistivity while in place. Such changes can be observed by incorporating the inductor(s) in a resonant circuit and determining the frequency of oscillation or by measuring the voltage induced on a coupled inductive element. The present invention is useful for determining the critical temperature and superconducting transition width for superconducting samples.

  9. Optimization of the Transport Shield for Neutrinoless Double Beta-decay Enriched Germanium

    SciTech Connect (OSTI)

    Aguayo Navarrete, Estanislao; Kouzes, Richard T.; Orrell, John L.; Reid, Douglas J.; Fast, James E.

    2012-04-15

    This document presents results of an investigation of the material and geometry choice for the transport shield of germanium, the active detector material used in 76Ge neutrinoless double beta decay searches. The objective of this work is to select the optimal material and geometry to minimize cosmogenic production of radioactive isotopes in the germanium material. The design of such a shield is based on the calculation of the cosmogenic production rate of isotopes that are known to cause interfering backgrounds in 76Ge neutrinoless double beta decay searches.

  10. Experimental Test of Self-Shielding in VUV Photodissociation of CO

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Experimental Test of Self-Shielding in VUV Photodissociation of CO Experimental Test of Self-Shielding in VUV Photodissociation of CO Print Wednesday, 25 March 2009 00:00 One way to test models of the solar system's formation is to compare the isotopic abundances of the elements found in its constituent bodies. A case in point is oxygen with three stable isotopes dominated by oxygen-16, with minute fractions of oxygen-17 and oxygen-18. Primitive objects whose formation predates the Earth's, such

  11. Graphite-ceramic rf Faraday-thermal shield and plasma limiter

    DOE Patents [OSTI]

    Hwang, D.L.Q.; Hosea, J.C.

    1983-05-05

    The present invention is directed to a brazing procedure for joining a ceramic or glass material (e.g., Al/sub 2/O/sub 3/ or Macor) to graphite. In particular, the present invention is directed to a novel brazing procedure for the production of a brazed ceramic graphite product useful as a Faraday shield. The brazed ceramic graphite Faraday shield of the present invention may be used in Magnetic Fusion Devices (e.g., Princeton Large Torus Tokamak) or other high temperature resistant apparatus.

  12. In situ reactor

    DOE Patents [OSTI]

    Radtke, Corey William; Blackwelder, David Bradley

    2004-01-27

    An in situ reactor for use in a geological strata, is described and which includes a liner defining a centrally disposed passageway and which is placed in a borehole formed in the geological strata; and a sampling conduit is received within the passageway defined by the liner and which receives a geological specimen which is derived from the geological strata, and wherein the sampling conduit is in fluid communication with the passageway defined by the liner.

  13. Uncertainty and sensitivity analysis of early exposure results with the MACCS Reactor Accident Consequence Model

    SciTech Connect (OSTI)

    Helton, J.C.; Johnson, J.D.; McKay, M.D.; Shiver, A.W.; Sprung, J.L.

    1995-01-01

    Uncertainty and sensitivity analysis techniques based on Latin hypercube sampling, partial correlation analysis and stepwise regression analysis are used in an investigation with the MACCS model of the early health effects associated with a severe accident at a nuclear power station. The primary purpose of this study is to provide guidance on the variables to be considered in future review work to reduce the uncertainty in the important variables used in the calculation of reactor accident consequences. The effects of 34 imprecisely known input variables on the following reactor accident consequences are studied: number of early fatalities, number of cases of prodromal vomiting, population dose within 10 mi of the reactor, population dose within 1000 mi of the reactor, individual early fatality probability within 1 mi of the reactor, and maximum early fatality distance. When the predicted variables are considered collectively, the following input variables were found to be the dominant contributors to uncertainty: scaling factor for horizontal dispersion, dry deposition velocity, inhalation protection factor for nonevacuees, groundshine shielding factor for nonevacuees, early fatality hazard function alpha value for bone marrow exposure, and scaling factor for vertical dispersion.

  14. NUCLEAR REACTOR CORE DESIGN

    DOE Patents [OSTI]

    Mahlmeister, J.E.; Peck, W.S.; Haberer, W.V.; Williams, A.C.

    1960-03-22

    An improved core design for a sodium-cooled, graphitemoderated nuclear reactor is described. The improved reactor core comprises a number of blocks of moderator material, each block being in the shape of a regular prism. A number of channels, extending the length of each block, are disposed around the periphery. When several blocks are placed in contact to form the reactor core, the channels in adjacent blocks correspond with each other to form closed conduits extending the length of the core. Fuel element clusters are disposed in these closed conduits, and liquid coolant is forced through the annulus between the fuel cluster and the inner surface of the conduit. In a preferred embodiment of the invention, the moderator blocks are in the form of hexagonal prisms with longitudinal channels cut into the corners of the hexagon. The main advantage of an "edge-loaded" moderator block is that fewer thermal neutrons are absorbed by the moderator cladding, as compared with a conventional centrally loaded moderator block.

  15. Cooling system for electronic components

    DOE Patents [OSTI]

    Anderl, William James; Colgan, Evan George; Gerken, James Dorance; Marroquin, Christopher Michael; Tian, Shurong

    2015-12-15

    Embodiments of the present invention provide for non interruptive fluid cooling of an electronic enclosure. One or more electronic component packages may be removable from a circuit card having a fluid flow system. When installed, the electronic component packages are coincident to and in a thermal relationship with the fluid flow system. If a particular electronic component package becomes non-functional, it may be removed from the electronic enclosure without affecting either the fluid flow system or other neighboring electronic component packages.

  16. Containment at the Source during Waste Volume Reduction of Large Radioactive Components Using Oxylance High-Temperature Cutting Equipment - 13595

    SciTech Connect (OSTI)

    Keeney, G. Neil

    2013-07-01

    As a waste-volume reduction and management technique, highly contaminated Control Element Drive Mechanism (CEDM) housings were severed from the Reactor Pressure Vessel Head (RPVH) inside the San Onofre Unit 2 primary containment utilizing Oxylance high-temperature cutting equipment and techniques. Presented are relevant data concerning: - Radiological profiles of the RPVH and individual CEDMs; - Design overviews of the engineering controls and the specialized confinement housings; - Utilization of specialized shielding; - Observations of apparent metallurgical-contamination coalescence phenomena at high temperatures resulting in positive control over loose-surface contamination conditions; - General results of radiological and industrial hygiene air sampling and monitoring; - Collective dose and personnel contamination event statistics; - Lessons learned. (author)

  17. Independent Confirmatory Survey Report for the University of Arizona Nuclear Reactor Laboratory, Tucson, Arizona

    SciTech Connect (OSTI)

    Nick A. Altic

    2011-11-11

    The University of Arizona (University) research reactor is a TRIGA swimming pool type reactor designed by General Atomics and constructed at the University in 1958. The reactor first went into operation in December of 1958 under U.S. Nuclear Regulatory Commission (NRC) license R-52 until final shut down on May 18, 2010. Initial site characterization activities were conducted in February 2009 during ongoing reactor operations to assess the radiological status of the Nuclear Reactor Laboratory (NRL) excluding the reactor tank, associated components, and operating systems. Additional post-shutdown characterization activities were performed to complete characterization activities as well as verify assumptions made in the Decommissioning Plan (DP) that were based on a separate activation analysis (ESI 2009 and WMG 2009). Final status survey (FSS) activities began shortly after the issuance of the FSS plan in May 2011. The contractor completed measurement and sampling activities during the week of August 29, 2011.

  18. Comparative assessment of nuclear fuel cycles. Light-water reactor once-through, classical fast breeder reactor, and symbiotic fast breeder reactor cycles

    SciTech Connect (OSTI)

    Hardie, R.W.; Barrett, R.J.; Freiwald, J.G.

    1980-06-01

    The object of the Alternative Nuclear Fuel Cycle Study is to perform comparative assessments of nuclear power systems. There are two important features of this study. First, this evaluation attempts to encompass the complete, integrated fuel cycle from mining of uranium ore to disposal of waste rather than isolated components. Second, it compares several aspects of each cycle - energy use, economics, technological status, proliferation, public safety, and commercial potential - instead of concentrating on one or two assessment areas. This report presents assessment results for three fuel cycles. These are the light-water reactor once-through cycle, the fast breeder reactor on the classical plutonium cycle, and the fast breeder reactor on a symbiotic cycle using plutonium and /sup 233/U as fissile fuels. The report also contains a description of the methodology used in this assessment. Subsequent reports will present results for additional fuel cycles.

  19. Thermionic Reactor Design Studies

    SciTech Connect (OSTI)

    Schock, Alfred

    1994-06-01

    During the 1960's and early 70's the author performed extensive design studies, analyses, and tests aimed at thermionic reactor concepts that differed significantly from those pursued by other investigators. Those studies, like most others under Atomic Energy Commission (AEC and DOE) and the National Aeronautics and Space Administration (NASA) sponsorship, were terminated in the early 1970's. Some of this work was previously published, but much of it was never made available in the open literature. U.S. interest in thermionic reactors resumed in the early 80's, and was greatly intensified by reports about Soviet ground and flight tests in the late 80's. This recent interest resulted in renewed U.S. thermionic reactor development programs, primarily under Department of Defense (DOD) and Department of Energy (DOE) sponsorship. Since most current investigators have not had an opportunity to study all of the author's previous work, a review of the highlights of that work may be of value to them. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling. Where the author's concepts differed from the later Topaz-2 design was in the relative location of the emitter and the collector. Placing the fueled emitter on the outside of the cylindrical diodes permits much higher axial conductances to reduce ohmic losses in the electrodes of full-core-height diodes. Moreover, placing the fuel on the outside of the diode makes possible reactors with much higher fuel volume fractions, which enable power-flattened fast reactors scalable to very low power levels without the need for life-limiting hydride moderators or the use of efficiency-limiting driver fuel. In addition, with the fuel on the outside its swelling does not increase the emitter diameter or reduce the interelectrode gap. This should permit long lifetimes even with closer spacings, which can significantly improve the system efficiences. This was confirmed by coupled neutronic, thermal, thermionic, and electrical system analyses - some of which are presented in this paper - and by subsequent experiments. A companion paper presented next describes the fabrication and testing of full-scale converter elements, both fueled and unfueled, and summarizes the test results obtained. There is a duplicate copy in the file.

  20. Advanced Reactor Technology Documents | Department of Energy

    Energy Savers [EERE]

    Nuclear Reactor Technologies » Advanced Reactor Technologies » Advanced Reactor Technology Documents Advanced Reactor Technology Documents January 30, 2013 Advanced Reactor Concepts Technical Review Panel Report This report documents the establishment of a technical review process and the findings of the Advanced Reactor Concepts (ARC) Technical Review Panel (TRP).1 The intent of the process is to identify R&D needs for viable advanced reactor concepts in order to inform DOE-NE R&D

  1. F Reactor Inspection | Department of Energy

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

    F Reactor Inspection F Reactor Inspection Addthis Description Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor last week before stepping inside the reactor to complete its periodic inspection. This is the first time the Department of Energy (DOE) has had the reactor open since 2008. The F Reactor is one of nine reactors along the Columbia River at the Department's Hanford Site in southeastern Washington State, where environmental cleanup has

  2. Emulation of reactor irradiation damage using ion beams

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Was, G. S.; Jiao, Z.; Getto, E.; Sun, K.; Monterrosa, A. M.; Maloy, S. A.; Anderoglu, O.; Sencer, B. H.; Hackett, M.

    2014-06-14

    The continued operation of existing light water nuclear reactors and the development of advanced nuclear reactor depend heavily on understanding how damage by radiation to levels degrades materials that serve as the structural components in reactor cores. The first high dose ion irradiation experiments on a ferritic-martensitic steel showing that ion irradiation closely emulates the full radiation damage microstructure created in-reactor are described. Ferritic-martensitic alloy HT9 (heat 84425) in the form of a hexagonal fuel bundle duct (ACO-3) accumulated 155 dpa at an average temperature of 443°C in the Fast Flux Test Facility (FFTF). Using invariance theory as a guide,more » irradiation of the same heat was conducted using self-ions (Fe++) at 5 MeV at a temperature of 460°C and to a dose of 188 displacements per atom. The void swelling was nearly identical between the two irradiation and the size and density of precipitates and loops following ion irradiation are within a factor of two of those for neutron irradiation. The level of agreement across all of the principal microstructure changes between ion and reactor irradiation establishes the capability of tailoring ion irradiation to emulate the reactor-irradiated microstructure.« less

  3. Emulation of reactor irradiation damage using ion beams

    SciTech Connect (OSTI)

    Was, G. S.; Jiao, Z.; Getto, E.; Sun, K.; Monterrosa, A. M.; Maloy, S. A.; Anderoglu, O.; Sencer, B. H.; Hackett, M.

    2014-06-14

    The continued operation of existing light water nuclear reactors and the development of advanced nuclear reactor depend heavily on understanding how damage by radiation to levels degrades materials that serve as the structural components in reactor cores. The first high dose ion irradiation experiments on a ferritic-martensitic steel showing that ion irradiation closely emulates the full radiation damage microstructure created in-reactor are described. Ferritic-martensitic alloy HT9 (heat 84425) in the form of a hexagonal fuel bundle duct (ACO-3) accumulated 155 dpa at an average temperature of 443C in the Fast Flux Test Facility (FFTF). Using invariance theory as a guide, irradiation of the same heat was conducted using self-ions (Fe++) at 5 MeV at a temperature of 460C and to a dose of 188 displacements per atom. The void swelling was nearly identical between the two irradiation and the size and density of precipitates and loops following ion irradiation are within a factor of two of those for neutron irradiation. The level of agreement across all of the principal microstructure changes between ion and reactor irradiation establishes the capability of tailoring ion irradiation to emulate the reactor-irradiated microstructure.

  4. Emulation of reactor irradiation damage using ion beams

    SciTech Connect (OSTI)

    G. S. Was; Z. Jiao; E. Beckett; A. M. Monterrosa; O. Anderoglu; B. H. Sencer; M. Hackett

    2014-10-01

    The continued operation of existing light water nuclear reactors and the development of advanced nuclear reactor depend heavily on understanding how damage by radiation to levels degrades materials that serve as the structural components in reactor cores. The first high dose ion irradiation experiments on a ferritic-martensitic steel showing that ion irradiation closely emulates the full radiation damage microstructure created in-reactor are described. Ferritic-martensitic alloy HT9 (heat 84425) in the form of a hexagonal fuel bundle duct (ACO-3) accumulated 155 dpa at an average temperature of 443C in the Fast Flux Test Facility (FFTF). Using invariance theory as a guide, irradiation of the same heat was conducted using self-ions (Fe++) at 5 MeV at a temperature of 460C and to a dose of 188 displacements per atom. The void swelling was nearly identical between the two irradiations and the size and density of precipitates and loops following ion irradiation are within a factor of two of those for neutron irradiation. The level of agreement across all of the principal microstructure changes between ion and reactor irradiations establishes the capability of tailoring ion irradiations to emulate the reactor-irradiated microstructure.

  5. Extended Sleeve Products Allow Control and Monitoring of Process Fluid Flows Inside Shielding, Behind Walls and Beneath Floors - 13041

    SciTech Connect (OSTI)

    Abbott, Mark W.

    2013-07-01

    Throughout power generation, delivery and waste remediation, the ability to control process streams in difficult or impossible locations becomes increasingly necessary as the complexity of processes increases. Example applications include radioactive environments, inside concrete installations, buried in dirt, or inside a shielded or insulated pipe. In these situations, it is necessary to implement innovative solutions to tackle such issues as valve maintenance, valve control from remote locations, equipment cleaning in hazardous environments, and flow stream analysis. The Extended Sleeve family of products provides a scalable solution to tackle some of the most challenging applications in hazardous environments which require flow stream control and monitoring. The Extended Sleeve family of products is defined in three groups: Extended Sleeve (ESV), Extended Bonnet (EBV) and Instrument Enclosure (IE). Each of the products provides a variation on the same requirements: to provide access to the internals of a valve, or to monitor the fluid passing through the pipeline through shielding around the process pipe. The shielding can be as simple as a grout filled pipe covering a process pipe or as complex as a concrete deck protecting a room in which the valves and pipes pass through at varying elevations. Extended Sleeves are available between roughly 30 inches and 18 feet of distance between the pipeline centerline and the top of the surface to which it mounts. The Extended Sleeve provides features such as 1.5 inches of adjustment between the pipeline and deck location, internal flush capabilities, automatic alignment of the internal components during assembly and integrated actuator mounting pads. The Extended Bonnet is a shorter fixed height version of the Extended Sleeve which has a removable deck flange to facilitate installation through walls, and is delivered fully assembled. The Instrument Enclosure utilizes many of the same components as an Extended Sleeve, yet allows the installation of process monitoring instruments, such as a turbidity meter to be placed in the flow stream. The basis of the design is a valve body, which, rather than having a directly mounted bonnet has lengths of concentric pipe added, which move the bonnet away from the valve body. The pipe is conceptually similar to an oil field well, with the various strings of casing, and tubing installed. Each concentric pipe provides a required function, such as the outermost pipes, the valve sleeve and penetration sleeve, which provide structural support to the deck flange. For plug valve based designs, the next inner pipe provides compression on the environmental seals at the top of the body to bonnet joint, followed by the innermost pipe which provides rotation of the plug, in the same manner as an extended stem. Ball valve ESVs have an additional pipe to provide compressive loading on the stem packing. Due to the availability of standard pipe grades and weights, the product can be configured to fit a wide array of valve sizes, and application lengths, with current designs as short as seven inches and as tall as 18 feet. Central to the design is the requirement for no special tools or downhole tools to remove parts or configure the product. Off the shelf wrenches, sockets or other hand tools are all that is required. Compared to other products historically available, this design offers a lightweight option, which, while not as rigidly stiff, can deflect compliantly under extreme seismic loading, rather than break. Application conditions vary widely, as the base product is 316 and 304 stainless steel, but utilizes 17-4PH, and other allows as needed based on the temperature range and mechanical requirements. Existing designs are installed in applications as hot as 1400 deg. F, at low pressure, and separately in highly radioactive environments. The selection of plug versus ball valve, metal versus soft seats, and the material of the seals and seats is all dependent on the application requirements. The design of the Extended Sleeve family of products provid

  6. Radiation Damage In Reactor Cavity Concrete

    SciTech Connect (OSTI)

    Field, Kevin G; Le Pape, Yann; Naus, Dan J; Remec, Igor; Busby, Jeremy T; Rosseel, Thomas M; Wall, Dr. James Joseph

    2015-01-01

    License renewal up to 60 years and the possibility of subsequent license renewal to 80 years has established a renewed focus on long-term aging of nuclear generating stations materials, and recently, on concrete. Large irreplaceable sections of most nuclear generating stations include concrete. The Expanded Materials Degradation Analysis (EMDA), jointly performed by the Department of Energy, the Nuclear Regulatory Commission and Industry, identified the urgent need to develop a consistent knowledge base on irradiation effects in concrete. Much of the historical mechanical performance data of irradiated concrete does not accurately reflect typical radiation conditions in NPPs or conditions out to 60 or 80 years of radiation exposure. To address these potential gaps in the knowledge base, The Electric Power Research Institute and Oak Ridge National Laboratory are working to disposition radiation damage as a degradation mechanism. This paper outlines the research program within this pathway including: (i) defining the upper bound of the neutron and gamma dose levels expected in the biological shield concrete for extended operation (80 years of operation and beyond), (ii) determining the effects of neutron and gamma irradiation as well as extended time at temperature on concrete, (iii) evaluating opportunities to irradiate prototypical concrete under accelerated neutron and gamma dose levels to establish a conservative bound and share data obtained from different flux, temperature, and fluence levels, (iv) evaluating opportunities to harvest and test irradiated concrete from international NPPs, (v) developing cooperative test programs to improve confidence in the results from the various concretes and research reactors, (vi) furthering the understanding of the effects of radiation on concrete (see companion paper) and (vii) establishing an international collaborative research and information exchange effort to leverage capabilities and knowledge.

  7. Above-ground Antineutrino Detection for Nuclear Reactor Monitoring

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Sweany, Melinda; Brennan, James S.; Cabrera-Palmer, Belkis; Kiff, Scott D.; Reyna, David; Throckmorton, Daniel J.

    2014-08-01

    Antineutrino monitoring of nuclear reactors has been demonstrated many times, however the technique has not as of yet been developed into a useful capability for treaty verification purposes. The most notable drawback is the current requirement that detectors be deployed underground, with at least several meters-water-equivalent of shielding from cosmic radiation. In addition, the deployment of liquid-based detector media presents a challenge in reactor facilities. We are currently developing a detector system that has the potential to operate above ground and circumvent deployment problems associated with a liquid detection media: the system is composed of segments of plastic scintillator surroundedmore » by 6LiF/ZnS:Ag. ZnS:Ag is a radio-luminescent phosphor used to detect the neutron capture products of lithium-6. Because of its long decay time compared to standard plastic scintillators, pulse-shape discrimination can be used to distinguish positron and neutron interactions resulting from the inverse beta decay (IBD) of antineutrinos within the detector volume, reducing both accidental and correlated backgrounds. Segmentation further reduces backgrounds by identifying the positron’s annihilation gammas, which are absent for most correlated and uncorrelated backgrounds. This work explores different configurations in order to maximize the size of the detector segments without reducing the intrinsic neutron detection efficiency. We believe this technology will ultimately be applicable to potential safeguards scenarios such as those recently described.« less

  8. Pressurized-water reactor internals aging degradation study. Phase 1

    SciTech Connect (OSTI)

    Luk, K.H.

    1993-09-01

    This report documents the results of a Phase I study on the effects of aging degradations on pr internals. Primary stressers for internals an generated by the primary coolant flow in the they include unsteady hydrodynamic forces and pump-generated pressure pulsations. Other stressors are applied loads, manufacturing processes, impurities in the coolant and exposures to fast neutron fluxes. A survey of reported aging-related failure information indicates that fatigue, stress corrosion cracking (SCC) and mechanical wear are the three major aging-related degradation mechanisms for PWR internals. Significant reported failures include thermal shield flow-induced vibration problems, SCC in guide tube support pins and core support structure bolts, fatigue-induced core baffle water-jet impingement problems and excess wear in flux thimbles. Many of the reported problems have been resolved by accepted engineering practices. Uncertainties remain in the assessment of long-term neutron irradiation effects and environmental factors in high-cycle fatigue failures. Reactor internals are examined by visual inspections and the technique is access limited. Improved inspection methods, especially one with an early failure detection capability, can enhance the safety and efficiency of reactor operations.

  9. Above-ground Antineutrino Detection for Nuclear Reactor Monitoring

    SciTech Connect (OSTI)

    Sweany, Melinda; Brennan, James S.; Cabrera-Palmer, Belkis; Kiff, Scott D.; Reyna, David; Throckmorton, Daniel J.

    2014-08-01

    Antineutrino monitoring of nuclear reactors has been demonstrated many times, however the technique has not as of yet been developed into a useful capability for treaty verification purposes. The most notable drawback is the current requirement that detectors be deployed underground, with at least several meters-water-equivalent of shielding from cosmic radiation. In addition, the deployment of liquid-based detector media presents a challenge in reactor facilities. We are currently developing a detector system that has the potential to operate above ground and circumvent deployment problems associated with a liquid detection media: the system is composed of segments of plastic scintillator surrounded by 6LiF/ZnS:Ag. ZnS:Ag is a radio-luminescent phosphor used to detect the neutron capture products of lithium-6. Because of its long decay time compared to standard plastic scintillators, pulse-shape discrimination can be used to distinguish positron and neutron interactions resulting from the inverse beta decay (IBD) of antineutrinos within the detector volume, reducing both accidental and correlated backgrounds. Segmentation further reduces backgrounds by identifying the positrons annihilation gammas, which are absent for most correlated and uncorrelated backgrounds. This work explores different configurations in order to maximize the size of the detector segments without reducing the intrinsic neutron detection efficiency. We believe this technology will ultimately be applicable to potential safeguards scenarios such as those recently described.

  10. System Study: Reactor Core Isolation Cooling 19982012

    SciTech Connect (OSTI)

    T. E. Wierman

    2013-10-01

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2012 for selected components were obtained from the Equipment Performance and Information Exchange (EPIX). The unreliability results are trended for the most recent 10 year period while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing trend was identified in the HPCI results. Statistically significant decreasing trends were identified for RCIC start-only and 8-hour trends.

  11. The modular high-temperature gas-cooled reactor (MHTGR)

    SciTech Connect (OSTI)

    Neylan, A.J.

    1986-10-01

    The MHTGR is an advanced reactor concept being developed in the USA under a cooperative program involving the US Government, the nuclear industry and the utilities. The design utilizes basic HTGR features of ceramic fuel, helium coolant and a graphite moderator. However the specific size and configuration are selected to utilize the inherently safe characteristics associated with these standard features coupled with passive safety systems to provide a significantly higher margin of safety and investment protection than current generation reactors. Evacuation or sheltering of the public is not required. The major components of the nuclear steam supply, with special emphasis on the core, are described. Safety assessments of the concept are discussed.

  12. NON-NRC FUNDED RELAP5-3D VERSION 4.x.x SOFTWARE REACTOR EXCURSION AND LEAK ANALYSIS PACKAGE - THREE DIMENSIONAL

    SciTech Connect (OSTI)

    2012-03-26

    The RELAP5-3D Version 3.x code has been developed for best-estimate transient simulation of nuclear reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents and operational transients such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal hydraulic systems including pressurized water reactors, boiling water reactors, Soviet-designed reactors, heavy water reactors, gas-cooled reactors, liquid metal and molten salt cooled reactors, and even fusion reactors. Numerical models include multi-dimensional hydrodynamics, 1- and 2-D heat transfer in metal walls, 0-, 1-, 2-, and 3-D neutron kinetics, trips, and control systems. Secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems.

  13. NON-NRC FUNDED RELAP5-3D VERSION 4.x.x SOFTWARE REACTOR EXCURSION AND LEAK ANALYSIS PACKAGE - THREE DIMENSIONAL

    Energy Science and Technology Software Center (OSTI)

    2012-03-26

    The RELAP5-3D Version 3.x code has been developed for best-estimate transient simulation of nuclear reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents and operational transients such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal hydraulic systems including pressurized watermore » reactors, boiling water reactors, Soviet-designed reactors, heavy water reactors, gas-cooled reactors, liquid metal and molten salt cooled reactors, and even fusion reactors. Numerical models include multi-dimensional hydrodynamics, 1- and 2-D heat transfer in metal walls, 0-, 1-, 2-, and 3-D neutron kinetics, trips, and control systems. Secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems.« less

  14. Continuous Energy, Multi-Dimensional Transport Calculations for Problem Dependent Resonance Self-Shielding

    SciTech Connect (OSTI)

    T. Downar

    2009-03-31

    The overall objective of the work here has been to eliminate the approximations used in current resonance treatments by developing continuous energy multi-dimensional transport calculations for problem dependent self-shielding calculations. The work here builds on the existing resonance treatment capabilities in the ORNL SCALE code system.

  15. Neutron and gamma radiation shielding material, structure, and process of making structure

    DOE Patents [OSTI]

    Hondorp, Hugh L. (Princeton Junction, NJ)

    1984-01-01

    The present invention is directed to a novel neutron and gamma radiation elding material consisting of 95 to 97 percent by weight SiO.sub.2 and 5 to 3 percent by weight sodium silicate. In addition, the method of using this composition to provide a continuous neutron and gamma radiation shielding structure is disclosed.

  16. Shielding gas selection for increased weld penetration and productivity in GTA welding

    SciTech Connect (OSTI)

    Leinonen, J.I.

    1996-12-31

    The effects of hydrogen and helium additions to the argon shielding gas on GTA weld pool profiles in the case of two austenitic stainless steel sheets 3 mm thick are investigated here in detail. One of the test steels shows good weldability, with a relatively deep, narrow weld pool profile, but the other is poorly weldable, with a shallow, wide weld pool when argon shielding gas is used. Bead-on-plate test welds were produced with arc shields of argon, argon with hydrogen additions of 2 to 18.2% and argon with helium additions of 20 to 80%. The hydrogen additions increases the depth of weld penetration in both test steels, but productivity with respect to maximum welding speed can be improved to an accepted level only with steel sheets of good weldability in terms of a relatively high depth/width (D/W) ratio. The depth of penetration in the test steel of good weldability increased somewhat with helium additions and the D/W ratio remained unchanged, while these parameters increased markedly in the poorly weldable steel when a He-20% Ar shielding gas was used and resembled those of the more weldable steel.

  17. Shielded metal arc welding of AISI 4340 alloy steel. Welding procedure specification

    SciTech Connect (OSTI)

    Wodtke, C.H.; Frizzell, D.R.; Plunkett, W.A.

    1985-08-01

    Procedure WPS-115 is qualified under Section IX of the ASME Boiler and Pressure Vessel Code for shielded metal arc welding of AISI 4340 alloy steel, in thickness range 0.187 to 2 inch; filler metal is E7018 (F-4, A-1).

  18. Fast quench reactor and method

    DOE Patents [OSTI]

    Detering, B.A.; Donaldson, A.D.; Fincke, J.R.; Kong, P.C.

    1998-05-12

    A fast quench reactor includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This ``freezes`` the desired end product(s) in the heated equilibrium reaction stage. 7 figs.

  19. Massive Hanford Test Reactor Removed - Plutonium Recycle Test...

    Office of Environmental Management (EM)

    Massive Hanford Test Reactor Removed - Plutonium Recycle Test Reactor removed from Hanford's 300 Area Massive Hanford Test Reactor Removed - Plutonium Recycle Test Reactor removed ...

  20. Small Modular Nuclear Reactors: Parametric Modeling of Integrated...

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

    1 Small Modular Nuclear Reactors: Parametric Modeling of Integrated Reactor Vessel ... PDF icon Small Modular Nuclear Reactors: Parametric Modeling of Integrated Reactor Vessel ...

  1. Secondary Startup Neutron Sources as a Source of Tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS)

    SciTech Connect (OSTI)

    Shaver, Mark W.; Lanning, Donald D.

    2010-02-01

    The hypothesis of this paper is that the Zircaloy clad fuel source is minimal and that secondary startup neutron sources are the significant contributors of the tritium in the RCS that was previously assigned to release from fuel. Currently there are large uncertainties in the attribution of tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS). The measured amount of tritium in the coolant cannot be separated out empirically into its individual sources. Therefore, to quantify individual contributors, all sources of tritium in the RCS of a PWR must be understood theoretically and verified by the sum of the individual components equaling the measured values.

  2. Progress Update: Reactor Disassembly Grouting

    ScienceCinema (OSTI)

    Cody, Tom

    2012-06-14

    Grouting the P&R reactors in order to remove these basins as an environmental threat. This will end the Cold War legacy and end the environmental footprint.

  3. Neutrino oscillation studies with reactors

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Vogel, P.; Wen, L.J.; Zhang, C.

    2015-04-27

    Nuclear reactors are one of the most intense, pure, controllable, cost-effective and well-understood sources of neutrinos. Reactors have played a major role in the study of neutrino oscillations, a phenomenon that indicates that neutrinos have mass and that neutrino flavours are quantum mechanical mixtures. Over the past several decades, reactors were used in the discovery of neutrinos, were crucial in solving the solar neutrino puzzle, and allowed the determination of the smallest mixing angle θ13. In the near future, reactors will help to determine the neutrino mass hierarchy and to solve the puzzling issue of sterile neutrinos.

  4. Daya Bay Reactor Neutrino Experiment

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Ao Nuclear Power Plant reactors. The experiment is being built by blasting three kilometers of tunnel through the granite rock under the mountains where the power plants are...

  5. Reactor Materials Newsletter- Issue 1

    Broader source: Energy.gov [DOE]

    The Reactor Materials (RM) newsletter includes information about key nuclear materials programs, results from ongoing projects across the Office of Nuclear Energy, and other relevant information.

  6. Evaluation of a Method for Remote Detection of Fuel Relocation Outside the Original Core Volumes of Fukushima Reactor Units 1-3

    SciTech Connect (OSTI)

    Douglas W. Akers; Edwin A. Harvego

    2012-08-01

    This paper presents the results of a study to evaluate the feasibility of remotely detecting and quantifying fuel relocation from the core to the lower head, and to regions outside the reactor vessel primary containment of the Fukushima 1-3 reactors. The goals of this study were to determine measurement conditions and requirements, and to perform initial radiation transport sensitivity analyses for several potential measurement locations inside the reactor building. The radiation transport sensitivity analyses were performed based on reactor design information for boiling water reactors (BWRs) similar to the Fukushima reactors, ORIGEN2 analyses of 3-cycle BWR fuel inventories, and data on previously molten fuel characteristics from TMI- 2. A 100 kg mass of previously molten fuel material located on the lower head of the reactor vessel was chosen as a fuel interrogation sensitivity target. Two measurement locations were chosen for the transport analyses, one inside the drywell and one outside the concrete biological shield surrounding the drywell. Results of these initial radiation transport analyses indicate that the 100 kg of previously molten fuel material may be detectable at the measurement location inside the drywell, but that it is highly unlikely that any amount of fuel material inside the RPV will be detectable from a location outside the concrete biological shield surrounding the drywell. Three additional fuel relocation scenarios were also analyzed to assess detection sensitivity for varying amount of relocated material in the lower head of the reactor vessel, in the control rods perpendicular to the detector system, and on the lower head of the drywell. Results of these analyses along with an assessment of background radiation effects and a discussion of measurement issues, such as the detector/collimator design, are included in the paper.

  7. Biparticle fluidized bed reactor

    DOE Patents [OSTI]

    Scott, C.D.

    1993-12-14

    A fluidized bed reactor system which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase is described. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves. 3 figures.

  8. FAST NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Snell, A.H.

    1957-12-01

    This patent relates to a reactor and process for carrying out a controlled fast neutron chain reaction. A cubical reactive mass, weighing at least 920 metric tons, of uranium metal containing predominantly U/sup 238/ and having a U/sup 235/ content of at least 7.63% is assembled and the maximum neutron reproduction ratio is limited to not substantially over 1.01 by insertion and removal of a varying amount of boron, the reactive mass being substantially freed of moderator.

  9. Biparticle fluidized bed reactor

    DOE Patents [OSTI]

    Scott, C.D.; Marasco, J.A.

    1996-02-27

    A fluidized bed reactor system is described which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary and tertiary particulate phases, continuously introduced and removed simultaneously in the cocurrent and countercurrent mode, act in a role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Means for introducing and removing the sorbent phases include feed screw mechanisms and multivane slurry valves. 3 figs.

  10. Reactor refueling containment system

    DOE Patents [OSTI]

    Gillett, J.E.; Meuschke, R.E.

    1995-05-02

    A method of refueling a nuclear reactor is disclosed whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced. 2 figs.

  11. Biparticle fluidized bed reactor

    DOE Patents [OSTI]

    Scott, C.D.; Marasco, J.A.

    1995-04-25

    A fluidized bed reactor system utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves. 3 figs.

  12. Biparticle fluidized bed reactor

    DOE Patents [OSTI]

    Scott, Charles D. (Oak Ridge, TN); Marasco, Joseph A. (Kingston, TN)

    1996-01-01

    A fluidized bed reactor system which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary and tertiary particulate phases, continuously introduced and removed simultaneously in the cocurrent and countercurrent mode, act in a role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Means for introducing and removing the sorbent phases include feed screw mechanisms and multivane slurry valves.

  13. Reactor refueling containment system

    DOE Patents [OSTI]

    Gillett, James E. (Greensburg, PA); Meuschke, Robert E. (Pittsburgh, PA)

    1995-01-01

    A method of refueling a nuclear reactor whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced.

  14. Nuclear reactor control apparatus

    DOE Patents [OSTI]

    Sridhar, Bettadapur N. (Cupertino, CA)

    1983-11-01

    Nuclear reactor core safety rod release apparatus comprises a control rod having a detent notch in the form of an annular peripheral recess at its upper end, a control rod support tube for raising and lowering the control rod under normal conditions, latches pivotally mounted on the control support tube with free ends thereof normally disposed in the recess in the control rod, and cam means for pivoting the latches out of the recess in the control rod when a scram condition occurs. One embodiment of the invention comprises an additional magnetically-operated latch for releasing the control rod under two different conditions, one involving seismic shock.

  15. Nuclear reactor fuel element

    DOE Patents [OSTI]

    Johnson, Carl E. (Elk Grove, IL); Crouthamel, Carl E. (Richland, WA)

    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.

  16. Reactor coolant pump flywheel

    DOE Patents [OSTI]

    Finegan, John Raymond; Kreke, Francis Joseph; Casamassa, John Joseph

    2013-11-26

    A flywheel for a pump, and in particular a flywheel having a number of high density segments for use in a nuclear reactor coolant pump. The flywheel includes an inner member and an outer member. A number of high density segments are provided between the inner and outer members. The high density segments may be formed from a tungsten based alloy. A preselected gap is provided between each of the number of high density segments. The gap accommodates thermal expansion of each of the number of segments and resists the hoop stress effect/keystoning of the segments.

  17. High flux reactor

    DOE Patents [OSTI]

    Lake, James A.; Heath, Russell L.; Liebenthal, John L.; DeBoisblanc, Deslonde R.; Leyse, Carl F.; Parsons, Kent; Ryskamp, John M.; Wadkins, Robert P.; Harker, Yale D.; Fillmore, Gary N.; Oh, Chang H.

    1988-01-01

    A high flux reactor is comprised of a core which is divided into two symetric segments housed in a pressure vessel. The core segments include at least one radial fuel plate. The spacing between the plates functions as a coolant flow channel. The core segments are spaced axially apart such that a coolant mixing plenum is formed between them. A channel is provided such that a portion of the coolant bypasses the first core section and goes directly into the mixing plenum. The outlet coolant from the first core segment is mixed with the bypass coolant resulting in a lower inlet temperature to the lower core segment.

  18. Biparticle fluidized bed reactor

    DOE Patents [OSTI]

    Scott, Charles D. (Oak Ridge, TN); Marasco, Joseph A. (Kingston, TN)

    1995-01-01

    A fluidized bed reactor system utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves.

  19. Biparticle fluidized bed reactor

    DOE Patents [OSTI]

    Scott, Charles D. (Oak Ridge, TN)

    1993-01-01

    A fluidized bed reactor system which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves.

  20. In-Reactor Experiment

    Office of Environmental Management (EM)

    Update on the TMIST-3 In-Reactor Experiment Tritium Release and Speciation from LiAlO 2 and LiAlO 2 /Zr Cermets DJ SENOR 1 , WG LUSCHER 1 , AND KK CLAYTON 2 1 Pacific Northwest National Laboratory, 2 Idaho National Laboratory Tritium Focus Group Meeting, Princeton, NJ 5 May 2015 1 PNNL-SA-109847 Tritium Production Enterprise: Background Tritium is required for US nuclear weapons stockpile Tritium has a 12.3 year half-life and must be replenished 1988: DOE ceased production of tritium at SRS

  1. Enabling Technologies for Ceramic Hot Section Components

    SciTech Connect (OSTI)

    Venkat Vedula; Tania Bhatia

    2009-04-30

    Silicon-based ceramics are attractive materials for use in gas turbine engine hot sections due to their high temperature mechanical and physical properties as well as lower density than metals. The advantages of utilizing ceramic hot section components include weight reduction, and improved efficiency as well as enhanced power output and lower emissions as a result of reducing or eliminating cooling. Potential gas turbine ceramic components for industrial, commercial and/or military high temperature turbine applications include combustor liners, vanes, rotors, and shrouds. These components require materials that can withstand high temperatures and pressures for long duration under steam-rich environments. For Navy applications, ceramic hot section components have the potential to increase the operation range. The amount of weight reduced by utilizing a lighter gas turbine can be used to increase fuel storage capacity while a more efficient gas turbine consumes less fuel. Both improvements enable a longer operation range for Navy ships and aircraft. Ceramic hot section components will also be beneficial to the Navy's Growth Joint Strike Fighter (JSF) and VAATE (Versatile Affordable Advanced Turbine Engines) initiatives in terms of reduced weight, cooling air savings, and capability/cost index (CCI). For DOE applications, ceramic hot section components provide an avenue to achieve low emissions while improving efficiency. Combustors made of ceramic material can withstand higher wall temperatures and require less cooling air. Ability of the ceramics to withstand high temperatures enables novel combustor designs that have reduced NO{sub x}, smoke and CO levels. In the turbine section, ceramic vanes and blades do not require sophisticated cooling schemes currently used for metal components. The saved cooling air could be used to further improve efficiency and power output. The objectives of this contract were to develop technologies critical for ceramic hot section components for gas turbine engines. Significant technical progress has been made towards maturation of the EBC and CMC technologies for incorporation into gas turbine engine hot-section. Promising EBC candidates for longer life and/or higher temperature applications relative to current state of the art BSAS-based EBCs have been identified. These next generation coating systems have been scaled-up from coupons to components and are currently being field tested in Solar Centaur 50S engine. CMC combustor liners were designed, fabricated and tested in a FT8 sector rig to demonstrate the benefits of a high temperature material system. Pretest predictions made through the use of perfectly stirred reactor models showed a 2-3x benefit in CO emissions for CMC versus metallic liners. The sector-rig test validated the pretest predictions with >2x benefit in CO at the same NOx levels at various load conditions. The CMC liners also survived several trip shut downs thereby validating the CMC design methodology. Significant technical progress has been made towards incorporation of ceramic matrix composites (CMC) and environmental barrier coatings (EBC) technologies into gas turbine engine hot-section. The second phase of the program focused on the demonstration of a reverse flow annular CMC combustor. This has included overcoming the challenges of design and fabrication of CMCs into 'complex' shapes; developing processing to apply EBCs to 'engine hardware'; testing of an advanced combustor enabled by CMCs in a PW206 rig; and the validation of performance benefits against a metal baseline. The rig test validated many of the pretest predictions with a 40-50% reduction in pattern factor compared to the baseline and reductions in NOx levels at maximum power conditions. The next steps are to develop an understanding of the life limiting mechanisms in EBC and CMC materials, developing a design system for EBC coated CMCs and durability testing in an engine environment.

  2. Advanced Nuclear Technology: Advanced Light Water Reactors Utility Requirements Document Small Modular Reactors Inclusion Summary

    Broader source: Energy.gov [DOE]

    Advanced Nuclear Technology: Advanced Light Water Reactors Utility Requirements Document Small Modular Reactors Inclusion Summary November 2014

  3. Foreign Research Reactor/Domestic Research Reactor Receipt Coordinator,

    National Nuclear Security Administration (NNSA)

    Savannah River Nuclear Solutions | National Nuclear Security Administration Foreign Research Reactor/Domestic Research Reactor Receipt Coordinator, Savannah River Nuclear Solutions | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Countering Nuclear Terrorism About Our Programs Our History Who We Are Our Leadership Our

  4. Multi-tipped optical component

    DOE Patents [OSTI]

    D'Urso, Brian R.; Simpson, John T.

    2010-04-13

    An optical component has a plurality of parallel noncontiguous optical conduits of at least one protrusive phase material embedded in a recessive phase material that acts as a support structure. The optical conduits extend from a proximal surface to a distal surface of the optical component. The distal surface has a plurality of spaced apart surface features of the protrusive phase material. Each independent optical conduits act as waveguides for a wavelength or range of wavelengths. The optical component can be formed such that the protruding surface features at the distal end of the component form an ordered array. An optical instrument can include the optical component in conjunction with a light source for illuminating a sample and a detector in optical communication optical component via the optical conduits.

  5. Multi-component assembly casting

    DOE Patents [OSTI]

    James, Allister W.

    2015-10-13

    Multi-component vane segment and method for forming the same. Assembly includes: positioning a pre-formed airfoil component (12) and a preformed shroud heat resistant material (18) in a mold, wherein the airfoil component (12) and the shroud heat resistant material (18) each comprises an interlocking feature (24); preheating the mold; introducing molten structural material (46) into the mold; and solidifying the molten structural material such that it interlocks the pre-formed airfoil component (12) with respect to the preformed shroud heat resistant material (18) and is effective to provide structural support for the shroud heat resistant material (18). Surfaces between the airfoil component (12) and the structural material (46), between the airfoil component (12) and the shroud heat resistant material (18), and between the shroud heat resistant material (18) and the structural material (46) are free of metallurgical bonds.

  6. Fast Reactor Fuel Type and Reactor Safety Performance

    SciTech Connect (OSTI)

    R. Wigeland; J. Cahalan

    2009-09-01

    Fast Reactor Fuel Type and Reactor Safety Performance R. Wigeland , Idaho National Laboratory J. Cahalan, Argonne National Laboratory The sodium-cooled fast neutron reactor is currently being evaluated for the efficient transmutation of the highly-hazardous, long-lived, transuranic elements that are present in spent nuclear fuel. One of the fundamental choices that will be made is the selection of the fuel type for the fast reactor, whether oxide, metal, carbide, nitride, etc. It is likely that a decision on the fuel type will need to be made before many of the related technologies and facilities can be selected, from fuel fabrication to spent fuel reprocessing. A decision on fuel type should consider all impacts on the fast reactor system, including safety. Past work has demonstrated that the choice of fuel type may have a significant impact on the severity of consequences arising from accidents, especially for severe accidents of low probability. In this paper, the response of sodium-cooled fast reactors is discussed for both oxide and metal fuel types, highlighting the similarities and differences in reactor response and accident consequences. Any fast reactor facility must be designed to be able to successfully prevent, mitigate, or accommodate all consequences of potential events, including accidents. This is typically accomplished by using multiple barriers to the release of radiation, including the cladding on the fuel, the intact primary cooling system, and most visibly the reactor containment building. More recently, this has also included the use of inherent safety concepts to reduce or eliminate the potential for serious damage in some cases. Past experience with oxide and metal fuel has demonstrated that both fuel types are suitable for use as fuel in a sodium-cooled fast reactor. However, safety analyses for these two fuel types have also shown that there can be substantial differences in accident consequences due to the neutronic and thermophysical properties of the fuel and their compatibility with the reactor coolant, with corresponding differences in the challenges presented to the reactor developers. Accident phenomena are discussed for the sodium-cooled fast reactor based on the mechanistic progression of conditions from accident initiation to accident termination, whether a benign state is achieved or more severe consequences are expected. General principles connecting accident phenomena and fuel properties are developed from the oxide and metal fuel safety analyses, providing guidelines that can be used as part of the evaluation for selection of fuel type for the sodium-cooled fast reactor.

  7. Reactor refueling machine simulator

    SciTech Connect (OSTI)

    Rohosky, T.L.; Swidwa, K.J.

    1987-10-13

    This patent describes in combination: a nuclear reactor; a refueling machine having a bridge, trolley and hoist each driven by a separate motor having feedback means for generating a feedback signal indicative of movement thereof. The motors are operable to position the refueling machine over the nuclear reactor for refueling the same. The refueling machine also has a removable control console including means for selectively generating separate motor signals for operating the bridge, trolley and hoist motors and for processing the feedback signals to generate an indication of the positions thereof, separate output leads connecting each of the motor signals to the respective refueling machine motor, and separate input leads for connecting each of the feedback means to the console; and a portable simulator unit comprising: a single simulator motor; a single simulator feedback signal generator connected to the simulator motor for generating a simulator feedback signal in response to operation of the simulator motor; means for selectively connecting the output leads of the console to the simulator unit in place of the refueling machine motors, and for connecting the console input leads to the simulator unit in place of the refueling machine motor feedback means; and means for driving the single simulator motor in response to any of the bridge, trolley or hoist motor signals generated by the console and means for applying the simulator feedback signal to the console input lead associated with the motor signal being generated by the control console.

  8. Burnup concept for a long-life fast reactor core using MCNPX.

    SciTech Connect (OSTI)

    Holschuh, Thomas Vernon,; Lewis, Tom Goslee,; Parma, Edward J.,

    2013-02-01

    This report describes a reactor design with a burnup concept for a long-life fast reactor core that was evaluated using Monte Carlo N-Particle eXtended (MCNPX). The current trend in advanced reactor design is the concept of a small modular reactor (SMR). However, very few of the SMR designs attempt to substantially increase the lifetime of a reactor core, especially without zone loading, fuel reshuffling, or other artificial mechanisms in the core that %E2%80%9Cflatten%E2%80%9D the power profile, including non-uniform cooling, non-uniform moderation, or strategic poison placement. Historically, the limitations of computing capabilities have prevented acceptable margins in the temporal component of the spatial excess reactivity in a reactor design, due primarily to the error in burnup calculations. This research was performed as an initial scoping analysis into the concept of a long-life fast reactor. It can be shown that a long-life fast reactor concept can be modeled using MCNPX to predict burnup and neutronics behavior. The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional Light Water Reactors (LWRs) or other SMR designs. For the purpose of this study, a single core design was investigated: a relatively small reactor core, yielding a medium amount of power (~200 to 400 MWth). The results of this scoping analysis were successful in providing a preliminary reactor design involving metal U-235/U-238 fuel with HT-9 fuel cladding and sodium coolant at a 20% volume fraction.

  9. Advances in process intensification through multifunctional reactor engineering.

    SciTech Connect (OSTI)

    Cooper, Marcia A.; Miller, James Edward; O'Hern, Timothy John; Gill, Walter; Evans, Lindsey R.

    2011-02-01

    A multifunctional reactor is a chemical engineering device that exploits enhanced heat and mass transfer to promote production of a desired chemical, combining more than one unit operation in a single system. The main component of the reactor system under study here is a vertical column containing packing material through which liquid(s) and gas flow cocurrently downward. Under certain conditions, a range of hydrodynamic regimes can be achieved within the column that can either enhance or inhibit a desired chemical reaction. To study such reactors in a controlled laboratory environment, two experimental facilities were constructed at Sandia National Laboratories. One experiment, referred to as the Two-Phase Experiment, operates with two phases (air and water). The second experiment, referred to as the Three-Phase Experiment, operates with three phases (immiscible organic liquid and aqueous liquid, and nitrogen). This report describes the motivation, design, construction, operational hazards, and operation of the both of these experiments. Data and conclusions are included.

  10. Fuel Summary Report: Shippingport Light Water Breeder Reactor - Rev. 2

    SciTech Connect (OSTI)

    Olson, Gail Lynn; Mc Cardell, Richard Keith; Illum, Douglas Brent

    2002-09-01

    The Shippingport Light Water Breeder Reactor (LWBR) was developed by Bettis Atomic Power Laboratory to demonstrate the potential of a water-cooled, thorium oxide fuel cycle breeder reactor. The LWBR core operated from 1977-82 without major incident. The fuel and fuel components suffered minimal damage during operation, and the reactor testing was deemed successful. Extensive destructive and nondestructive postirradiation examinations confirmed that the fuel was in good condition with minimal amounts of cladding deformities and fuel pellet cracks. Fuel was placed in wet storage upon arrival at the Expended Core Facility, then dried and sent to the Idaho Nuclear Technology and Engineering Center for underground dry storage. It is likely that the fuel remains in good condition at its current underground dry storage location at the Idaho Nuclear Technology and Engineering Center. Reports show no indication of damage to the core associated with shipping, loading, or storage.

  11. A Compact Reactor Gate Discharge Monitor for Spent Fuel.

    SciTech Connect (OSTI)

    Franco, J. B.; Menlove, Howard O.; Eccleston, G. W.; Miller, M. C.

    2005-01-01

    This paper presents a new design for a reactor gate discharge monitor that has evolved from the baseline discharge monitors used at the Fugen and Tokai-1 reactors in Japan. The main design innovation is the ability to determine direction-of-motion of spent fuel using a single sensor module, as opposed to the two modules used in both baseline design systems. Use of a single module reduces the final system complexity and weight significantly without compromising functionality. The reactor gate discharge monitor uses standard International Atomic Energy Agency (IAEA) hardware and software components. The requirements to determine direction-of-motion from a single module precipitated several development efforts described herein in both the MiniGRAND data acquisition hardware and in the uninterruptible power supply source.

  12. Spent nuclear fuel discharges from US reactors 1993

    SciTech Connect (OSTI)

    Not Available

    1995-02-01

    The Energy Information Administration (EIA) of the U.S. Department of Energy (DOE) administers the Nuclear Fuel Data Survey, Form RW-859. This form is used to collect data on fuel assemblies irradiated at commercial nuclear reactors operating in the United States, and the current inventories and storage capacities of those reactors. These data are important to the design and operation of the equipment and facilities that DOE will use for the future acceptance, transportation, and disposal of spent fuels. The data collected and presented identifies trends in burnup, enrichment, and spent nuclear fuel discharged form commercial light-water reactor as of December 31, 1993. The document covers not only spent nuclear fuel discharges; but also site capacities and inventories; canisters and nonfuel components; and assembly type characteristics.

  13. Health Monitoring to Support Advanced Small Modular Reactors

    SciTech Connect (OSTI)

    Coble, Jamie B.; Meyer, Ryan M.; Ramuhalli, Pradeep

    2013-08-01

    Advanced small modular reactors (aSMRs) are based on advanced reactor concepts, some of which were promoted by the Generation IV International Forum, and are being considered for diverse missions including desalination of water, production of hydrogen, etc. While the existing fleet of commercial nuclear reactors provides baseload electricity, it is conceivable that aSMRs could be implemented for both baseload and load following applications. The effect of diverse operating missions and unit modularity on plant operations and maintenance (O&M) is not fully understood and limiting these costs will be essential to successful deployment of aSMRs. Integrated health monitoring concepts are proposed to support the safe and affordable operation of aSMRs over their lifetime by enabling management of significant in-vessel and in-containment active and passive components.

  14. NEUTRON RADIOGRAPHY (NRAD) REACTOR 64-ELEMENT CORE UPGRADE

    SciTech Connect (OSTI)

    John D. Bess

    2014-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA (registered) (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The interim critical configuration developed during the core upgrade, which contains only 62 fuel elements, has been evaluated as an acceptable benchmark experiment. The final 64-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has also been evaluated as an acceptable benchmark experiment. Calculated eigenvalues differ significantly (approximately +/-1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  15. Medium-size high-temperature gas-cooled reactor

    SciTech Connect (OSTI)

    Peinado, C.O.; Koutz, S.L.

    1980-08-01

    This report summarizes high-temperature gas-cooled reactor (HTGR) experience for the 40-MW(e) Peach Bottom Nuclear Generating Station of Philadelphia Electric Company and the 330-MW(e) Fort St. Vrain Nuclear Generating Station of the Public Service Company of Colorado. Both reactors are graphite moderated and helium cooled, operating at approx. 760/sup 0/C (1400/sup 0/F) and using the uranium/thorium fuel cycle. The plants have demonstrated the inherent safety characteristics, the low activation of components, and the high efficiency associated with the HTGR concept. This experience has been translated into the conceptual design of a medium-sized 1170-MW(t) HTGR for generation of 450 MW of electric power. The concept incorporates inherent HTGR safety characteristics (a multiply redundant prestressed concrete reactor vessel (PCRV), a graphite core, and an inert single-phase coolant) and engineered safety features (core auxiliary cooling, relief valve, and steam generator dump systems).

  16. Decontamination and decommissioning of the Argonne Thermal Source Reactor at Argonne National Laboratory - East project final report.

    SciTech Connect (OSTI)

    Fellhauer, C.; Garlock, G.; Mathiesen, J.

    1998-12-02

    The ATSR D&D Project was directed toward the following goals: (1) Removal of radioactive and hazardous materials associated with the ATSR Reactor facility; (2) Decontamination of the ATSR Reactor facility to unrestricted use levels; and (3)Documentation of all project activities affecting quality (i.e., waste packaging, instrument calibration, audit results, and personnel exposure). These goals had been set in order to eliminate the radiological and hazardous safety concerns inherent in the ATSR Reactor facility and to allow, upon completion of the project, unescorted and unmonitored access to the area. The reactor aluminum, reactor lead, graphite piles in room E-111, and the contaminated concrete in room E-102 were the primary areas of concern. NES, Incorporated (Danbury, CT) characterized the ATSR Reactor facility from January to March 1998. The characterization identified a total of thirteen radionuclides, with a total activity of 64.84 mCi (2.4 GBq). The primary radionuclides of concern were Co{sup 60}, Eu{sup 152}, Cs{sup 137}, and U{sup 238}. No additional radionuclides were identified during the D&D of the facility. The highest dose rates observed during the project were associated with the reactor tank and shield tank. Contact radiation levels of 30 mrem/hr (0.3 mSv/hr) were measured on reactor internals during dismantlement of the reactor. A level of 3 mrem/hr (0.03 mSv/hr) was observed in a small area (hot spot) in room E-102. DOE Order 5480.2A establishes the maximum whole body exposure for occupational workers at 5 rem/yr (50 mSv/yr); the administrative limit at ANL-E is 1 rem/yr (10 mSv/yr).

  17. Fast reactors and nuclear nonproliferation

    SciTech Connect (OSTI)

    Avrorin, E.N.; Rachkov, V.I.; Chebeskov, A.N.

    2013-07-01

    Problems are discussed with regard to nuclear fuel cycle resistance in fast reactors to nuclear proliferation risk due to the potential for use in military programs of the knowledge, technologies and materials gained from peaceful nuclear power applications. Advantages are addressed for fast reactors in the creation of a more reliable mode of nonproliferation in the closed nuclear fuel cycle in comparison with the existing fully open and partially closed fuel cycles of thermal reactors. Advantages and shortcomings are also discussed from the point of view of nonproliferation from the start with fast reactors using plutonium of thermal reactor spent fuel and enriched uranium fuel to the gradual transition using their own plutonium as fuel. (authors)

  18. Solvent refined coal reactor quench system

    DOE Patents [OSTI]

    Thorogood, R.M.

    1983-11-08

    There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream. 1 fig.

  19. Reactor Engineering Design | netl.doe.gov

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Reactor Engineering Design The Reactor Engineering Design Key Technology will focus on control of chemical reactions with unprecedented precision in increasingly modular and efficient reactors, allowing for smaller reactors and streamlined processes that will convert coal into valuable products at low cost and with high energy efficiency. Here, the specific emphasis will be reactors enabling conversion of coal-biomass to liquid fuels, Novel reactors, advanced manufacturing, etc. will be

  20. Solvent refined coal reactor quench system

    DOE Patents [OSTI]

    Thorogood, Robert M.

    1983-01-01

    There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream.

  1. Heat treating of manufactured components

    DOE Patents [OSTI]

    Ripley, Edward B. (Knoxville, TN)

    2012-05-22

    An apparatus for heat treating manufactured components using microwave energy and microwave susceptor material is disclosed. The system typically includes an insulating vessel placed within a microwave applicator chamber. A moderating material is positioned inside the insulating vessel so that a substantial portion of the exterior surface of each component for heat treating is in contact with the moderating material.

  2. A Conceptual Multi-Megawatt System Based on a Tungsten CERMET Reactor

    SciTech Connect (OSTI)

    Jonathan A. Webb; Brian Gross

    2011-02-01

    Abstract. A conceptual reactor system to support Multi-Megawatt Nuclear Electric Propulsion is investigated within this paper. The reactor system consists of a helium cooled Tungsten-UN fission core, surrounded by a beryllium neutron reflector and 13 B4C control drums coupled to a high temperature Brayton power conversion system. Excess heat is rejected via carbon reinforced heat pipe radiators and the gamma and neutron flux is attenuated via segmented shielding consisting of lithium hydride and tungsten layers. Turbine inlet temperatures ranging from 1300 K to 1500 K are investigated for their effects on specific powers and net electrical outputs ranging from 1 MW to 100 MW. The reactor system is estimated to have a mass, which ranges from 15 Mt at 1 MWe and a turbine inlet temperature of 1500 K to 1200 Mt at 100 MWe and a turbine temperature of 1300 K. The reactor systems specific mass ranges from 32 kg/kWe at a turbine inlet temperature of 1300 K and a power of 1 MWe to 9.5 kg/kW at a turbine temperature of 1500 K and a power of 100 MWe.

  3. MOOSE simulating nuclear reactor CRUD buildup

    ScienceCinema (OSTI)

    None

    2014-07-21

    This simulation uses multiple physical models to show how the buildup of boron deposits on reactor fuel can affect performance and the reactor's power profile.

  4. MOOSE simulating nuclear reactor CRUD buildup

    SciTech Connect (OSTI)

    2014-02-06

    This simulation uses multiple physical models to show how the buildup of boron deposits on reactor fuel can affect performance and the reactor's power profile.

  5. Energy Department Announces Small Modular Reactor Technology...

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

    ... of nuclear reactors by providing more than 200 million through a cost-share agreement to support the licensing reviews for Westinghouse's AP1000 reactor design certification. ...

  6. NUCLEAR REACTOR CONTROL SYSTEM

    DOE Patents [OSTI]

    Howard, D.F.; Motta, E.E.

    1961-06-27

    A method for controlling the excess reactivity in a nuclear reactor throughout the core life while maintaining the neutron flux distribution at the desired level is described. The control unit embodies a container having two electrodes of different surface area immersed in an electrolytic solution of a good neutron sbsorbing metal ion such as boron, gadolinium, or cadmium. Initially, the neutron absorber is plated on the larger electrode to control the greater neutron flux of a freshly refueled core. As the fuel burns up, the excess reactivity decreases and the neutron absorber is then plated onto the smaller electrode so that the number of neutrons absorbed also decreases. The excess reactivity in the core may thus be maintained without the introduction of serious perturbations in the neutron flux distributibn.

  7. GAS COOLED NUCLEAR REACTORS

    DOE Patents [OSTI]

    Long, E.; Rodwell, W.

    1958-06-10

    A gas-cooled nuclear reactor consisting of a graphite reacting core and reflector structure supported in a containing vessel is described. A gas sealing means is included for sealing between the walls of the graphite structure and containing vessel to prevent the gas coolant by-passing the reacting core. The reacting core is a multi-sided right prismatic structure having a pair of parallel slots around its periphery. The containing vessel is cylindrical and has a rib on its internal surface which supports two continuous ring shaped flexible web members with their radially innermost ends in sealing engagement within the radially outermost portion of the slots. The core structure is supported on ball bearings. This design permits thermal expansion of the core stracture and vessel while maintainirg a peripheral seal between the tvo elements.

  8. Nuclear reactor control

    DOE Patents [OSTI]

    Cawley, William E. (Phoenix, AZ); Warnick, Robert F. (Pasco, WA)

    1982-01-01

    1. In a nuclear reactor incorporating a plurality of columns of tubular fuel elements disposed in horizontal tubes in a mass of graphite wherein water flows through the tubes to cool the fuel elements, the improvement comprising at least one control column disposed in a horizontal tube including fewer fuel elements than in a normal column of fuel elements and tubular control elements disposed at both ends of said control column, and means for varying the horizontal displacement of the control column comprising a winch at the upstream end of the control column and a cable extending through the fuel and control elements and attached to the element at the downstream end of the column.

  9. Nuclear reactor control apparatus

    DOE Patents [OSTI]

    Sridhar, Bettadapur N. (Cupertino, CA)

    1983-10-25

    Nuclear reactor safety rod release apparatus comprises a ring which carries detents normally positioned in an annular recess in outer side of the rod, the ring being held against the lower end of a drive shaft by magnetic force exerted by a solenoid carried by the drive shaft. When the solenoid is de-energized, the detent-carrying ring drops until the detents contact a cam surface associated with the lower end of the drive shaft, at which point the detents are cammed out of the recess in the safety rod to release the rod from the drive shaft. In preferred embodiments of the invention, an additional latch is provided to release a lower portion of a safety rod under conditions that may interfere with movement of the entire rod.

  10. Structural Analysis of Thermal Shields During a Quench of a Torus Magnet for the 12 GeV Upgrade

    SciTech Connect (OSTI)

    Pastor, Orlando; Willard, Thomas; Ghoshal, Probir K.; Kashy, David H.; Wiseman, Mark A.; Kashikhin, V.; Young, Glenn R.; Elouadrhiri, Latifa; Rode, Claus H.

    2015-06-01

    A toroidal magnet system consisting of six superconducting coils is being built for the Jefferson Lab 12- GeV accelerator upgrade project. This paper details the analysis of eddy current effects during a quench event on the aluminum thermal shield. The shield has been analyzed for mechanical stresses induced as a result of a coil quench as well as a fast discharge of the complete magnet system. The shield has been designed to reduce the eddy current effects and result in stresses within allowable limits.

  11. Foundational development of an advanced nuclear reactor integrated safety code.

    SciTech Connect (OSTI)

    Clarno, Kevin; Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth; Hooper, Russell Warren; Humphries, Larry LaRon

    2010-02-01

    This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

  12. Apparatus for remotely handling components

    DOE Patents [OSTI]

    Szkrybalo, Gregory A.; Griffin, Donald L.

    1994-01-01

    The inventive apparatus for remotely handling bar-like components which define a longitudinal direction includes a gripper mechanism for gripping the component including first and second gripper members longitudinally fixedly spaced from each other and oriented parallel to each other in planes transverse to the longitudinal direction. Each gripper member includes a jaw having at least one V-groove with opposing surfaces intersecting at a base and extending radially relative to the longitudinal direction for receiving the component in an open end between the opposing surfaces. The V-grooves on the jaw plate of the first and second gripper members are aligned in the longitudinal direction to support the component in the first and second gripper members. A jaw is rotatably mounted on and a part of each of the first and second gripper members for selectively assuming a retracted mode in which the open end of the V-groove is unobstructed and active mode in which the jaw spans the open end of the V-groove in the first and second gripper members. The jaw has a locking surface for contacting the component in the active mode to secure the component between the locking surface of the jaw and the opposing surfaces of the V-groove. The locking surface has a plurality of stepped portions, each defining a progressively decreasing radial distance between the base of the V-groove and the stepped portion opposing the base to accommodate varying sizes of components.

  13. Process for producing an aggregate suitable for inclusion into a radiation shielding product

    DOE Patents [OSTI]

    Lessing, Paul A. (Idaho Falls, ID); Kong, Peter C. (Idaho Falls, ID)

    2000-01-01

    The present invention is directed to methods for converting depleted uranium hexafluoride to a stable depleted uranium silicide in a one-step reaction. Uranium silicide provides a stable aggregate material that can be added to concrete to increase the density of the concrete and, consequently, shield gamma radiation. As used herein, the term "uranium silicide" is defined as a compound generically having the formula U.sub.x Si.sub.y, wherein the x represents the molecules of uranium and the y represent the molecules of silicon. In accordance with the present invention, uranium hexafluoride is converted to a uranium silicide by contacting the uranium hexafluoride with a silicon-containing material at a temperature in a range between about 1450.degree. C. and about 1750.degree. C. The stable depleted uranium silicide is included as an aggregate in a radiation shielding product, such as a concrete product.

  14. Effects of SO/sub 2/ shielding gas additions on GTA weld shape

    SciTech Connect (OSTI)

    Heiple, C.R.; Burgardt, P.

    1985-06-01

    Substantial increases in GTA weld depth/width ratio resulted from small additions of sulfur dioxide (SO/sub 2/) to the torch shielding gas when welding two stainless steels. The improvement was demonstrated on both Types 304 and 21-6-9 austenitic stainless steels, but would be expected for iron-base alloys generally. The weld pool shape achieved was essentially independent of variations in both SO/sub 2/ content of the torch gas and base metal composition when SO/sub 2/ in the shielding gas was in the range of 500 to 1400 ppm. With 700 ppm SO/sub 2/ in the torch gas, less than 30 ppm sulfur was added to an autogenous weld bead. For alloys where this additional sulfur can be tolerated and appropriate measures can be taken to handle the toxic SO/sub 2/, this technique offers a promising way to improve GTA weld joint penetration while suppressing variable penetration.

  15. Analytical cell decontamination and shielding window refurbishment. Final report, March 1984-March 1985

    SciTech Connect (OSTI)

    Smokowski, R.T.

    1985-12-01

    This is a report on the decontamination and refurbishment of five inactive contaminated analytical cells and six zinc bromide filled shielding windows. The analytical cells became contaminated during the nuclear fuel reprocessing carried out by Nuclear Fuel Services from 1966 to 1972. The decontamination and decommissioning (D and D) work was performed in these cells to make them useful as laboratories in support of the West Valley Demonstration Project. To accomplish this objective, unnecessary equipment was removed from these cells. Necessary equipment and the interior of each cell were decontaminated and repaired. The shielding windows, essentially tanks holding zinc bromide, were drained and disassembled. The deteriorated, opaque zinc bromide was refined to optical clarity and returned to the tanks. All wastes generated in this operation were characterized and disposed of properly. All the decontamination and refurbishment was accomplished within 13 months. The Analytical Hot Cell has been turned over to Analytical Chemistry for the performance high-level waste (HLW) characterization analysis.

  16. Semi-flexible gas-insulated transmission line using electric field stress shields

    DOE Patents [OSTI]

    Cookson, Alan H. (Churchill Borough, PA); Dale, Steinar J. (Monroeville, PA); Bolin, Philip C. (Wilkins Township, Allegheny County, PA)

    1982-12-28

    A gas-insulated transmission line includes an outer sheath, an inner conductor, an insulating gas electrically insulating the inner conductor from the outer sheath, and insulating supports insulatably supporting the inner conductor within the outer sheath. The inner conductor is provided with flexibility by use of main conductor sections which are joined together through a conductor hub section and flexible flexing elements. Stress shields are provided to control the electric field at the locations of the conductor hub sections where the insulating supports are contacting the inner conductor. The flexing elements and the stress shields may also be utilized in connection with a plug and socket arrangement for providing electrical connection between main conductor sections.

  17. Semi-flexible gas-insulated transmission line using electric field stress shields

    DOE Patents [OSTI]

    Cookson, A.H.; Dale, S.J.; Bolin, P.C.

    1982-12-28

    A gas-insulated transmission line includes an outer sheath, an inner conductor, an insulating gas electrically insulating the inner conductor from the outer sheath, and insulating supports insulatably supporting the inner conductor within the outer sheath. The inner conductor is provided with flexibility by use of main conductor sections which are joined together through a conductor hub section and flexible flexing elements. Stress shields are provided to control the electric field at the locations of the conductor hub sections where the insulating supports are contacting the inner conductor. The flexing elements and the stress shields may also be utilized in connection with a plug and socket arrangement for providing electrical connection between main conductor sections. 10 figs.

  18. A magnetically shielded room with ultra low residual field and gradient

    SciTech Connect (OSTI)

    Altarev, I.; Chesnevskaya, S.; Gutsmiedl, E.; Kuchler, F.; Lins, T.; Marino, M.; McAndrew, J.; Niessen, B.; Paul, S.; Petzoldt, G.; Singh, J.; Stoepler, R.; Stuiber, S.; Sturm, M.; Taubenheim, B. [Physikdepartment, Technische Universitt Mnchen, D-85748 Garching (Germany); Babcock, E. [Jlich Center for Neutron Science, Lichtenbergstrasse 1, D-85748 Garching (Germany); Beck, D.; Sharma, S. [Physics Department, University of Illinois at Urbana-Champaign, Urbana, Illinois 61801 (United States); Burghoff, M.; Fan, I. [Physikalisch-Technische Bundesanstalt Berlin, D-10587 Berlin (Germany); and others

    2014-07-15

    A versatile and portable magnetically shielded room with a field of (700200) pT within a central volume of 1 m 1 m 1 m and a field gradient less than 300 pT/m, achieved without any external field stabilization or compensation, is described. This performance represents more than a hundredfold improvement of the state of the art for a two-layer magnetic shield and provides an environment suitable for a next generation of precision experiments in fundamental physics at low energies; in particular, searches for electric dipole moments of fundamental systems and tests of Lorentz-invariance based on spin-precession experiments. Studies of the residual fields and their sources enable improved design of future ultra-low gradient environments and experimental apparatus. This has implications for developments of magnetometry beyond the femto-Tesla scale in, for example, biomagnetism, geosciences, and security applications and in general low-field nuclear magnetic resonance (NMR) measurements.

  19. Nuclear Rocket Test Facility Decommissioning Including Controlled Explosive Demolition of a Neutron-Activated Shield Wall

    SciTech Connect (OSTI)

    Michael Kruzic

    2007-09-01

    Located in Area 25 of the Nevada Test Site, the Test Cell A Facility was used in the 1960s for the testing of nuclear rocket engines, as part of the Nuclear Rocket Development Program. The facility was decontaminated and decommissioned (D&D) in 2005 using the Streamlined Approach For Environmental Restoration (SAFER) process, under the Federal Facilities Agreement and Consent Order (FFACO). Utilities and process piping were verified void of contents, hazardous materials were removed, concrete with removable contamination decontaminated, large sections mechanically demolished, and the remaining five-foot, five-inch thick radiologically-activated reinforced concrete shield wall demolished using open-air controlled explosive demolition (CED). CED of the shield wall was closely monitored and resulted in no radiological exposure or atmospheric release.

  20. Beryllium processing technology review for applications in plasma-facing components

    SciTech Connect (OSTI)

    Castro, R.G.; Jacobson, L.A.; Stanek, P.W.

    1993-07-01

    Materials research and development activities for the International Thermonuclear Experimental Reactor (ITER), i.e., the next generation fusion reactor, are investigating beryllium as the first-wall containment material for the reactor. Important in the selection of beryllium is the ability to process, fabricate and repair beryllium first-wall components using existing technologies. Two issues that will need to be addressed during the engineering design activity will be the bonding of beryllium tiles in high-heat-flux areas of the reactor, and the in situ repair of damaged beryllium tiles. The following review summarizes the current technology associated with welding and joining of beryllium to itself and other materials, and the state-of-the-art in plasma-spray technology as an in situ repair technique for damaged beryllium tiles. In addition, a review of the current status of beryllium technology in the former Soviet Union is also included.