National Library of Energy BETA

Sample records for reactor dome removal

  1. Heavy Water Test Reactor Dome Removal

    SciTech Connect (OSTI)

    2011-01-01

    A high speed look at the removal of the Heavy Water Test Reactor Dome Removal. A project sponsored by the Recovery Act on the Savannah River Site.

  2. Recovery Act Funds Test Reactor Dome Removal in Historic D&D Project |

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

    Department of Energy Funds Test Reactor Dome Removal in Historic D&D Project Recovery Act Funds Test Reactor Dome Removal in Historic D&D Project February 1, 2011 - 12:00pm Addthis Media Contacts Jim Giusti, DOE (803) 952-7697 james-r.giusti@srs.gov Paivi Nettamo, SRNS (803) 646-6075 paivi.nettamo@srs.gov AIKEN, S.C. - The landscape of the Savannah River Site (SRS) is a little flatter and a little less colorful with the removal today of the 75-foot-tall rusty-orange dome from the

  3. Savannah River Site Removes Dome, Opening Reactor for Recovery Act Decommissioning

    Broader source: Energy.gov [DOE]

    American Recovery and Reinvestment Act workers achieved a significant milestone in the decommissioning of a Cold War reactor at the Savannah River Site this month after they safely removed its...

  4. C-105 heel pit removed and C-105 dome cut paves way for new retrieval...

    Office of Scientific and Technical Information (OSTI)

    Program Document: C-105 heel pit removed and C-105 dome cut paves way for new retrieval technology Citation Details In-Document Search Title: C-105 heel pit removed and C-105 dome ...

  5. 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 ...

  6. Reactor for removing ammonia

    DOE Patents [OSTI]

    Luo, Weifang; Stewart, Kenneth D.

    2009-11-17

    Disclosed is a device for removing trace amounts of ammonia from a stream of gas, particularly hydrogen gas, prepared by a reformation apparatus. The apparatus is used to prevent PEM "poisoning" in a fuel cell receiving the incoming hydrogen stream.

  7. Massive Hanford Test Reactor Removed - Plutonium Recycle Test Reactor

    Office of Environmental Management (EM)

    removed from Hanford's 300 Area | Department of Energy 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 from Hanford's 300 Area January 22, 2014 - 12:00pm Addthis Media Contacts Cameron Hardy, DOE 509-376-5365 Cameron.Hardy@re.doe.gov Mark McKenna, Washington Closure 509-372-9032 media@wch-rcc.com RICHLAND, WA - Hanford's River Corridor contractor, Washington

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

    SciTech Connect (OSTI)

    Austin, W.; Brinkley, D.

    2010-05-05

    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).

  9. WCH Removes Massive Test Reactor | Department of Energy

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

    WCH Removes Massive Test Reactor WCH Removes Massive Test Reactor Addthis Description Hanford's River Corridor contractor, Washington Closure Hanford, has met a significant cleanup challenge on the U.S. Department of Energy's (DOE) Hanford Site by removing a 1,082-ton nuclear test reactor from the 300 Area.

  10. Fast Flux Test Facility Reactor Vessel Removal Study

    SciTech Connect (OSTI)

    BOWMAN, B.R.

    2002-10-23

    This study assesses the feasibility of removing the FFTF reactor vessel from its current location in the reactor cavity inside the Containment vessel to a transporter for relocation to a burial pit in the 200 Area.

  11. C-105 heel pit removed and C-105 dome cut paves way for new retrieval technology

    SciTech Connect (OSTI)

    Mackey, Thomas C.; Sutey, Michael J.

    2013-06-10

    For just the second time, crews have cut a hole in the top of an active radioactive waste storage tank at Hanford. Workers began cutting a 55-inch hole in the top of Tank C-105 last Tuesday night on graveyard shift, completing the cut early Wednesday. The hole will allow for installation of the Mobile Arm Retrieval System (MARS) Vacuum into the tank. The cut was made through 17 inches of concrete and rebar using the newly developed rotary-core cutting system, which uses a laser-guided steel canister with teeth on the bottom to drill a round hole into the tank dome. The project was completed safely and successfully in a high-rad area without contamination or significant dose to workers.

  12. Dome load control and crane land path evaluation for Tank 241-SY-101 during hydrogen mitigation pump removal and installation

    SciTech Connect (OSTI)

    Weis, M.P.; Lawler, D.M.

    1994-08-01

    This report revisits and consolidates two analyses previously performed for the installation of the Hydrogen Mitigation Pump (HMT) pump. The first report determines, as a function of the crane-imposed dome load, the point to which the crane can encroach into the exclusion zone without exceeding the 50-ton limit. The second performs a load evaluation for the crane and the components in the load path (crane lift accessories and pump). In doing so, it determines the weakest component in the load path and the effect of this component on the allowable encroachment distance. Furthermore, the second report sets operational limits on the allowable load decrease (unload) during installation in the event the pump sticks in the riser. The analysis presented here expands on the latter subject by setting an operational limit on the amount of allowable load increase (overload) during pump removal in the event the pump sticks in the riser.

  13. Control of reactor coolant flow path during reactor decay heat removal

    DOE Patents [OSTI]

    Hunsbedt, Anstein N.

    1988-01-01

    An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

  14. Emergency heat removal system for a nuclear reactor

    DOE Patents [OSTI]

    Dunckel, Thomas L.

    1976-01-01

    A heat removal system for nuclear reactors serving as a supplement to an Emergency Core Cooling System (ECCS) during a Loss of Coolant Accident (LOCA) comprises a plurality of heat pipes having one end in heat transfer relationship with either the reactor pressure vessel, the core support grid structure or other in-core components and the opposite end located in heat transfer relationship with a heat exchanger having heat transfer fluid therein. The heat exchanger is located external to the pressure vessel whereby excessive core heat is transferred from the above reactor components and dissipated within the heat exchanger fluid.

  15. Savannah River Site Removes Dome, Opening Reactor for Recovery Act Decommissioning

    Office of Environmental Management (EM)

    Rating | Department of Energy Liquid Waste Contractor Earns Excellent Performance Rating Savannah River Site Liquid Waste Contractor Earns Excellent Performance Rating February 11, 2016 - 12:35pm Addthis SRR workers oversaw placement of nearly 6,100 cubic yards of grout into Tank 16 from June to September 2015, achieving operational closure ahead of the October 2015 scheduled deadline, and making it the seventh tank closed at SRS. SRR workers oversaw placement of nearly 6,100 cubic yards of

  16. Detroit Edison's Fermi 1 - Preparation for Reactor Removal

    SciTech Connect (OSTI)

    Swindle, Danny [Sargent and Lundy Engineers, LLC, 55 E. Monroe Street, Chicago, IL 60603 (United States)

    2008-01-15

    This paper is intended to provide information about the ongoing decommissioning tasks at Detroit Edison's Fermi 1 plant, and in particular, the work being performed to prepare the reactor for removal and disposal. In 1972 Fermi 1 was shutdown and the fuel returned to the Atomic Energy Commission. By the end of 1975, a retirement plan was prepared, the bulk sodium removed, and the plant placed in a safe store condition. The plant systems were left isolated with the sodium containing systems inert with carbon dioxide in an attempt to form a carbonate layer, thus passivating the underlying reactive sodium. In 1996, Detroit Edison determined to evaluate the condition of the plant and to make recommendations in relation to the Fermi 1 future plans. At the end of 1997 approval was obtained to remove the bulk asbestos and residual alkali-metals (i.e., sodium and sodium potassium (NaK)). In 2000, full nuclear decommissioning of the plant was approved. To date, the bulk asbestos insulation has been removed, and the only NaK remaining is located in six capillary instrument tubes. The remaining sodium is contained within the reactor, two of the three primary loops, and miscellaneous removed pipes and equipment to be processed. The preferred method for removing or reacting sodium at Fermi 1 is by injecting superheated steam into a heated, nitrogen inert system. The byproducts of this reaction are caustic sodium hydroxide, hydrogen gas, and heat. The decision was made to separate the three primary loops from the reactor for better control prior to processing each loop and the reactor separately. The first loop has already been processed. The main focus is now to process the reactor to allow removal and disposal of the Class C waste prior to the anticipated June 2008 closure of the Barnwell radioactive waste disposal facility located in South Carolina. Lessons learnt are summarized and concern: the realistic schedule and adherence to the schedule, time estimates, personnel

  17. Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors (Journal...

    Office of Scientific and Technical Information (OSTI)

    Journal Article: Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors Citation Details In-Document Search Title: Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors The safety ...

  18. Removable check valve for use in a nuclear reactor

    DOE Patents [OSTI]

    Dunn, Charlton (Calabasas, CA); Gutzmann, Edward A. (Simi Valley, CA)

    1988-01-01

    A removable check valve for interconnecting the discharge duct of a pump and an inlet coolant duct of a reactor core in a pool-type nuclear reactor. A manifold assembly is provided having an outer periphery affixed to and in fluid communication with the discharge duct of the pump and has an inner periphery having at least one opening therethrough. A housing containing a check valve is located within the inner periphery of the manifold. The upper end of the housing has an opening in alignment with the opening in the manifold assembly, and seals are provided above and below the openings. The lower end of the housing is adapted for fluid communication with the inlet duct of the reactor core.

  19. Alternative cooling resource for removing the residual heat of reactor

    SciTech Connect (OSTI)

    Park, H. C.; Lee, J. H.; Lee, D. S.; Jung, C. Y.; Choi, K. Y.

    2012-07-01

    The Recirculated Cooling Water (RCW) system of a Candu reactor is a closed cooling system which delivers demineralized water to coolers and components in the Service Building, the Reactor Building, and the Turbine Building and the recirculated cooling water is designed to be cooled by the Raw Service Water (RSW). During the period of scheduled outage, the RCW system provides cooling water to the heat exchangers of the Shutdown Cooling System (SDCS) in order to remove the residual heat of the reactor, so the RCW heat exchangers have to operate at all times. This makes it very hard to replace the inlet and outlet valves of the RCW heat exchangers because the replacement work requires the isolation of the RCW. A task force was formed to prepare a plan to substitute the recirculated water with the chilled water system in order to cool the SDCS heat exchangers. A verification test conducted in 2007 proved that alternative cooling was possible for the removal of the residual heat of the reactor and in 2008 the replacement of inlet and outlet valves of the RCW heat exchangers for both Wolsong unit 3 and 4 were successfully completed. (authors)

  20. TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 2 DF WASTE LINE REMOVAL, BNL

    SciTech Connect (OSTI)

    P.C. Weaver

    2010-07-09

    5098-SR-02-0 PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 2 DF WASTE LINE REMOVAL, BROOKHAVEN NATIONAL LABORATORY

  1. Process and apparatus for adding and removing particles from pressurized reactors

    DOE Patents [OSTI]

    Milligan, John D.

    1983-01-01

    A method for adding and removing fine particles from a pressurized reactor is provided, which comprises connecting the reactor to a container, sealing the container from the reactor, filling the container with particles and a liquid material compatible with the reactants, pressurizing the container to substantially the reactor pressure, removing the seal between the reactor and the container, permitting particles to fall into or out of the reactor, and resealing the container from the reactor. An apparatus for adding and removing particles is also disclosed.

  2. System for fuel rod removal from a reactor module

    DOE Patents [OSTI]

    Matchett, Richard L.; Roof, David R.; Kikta, Thomas J.; Wilczynski, Rosemarie; Nilsen, Roy J.; Bacvinskas, William S.; Fodor, George

    1990-01-01

    A robotic system for remote underwater withdrawal of the fuel rods from fuel modules of a light water breeder reactor includes a collet/grapple assembly for gripping and removing fuel rods in each module, which is positioned by use of a winch and a radial support means attached to a vertical support tube which is mounted over the fuel module. A programmable logic controller in conjunction with a microcomputer, provides control for the accurate positioning and pulling force of the rod grapple assembly. Closed circuit television cameras are provided which aid in operator interface with the robotic system.

  3. System for fuel rod removal from a reactor module

    DOE Patents [OSTI]

    Matchett, R.L.; Fodor, G.; Kikta, T.J.; Bacvinsicas, W.S.; Roof, D.R.; Nilsen, R.J.; Wilczynski, R.

    1988-07-28

    A robotic system for remote underwater withdrawal of the fuel rods from fuel modules of a light water breeder reactor includes a collet/grapple assembly for gripping and removing fuel rods in each module, which is positioned by use of a winch and a radial support means attached to a vertical support tube which is mounted over the fuel module. A programmable logic controller in conjunction with a microcomputer, provides control for the accurate positioning and pulling force of the rod grapple assembly. Closed circuit television cameras are provided which aid in operator interface with the robotic system. 7 figs.

  4. Method for removing cesium from a nuclear reactor coolant

    DOE Patents [OSTI]

    Colburn, Richard P. (Pasco, WA)

    1986-01-01

    A method of and system for removing cesium from a liquid metal reactor coolant including a carbon packing trap in the primary coolant system for absorbing a major portion of the radioactive cesium from the coolant flowing therethrough at a reduced temperature. A regeneration subloop system having a secondary carbon packing trap is selectively connected to the primary system for isolating the main trap therefrom and connecting it to the regeneration system. Increasing the temperature of the sodium flowing through the primary trap diffuses a portion of the cesium

  5. Method for removing cesium from a nuclear reactor coolant

    DOE Patents [OSTI]

    Colburn, R.P.

    1983-08-10

    A method of and system for removing cesium from a liquid metal reactor coolant including a carbon packing trap in the primary coolant system for absorbing a major portion of the radioactive cesium from the coolant flowing therethrough at a reduced temperature. A regeneration subloop system having a secondary carbon packing trap is selectively connected to the primary system for isolating the main trap therefrom and connecting it to the regeneration system. Increasing the temperature of the sodium flowing through the primary trap diffuses a portion of the cesium inventory thereof further into the carbon matrix while simultaneously redispersing a portion into the regeneration system for absorption at a reduced temperature by the secondary trap.

  6. 100-KE REACTOR CORE REMOVAL PROJECT ALTERNATIVE ANALYSIS WORKSHOP REPORT

    SciTech Connect (OSTI)

    HARRINGTON RA

    2010-01-15

    On December 15-16, 2009, a 100-KE Reactor Core Removal Project Alternative Analysis Workshop was conducted at the Washington State University Consolidated Information Center, Room 214. Colburn Kennedy, Project Director, CH2M HILL Plateau Remediation Company (CHPRC) requested the workshop and Richard Harrington provided facilitation. The purpose of the session was to select the preferred Bio Shield Alternative, for integration with the Thermal Shield and Core Removal and develop the path forward to proceed with project delivery. Prior to this workshop, the S.A. Robotics (SAR) Obstruction Removal Alternatives Analysis (565-DLV-062) report was issued, for use prior to and throughout the session, to all the team members. The multidisciplinary team consisted ofrepresentatives from 100-KE Project Management, Engineering, Radcon, Nuclear Safety, Fire Protection, Crane/Rigging, SAR Project Engineering, the Department of Energy Richland Field Office, Environmental Protection Agency, Washington State Department of Ecology, Defense Nuclear Facility Safety Board, and Deactivation and Decommission subject matter experts from corporate CH2M HILL and Lucas. Appendix D contains the workshop agenda, guidelines and expectations, opening remarks, and attendance roster going into followed throughout the workshop. The team was successful in selecting the preferred alternative and developing an eight-point path forward action plan to proceed with conceptual design. Conventional Demolition was selected as the preferred alternative over two other alternatives: Diamond Wire with Options, and Harmonic Delamination with Conventional Demolition. The teams preferred alternative aligned with the SAR Obstruction Removal Alternative Analysis report conclusion. However, the team identified several Path Forward actions, in Appendix A, which upon completion will solidify and potentially enhance the Conventional Demolition alternative with multiple options and approaches to achieve project delivery

  7. Lava Dome | Open Energy Information

    Open Energy Info (EERE)

    Horst and Graben Shield Volcano Flat Lava Dome Stratovolcano Cinder Cone Caldera Depression Resurgent Dome Complex "Volcanic or lava domes are formed by relatively small,...

  8. Passive decay heat removal system for water-cooled nuclear reactors

    DOE Patents [OSTI]

    Forsberg, Charles W.

    1991-01-01

    A passive decay-heat removal system for a water-cooled nuclear reactor employs a closed heat transfer loop having heat-exchanging coils inside an open-topped, insulated box located inside the reactor vessel, below its normal water level, in communication with a condenser located outside of containment and exposed to the atmosphere. The heat transfer loop is located such that the evaporator is in a position where, when the water level drops in the reactor, it will become exposed to steam. Vapor produced in the evaporator passes upward to the condenser above the normal water level. In operation, condensation in the condenser removes heat from the system, and the condensed liquid is returned to the evaporator. The system is disposed such that during normal reactor operations where the water level is at its usual position, very little heat will be removed from the system, but during emergency, low water level conditions, substantial amounts of decay heat will be removed.

  9. Method and apparatus for removing iodine from a nuclear reactor coolant

    DOE Patents [OSTI]

    Cooper, Martin H.

    1980-01-01

    A method and apparatus for removing iodine-131 and iodine-125 from a liquid sodium reactor coolant. Non-radioactive iodine is dissolved in hot liquid sodium to increase the total iodine concentration. Subsequent precipitation of the iodine in a cold trap removes both the radioactive iodine isotopes as well as the non-radioactive iodine.

  10. 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.

  11. 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.

  12. NNSA Completes Conversion of the Budapest Research Reactor and Removal of

    National Nuclear Security Administration (NNSA)

    All Fresh HEU in Hungary | National Nuclear Security Administration | (NNSA) Completes Conversion of the Budapest Research Reactor and Removal of All Fresh HEU in Hungary September 15, 2009 WASHINGTON, D.C. - This week, the National Nuclear Security Administration (NNSA), in cooperation with KFKI Atomic Energy Research Institute, successfully converted the Budapest Research Reactor (BRR) from the use of highly enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. The BRR conversion

  13. 30th Anniversary of the Experimental Breeder Reactor-II Heat Removal

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

    Tests | Department of Energy 30th Anniversary of the Experimental Breeder Reactor-II Heat Removal Tests 30th Anniversary of the Experimental Breeder Reactor-II Heat Removal Tests April 12, 2016 - 9:46am Addthis John Kotek John Kotek Acting Assistant Secretary for the Office of Nuclear Energy Thirty years ago this month, two events happened that had profound and lasting impacts on energy and environmental issues in the U.S. and around the world. One overshadowed the other in public awareness,

  14. Nuclear reactor with makeup water assist from residual heat removal system

    DOE Patents [OSTI]

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

    1993-12-07

    A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path. 2 figures.

  15. Nuclear reactor with makeup water assist from residual heat removal system

    DOE Patents [OSTI]

    Corletti, Michael M.; Schulz, Terry L.

    1993-01-01

    A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path.

  16. Dome Tech | Open Energy Information

    Open Energy Info (EERE)

    Dome Tech Jump to: navigation, search Name: Dome-Tech Place: Edison, New Jersey Zip: 8837 Sector: Services Product: Edison-based provider of services in engineering, energy...

  17. Emergency Decay Heat Removal in a GEN-IV Gas-Cooled Fast Reactor

    SciTech Connect (OSTI)

    Cheng, Lap Y.; Ludewig, Hans; Jo, Jae [Brookhaven National Laboratory, P.O. Box 5000, Upton, NY 11973-5000 (United States)

    2006-07-01

    A series of transient analyses using the system code RELAP5-3d has been performed to confirm the efficacy of a proposed hybrid active/passive combination approach to the decay heat removal for an advanced 2400 MWt GEN-IV gas-cooled fast reactor. The accident sequence of interest is a station blackout simultaneous with a small break (10 sq.inch/0.645 m{sup 2}) in the reactor vessel. The analyses cover the three phases of decay heat removal in a depressurization accident: (1) forced flow cooling by the power conversion unit (PCU) coast down, (2) active forced flow cooling by a battery powered blower, and (3) passive cooling by natural circulation. The blower is part of an emergency cooling system (ECS) that by design is to sustain passive decay heat removal via natural circulation cooling 24 hours after shutdown. The RELAP5 model includes the helium-cooled reactor, the ECS (primary and secondary side), the PCU with all the rotating machinery (turbine and compressors) and the heat transfer components (recuperator, pre-cooler and inter-cooler), and the guard containment that surrounds the reactor and the PCU. The transient analysis has demonstrated the effectiveness of passive decay heat removal by natural circulation cooling when the guard containment pressure is maintained at or above 800 kPa. (authors)

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

    DOE Patents [OSTI]

    Pennell, William E.; Kalinowski, Joseph E.; Waldby, Robert N.; Rylatt, John A.; Swenson, Daniel V.

    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.

  19. Organic and nitrogen removal from landfill leachate in aerobic granular sludge sequencing batch reactors

    SciTech Connect (OSTI)

    Wei Yanjie; Ji Min; Li Ruying; Qin Feifei

    2012-03-15

    Highlights: Black-Right-Pointing-Pointer Aerobic granular sludge SBR was used to treat real landfill leachate. Black-Right-Pointing-Pointer COD removal was analyzed kinetically using a modified model. Black-Right-Pointing-Pointer Characteristics of nitrogen removal at different ammonium inputs were explored. Black-Right-Pointing-Pointer DO variations were consistent with the GSBR performances at low ammonium inputs. - Abstract: Granule sequencing batch reactors (GSBR) were established for landfill leachate treatment, and the COD removal was analyzed kinetically using a modified model. Results showed that COD removal rate decreased as influent ammonium concentration increasing. Characteristics of nitrogen removal at different influent ammonium levels were also studied. When the ammonium concentration in the landfill leachate was 366 mg L{sup -1}, the dominant nitrogen removal process in the GSBR was simultaneous nitrification and denitrification (SND). Under the ammonium concentration of 788 mg L{sup -1}, nitrite accumulation occurred and the accumulated nitrite was reduced to nitrogen gas by the shortcut denitrification process. When the influent ammonium increased to a higher level of 1105 mg L{sup -1}, accumulation of nitrite and nitrate lasted in the whole cycle, and the removal efficiencies of total nitrogen and ammonium decreased to only 35.0% and 39.3%, respectively. Results also showed that DO was a useful process controlling parameter for the organics and nitrogen removal at low ammonium input.

  20. Experimental and analytical studies of passive shutdown heat removal from advanced LMRs (liquid metal reactors)

    SciTech Connect (OSTI)

    Pedersen, D.; Heineman, J.; Stewart, R.; Anderson, T.; Lottes, P.; Tessier, J.

    1988-01-01

    A facility designed and constructed to demonstrate the viability of natural convection passive heat removal systems as a key feature of innovative LMR Shutdown Heat Removal (SHR) systems is in operation at Argonne National Laboratory (ANL). This Natural Convection Shutdown Heat Removal Test Facility (NSTF) has investigated the heat transfer performance of the GE/PRISM passive design. This initial series of experiments simulates the air-side geometry of the PRISM Radiant Reactor Vessel Auxiliary Cooling System (RVACS). The NSTF operates in either a uniform heat flux mode and a uniform temperature mode at the air/guard vessel interface. Analysis of the RVACS performance data indicates excellent agreement with pretest analytical predictions. Correlation analysis presents the heat transfer data in a form suitable for use in LMR design and verification of analytical studies.

  1. The Gas-Cooled Fast Reactor: Report on Safety System Design for Decay Heat Removal

    SciTech Connect (OSTI)

    K. D. Weaver; T. Marshall; T. Y. C. Wei; E. E. Feldman; M. J. Driscoll; H. Ludewig

    2003-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radiotoxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. This report addresses/discusses the decay heat removal options available to the GFR, and the current solutions. While it is possible to design a GFR with complete passive safety (i.e., reliance solely on conductive and radiative heat transfer for decay heat removal), it has been shown that the low power density results in unacceptable fuel cycle costs for the GFR. However, increasing power density results in higher decay heat rates, and the attendant temperature increase in the fuel and core. Use of active movers, or blowers/fans, is possible during accident conditions, which only requires 3% of nominal flow to remove the decay heat. Unfortunately, this requires reliance on active systems. In order to incorporate passive systems, innovative designs have been studied, and a mix of passive and active systems appears to meet the requirements for decay heat removal during accident conditions.

  2. Cyclic process for producing methane in a tubular reactor with effective heat removal

    DOE Patents [OSTI]

    Frost, Albert C.; Yang, Chang-Lee

    1986-01-01

    Carbon monoxide-containing gas streams are converted to methane by a cyclic, essentially two-step process in which said carbon monoxide is disproportionated to form carbon dioxide and active surface carbon deposited on the surface of a catalyst, and said carbon is reacted with steam to form product methane and by-product carbon dioxide. The exothermic heat of reaction generated in each step is effectively removed during each complete cycle so as to avoid a build up of heat from cycle-to-cycle, with particularly advantageous techniques being employed for fixed bed, tubular and fluidized bed reactor operations.

  3. Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors

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

    Cheng, Lap-Yan; Wei, Thomas Y. C.

    2009-01-01

    The safety goal of the current designs of advanced high-temperature thermal gas-cooled reactors (HTRs) is that no core meltdown would occur in a depressurization event with a combination of concurrent safety system failures. This study focused on the analysis of passive decay heat removal (DHR) in a GEN IV direct-cycle gas-cooled fast reactor (GFR) which is based on the technology developments of the HTRs. Given the different criteria and design characteristics of the GFR, an approach different from that taken for the HTRs for passive DHR would have to be explored. Different design options based on maintaining core flow weremore » evaluated by performing transient analysis of a depressurization accident using the system code RELAP5-3D. The study also reviewed the conceptual design of autonomous systems for shutdown decay heat removal and recommends that future work in this area should be focused on the potential for Brayton cycle DHRs.« less

  4. PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 3 TRENCH 1, BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK

    SciTech Connect (OSTI)

    E.M. Harpenau

    2010-12-15

    5098-SR-05-0 PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 3 TRENCH 1 BROOKHAVEN NATIONAL LABORATORY

  5. PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 3 TRENCH 5, BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK

    SciTech Connect (OSTI)

    P.C. Weaver

    2010-11-03

    5098-SR-04-0 PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 3 TRENCH 5, BROOKHAVEN NATIONAL LABORATORY

  6. Resurgent Dome Complex | Open Energy Information

    Open Energy Info (EERE)

    the formation of a resurgent dome. http:www.iub.edusierrapapers2012pardoski.html Resurgent domes are encountered near the center of many caldera depressions, and form...

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

    SciTech Connect (OSTI)

    Lata

    1996-09-26

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

  8. A vacuum disengager for tritium removal from HYLIFE-II Reactor Flibe

    SciTech Connect (OSTI)

    Dolan, T.J.; Longhurst, G.R.; Garcia-Otero, E.

    1992-09-01

    We have designed a vacuum disengager system to remove tritium from the Flibe (Li{sub 2}BeF{sub 4}) molten salt coolant of the HYLIFE-II fusion reactor. There is a two-stage vacuum disengager in each of three intermediate heat exchanger (IHX) loops. Each stage consists of a vacuum chamber 4 m in diameter and 7 m tall. As 0.2 mm diameter molten salt droplets fall vertically downward into the vacuum, most of the tritium diffuses out of the droplets and is pumped away. A fraction {Phi} {approximately}10{sup {minus}5} of the 8.6 MCi/day tritium source (from breeding in the Flibe and from unburned fuel) remains in the Flibe as it leaves the vacuum disengagers, and about 21% of that permeates into the intermediate coolant loop, so about 20 Ci/day leak into the steam system. With Flibe primary coolant and a vacuum disengager, it appears that an intermediate coolant loop is not needed to prevent tritium from leaking into the steam system. An experiment is needed to demonstrate Flibe vacuum disengager operation.

  9. Removal of 1,082-Ton Reactor Among Richland Operations Office...

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

    from groundwater across the site ahead of schedule and pumped a record volume of water through treatment facilities to remove contamination, with more than 130 tons of...

  10. Analytical studies on the impact of using repeated-rib roughness in LMR (Liquid Metal Reactor) decay heat removal systems

    SciTech Connect (OSTI)

    Obot, N.T.; Tessier, J.H.; Pedersen, D.R.

    1988-01-01

    A numerical study was carried out to determine the effects of roughness on the thermal performance of Liquid Metal Reactor (LMR) decay heat removal systems for a range of possible design configurations and operating conditions. The ranges covered for relative rib height (e/D/sub h/), relative pitch (p/e) and flow attack angle were 0.026--0.103, 5--20 and 0--90 degrees, successively. The heat flux was varied between 1.1 and 21.5 kW/m/sup 2/ (0.1 and 2.0 kW/ft/sup 2/). Calculations were made for three cases: smooth duct with no ribs, ribs on both the guard vessel and collector wall, and ribs on the collector wall only. The results indicate that significant benefits, amounting to nearly two-fold reductions in guard vessel and collector wall temperatures, can be realized by placing repeated ribs on both the guard vessel and the collector wall. The magnitudes of the reduction in the reactor vessel temperature are considerably smaller. In general, the level of improvement, be it with respect to temperature or heat flux, is only mildly affected by changes in rib height or pitch but exhibits greater sensitivity to the assumed value for the system form loss. When the ribs are placed only on the collector wall, the heat removal capability is substantially reduced.

  11. North Dome decision expected soon

    SciTech Connect (OSTI)

    Not Available

    1981-08-01

    Decisions soon will be made which will set in motion the development of Qatar's huge North Dome gas field. The government and state company, Qatar General Petroleum Corp. (QGPC) is studying the results of 2 feasibility studies on the economics of LNG export, although initially North Dome exploitation will be aimed at the domestic market. Decisions on the nature and timing of the North Dome development are the most important that have had to be faced in the short 10-yr history of the small Gulf state. The country's oil production is currently running at approximately 500,000 bpd, with 270,000 bpd originating from 3 offshore fields. Output is expected to decline through 1990, and it generally is accepted that there is little likelihood of further major crude discoveries. Therefore, Qatar has to begin an adjustment from an economy based on oil to one based on gas, while adhering to the underlying tenets of long-term conservation and industrial diversification.

  12. Systems and methods for retaining and removing irradiation targets in a nuclear reactor

    SciTech Connect (OSTI)

    Runkle, Gary A.; Matsumoto, Jack T.; Dayal, Yogeshwar; Heinold, Mark R.

    2015-12-08

    A retainer is placed on a conduit to control movement of objects within the conduit in access-restricted areas. Retainers can prevent or allow movement in the conduit in a discriminatory fashion. A fork with variable-spacing between prongs can be a retainer and be extended or collapsed with respect to the conduit to change the size of the conduit. Different objects of different sizes may thus react to the fork differently, some passing and some being blocked. Retainers can be installed in inaccessible areas and allow selective movement in remote portions of conduit where users cannot directly interface, including below nuclear reactors. Position detectors can monitor the movement of objects through the conduit remotely as well, permitting engagement of a desired level of restriction and object movement. Retainers are useable in a variety of nuclear power plants and with irradiation target delivery, harvesting, driving, and other remote handling or robotic systems.

  13. ALARA Controls and the Radiological Lessons Learned During the Uranium Fuel Removal Projects at the Molten Salt Reactor Experiment

    SciTech Connect (OSTI)

    Gilliam, B. J.; Chapman, J. A.; Jugan, M. R.

    2002-02-26

    The removal of uranium-233 (233 U) from the auxiliary charcoal bed (ACB) of the Molten Salt Reactor Experiment (MSRE), performed from January through May 2001, created both unique radiological challenges and widely-applicable lessons learned. In addition to the criticality concerns and alpha contamination, 233U has an associated intense gamma photon from the cocontaminant uranium-232 (232U) decaying to thallium-208 (208Tl). Therefore, rigorous contamination controls and significant shielding were implemented. Extensive, timed mock-up training was also imperative to minimize individual and collective personnel exposures. Back-up shielding and containment techniques (that had been previously developed for defense in depth) were used successfully to control significant, changed conditions. Additional controls were placed on tests and on recovery designs to assure a higher level of safety throughout the removal operations. This paper delineates the manner in which each difficulty was solved, while relating the relevance of the results and the methodology to other projects with high dose-rate, highly-contaminated ionizing radiation hazards. Because of the distinctive features of and current interest in molten salt technology, a brief overview is provided. Also presented is the detailed, practical application of radiological controls integrated into, rather than added after, each evolution of the project--thus demonstrating the broad-based benefits of radiological engineering and ALARA reviews. The resolution of the serious contamination-control problems caused by unexpected uranium hexafluoride (UF6) gaseous diffusion is also explicated. Several tables and figures document the preparations, equipment and operations. A comparison of the pre-job dose calculations for the various functions of the uranium deposit removal (UDR) and the post-job dose-rate data are included in the conclusion.

  14. Environmental assessment: Richton Dome Site, Mississippi

    SciTech Connect (OSTI)

    none,

    1986-05-01

    In February 1983, the US Department of Energy (DOE) identified the Richton Dome site in Mississippi as one of the nine potentially acceptable sites for a mined geologic repository for spent nuclear fuel and high-level radioactive waste. To determine their suitability, the Richton Dome site and the eight other potentially acceptable sites have been evaluated in accordance with the DOE's General Guidelines for the Recommendation of Sites for the Nuclear Waste Repositories. These evaluations were reported in draft environmental assessments (EAs), which were issued for public review and comment. After considering the comments received on the draft EAs, the DOE prepared the final EAs. The site is in the Gulf interior region, which is one of five distinct geohydrologic settings considered for the first repository. This setting contains two other potentially acceptable sites--the Cypress Creek Dome site in Mississippi and the Vacherie Dome site in Louisiana. Although the Cypress Creek Dome and the Vacherie Dome sites are suitable for site characterization, the DOE has concluded that the Richton Dome site is the preferred site in the Gulf interior region. On the basis of the evaluations reported in this EA, the DOE has found that the Richton Dome site is not disqualified under the guidelines.

  15. Environmental assessment: Richton Dome site, Mississippi

    SciTech Connect (OSTI)

    none,

    1986-05-01

    In February 1983, the US Department of Energy (DOE) identified the Richton Dome site in Mississippi as one of the nine potentially acceptable sites for a mined geologic repository for spent nuclear fuel and high-level radioactive waste. To determine their suitability, the Richton Dome site and the eight other potentially acceptable sites have been evaluated in accordance with the DOE's General Guidelines for the Recommendation of Sites for the Nuclear Waste Repositories. These evaluations were reported in draft environmental assessments (EAs), which were issued for public review and comment. After considering the comments received on the draft EAs, the DOE prepared the final EAs. The site is in the Gulf interior region, which is one of five distinct geohydrologic settings considered for the first repository. This setting contains two other potentially acceptable sites--the Cypress Creek Dome site in Mississippi and the Vacherie Dome site in Louisiana. Although the Cypress Creek Dome and the Vacherie Dome sites are suitable for site characterization, the DOE has concluded that the Richton Dome site is the preferred site in the Gulf interior region. On the basis of the evaluations reported in this EA, the DOE has found that the Richton Dome site is not disqualified under the guidelines.

  16. REACTOR

    DOE Patents [OSTI]

    Christy, R.F.

    1961-07-25

    A means is described for co-relating the essential physical requirements of a fission chain reaction in order that practical, compact, and easily controllable reactors can be built. These objects are obtained by employing a composition of fissionsble isotope and moderator in fluid form in which the amount of fissionsble isotcpe present governs the reaction. The size of the reactor is no longer a critical factor, the new criterion being the concentration of the fissionable isotope.

  17. REACTOR

    DOE Patents [OSTI]

    Szilard, L.

    1963-09-10

    A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

  18. 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

  19. REACTOR

    DOE Patents [OSTI]

    Roman, W.G.

    1961-06-27

    A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

  20. 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)

  1. Effect of fins and repeated-rib roughness on the performance characteristics of a reactor vessel air cooling system for LMFBR shutdown heat removal

    SciTech Connect (OSTI)

    Cheung, F.B.; Chawla, T.C.; Pedersen, D.R.; Tessier, J.H.; Webb, R.L.

    1986-01-01

    The use of a totally passive cooling system for shutdown heat removal that rejects heat from the reactor vessel by radiation to the guard vessel and from the guard vessel to a circulating air stream driven by natural convection is a key feature of the US Department of Energy's liquid-metal reactor advanced design study concepts. General Electric refers to the system as the Reactor Vessel Auxiliary Cooling System (RVACS) and Rockwell International as the Reactor Auxiliary Cooling System (RACS). The circulating air stream is contained in the annular passage formed with guard vessel wall and the duct wall surrounding the guard vessel. Specifically, the RVACS/RACS is designed to assure adequate cooling of the reactor vessel under abnormal operational conditions associated with loss of heat removal through the normal heat transport path via the steam generator system or the DRACS, if available. To enhance the heat transfer, longitudinal radial fins or repeated ribs can be attached to the duct wall and/or the guard vessel. The purpose of the present paper is to summarize the status of the analytical work on the development of an optimum design configuration for the RVACS/RACS.

  2. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Daniels, F.

    1959-10-27

    A reactor in which at least a portion of the moderator is in the form of movable refractory balls is described. In addition to their moderating capacity, these balls may serve as carriers for fissionable material or fertile material, or may serve in a coolant capacity to remove heat from the reactor. A pneumatic system is used to circulate the balls through the reactor.

  3. Reactor pressure vessel vented head

    DOE Patents [OSTI]

    Sawabe, James K.

    1994-01-11

    A head for closing a nuclear reactor pressure vessel shell includes an arcuate dome having an integral head flange which includes a mating surface for sealingly mating with the shell upon assembly therewith. The head flange includes an internal passage extending therethrough with a first port being disposed on the head mating surface. A vent line includes a proximal end disposed in flow communication with the head internal passage, and a distal end disposed in flow communication with the inside of the dome for channeling a fluid therethrough. The vent line is fixedly joined to the dome and is carried therewith when the head is assembled to and disassembled from the shell.

  4. Reactor

    DOE Patents [OSTI]

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  5. Environmental assessment, Richton Dome site, Mississippi (US)

    SciTech Connect (OSTI)

    none,

    1986-05-01

    The Nuclear Waste Policy Act of 1982 (42 USC Sections 10101-10226) requires the environmental assessment of a potential site to include a statement of the basis for the nomination of a site as suitable for characterization. Volume 2 of this environmental assessment provides a detailed evaluation of the Richton Dome Site and its suitability as the site for a radioactive waste disposal facility under DOE siting guidelines, as well as a comparison of the Richton Dome site with other proposed sites. Evaluation of the Richton Dome site is based on the reference repository design, but the evaluation will not change if based on the Mission Plan repository concept. The comparative evaluation of proposed sites is required under DOE guidelines, but is not intended to directly support the subsequent recommendation of three sites for characterization as candidate sites. 428 refs., 24 figs., 62 tabs. (MHB)

  6. Analysis of palladium coatings to remove hydrogen isotopes from zirconium fuel rods in Canada deuterium uranium-pressurized heavy water reactors; Thermal and neutron diffusion effects

    SciTech Connect (OSTI)

    Stokes, C.L.; Buxbaum, R.E. )

    1992-05-01

    This paper reports that, in pressurized heavy water nuclear reactors of the type standardly used in Canada (Canada deuterium uranium-pressurized heavy water reactors), the zirconium alloy pressure tubes of the core absorb deuterium produced by corrosion reactions. This deuterium weakens the tubes through hydrogen embrittlement. Thin palladium coatings on the outside of the zirconium are analyzed as a method for deuterium removal. This coating is expected to catalyze the reaction D{sub 2} + 1/2O{sub 2} {r reversible} D{sub 2}O when O{sub 2} is added to the annular (insulating) gas in the tubes. Major reductions in the deuterium concentration and, hence, hydrogen embrittlement are predicted. Potential problems such as plating the tube geometry, neutron absorption, catalyst deactivation, radioactive waste production, and oxygen corrosion are shown to be manageable. Also, a simple set of equations are derived to calculate the effect on diffusion caused by neutron interactions. Based on calculations of ordinary and neutron flux induced diffusion, a palladium coating of 1 {times} 10{sup {minus}6} m is recommended. This would cost approximately $60,000 per reactor unit and should more than double reactor lifetime. Similar coatings and similar interdiffusion calculations might have broad applications.

  7. Development of 1000 MWe Advanced Boiling Water Reactor

    SciTech Connect (OSTI)

    Kazuo Hisajima; Ken Uchida; Keiji Matsumoto; Koichi Kondo; Shigeki Yokoyama; Takuya Miyagawa [Toshiba Corporation (Japan)

    2006-07-01

    1000 MWe Advanced Boiling Water Reactor has only two main steam lines and six reactor internal pumps, whereas 1350 MWe ABWR has four main steam lines and ten reactor internal pumps. In order to confirm how the differences affect hydrodynamic conditions in the dome and lower plenum of the reactor pressure vessel, fluid analyses have been performed. The results indicate that there is not substantial difference between 1000 MWe ABWR and 1350 MWe ABWR. The primary containment vessel of the ABWR consists of the drywell and suppression chamber. The suppression chamber stores water to suppress pressure increase in the primary containment vessel and to be used as the source of water for the emergency core cooling system following a loss-of-coolant accident. Because the reactor pressure vessel of 1000 MWe ABWR is smaller than that of 1350 MWe ABWR, there is room to reduce the size of the primary containment vessel. It has been confirmed feasible to reduce inner diameter of the primary containment vessel from 29 m of 1350 MWe ABWR to 26.5 m. From an economic viewpoint, a shorter outage that results in higher availability of the plant is preferable. In order to achieve 20-day outage that results in 97% of availability, improvement of the systems for removal of decay heat is introduced that enables to stop all the safety-related decay heat removal systems except at the beginning of an outage. (authors)

  8. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Anderson, H.L.

    1960-09-20

    A nuclear reactor is described comprising fissionable material dispersed in graphite blocks, helium filling the voids of the blocks and the spaces therebetween, and means other than the helium in thermal conductive contact with the graphite for removing heat.

  9. Suitability of Palestine salt dome, Anderson Co. , Texas for disposal of high-level radioactive waste

    SciTech Connect (OSTI)

    Patchick, P.F.

    1980-01-01

    The suitability of Palestine salt dome, in Anderson County, Texas, is in serious doubt for a repository to isolate high-level nuclear waste because of abandoned salt brining operations. The random geographic and spatial occurrence of 15 collapse sinks over the dome may prevent safe construction of the necessary surface installations for a repository. The dissolution of salt between the caprock and dome, from at least 15 brine wells up to 500 feet deep, may permit increased rates of salt dissolution long into future geologic time. The subsurface dissolution is occurring at a rate difficult, if not impossible, to assess or to calculate. It cannot be shown that this dissolution rate is insignificant to the integrity of a future repository or to ancillary features. The most recent significant collapse was 36 feet in diameter and took place in 1972. The other collapses ranged from 27 to 105 feet in diameter and from 1.5 to more than 15 feet in depth. ONWI recommends that this dome be removed from consideration as a candidate site.

  10. Integrated Removal of NOx with Carbon Monoxide as Reductant, and Capture of Mercury in a Low Temperature Selective Catalytic and Adsorptive Reactor

    SciTech Connect (OSTI)

    Neville Pinto; Panagiotis Smirniotis; Stephen Thiel

    2010-08-31

    Coal will likely continue to be a dominant component of power generation in the foreseeable future. This project addresses the issue of environmental compliance for two important pollutants: NO{sub x} and mercury. Integration of emission control units is in principle possible through a Low Temperature Selective Catalytic and Adsorptive Reactor (LTSCAR) in which NO{sub x} removal is achieved in a traditional SCR mode but at low temperature, and, uniquely, using carbon monoxide as a reductant. The capture of mercury is integrated into the same process unit. Such an arrangement would reduce mercury removal costs significantly, and provide improved control for the ultimate disposal of mercury. The work completed in this project demonstrates that the use of CO as a reductant in LTSCR is technically feasible using supported manganese oxide catalysts, that the simultaneous warm-gas capture of elemental and oxidized mercury is technically feasible using both nanostructured chelating adsorbents and ceria-titania-based materials, and that integrated removal of mercury and NO{sub x} is technically feasible using ceria-titania-based materials.

  11. Simulation of decay heat removal by natural convection in a pool type fast reactor model-ramona-with coupled 1D/2D thermal hydraulic code system

    SciTech Connect (OSTI)

    Kasinathan, N.; Rajakumar, A.; Vaidyanathan, G.; Chetal, S.C.

    1995-09-01

    Post shutdown decay heat removal is an important safety requirement in any nuclear system. In order to improve the reliability of this function, Liquid metal (sodium) cooled fast breeder reactors (LMFBR) are equipped with redundant hot pool dipped immersion coolers connected to natural draught air cooled heat exchangers through intermediate sodium circuits. During decay heat removal, flow through the core, immersion cooler primary side and in the intermediate sodium circuits are also through natural convection. In order to establish the viability and validate computer codes used in making predictions, a 1:20 scale experimental model called RAMONA with water as coolant has been built and experimental simulation of decay heat removal situation has been performed at KfK Karlsruhe. Results of two such experiments have been compiled and published as benchmarks. This paper brings out the results of the numerical simulation of one of the benchmark case through a 1D/2D coupled code system, DHDYN-1D/THYC-2D and the salient features of the comparisons. Brief description of the formulations of the codes are also included.

  12. Power Burst Facility (PBF) Reactor Reactor Decommissioning

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

    Click here to view Click here to view Reactor Decommissioning Click on an image to enlarge A crane removes the reactor vessel from the Power Burst Facility (top), then places it ...

  13. Geohydrology of the Keechi, Mount Sylvan, Oakwood, and Palestine salt domes in the northeast Texas salt-dome basin

    SciTech Connect (OSTI)

    Carr, J.E.; Halasz, S.J.; Peters, H.B.

    1980-01-01

    The salt within these domes has penetrated as much as 20,000 feet of Mesozoic and Cenozoic strata, and presently extends to within 120 to 800 feet of the land surface. The salt penetrates or closely underlies major freshwater and salinewater aquifers within the basin. To provide a safe repository for radioactive wastes within one or more of these domes, a thorough understanding of the geohydrology needs to be obtained, and the hydrologic stability of the domes needs to be established for the expected life of the storage facility. Dissolution may exist at all four candidate salt domes, possibly through contact with Cretaceous or Tertiary aquifers, or through fault systems in the vicinity of the domes. Strata overlying and surrounding Palestine and Keechi Salt Domes have been arched into steeply-dipping folds that are complexly faulted. Similar conditions exist at Oakwood and Mount Sylvan Domes, except that the Tertiary strata have been only moderately disturbed. Additional problems concerning the hydrologic stability of Oakwood and Palestine Salt Domes have resulted from the disposal of oil-field salinewater in the cap rock at the Oakwood Dome and previous solution mining of salt at the Palestine Dome.

  14. Packed-Bed Reactor Study of NETL Sample 196c for the Removal of Carbon Dioxide from Simulated Flue Gas Mixture

    SciTech Connect (OSTI)

    Hoffman, James S.; Hammache, Sonia; Gray, McMahan L.; Fauth Daniel J.; Pennline, Henry W.

    2012-04-24

    An amine-based solid sorbent process to remove CO2 from flue gas has been investigated. The sorbent consists of polyethylenimine (PEI) immobilized onto silica (SiO2) support. Experiments were conducted in a packed-bed reactor and exit gas composition was monitored using mass spectrometry. The effects of feed gas composition (CO2 and H2O), temperature, and simulated steam regeneration were examined for both the silica support as well as the PEI-based sorbent. The artifact of the empty reactor was also quantified. Sorbent CO2 capacity loading was compared to thermogravimetric (TGA) results to further characterize adsorption isotherms and better define CO2 working capacity. Sorbent stability was monitored by periodically repeating baseline conditions throughout the parametric testing and replacing with fresh sorbent as needed. The concept of the Basic Immobilized Amine Sorbent (BIAS) Process using this sorbent within a system where sorbent continuously flows between the absorber and regenerator was introduced. The basic tenet is to manipulate or control the level of moisture on the sorbent as it travels around the sorbent circulation path between absorption and regeneration stages to minimize its effect on regeneration heat duty.

  15. Identifying suitable "piercement" salt domes for nuclear waste storage sites

    SciTech Connect (OSTI)

    Kehle, R.

    1980-08-01

    Piercement salt domes of the northern interior salt basins of the Gulf of Mexico are being considered as permanent storage sites for both nuclear and chemically toxic wastes. The suitable domes are stable and inactive, having reached their final evolutionary configuration at least 30 million years ago. They are buried to depths far below the level to which erosion will penetrate during the prescribed storage period and are not subject to possible future reactivation. The salt cores of these domes are themselves impermeable, permitting neither the entry nor exit of ground water or other unwanted materials. In part, a stable dome may be recognized by its present geometric configuration, but conclusive proof depends on establishing its evolutionary state. The evolutionary state of a dome is obtained by reconstructing the growth history of the dome as revealed by the configuration of sedimentary strata in a large area (commonly 3,000 square miles or more) surrounding the dome. A high quality, multifold CDP reflection seismic profile across a candidate dome will provide much of the necessary information when integrated with available subsurface control. Additional seismic profiles may be required to confirm an apparent configuration of the surrounding strata and an interpreted evolutionary history. High frequency seismic data collected in the near vicinity of a dome are also needed as a supplement to the CDP data to permit accurate depiction of the configuration of shallow strata. Such data must be tied to shallow drill hole control to confirm the geologic age at which dome growth ceased. If it is determined that a dome reached a terminal configuration many millions of years ago, such a dome is incapable of reactivation and thus constitutes a stable storage site for nuclear wastes.

  16. In Situ NDA Conformation Measurements Performed at Auxiliary Charcoal Bed and Other Main Charcoal Beds After Uranium Removal from Molten Salt Reactor Experiment ACB at Oak Ridge National Laboratory

    SciTech Connect (OSTI)

    Haghighi, M. H.; Kring, C. T.; McGehee, J. T.; Jugan, M. R.; Chapman, J.; Meyer, K. E.

    2002-02-26

    The Molten Salt Reactor Experiment (MSRE) site is located in Tennessee, on the U.S. Department of Energy (DOE) Oak Ridge Reservation (ORR). The MSRE was run by Oak Ridge National Laboratory (ORNL) to demonstrate the desirable features of the molten-salt concept in a practical reactor that could be operated safely and reliably. It introduced the idea of a homogeneous reactor using fuel salt media and graphite moderation for power and breeder reactors. The MSRE reactor and associated components are located in cells beneath the floor in the high-bay area of Building 7503. The reactor was operated from June 1965 to December 1969. When the reactor was shut down, fuel salt was drained from the reactor circuit to two drain tanks. A ''clean'' salt was then circulated through the reactor as a decontamination measure and drained to a third drain tank. When operations ceased, the fuel and flush salts were allowed to cool and solidify in the drain tanks. At shutdown, the MSRE facility complex was placed in a surveillance and maintenance program. Beginning in 1987, it was discovered that gaseous uranium (U-233/U-232) hexafluoride (UF6) had moved throughout the MSRE process systems. The UF6 had been generated when radiolysis in the fluorine salts caused the individual constituents to dissociate to their component atoms, including free fluorine. Some of the free fluorine combined with uranium fluorides (UF4) in the salt to produce UF6. UF6 is gaseous at slightly above ambient temperatures; thus, periodic heating of the fuel salts (which was intended to remedy the radiolysis problems) and simple diffusion had allowed the UF6 to move out of the salt and into the process systems of MSRE. One of the systems that UF6 migrated into due to this process was the offgas system which is vented to the MSRE main charcoal beds and MSRE auxiliary charcoal bed (ACB). Recently, the majority of the uranium laden-charcoal material residing within the ACB was safely and successfully removed using

  17. Inferences On The Hydrothermal System Beneath The Resurgent Dome...

    Open Energy Info (EERE)

    Inferences On The Hydrothermal System Beneath The Resurgent Dome In Long Valley Caldera, East-Central California, USA, From Recent Pumping Tests And Geochemical Sampling Jump to:...

  18. COMBINED ACTIVE/PASSIVE DECAY HEAT REMOVAL APPROACH FOR THE 24 MWt GAS-COOLED FAST REACTOR

    SciTech Connect (OSTI)

    CHENG,L.Y.; LUDEWIG, H.

    2007-06-01

    Decay heat removal at depressurized shutdown conditions has been regarded as one of the key areas where significant improvement in passive response was targeted for the GEN IV GFR over the GCFR designs of thirty years ago. It has been recognized that the poor heat transfer characteristics of gas coolant at lower pressures needed to be accommodated in the GEN IV design. The design envelope has therefore been extended to include a station blackout sequence simultaneous with a small break/leak. After an exploratory phase of scoping analysis in this project, together with CEA of France, it was decided that natural convection would be selected as the passive decay heat removal approach of preference. Furthermore, a double vessel/containment option, similar to the double vessel/guard vessel approach of the SFR, was selected as the means of design implementation to reduce the PRA risks of the depressurization accident. However additional calculations in conjunction with CEA showed that there was an economic penalty in terms of decay heat removal system heat exchanger size, elevation heights for thermal centers, and most of all in guard containment back pressure for complete reliance on natural convection only. The back pressure ranges complicated the design requirements for the guard containment. Recognizing that the definition of a loss-of-coolant-accident in the GFR is a misnomer, since gas coolant will always be present, and the availability of some driven blower would reduce fuel temperature transients significantly; it was decided instead to aim for a hybrid active/passive combination approach to the selected BDBA. Complete natural convection only would still be relied on for decay heat removal but only after the first twenty four hours after the initiation of the accident. During the first twenty four hour period an actively powered blower would be relied on to provide the emergency decay power removal. However the power requirements of the active blower

  19. SOLUTION MINING IN SALT DOMES OF THE GULF COAST EMBAYMENT

    SciTech Connect (OSTI)

    Griswold, G. B.

    1981-02-01

    Following a description of salt resources in the salt domes of the gulf coast embayment, mining, particularly solution mining, is described. A scenario is constructed which could lead to release of radioactive waste stored in a salt dome via inadvertent solution mining and the consequences of this scenario are analyzed.

  20. Recovery Act Workers Clear Reactor Shields from Brookhaven Lab

    Office of Environmental Management (EM)

    final stage of decommissioning a nuclear reactor after they recently removed thick steel ... the Brookhaven Graphite Research Reactor, the world's first reactor built solely ...

  1. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Breden, C.R.; Dietrich, J.R.

    1961-06-20

    A water-soluble non-volatile poison may be introduced into a reactor to nullify excess reactivity. The poison is removed by passing a side stream of the water containing the soluble poison to an evaporation chamber. The vapor phase is returned to the reactor to decrease the concentration of soluble poison and the liquid phase is returned to increase the concentration of soluble poison.

  2. Reactor pressure vessel vented head

    DOE Patents [OSTI]

    Sawabe, J.K.

    1994-01-11

    A head for closing a nuclear reactor pressure vessel shell includes an arcuate dome having an integral head flange which includes a mating surface for sealingly mating with the shell upon assembly therewith. The head flange includes an internal passage extending therethrough with a first port being disposed on the head mating surface. A vent line includes a proximal end disposed in flow communication with the head internal passage, and a distal end disposed in flow communication with the inside of the dome for channeling a fluid therethrough. The vent line is fixedly joined to the dome and is carried therewith when the head is assembled to and disassembled from the shell. 6 figures.

  3. Energy Department Sells Historic Teapot Dome Oilfield | Department...

    Energy Savers [EERE]

    to Stranded Oil Resources Corporation, a subsidiary of Alleghany Capital Corporation. ... Teapot Dome, consisting of 9,481 acres, was set aside as a naval oil reserve in 1915, and ...

  4. Internal Geology and Evolution of the Redondo Dome, Valles Caldera...

    Open Energy Info (EERE)

    dome. A comparison of the uplift with a model for formation of the laccoliths of the Henry Mountains indicated the magma was 4700 m thick, in line with the fact that the 3243 m...

  5. Phase Competition in Trisected Superconducting Dome (Journal Article) |

    Office of Scientific and Technical Information (OSTI)

    SciTech Connect Journal Article: Phase Competition in Trisected Superconducting Dome Citation Details In-Document Search Title: Phase Competition in Trisected Superconducting Dome Authors: Vishik, I.M. ; Hashimoto, M ; He, Rui-Hua ; Lee, Wei-Sheng ; Schmitt, Felix ; Lu, Donghui ; Moore, R.G. ; Zhang, C. ; Meevasana, W. ; Sasagawa, T. ; Uchida, S. ; Fujita, Kazuhiro ; Ishida, S. ; Ishikado, M. ; Yoshida, Yoshiyuki ; Eisaki, Hiroshi ; Hussain, Zaheed ; Devereaux, Thomas P. ; Shen, Zhi-Xun

  6. LANL demolishes first containment dome at disposal area

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

    LANL Demolishes First Containment Dome LANL demolishes first containment dome at disposal area It once housed thousands of drums of radioactive waste that have been shipped to the Waste Isolation Pilot Plant for disposal. September 30, 2009 Los Alamos National Laboratory sits on top of a once-remote mesa in northern New Mexico with the Jemez mountains as a backdrop to research and innovation covering multi-disciplines from bioscience, sustainable energy sources, to plasma physics and new

  7. Dome houses and energy conservation: an introductory bibliography. [38 references to dome efficiency

    SciTech Connect (OSTI)

    Not Available

    1983-01-01

    The appearance of geodesic domes in conventional neighborhoods is recent. The current popularity of these spherical designs is due to their energy efficiency. Some manufacturers have claimed over 40% efficiency improvement over conventional homes of the same size. A host of low utility bills across the country is now backing up these claims. This bibliography concentrates on the period from 1960 to the present, although there are a few entries from earlier periods. Most of the material is available in articles rather than books.

  8. 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.

  9. Subsidence at Boling salt dome: results of multiple resource production

    SciTech Connect (OSTI)

    Mullican, W.F. III

    1988-02-01

    Boling dome (Wharton and Fort Bend Counties) has experienced more overall subsidence and collapse than any other dome in Texas. These processes are directly related to production of sulfur and hydrocarbons from the southeastern quadrant of the dome. Greatest vertical movement due to subsidence and collapse is 35 ft (based on the Boling 7.5 min topographic map, last surveyed in 1953). Most of the subsidence (83%) is attributed to sulfur production, whereas only 11 to 12% can be linked to hydrocarbon production. Reservoir compaction is the dominant mechanism of land subsidence in areas of hydrocarbon production at Boling dome. Trough subsidence, chimneying, plug caving, and piping are the characteristic mechanisms over sulfur fields developed at the salt dome. The structural and hydrologic stability of the surface and subsurface at Boling dome is compromised by these active deformation processes. Damage to pipelines and well-casing strings may result in costly leaks which have the potential of being uncontrollable and catastrophic. Reduction in hydrologic stability may result if natural aquitards are breached and fresh water mixes with saline water or if hydrologic conduits to the diapir are opened, allowing unrestricted dissolution of the salt stock.

  10. NEUTRONIC REACTOR POWER PLANT

    DOE Patents [OSTI]

    Metcalf, H.E.

    1962-12-25

    This patent relates to a nuclear reactor power plant incorporating an air-cooled, beryllium oxide-moderated, pebble bed reactor. According to the invention means are provided for circulating a flow of air through tubes in the reactor to a turbine and for directing a sidestream of the circu1ating air through the pebble bed to remove fission products therefrom as well as assist in cooling the reactor. (AEC)

  11. Vertical Extraction Process Implemented at the 118-K-1 Burial Ground for Removal of Irradiated Reactor Debris from Silo Structures - 12431

    SciTech Connect (OSTI)

    Teachout, Douglas B.; Adamson, Clinton J.; Zacharias, Ames

    2012-07-01

    The primary objective of a remediation project is the safe extraction and disposition of diverse waste forms and materials. Remediation of a solid waste burial ground containing reactor hardware and irradiated debris involves handling waste with the potential to expose workers to significantly elevated dose rates. Therefore, a major challenge confronted by any remediation project is developing work processes that facilitate compliant waste management practices while at the same time implementing controls to protect personnel. Traditional burial ground remediation is accomplished using standard excavators to remove materials from trenches and other excavation configurations often times with minimal knowledge of waste that will be encountered at a specific location. In the case of the 118-K-1 burial ground the isotopic activity postulated in historic documents to be contained in vertical cylindrical silos was sufficient to create the potential for a significant radiation hazard to project personnel. Additionally, certain reported waste forms posed an unacceptably high potential to contaminate the surrounding environment and/or workers. Based on process knowledge, waste management requirements, historic document review, and a lack of characterization data it was determined that traditional excavation techniques applied to remediation of vertical silos would expose workers to unacceptable risk. The challenging task for the 118-K-1 burial ground remediation project team then became defining an acceptable replacement technology or modification of an existing technology to complete the silo remediation. Early characterization data provided a good tool for evaluating the location of potential high exposure rate items in the silos. Quantitative characterization was a different case and proved difficult because of the large diameter of the silos and the potential for variable density of attenuating soils and waste forms in the silo. Consequently, the most relevant

  12. Chris Bergren Director, Environment ...

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

    Reactor Dome Removal Reactor Vessel Removal Demolition of Dome After 3 K-Area Cooling Tower Implosion K-Area Cooling Tower at Completion D & D of K-Area Cooling Tower 4 In...

  13. Geodesic-dome tank roof cuts water contamination, vapor losses

    SciTech Connect (OSTI)

    Barrett, A.E. )

    1989-07-10

    Colonial Pipeline Co. has established an ongoing program for using geodesic-dome roofs on tanks in liquid petroleum-product service. As its standard, Colonial adopted geodesicodone roofs, in conjunction with internal floating decks, to replace worn external floating roofs on existing tanks used in gasoline service and for use on new tanks in all types of product service. Geodesic domes are clear-span structures requiring no internal-support columns. This feature allows the associated use of a floating deck that is as vapor tight as is possible to construct. Further, geodesic domes can practically eliminate rainwater contamination, eliminate wind-generated vapor losses, and greatly reduce filling losses associated with conventional external floating roofs.

  14. FLAMMABLE GAS DIFFUSION THROUGH SINGLE SHELL TANK (SST) DOMES

    SciTech Connect (OSTI)

    MEACHAM, J.E.

    2003-11-10

    This report quantified potential hydrogen diffusion through Hanford Site Single-Shell tank (SST) domes if the SSTs were hypothetically sealed airtight. Results showed that diffusion would keep headspace flammable gas concentrations below the lower flammability limit in the 241-AX and 241-SX SST. The purpose of this document is to quantify the amount of hydrogen that could diffuse through the domes of the SSTs if they were hypothetically sealed airtight. Diffusion is assumed to be the only mechanism available to reduce flammable gas concentrations. The scope of this report is limited to the 149 SSTs.

  15. Progress Update: Reactor Disassembly Grouting

    SciTech Connect (OSTI)

    Cody, Tom

    2010-01-01

    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.

  16. 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.

  17. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Wigner, E.P.; Young, G.J.

    1958-10-14

    A method is presented for loading and unloading rod type fuel elements of a neutronic reactor of the heterogeneous, solld moderator, liquid cooled type. In the embodiment illustrated, the fuel rods are disposed in vertical coolant channels in the reactor core. The fuel rods are loaded and unloaded through the upper openings of the channels which are immersed in the coolant liquid, such as water. Unloading is accomplished by means of a coffer dam assembly having an outer sleeve which is placed in sealing relation around the upper opening. A radiation shield sleeve is disposed in and reciprocable through the coffer dam sleeve. A fuel rod engaging member operates through the axial bore in the radiation shield sleeve to withdraw the fuel rod from its position in the reactor coolant channel into the shield, the shield snd rod then being removed. Loading is accomplished in the reverse procedure.

  18. Let's Try That Again: Selling the Teapot Dome Oil Field | Department of

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

    Energy Let's Try That Again: Selling the Teapot Dome Oil Field Let's Try That Again: Selling the Teapot Dome Oil Field January 30, 2015 - 11:28am Addthis A solitary oil pump at the Teapot Dome Oilfield in Wyoming. | Department of Energy photo. A solitary oil pump at the Teapot Dome Oilfield in Wyoming. | Department of Energy photo. Allison Lantero Allison Lantero Digital Content Specialist, Office of Public Affairs In 1922, President Warren Harding's Interior Secretary Albert Fall found

  19. 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

  20. Nuclear reactor

    DOE Patents [OSTI]

    Pennell, William E.; Rowan, William J.

    1977-01-01

    A nuclear reactor in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assmblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters and the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters and the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance.

  1. Geological evaluation of Gulf Coast salt domes: overall assessment of the Gulf Interior Region

    SciTech Connect (OSTI)

    1981-10-01

    The three major phases in site characterization and selection are regional studies, area studies, and location studies. This report characterizes regional geologic aspects of the Gulf Coast salt dome basins. It includes general information from published sources on the regional geology; the tectonic, domal, and hydrologic stability; and a brief description the salt domes to be investigated. After a screening exercise, eight domes were chosen for further characterization: Keechi, Oakwood, and Palestine Domes in Texas; Vacherie and Rayburn's domes in North Louisiana; and Cypress Creek and Richton domes in Mississippi. A general description of each, maps of the location, property ownership, and surface geology, and a geologic cross section were presented for each dome.

  2. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Treshow, M.

    1958-08-19

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

  3. Brookhaven Lab Completes Decommissioning of Graphite Research Reactor:

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

    Reactor core and associated structures successfully removed; waste shipped offsite for disposal | Department of Energy Brookhaven Lab Completes Decommissioning of Graphite Research Reactor: Reactor core and associated structures successfully removed; waste shipped offsite for disposal Brookhaven Lab Completes Decommissioning of Graphite Research Reactor: Reactor core and associated structures successfully removed; waste shipped offsite for disposal September 1, 2012 - 12:00pm Addthis The

  4. Conceptual model for regional radionuclide transport from a salt dome repository: a technical memorandum

    SciTech Connect (OSTI)

    Kier, R.S.; Showalter, P.A.; Dettinger, M.D.

    1980-05-30

    Disposal of high-level radioactive wastes is a major environmental problem influencing further development of nuclear energy in this country. Salt domes in the Gulf Coast Basin are being investigated as repository sites. A major concern is geologic and hydrologic stability of candidate domes and potential transport of radionuclides by groundwater to the biosphere prior to their degradation to harmless levels of activity. This report conceptualizes a regional geohydrologic model for transport of radionuclides from a salt dome repository. The model considers transport pathways and the physical and chemical changes that would occur through time prior to the radionuclides reaching the biosphere. Necessary, but unknown inputs to the regional model involve entry and movement of fluids through the repository dome and across the dome-country rock interface and the effect on the dome and surrounding strata of heat generated by the radioactive wastes.

  5. DOE - Office of Legacy Management -- Tatum Salt Dome Test Site - MS 01

    Office of Legacy Management (LM)

    Tatum Salt Dome Test Site - MS 01 Site ID (CSD Index Number): MS.01 Site Name: Tatum Salt Dome Test Site Site Summary: Site Link: http://www.lm.doe.gov/salmon/Sites.aspx External Site Link: Alternate Name(s): Tatum Salt Dome Test Site Alternate Name Documents: Location: Salmon, Mississippi Location Documents: Historical Operations (describe contaminants): Underground nuclear test site Historical Operations Documents: Eligibility Determination: Remediated by DOE Eligibility Determination

  6. Draft environmental assessment: Richton Dome site, Mississippi. Nuclear Waste Policy Act (Section 112). [Contains Glossary

    SciTech Connect (OSTI)

    Not Available

    1984-12-01

    In February 1983, the US Department of Energy identified the Richton dome site as one of the nine potentially acceptable sites for a mined geo

  7. Field Survey of Cactus Crater Storage Facility (Runit Dome)

    SciTech Connect (OSTI)

    Douglas Miller, Terence Holland

    2008-10-31

    The US Department of Energy, Office of Health and Safety (DOE/HS-10), requested that National Security Technologies, LLC, Environmental Management directorate (NSTec/EM) perform a field survey of the Cactus Crater Storage Facility (Runit Dome), similar to past surveys conducted at their request. This field survey was conducted in conjunction with a Lawrence Livermore National Laboratory (LLNL) mission on Runit Island in the Enewetak Atoll in the Republic of the Marshall Islands (RMI). The survey was strictly a visual survey, backed up by digital photos and a written description of the current condition.

  8. Steam turbine: Alternative emergency drive for the secure removal of residual heat from the core of light water reactors in ultimate emergency situation

    SciTech Connect (OSTI)

    Souza Dos Santos, R.

    2012-07-01

    In 2011 the nuclear power generation has suffered an extreme probation. That could be the meaning of what happened in Fukushima Nuclear Power Plants. In those plants, an earthquake of 8.9 on the Richter scale was recorded. The quake intensity was above the trip point of shutting down the plants. Since heat still continued to be generated, the procedure to cooling the reactor was started. One hour after the earthquake, a tsunami rocked the Fukushima shore, degrading all cooling system of plants. Since the earthquake time, the plant had lost external electricity, impacting the pumping working, drive by electric engine. When operable, the BWR plants responded the management of steam. However, the lack of electricity had degraded the plant maneuvers. In this paper we have presented a scheme to use the steam as an alternative drive to maintain operable the cooling system of nuclear power plant. This scheme adds more reliability and robustness to the cooling systems. Additionally, we purposed a solution to the cooling in case of lacking water for the condenser system. In our approach, steam driven turbines substitute electric engines in the ultimate emergency cooling system. (authors)

  9. material removal

    National Nuclear Security Administration (NNSA)

    %2A en Nuclear Material Removal http:www.nnsa.energy.govaboutusourprogramsdnnm3remove

    Pag...

  10. material removal

    National Nuclear Security Administration (NNSA)

    %2A en Nuclear Material Removal http:nnsa.energy.govaboutusourprogramsdnnm3remove

    Page...

  11. Actinide removal from spent salts

    DOE Patents [OSTI]

    Hsu, Peter C.; von Holtz, Erica H.; Hipple, David L.; Summers, Leslie J.; Adamson, Martyn G.

    2002-01-01

    A method for removing actinide contaminants (uranium and thorium) from the spent salt of a molten salt oxidation (MSO) reactor is described. Spent salt is removed from the reactor and analyzed to determine the contaminants present and the carbonate concentration. The salt is dissolved in water, and one or more reagents are added to precipitate the thorium as thorium oxide and/or the uranium as either uranium oxide or as a diuranate salt. The precipitated materials are filtered, dried and packaged for disposal as radioactive waste. About 90% of the thorium and/or uranium present is removed by filtration. After filtration, salt solutions having a carbonate concentration >20% can be dried and returned to the reactor for re-use. Salt solutions containing a carbonate concentration <20% require further clean-up using an ion exchange column, which yields salt solutions that contain less than 0.1 ppm of thorium or uranium.

  12. Metals removal from spent salts

    DOE Patents [OSTI]

    Hsu, Peter C.; Von Holtz, Erica H.; Hipple, David L.; Summers, Leslie J.; Brummond, William A.; Adamson, Martyn G.

    2002-01-01

    A method and apparatus for removing metal contaminants from the spent salt of a molten salt oxidation (MSO) reactor is described. Spent salt is removed from the reactor and analyzed to determine the contaminants present and the carbonate concentration. The salt is dissolved in water, and one or more reagents may be added to precipitate the metal oxide and/or the metal as either metal oxide, metal hydroxide, or as a salt. The precipitated materials are filtered, dried and packaged for disposal as waste or can be immobilized as ceramic pellets. More than about 90% of the metals and mineral residues (ashes) present are removed by filtration. After filtration, salt solutions having a carbonate concentration >20% can be spray-dried and returned to the reactor for re-use. Salt solutions containing a carbonate concentration <20% require further clean-up using an ion exchange column, which yields salt solutions that contain less than 1.0 ppm of contaminants.

  13. Hanford Deep Dig Removes Contaminated Soil | Department of Energy

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

    An aerial view of Hanfords D Area shows the D Reactor (lower left) and DR Reactor. Workers are digging 85 feet to groundwater at two sites there to remove chromium ...

  14. Sensitivity of storage field performance to geologic and cavern design parameters in salt domes.

    SciTech Connect (OSTI)

    Ehgartner, Brian L.; Park, Byoung Yoon

    2009-03-01

    A sensitivity study was performed utilizing a three dimensional finite element model to assess allowable cavern field sizes for strategic petroleum reserve salt domes. A potential exists for tensile fracturing and dilatancy damage to salt that can compromise the integrity of a cavern field in situations where high extraction ratios exist. The effects of salt creep rate, depth of salt dome top, dome size, caprock thickness, elastic moduli of caprock and surrounding rock, lateral stress ratio of surrounding rock, cavern size, depth of cavern, and number of caverns are examined numerically. As a result, a correlation table between the parameters and the impact on the performance of storage field was established. In general, slower salt creep rates, deeper depth of salt dome top, larger elastic moduli of caprock and surrounding rock, and a smaller radius of cavern are better for structural performance of the salt dome.

  15. Nuclear reactor reflector

    DOE Patents [OSTI]

    Hopkins, Ronald J.; Land, John T.; Misvel, Michael C.

    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.

  16. 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.

  17. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Daniels, F.

    1962-12-18

    A power plant is described comprising a turbine and employing round cylindrical fuel rods formed of BeO and UO/sub 2/ and stacks of hexagonal moderator blocks of BeO provided with passages that loosely receive the fuel rods so that coolant may flow through the passages over the fuels to remove heat. The coolant may be helium or steam and fiows through at least one more heat exchanger for producing vapor from a body of fluid separate from the coolant, which fluid is to drive the turbine for generating electricity. By this arrangement the turbine and directly associated parts are free of particles and radiations emanating from the reactor. (AEC)

  18. NNSA Partnership Successfully Removes All Remaining HEU from...

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

    ... research reactors and medical isotope production facilities to the use of LEU, removing excess HEU and separated plutonium, and dispositioning HEU and plutonium domestically. ...

  19. Selective purge for hydrogenation reactor recycle loop

    DOE Patents [OSTI]

    Baker, Richard W.; Lokhandwala, Kaaeid A.

    2001-01-01

    Processes and apparatus for providing improved contaminant removal and hydrogen recovery in hydrogenation reactors, particularly in refineries and petrochemical plants. The improved contaminant removal is achieved by selective purging, by passing gases in the hydrogenation reactor recycle loop or purge stream across membranes selective in favor of the contaminant over hydrogen.

  20. NEUTRONIC REACTOR FUEL PUMP

    DOE Patents [OSTI]

    Cobb, W.G.

    1959-06-01

    A reactor fuel pump is described which offers long life, low susceptibility to radiation damage, and gaseous fission product removal. An inert-gas lubricated bearing supports a journal on one end of the drive shsft. The other end has an impeller and expansion chamber which effect pumping and gas- liquid separation. (T.R.H.)

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. Water-quality data for aquifers, streams, and lakes in the vicinity of Keechi, Mount Sylvan, Oakwood, and Palestine salt domes, northeast Texas salt-dome basin

    SciTech Connect (OSTI)

    Carr, J.E.; Halasz, S.J.; Liscum, F.

    1980-11-01

    This report contains water-quality data for aquifers, streams, and lakes in the vicinity of Keechi, Mount Sylvan, Oakwood, and Palestine Salt Domes in the northeast Texas salt-dome basin. Water-quality data were compiled for aquifers in the Wilcox Group, the Carrizo Sand, and the Queen City Sand. The data include analyses for dissolved solids, pH, temperature, hardness, calcium, magnesium, sodium, bicarbonate, chloride, and sulfate. Water-quality and streamflow data were obtained from 63 surface-water sites in the vicinity of the domes. These data include water discharge, specific conductance, pH, water temperature, and dissolved oxygen. Samples were collected at selected sites for analysis of principal and selected minor dissolved constituents.

  8. RADIATION FACILITY FOR NUCLEAR REACTORS

    DOE Patents [OSTI]

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

    1961-12-12

    A radiation facility is designed for irradiating samples in close proximity to the core of a nuclear reactor. The facility comprises essentially a tubular member extending through the biological shield of the reactor and containing a manipulatable rod having the sample carrier at its inner end, the carrier being longitudinally movable from a position in close proximity to the reactor core to a position between the inner and outer faces of the shield. Shield plugs are provided within the tubular member to prevent direct radiation from the core emanating therethrough. In this device, samples may be inserted or removed during normal operation of the reactor without exposing personnel to direct radiation from the reactor core. A storage chamber is also provided within the radiation facility to contain an irradiated sample during the period of time required to reduce the radioactivity enough to permit removal of the sample for external handling. (AEC)

  9. The Thermal Regime In The Resurgent Dome Of Long Valley Caldera...

    Open Energy Info (EERE)

    Jump to: navigation, search OpenEI Reference LibraryAdd to library Journal Article: The Thermal Regime In The Resurgent Dome Of Long Valley Caldera, California- Inferences From...

  10. EIS-0010: Strategic Petroleum Reserves, Sulphur Mines Salt Dome, Calcasieu Parish, Louisiana

    Broader source: Energy.gov [DOE]

    The Strategic Petroleum Reserves prepared this EIS to assess the environmental impacts of the proposed storage of 24 million barrels of crude oil at the Sulphur Mines salt dome located in Calcasieu Parish, Louisiana, including construction and operation impacts.

  11. The Thermal Regime in the Resurgent Dome of Long Valley Caldera...

    Open Energy Info (EERE)

    in the Resurgent Dome of Long Valley Caldera, California: Inferences from Precision Temperature Logs in Deep Wells Jump to: navigation, search OpenEI Reference LibraryAdd to...

  12. Fluid Flow In The Resurgent Dome Of Long Valley Caldera- Implications...

    Open Energy Info (EERE)

    Flow In The Resurgent Dome Of Long Valley Caldera- Implications From Thermal Data And Deep Electrical Sounding Jump to: navigation, search OpenEI Reference LibraryAdd to library...

  13. 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.

  14. METHOD AND APPARATUS FOR CONTROLLING DIRECT-CYCLE NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Reed, G.A.

    1961-01-10

    A control arrangement is offered for a boiling-water reactor. Boric acid is maintained in the water in the reactor and the amount in the reactor is controlled by continuously removing a portion of the water from the reactor, concentrating the boric acid by evaporating the water therefrom, returning a controlled amount of the acid to the reactor, and simultaneously controlling the water level by varying the rate of spent steam return to the reactor.

  15. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Fermi, E.; Zinn, W.H.; Anderson, H.L.

    1958-09-16

    Means are presenied for increasing the reproduction ratio of a gaphite- moderated neutronic reactor by diminishing the neutron loss due to absorption or capture by gaseous impurities within the reactor. This means comprised of a fluid-tight casing or envelope completely enclosing the reactor and provided with a valve through which the casing, and thereby the reactor, may be evacuated of atmospheric air.

  16. WA_03_040_UNITED_TECHNOLOGIES_RESEARCH_CENTER_Waiver_of_Dome.pdf |

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

    Department of Energy 40_UNITED_TECHNOLOGIES_RESEARCH_CENTER_Waiver_of_Dome.pdf WA_03_040_UNITED_TECHNOLOGIES_RESEARCH_CENTER_Waiver_of_Dome.pdf (705.58 KB) More Documents & Publications WA_02_054_ADVANCED_TECHNLOGY_MATERIALS_Waiver_of_Domestic_an.pdf WA_02_038_UNITED_TECHNOLOGIES_CORP_Waiver_of_Domestic_and_Fo.pdf Advance Patent Waiver W(A)2006-021

  17. Domed community and several alternatives for Winooski, Vermont: the environmental, organizational, and energy conservation issues

    SciTech Connect (OSTI)

    Wendt, R.L.

    1980-01-01

    The environmental, organizational, and energy conservation issues related to a domed structure enveloping Winooski, Vermont, are discussed. Alternative means of accomplishing energy conservation will be addressed. These include retrofitting of existing structures, replacement with state-of-the-art structures, the use of planting shelter-belts, redevelopment to an earth-sheltered community, and redevelopment to a composite domed neighborhood and earth-sheltered community. The assets and liabilities of each alternative are addressed.

  18. 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.

  19. Liquid metal cooled nuclear reactor plant system

    DOE Patents [OSTI]

    Hunsbedt, Anstein; Boardman, Charles E.

    1993-01-01

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

  20. A user`s perspective on aluminum dome roofs for aboveground tanks

    SciTech Connect (OSTI)

    Myers, P.E.

    1995-12-31

    There is a trend in the petroleum industry to install aluminum dome roofs on storage tanks of all kinds. Although most dome roofs have been installed on floating roof tanks, there is a trend to install them on fixed roof tanks as well, substituting the familiar shallow fixed cone roof with a geodesic dome. In part, this trend has been caused by EPA requirements causing a greater number of closed tanks to be vented to vapor recovery or vapor destruction systems. Both the aluminum roof manufacturing community and the user have moved into a whole new set of problems associated with the change in dome roof applications from atmospheric to those requiring internal pressure. New problems are just now being dealt with and solved because cost factors tend to make the aluminum dome an economic solution for many cases where sealed tank systems must be used. Because of the increased numbers of geodesic domes as either an alternative to a fixed cone roof tank or as a way to convert an external floating roof tank to an internal floating roof tank or as their potential to serve as tools in the environmental arena, it is the intent of this paper to examine them from the user`s perspective. In addition, some areas of research that should resolve some reliability and safety issues are presented for consideration and research by not only manufacturers but the users as well.

  1. Final report on decommissioning boreholes and wellsite restoration, Gulf Coast Interior Salt Domes of Mississippi

    SciTech Connect (OSTI)

    Not Available

    1989-04-01

    In 1978, eight salt domes in Texas, Louisiana, and Mississippi were identified for study as potential locations for a nuclear waste repository as part of the National Waste Terminal Storage (NWTS) program. Three domes were selected in Mississippi for ``area characterization`` phase study as follows: Lampton Dome near Columbia, Cypress Creek Dome near New Augusta, and Richton Dome near Richton. The purpose of the studies was to acquire geologic and geohydrologic information from shallow and deep drilling investigations to enable selection of sites suitable for more intensive study. Eleven deep well sites were selected for multiple-well installations to acquire information on the lithologic and hydraulic properties of regional aquifers. In 1986, the Gulf Coast salt domes were eliminated from further consideration for repository development by the selection of three candidate sites in other regions of the country. In 1987, well plugging and restoration of these deferred sites became a closeout activity. The primary objectives of this activity are to plug and abandon all wells and boreholes in accordance with state regulations, restore all drilling sites to as near original condition as feasible, and convey to landowners any wells on their property that they choose to maintain. This report describes the activities undertaken to accomplish these objectives, as outlines in Activity Plan 1--2, ``Activity Plan for Well Plugging and Site Restoration of Test Hole Sites in Mississippi.``

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

    DOE Patents [OSTI]

    Hunsbedt, Anstein; Busboom, Herbert J.

    1991-01-01

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

  3. F Reactor Inspection

    SciTech Connect (OSTI)

    Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

    2014-10-29

    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."

  4. 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."

  5. Plasma generators, reactor systems and related methods - Energy...

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

    and associated systems and methods are provided in accordance with the present invention. A plasma reactor may include multiple sections or modules which are removably...

  6. Design of megawatt power level heat pipe reactors (Technical...

    Office of Scientific and Technical Information (OSTI)

    pipe reactors An important niche for nuclear energy is the need for power at remote locations removed from a reliable electrical grid. Nuclear energy has potential applications at ...

  7. BOILING REACTORS

    DOE Patents [OSTI]

    Untermyer, S.

    1962-04-10

    A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

  8. Mexico HEU Removal: Fact Sheet | National Nuclear Security Administration |

    National Nuclear Security Administration (NNSA)

    (NNSA) Mexico HEU Removal: Fact Sheet March 26, 2012 At the 2012 Nuclear Security Summit, the United States, Mexico and Canada announced the successful removal of HEU from Mexico and conversion of the TrigaII Research Reactor to LEU. The HEU removal and reactor conversion were completed with international cooperation from Canada, Mexico and the United States and was supported by the IAEA. The National Nuclear Security Administration (NNSA) led the US government team in executing this mission

  9. Degradation of dome cutting minerals in Hanford waste

    SciTech Connect (OSTI)

    Reynolds, Jacob G.; Huber, Heinz J.; Cooke, Gary A.

    2013-01-11

    At the Hanford Tank Farms, recent changes in retrieval technology require cutting new risers in several single-shell tanks. The Hanford Tank Farm Operator is using water jet technology with abrasive silicate minerals such as garnet or olivine to cut through the concrete and rebar dome. The abrasiveness of these minerals, which become part of the high-level waste stream, may enhance the erosion of waste processing equipment. However, garnet and olivine are not thermodynamically stable in Hanford waste, slowly degrading over time. How likely these materials are to dissolve completely in the waste before the waste is processed in the Waste Treatment and Immobilization Plant can be evaluated using theoretical analysis for olivine and collected direct experimental evidence for garnet. Based on an extensive literature study, a large number of primary silicates decompose into sodalite and cancrinite when exposed to Hanford waste. Given sufficient time, the sodalite also degrades into cancrinite. Even though cancrinite has not been directly added to any Hanford tanks during process times, it is the most common silicate observed in current Hanford waste. By analogy, olivine and garnet are expected to ultimately also decompose into cancrinite. Garnet used in a concrete cutting demonstration was immersed in a simulated supernate representing the estimated composition of the liquid retrieving waste from Hanford tank 241-C-107 at both ambient and elevated temperatures. This simulant was amended with extra NaOH to determine if adding caustic would help enhance the degradation rate of garnet. The results showed that the garnet degradation rate was highest at the highest NaOH concentration and temperature. At the end of 12 weeks, however, the garnet grains were mostly intact, even when immersed in 2 molar NaOH at 80 deg C. Cancrinite was identified as the degradation product on the surface of the garnet grains. In the case of olivine, the rate of degradation in the high-pH regimes

  10. 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.

  11. Method for removing sulfur oxides from a hot gas

    SciTech Connect (OSTI)

    Morris, W.P.; Hurst, T.B.

    1984-06-05

    An improved method for removing sulfur oxides from a hot gas by introducing the gas into a first compartment of a spray drying reactor chamber for settleable particulate removal, by then directing the gas to a second compartment of the reactor chamber wherein the gas is contacted with an atomized alkali slurry for sulfur oxide removal by formation of a dry mixture of sulfite and sulfate compounds, by removing a portion of the dry mixture from the gas in the second compartment and by passing the gas from the second compartment to a dry particle collection zone for removal of substantially all of the remaining gas entrained dry mixture.

  12. Under the (Heat) Dome: Staying Cool and Efficient on the Hottest Days |

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

    Department of Energy Under the (Heat) Dome: Staying Cool and Efficient on the Hottest Days Under the (Heat) Dome: Staying Cool and Efficient on the Hottest Days July 21, 2016 - 10:47pm Addthis Prepare to keep yourself--and your pets--cool, healthy, and comfortable during extreme heat. | FEMA News Photo Prepare to keep yourself--and your pets--cool, healthy, and comfortable during extreme heat. | FEMA News Photo Allison Casey Senior Communicator, NREL We've been talking home cooling on

  13. Gulf Coast Salt Domes geologic Area Characterization Report, East Texas Study Area. Volume II. Technical report. [Contains glossary of geological terms; Oakwood, Keechi, and Palestine domes

    SciTech Connect (OSTI)

    Not Available

    1982-07-01

    The East Texas Area Characterization Report (ACR) is a compilation of data gathered during the Area Characterization phase of the Department of Energy's National Waste Terminal Storage program in salt. The characterization of Gulf Coast Salt Domes as a potential site for storage of nuclear waste is an ongoing process. This report summarizes investigations covering an area of approximately 2590 km/sup 2/ (1000 mi/sup 2/). Data on Oakwood, Keechi, and Palestine Domes are given. Subsequent phases of the program will focus on smaller land areas and fewer specific salt domes, with progressively more detailed investigations, possibly culminating with a license application to the Nuclear Regulatory Commission. The data in this report are a result of drilling and sampling, geophysical and geologic field work, and intensive literature review. The ACR contains text discussing data usage, interpretations, results and conclusions based on available geologic and hydrologic data, and figures including diagrams showing data point locations, geologic and hydrologic maps, geologic cross sections, and other geologic and hydrologic information. An appendix contains raw data gathered during this phase of the project and used in the preparation of these reports.

  14. Evaluation of in-vessel corium retention through external reactor vessel cooling for integral reactor

    SciTech Connect (OSTI)

    Park, R. J.; Lee, J. R.; Kim, S. B.; Jin, Y.; Kim, H. Y.

    2012-07-01

    In-vessel corium retention through external reactor vessel cooling (IVR-ERVC) for a small integral reactor has been evaluated to determine the thermal margin for the prevention of a reactor vessel failure. A thermal load analysis from the corium pool to the outer reactor vessel wall in the lower plenum of the reactor vessel has been performed to determine the heat flux distribution. The critical heat flux (CHF) on the outer reactor vessel wall has been determined to fix the maximum heat removal rate through the external coolant between the outer reactor vessel and the insulation of the reactor vessel. Finally, the thermal margin has been evaluated by comparison of the thermal load with the maximum heat removal rate of the CHF on the outer reactor vessel wall. The maximum heat flux from the corium pool to the outer reactor vessel is estimated at approximately 0.25 MW/m{sup 2} in the metallic layer because of the focusing effect. The CHF of the outer reactor vessel is approximately 1.1 MW/m{sup 2} because of a two phase natural circulation mass flow. Since the thermal margin for the IVR-ERVC is sufficient, the reactor vessel integrity is maintained during a severe accident of a small integral reactor. (authors)

  15. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Fraas, A.P.; Mills, C.B.

    1961-11-21

    A neutronic reactor in which neutron moderation is achieved primarily in its reflector is described. The reactor structure consists of a cylindrical central "island" of moderator and a spherical moderating reflector spaced therefrom, thereby providing an annular space. An essentially unmoderated liquid fuel is continuously passed through the annular space and undergoes fission while contained therein. The reactor, because of its small size, is particularly adapted for propulsion uses, including the propulsion of aircraft. (AEC)

  16. REACTOR COOLING

    DOE Patents [OSTI]

    Quackenbush, C.F.

    1959-09-29

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

  17. Log analysis of six boreholes in conjunction with geologic characterization above and on top of the Weeks Island salt dome

    SciTech Connect (OSTI)

    Sattler, A.R.

    1996-04-01

    Six boreholes were drilled during the geologic characterization and diagnostics of the Weeks Island sinkhole that is over the two-tiered salt mine which was converted for oil storage by the US Strategic Petroleum Reserve. These holes were drilled to provide for geologic characterization of the Weeks Island Salt Dome and its overburden in the immediate vicinity of the sinkhole (mainly through logs and core); to establish a crosswell configuration for seismic tomography; to establish locations for hydrocarbon detection and tracer injection; and to Provide direct observations of sinkhole geometry and material properties. Specific objectives of the logging program were to: (1) identify the top of and the physical state of the salt dome; (2) identify the water table; (3) obtain a relative salinity profile in the aquifer within the alluvium, which ranges from the water table directly to the top of the Weeks Island salt dome; and (4) identify a reflecting horizon seen on seismic profiles over this salt dome. Natural gamma, neutron, density, sonic, resistivity and caliper logs were run. Neutron and density logs were run from inside the well casing because of the extremely unstable condition of the deltaic alluvium overburden above the salt dome. The logging program provided important information about the salt dome and the overburden in that (1) the top of the salt dome was identified at {approximately}189 ft bgl (103 ft msl), and the top of the dome contains relatively few fractures; (2) the water table is approximately 1 ft msl, (3) this aquifer appears to become steadily more saline with depth; and (4) the water saturation of much of the alluvium over the salt dome is shown to be influenced by the prevalent heavy rainfall. This logging program, a part of the sinkhole diagnostics, provides unique information about this salt dome and the overburden.

  18. Liquid metal cooled nuclear reactors with passive cooling system

    DOE Patents [OSTI]

    Hunsbedt, Anstein; Fanning, Alan W.

    1991-01-01

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

  19. Workers Enter Cocooned F Reactor for Scheduled Inspection

    Broader source: Energy.gov [DOE]

    RICHLAND, Wash. – 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.

  20. Indirect passive cooling system for liquid metal cooled nuclear reactors

    DOE Patents [OSTI]

    Hunsbedt, Anstein; Boardman, Charles E.

    1990-01-01

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

  1. Passive cooling safety system for liquid metal cooled nuclear reactors

    DOE Patents [OSTI]

    Hunsbedt, Anstein; Boardman, Charles E.; Hui, Marvin M.; Berglund, Robert C.

    1991-01-01

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

  2. Spent Nuclear Fuel (SNF) Removal Campaign Plan

    SciTech Connect (OSTI)

    PAJUNEN, A.L.

    2000-08-07

    The overall operation of the Spent Nuclear Fuel Project will include fuel removal, sludge removal, debris removal, and deactivation transition activities. Figure 1-1 provides an overview of the current baseline operating schedule for project sub-systems, indicating that a majority of fuel removal activities are performed over an approximately three-and-one-half year time period. The purpose of this document is to describe the strategy for operating the fuel removal process systems. The campaign plan scope includes: (1) identifying a fuel selection sequence during fuel removal activities, (2) identifying MCOs that are subjected to extra testing (process validation) and monitoring, and (3) discussion of initial MCO loading and monitoring in the Canister Storage Building (CSB). The campaign plan is intended to integrate fuel selection requirements for handling special groups of fuel within the basin (e.g., single pass reactor fuel), process validation activities identified for process systems, and monitoring activities during storage.

  3. Packed fluidized bed blanket for fusion reactor

    DOE Patents [OSTI]

    Chi, John W. H.

    1984-01-01

    A packed fluidized bed blanket for a fusion reactor providing for efficient radiation absorption for energy recovery, efficient neutron absorption for nuclear transformations, ease of blanket removal, processing and replacement, and on-line fueling/refueling. The blanket of the reactor contains a bed of stationary particles during reactor operation, cooled by a radial flow of coolant. During fueling/refueling, an axial flow is introduced into the bed in stages at various axial locations to fluidize the bed. When desired, the fluidization flow can be used to remove particles from the blanket.

  4. Engineering Test Reactor (ETR) Vessel Relocated after 50 years.

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

    Printer Friendly Engineering Test Reactor (ETR) Vessel Relocated Engineering Test Reactor Vessel Pre-startup 1957 Click on image to enlarge. Image 1 of 5 Gantry jacks attached to ETR vessel. Initial lift starts. - Click on image to enlarge. Image 2 of 5 ETR vessel removed from substructure. Vessel lifted approximately 40 ft. - Click on image to enlarge. On Monday, September 24, 2007 the Engineering Test Reactor (ETR) vessel was removed from its location and delivered to the Idaho CERCLA Disposal

  5. Geologic technical assessment of the Chacahoula Salt Dome, Louisiana, for potential expansion of the U.S. strategic petroleum reserve.

    SciTech Connect (OSTI)

    Snider, Anna C.; Rautman, Christopher Arthur; Looff, Karl M.

    2006-03-01

    The Chacahoula salt dome, located in southern Louisiana, approximately 66 miles southwest of New Orleans, appears to be a suitable site for a 160-million-barrel-capacity expansion facility for the U.S. Strategic Petroleum Reserve, comprising sixteen 10-million barrel underground storage caverns. The overall salt dome appears to cover an area of some 1800 acres, or approximately 2.8 square miles, at a subsea elevation of 2000 ft, which is near the top of the salt stock. The shallowest known salt is present at 1116 ft, subsea. The crest of the salt dome is relatively flatlying, outward to an elevation of -4000 ft. Below this elevation, the flanks of the dome plunge steeply in all directions. The dome appears to comprise two separate spine complexes of quasi-independently moving salt. Two mapped areas of salt overhang, located on the eastern and southeastern flanks of the salt stock, are present below -8000 ft. These regions of overhang should present no particular design issues, as the conceptual design SPR caverns are located in the western portion of the dome. The proposed cavern field may be affected by a boundary shear zone, located between the two salt spines. However, the large size of the Chacahoula salt dome suggests that there is significant design flexibility to deal with such local geologic issues.

  6. Assessment of dome-fill technology and potential fill materials for the Hanford single-shell tanks

    SciTech Connect (OSTI)

    Smyth, J.D.; Shade, J.W.; Somasundaram, S.

    1992-05-01

    This study is part of a task that will identify dome-fill materials to stabilize and prevent the collapse of the structures of 149 single- shell tanks (SSTs). The SSTs were built at the Hanford Site in Washington State and used between 1944 and 1980 to store radioactive and other hazardous wastes. In addition to identifying suitable fill materials, this task will develop the technology and methods required to fill the tanks with the selected material. To date, basalt is the only candidate fill material with any testing conducted for its suitability as a dome-fill material. Sufficient data do not exist to select or eliminate basalt as a candidate material. This report documents a review of past dome-fill work at the Hanford Site and of other pertinent literature to establish a baseline for the dome-fill technology. In addition, the report identifies existing dome-fill technology, preliminary performance criteria for dome-fill technology development, potential testing strategies, and potential fill materials. As a part of this study, potential fill materials are qualitatively evaluated and a list of preliminary candidate fill materials is identified. Future work will further screen these materials. The dome-fill task work will ultimately contribute to the development of a final waste form package and the safe isolation of wastes from the Hanford Site SSTs.

  7. Experimental and analytical studies of passive shutdown heat removal systems

    SciTech Connect (OSTI)

    Pedersen, D.; Tessier, J.; Heineman, J.; Stewart, R.; Anderson, T.; August, C.; Chawla, T.; Cheung, F.B.; Despe, O.; Haupt, H.J.

    1987-01-01

    Using a naturally circulating air stream to remove shutdown decay heat from a nuclear reactor vessel is a key feature of advanced liquid metal reactor (LMR) concepts developed by potential vendors selected by the Department of Energy. General Electric and Rockwell International continue to develop innovative design concepts aimed at improving safety, lowering plant costs, simplifying plant operation, reducing construction times, and most of all, enhancing plant licensability. The reactor program at Argonne National Laboratory (ANL) provides technical support to both organizations. The method of shutdown heat removal proposed employs a totally passive cooling system that rejects heat from the reactor by radiation and natural convection to air. The system is inherently reliable since it is not subject failure modes associated with active decay cooling systems. The system is designed to assure adequate cooling of the reactor under abnormal operating conditions associated with loss of heat removal through other heat transport paths.

  8. CALANDRIA TYPE SODIUM GRAPHITE REACTOR

    DOE Patents [OSTI]

    Peterson, R.M.; Mahlmeister, J.E.; Vaughn, N.E.; Sanders, W.J.; Williams, A.C.

    1964-02-11

    A sodium graphite power reactor in which the unclad graphite moderator and fuel elements are contained within a core tank is described. The core tank is submersed in sodium within the reactor vessel. Extending longitudinally through the core thnk are process tubes with fuel elements positioned therein. A bellows sealing means allows axial expansion and construction of the tubes. Within the core tank, a leakage plenum is located below the graphite, and above the graphite is a gas space. A vent line regulates the gas pressure in the space, and another line removes sodium from the plenum. The sodium coolant flows from the lower reactor vessel through the annular space between the fuel elements and process tubes and out into the reactor vessel space above the core tank. From there, the heated coolant is drawn off through an outlet line and sent to the heat exchange. (AEC)

  9. Reactor core isolation cooling system

    DOE Patents [OSTI]

    Cooke, F.E.

    1992-12-08

    A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom. 1 figure.

  10. Reactor core isolation cooling system

    DOE Patents [OSTI]

    Cooke, Franklin E.

    1992-01-01

    A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom.

  11. 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.

  12. TA-2 Water Boiler Reactor Decommissioning Project

    SciTech Connect (OSTI)

    Durbin, M.E.; Montoya, G.M.

    1991-06-01

    This final report addresses the Phase 2 decommissioning of the Water Boiler Reactor, biological shield, other components within the biological shield, and piping pits in the floor of the reactor building. External structures and underground piping associated with the gaseous effluent (stack) line from Technical Area 2 (TA-2) Water Boiler Reactor were removed in 1985--1986 as Phase 1 of reactor decommissioning. The cost of Phase 2 was approximately $623K. The decommissioning operation produced 173 m{sup 3} of low-level solid radioactive waste and 35 m{sup 3} of mixed waste. 15 refs., 25 figs., 3 tabs.

  13. Heat dissipating nuclear reactor with metal liner

    DOE Patents [OSTI]

    Gluekler, Emil L.; Hunsbedt, Anstein; Lazarus, Jonathan D.

    1987-01-01

    Disclosed is a nuclear reactor containment including a reactor vessel disposed within a cavity with capability for complete inherent decay heat removal in the earth and surrounded by a cast steel containment member which surrounds the vessel. The member has a thick basemat in contact with metal pilings. The basemat rests on a bed of porous particulate material, into which water is fed to produce steam which is vented to the atmosphere. There is a gap between the reactor vessel and the steel containment member. The containment member holds any sodium or core debris escaping from the reactor vessel if the core melts and breaches the vessel.

  14. Heat dissipating nuclear reactor with metal liner

    DOE Patents [OSTI]

    Gluekler, E.L.; Hunsbedt, A.; Lazarus, J.D.

    1985-11-21

    A nuclear reactor containment including a reactor vessel disposed within a cavity with capability for complete inherent decay heat removal in the earth and surrounded by a cast steel containment member which surrounds the vessel is described in this disclosure. The member has a thick basemat in contact with metal pilings. The basemat rests on a bed of porous particulate material, into which water is fed to produce steam which is vented to the atmosphere. There is a gap between the reactor vessel and the steel containment member. The containment member holds any sodium or core debris escaping from the reactor vessel if the core melts and breaches the vessel.

  15. Reactor refueling containment system

    DOE Patents [OSTI]

    Gillett, James E.; Meuschke, Robert E.

    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.

  16. NEUTRONIC REACTOR CONSTRUCTION

    DOE Patents [OSTI]

    Vernon, H.C.; Goett, J.J.

    1958-09-01

    A cover device is described for the fuel element receiving tube of a neutronic reactor of the heterogeneous, water cooled type wherein said tubes are arranged in a moderator with their longitudinal axes vertical. The cover is provided with means to support a rod-type fuel element from the bottom thereof and means to lock the cover in place, the latter being adapted for remote operation. This cover device is easily removable and seals the opening in the upper end of the fuel tube against leakage of coolant.

  17. 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.

  18. 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.

  19. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Fermi, E.

    1960-04-01

    A nuclear reactor is described consisting of blocks of graphite arranged in layers, natural uranium bodies disposed in holes in alternate layers of graphite blocks, and coolant tubes disposed in the layers of graphite blocks which do not contain uranium.

  20. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Hurwitz, H. Jr.; Brooks, H.; Mannal, C.; Payne, J.H.; Luebke, E.A.

    1959-03-24

    A reactor of the heterogeneous, liquid cooled type is described. This reactor is comprised of a central region of a plurality of vertically disposed elongated tubes surrounded by a region of moderator material. The central region is comprised of a central core surrounded by a reflector region which is surrounded by a fast neutron absorber region, which in turn is surrounded by a slow neutron absorber region. Liquid sodium is used as the primary coolant and circulates through the core which contains the fuel elements. Control of the reactor is accomplished by varying the ability of the reflector region to reflect neutrons back into the core of the reactor. For this purpose the reflector is comprised of moderator and control elements having varying effects on reactivity, the control elements being arranged and actuated by groups to give regulation, shim, and safety control.

  1. 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.

  2. REACTOR SHIELD

    DOE Patents [OSTI]

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

    1959-02-17

    Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

  3. Reactor apparatus

    DOE Patents [OSTI]

    Echtler, J. Paul

    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.

  4. 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.

  5. Decay heat removal by natural convection - the RVACS system.

    SciTech Connect (OSTI)

    Tzanos, C. P.

    1999-08-17

    In conclusion, this work shows that for sodium coolant the reactor vessel auxiliary cooling system (RVACS) is an effective passive heat removal system if the reactor power does not exceed about 1600 MW(th). Its effectiveness is limited by the effective radiative heat transfer coefficient in the inner gap. In a lead cooled system, economic considerations may impose a lower limit.

  6. Fuel removal, transport, and storage

    SciTech Connect (OSTI)

    Reno, H.W.

    1986-01-01

    The March 1979 accident at Unit 2 of the Three Mile Island Nuclear Power Station (TMI-2) which damaged the core of the reactor resulted in numerous scientific and technical challenges. Some of those challenges involve removing the core debris from the reactor, packaging it into canisters, loading canisters into a rail cask, and transporting the debris to the Idaho National Engineering Laboratory (INEL) for storage, examination, and preparation for final disposal. This paper highlights how some challenges were resolved, including lessons learned and benefits derived therefrom. Key to some success at TMI was designing, testing, fabricating, and licensing two rail casks, which each provide double containment of the damaged fuel. 10 refs., 12 figs.

  7. 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.

  8. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Anderson, C.R.

    1962-07-24

    A fluidized bed nuclear reactor and a method of operating such a reactor are described. In the design means are provided for flowing a liquid moderator upwardly through the center of a bed of pellets of a nentron-fissionable material at such a rate as to obtain particulate fluidization while constraining the lower pontion of the bed into a conical shape. A smooth circulation of particles rising in the center and falling at the outside of the bed is thereby established. (AEC)

  9. 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.

  10. EIS-0029: Strategic Petroleum Reserve, Texoma Group Salt Domes, Cameron and Calcasieu Parishes, Louisiana and Jefferson County, TX

    Broader source: Energy.gov [DOE]

    The Strategic Petroleum Reserves developed this EIS to analyze the environmental impacts that could occur during site preparation and operation of oil storage facilities at each of four proposed candidate sites in the Texoma Group of salt domes.

  11. Nuclear reactor containment structure with continuous ring tunnel at grade

    DOE Patents [OSTI]

    Seidensticker, Ralph W.; Knawa, Robert L.; Cerutti, Bernard C.; Snyder, Charles R.; Husen, William C.; Coyer, Robert G.

    1977-01-01

    A nuclear reactor containment structure which includes a reinforced concrete shell, a hemispherical top dome, a steel liner, and a reinforced-concrete base slab supporting the concrete shell is constructed with a substantial proportion thereof below grade in an excavation made in solid rock with the concrete poured in contact with the rock and also includes a continuous, hollow, reinforced-concrete ring tunnel surrounding the concrete shell with its top at grade level, with one wall integral with the reinforced concrete shell, and with at least the base of the ring tunnel poured in contact with the rock.

  12. Continuous production of tritium in an isotope-production reactor with a separate circulation system

    DOE Patents [OSTI]

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

    1982-08-19

    A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium is allowed to flow through the reactor in separate loops in order to facilitate the production and removal of tritium.

  13. SO2 REMOVAL WITH COAL SCRUBBING

    SciTech Connect (OSTI)

    Eung Ha Cho; Hari Prashanth Sundaram; Aubrey L. Miller

    2001-07-01

    This project is based on an effective removal of sulfur dioxide from flue gas with coal as the scrubbing medium instead of lime, which is used in the conventional FGD processes. A laboratory study proves that coal scrubbing is an innovative technology that can be implemented into a commercial process in place of the conventional lime scrubbing flue gas desulfurization process. SO{sub 2} was removed from a gas stream using an apparatus, which consisted of a 1-liter stirred reactor immersed in a thermostated oil bath. The reactor contained 60 g of 35-65 mesh coal in 600 ml of water. The apparatus also had 2 bubblers connected to the outlet of the reactor, each containing 1500 ml of 1 molar NaOH solution. The flow rate of the gas was 30 ml/sec, temperature was varied from 21 C to 73 C. Oxygen concentration ranged from 3 to 20% while SO{sub 2} concentration, from 500 to 2000 ppm. SO{sub 2} recovery was determined by analyzing SO{sub 2} concentration in the liquid samples taken from the bubblers. The samples taken from the reactor were analyzed for iron concentrations, which were then used to calculate fractions of coal pyrite leached. It was found that SO{sub 2} removal was highly temperature sensitive, giving 13.1% recovery at 21 C and 99.2% recovery at 73 C after 4 hours. The removal of SO{sub 2} was accomplished by the catalysis of iron that was produced by leaching of coal pyrite with combination of SO{sub 2} and O{sub 2}. This leaching reaction was found to be controlled by chemical reaction with apparent activation energy of 11.6 kcal/mole. SO{sub 2} removal increased with increasing O{sub 2} concentration up to 10% and leveled off upon further increase. The effect of SO{sub 2} concentration on its removal was minimal.

  14. NEUTRONIC REACTOR

    DOE Patents [OSTI]

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

    1958-09-01

    This patent relates to neutronic reactors of the heterogeneous water cooled type, and in particular to a fuel element charging and discharging means therefor. In the embodiment illustrated the reactor contains horizontal, parallel coolant tubes in which the fuel elements are disposed. A loading cart containing a magnzine for holding a plurality of fuel elements operates along the face of the reactor at the inlet ends of the coolant tubes. The loading cart is equipped with a ram device for feeding fuel elements from the magazine through the inlot ends of the coolant tubes. Operating along the face adjacent the discharge ends of the tubes there is provided another cart means adapted to receive irradiated fuel elements as they are forced out of the discharge ends of the coolant tubes by the incoming new fuel elements. This cart is equipped with a tank coataining a coolant, such as water, into which the fuel elements fall, and a hydraulically operated plunger to hold the end of the fuel element being discharged. This inveation provides an apparatus whereby the fuel elements may be loaded into the reactor, irradiated therein, and unloaded from the reactor without stopping the fiow of the coolant and without danger to the operating personnel.

  15. 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.

  16. 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.

  17. POWER REACTOR

    DOE Patents [OSTI]

    Zinn, W.H.

    1958-07-01

    A fast nuclear reactor system ls described for producing power and radioactive isotopes. The reactor core is of the heterogeneous, fluid sealed type comprised of vertically arranged elongated tubular fuel elements having vertical coolant passages. The active portion is surrounded by a neutron reflector and a shield. The system includes pumps and heat exchangers for the primary and secondary coolant circuits. The core, primary coolant pump and primary heat exchanger are disposed within an irapenforate tank which is filled with the primary coolant, in this case a liquid metal such as Na or NaK, to completely submerge these elements. The tank is completely surrounded by a thick walled concrete shield. This reactor system utilizes enriched uranium or plutonium as the fissionable material, uranium or thorium as a diluent and thorium or uranium containing less than 0 7% of the U/sup 235/ isotope as a fertile material.

  18. REACTOR CONTROL

    DOE Patents [OSTI]

    Fortescue, P.; Nicoll, D.

    1962-04-24

    A control system employed with a high pressure gas cooled reactor in which a control rod is positioned for upward and downward movement into the neutron field from a position beneath the reactor is described. The control rod is positioned by a coupled piston cylinder releasably coupled to a power drive means and the pressurized coolant is directed against the lower side of the piston. The coolant pressure is offset by a higher fiuid pressure applied to the upper surface of the piston and means are provided for releasing the higher pressure on the upper side of the piston so that the pressure of the coolant drives the piston upwardly, forcing the coupled control rod into the ncutron field of the reactor. (AEC)

  19. Assessment of Effectiveness of Geologic Isolation Systems: REFERENCE SITE INITIAL ASSESSMENT FOR A SALT DOME REPOSITORY

    SciTech Connect (OSTI)

    Harwell, M. A.; Brandstetter, A.; Benson, G. L.; Bradley, D. J.; Serne, R. J.; Soldat, J. K; Cole, C. R.; Deutsch, W. J.; Gupta, S. K.; Harwell, C. C.; Napier, B. A.; Reisenauer, A. E.; Prater, L. S.; Simmons, C. S.; Strenge, D. L.; Washburn, J. F.; Zellmer, J. T.

    1982-06-01

    As a methodology demonstration for the Office of Nuclear Waste Isolation (ONWI), the Assessment of Effectiveness of Geologic Isolation Systems (AEGIS) Program conducted an initial reference site analysis of the long-term effectiveness of a salt dome repository. The Hainesville Salt Dome in Texas was chosen to be representative of the Gulf Coast interior salt domes; however, the Hainesville Site has been eliminated as a possible nuclear waste repository site. The data used for this exercise are not adequate for an actual assessment, nor have all the parametric analyses been made that would adequately characterize the response of the geosystem surrounding the repository. Additionally, because this was the first exercise of the complete AEGIS and WASTE Rock Interaction Technology (WRIT) methodology, this report provides the initial opportunity for the methodology, specifically applied to a site, to be reviewed by the community outside the AEGIS. The scenario evaluation, as a part of the methodology demonstration, involved consideration of a large variety of potentially disruptive phenomena, which alone or in concert could lead to a breach in a salt dome repository and to a subsequent transport of the radionuclides to the environment. Without waste- and repository-induced effects, no plausible natural geologic events or processes which would compromise the repository integrity could be envisioned over the one-million-year time frame after closure. Near-field (waste- and repository-induced) effects were excluded from consideration in this analysis, but they can be added in future analyses when that methodology development is more complete. The potential for consequential human intrusion into salt domes within a million-year time frame led to the consideration of a solution mining intrusion scenario. The AEGIS staff developed a specific human intrusion scenario at 100 years and 1000 years post-closure, which is one of a whole suite of possible scenarios. This scenario

  20. Assessment of Effectiveness of Geologic Isolation Systems: REFERENCE SITE INITIAL ASSESSMENT FOR A SALT DOME REPOSITORY

    SciTech Connect (OSTI)

    Harwell, M. A.; Brandstetter, A.; Benson, G. L.; Raymond, J. R.; Brandley, D. J.; Serne, R. J.; Soldat, J. K.; Cole, C. R.; Deutsch, W. J.; Gupta, S. K.; Harwell, C. C.; Napier, B. A.; Reisenauer, A. E.; Prater, L. S.; Simmons, C. S.; Strenge, D. L.; Washburn, J. F.; Zellmer, J. T.

    1982-06-01

    As a methodology demonstration for the Office of Nuclear Waste Isolation (ONWI), the Assessment of Effectiveness of Geologic Isolation Systems (AEGIS) Program conducted an initial reference site analysis of the long-term effectiveness of a salt dome repository. The Hainesville Salt Dome in Texas was chosen to be representative of the Gulf Coast interior salt domes; however, the Hainesville Site has been eliminated as a possible nuclear waste repository site. The data used for this exercise are not adequate for an actual assessment, nor have all the parametric analyses been made that would adequately characterize the response of the geosystem surrounding the repository. Additionally, because this was the first exercise of the complete AEGIS and WASTE Rock Interaction Technology (WRIT) methodology, this report provides the initial opportunity for the methodology, specifically applied to a site, to be reviewed by the community outside the AEGIS. The scenario evaluation, as a part of the methodology demonstration, involved consideration of a large variety of potentially disruptive phenomena, which alone or in concert could lead to a breach in a salt dome repository and to a subsequent transport of the radionuclides to the environment. Without waste- and repository-induced effects, no plausible natural geologic events or processes which would compromise the repository integrity could be envisioned over the one-million-year time frame after closure. Near-field (waste- and repository-induced) effects were excluded from consideration in this analysis, but they can be added in future analyses when that methodology development is more complete. The potential for consequential human intrusion into salt domes within a million-year time frame led to the consideration of a solution mining intrusion scenario. The AEGIS staff developed a specific human intrusion scenario at 100 years and 1000 years post-closure, which is one of a whole suite of possible scenarios. This scenario

  1. (Reactor dosimetry)

    SciTech Connect (OSTI)

    West, C.D.

    1990-09-13

    The lead in most aspects of research reactor design and use passed from the USA about 15 years ago, soon after the construction of the HFIR and HFBR. The Europeans have consistently upgraded and improved their existing facilities and have built new ones including the HFR at Grenoble and ORPHEE at Saclay. They studied ultra-high flux concepts ({approximately}10{sup 20}/m{sup {minus}2}{center dot}s{sup {minus}1}) about 10 years ago, and are in the design phase of a new, highly efficient medium flux reactor to be built at Garching, near Munich in Germany. A visit was made to Interatom, the firm -- the equivalent of the Architect/Engineer for the ANS project -- responsible, under contract to the Technical University of Munich, for the new Munich reactor design. There are many similarities to the ANS design, and we reviewed and discussed technical and safety aspects of the two reactors. A request was made for some new, hitherto proprietary, experimental data on reactor thermal hydraulics and cooling that will be very valuable to the ANS project. I presented a seminar on the ANS project. A visit was made to Kernforschungszentrum Karlsruhe and knowledge was gained from Dr. Kuchle, a true pioneer of ultra-high flux reactor concepts, of their work. Dr. Kuchle kindly reviewed the ANS reference core and cooling system design (with favorable conclusions). I then talked with researchers working on materials irradiation damage and activation of structural materials by neutron irradiation, both key issues for the ANS. I was shown some new techniques they have developed for testing materials irradiation effects at high fluences, in a short time, using accelerated particle beams.

  2. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Fermi, E.; Szilard, L.

    1957-09-24

    Reactors of the type employing plates of natural uranium in a moderator are discussed wherein the plates are um-formly disposed in parallel relationship to each other thereby separating the moderator material into distinct and individual layers. Each plate has an uninterrupted sunface area substantially equal to the cross-sectional area of the active portion of the reactor, the particular size of the plates and the volume ratio of moderator to uranium required to sustain a chain reaction being determinable from the known purity of these materials and other characteristics such as the predictable neutron losses due to the formation of radioactive elements of extremely high neutron capture cross section.

  3. NEUTRONIC REACTOR

    DOE Patents [OSTI]

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

    1958-04-15

    A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

  4. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Young, G.

    1963-01-01

    This patent covers a power-producing nuclear reactor in which fuel rods of slightly enriched U are moderated by heavy water and cooled by liquid metal. The fuel rods arranged parallel to one another in a circle are contained in a large outer closed-end conduit that extends into a tank containing the heavy water. Liquid metal is introduced into the large conduit by a small inner conduit that extends within the circle of fuel rods to a point near the lower closed end of the outer conduit. (AEC) Production Reactors

  5. Silica Scaling Removal Process

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

    Scaling Removal Process Scientists at Los Alamos National Laboratory have developed a novel technology to remove both dissolved and colloidal silica using small gel particles....

  6. Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites

    SciTech Connect (OSTI)

    Maheras, Steven J.; Best, Ralph E.; Ross, Steven B.; Buxton, Kenneth A.; England, Jeffery L.; McConnell, Paul E.

    2013-09-30

    This report fulfills the M2 milestone M2FT-13PN0912022, “Stranded Sites De-Inventorying Report.” In January 2013, the U.S. Department of Energy (DOE) issued the Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste (DOE 2013). Among the elements contained in this strategy is an initial focus on accepting used nuclear fuel from shutdown reactor sites. This focus is consistent with the recommendations of the Blue Ribbon Commission on America’s Nuclear Future, which identified removal of stranded used nuclear fuel at shutdown sites as a priority so that these sites may be completely decommissioned and put to other beneficial uses (BRC 2012). Shutdown sites are defined as those commercial nuclear power reactor sites where the nuclear power reactors have been shut down and the site has been decommissioned or is undergoing decommissioning. In this report, a preliminary evaluation of removing used nuclear fuel from 12 shutdown sites was conducted. The shutdown sites were Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, Zion, Crystal River, Kewaunee, and San Onofre. These sites have no other operating nuclear power reactors at their sites and have also notified the U.S. Nuclear Regulatory Commission that their reactors have permanently ceased power operations and that nuclear fuel has been permanently removed from their reactor vessels. Shutdown reactors at sites having other operating reactors are not included in this evaluation.

  7. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Starr, C.

    1963-01-01

    This patent relates to a combination useful in a nuclear reactor and is comprised of a casing, a mass of graphite irapregnated with U compounds in the casing, and at least one coolant tube extending through the casing. The coolant tube is spaced from the mass, and He is irtroduced irto the space between the mass and the coolant tube. (AEC)

  8. Neutronic reactor

    DOE Patents [OSTI]

    Wende, Charles W. J.; Babcock, Dale F.; Menegus, Robert L.

    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.

  9. 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.

  10. NEUTRONIC REACTOR

    DOE Patents [OSTI]

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

    1959-10-27

    BS>A reactor cooled by water, biphenyl, helium, or other fluid with provision made for replacing the fuel rods with the highest plutonium and fission product content without disassembling the entire core and for promptly cooling the rods after their replacement in order to prevent build-up of heat from fission product activity is described.

  11. Neutronic reactor

    DOE Patents [OSTI]

    Wende, Charles W. J.

    1976-08-17

    A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield.

  12. Neutronic reactor

    DOE Patents [OSTI]

    Carleton, John T.

    1977-01-25

    A graphite-moderated nuclear reactor includes channels between blocks of graphite and also includes spacer blocks between adjacent channeled blocks with an axis of extension normal to that of the axis of elongation of the channeled blocks to minimize changes in the physical properties of the graphite as a result of prolonged neutron bombardment.

  13. REGENERATION OF REACTOR FUEL ELEMENTS

    DOE Patents [OSTI]

    Roake, W.E.; Lyon, W.L.

    1960-03-29

    A process of concentrating by electrolysis the uraatum and/or plutonium of an aluminum alloy containing these actinides after the actinide has been partially consumed by neutron bombardment in a reactor is given. The alloy is made the anode in a system having an aluminum cathode and a cryolite electrolyte. Electrolysis from 22 to 28 ampere-hours removes a sufficient quantity of aluminum from the alloy to make it suitable for reuse.

  14. Reactor pressure vessel head vents and methods of using the same

    SciTech Connect (OSTI)

    Gels, John L; Keck, David J; Deaver, Gerald A

    2014-10-28

    Internal head vents are usable in nuclear reactors and include piping inside of the reactor pressure vessel with a vent in the reactor upper head. Piping extends downward from the upper head and passes outside of the reactor to permit the gas to escape or be forcibly vented outside of the reactor without external piping on the upper head. The piping may include upper and lowers section that removably mate where the upper head joins to the reactor pressure vessel. The removable mating may include a compressible bellows and corresponding funnel. The piping is fabricated of nuclear-reactor-safe materials, including carbon steel, stainless steel, and/or a Ni--Cr--Fe alloy. Methods install an internal head vent in a nuclear reactor by securing piping to an internal surface of an upper head of the nuclear reactor and/or securing piping to an internal surface of a reactor pressure vessel.

  15. Strategic Petroleum Reserve (SPR) additional geologic site characterization studies, Bryan Mound Salt Dome, Texas

    SciTech Connect (OSTI)

    Neal, J.T.; Magorian, T.R.; Ahmad, S.

    1994-11-01

    This report revises the original report that was published in 1980. Some of the topics covered in the earlier report were provisional and it is now practicable to reexamine them using new or revised geotechnical data and that obtained from SPR cavern operations, which involves 16 new caverns. Revised structure maps and sections show interpretative differences as compared with the 1980 report and more definition in the dome shape and caprock structural contours, especially a major southeast-northwest trending anomalous zone. The original interpretation was of westward tilt of the dome, this revision shows a tilt to the southeast, consistent with other gravity and seismic data. This interpretation refines the evaluation of additional cavern space, by adding more salt buffer and allowing several more caverns. Additional storage space is constrained on this nearly full dome because of low-lying peripheral wetlands, but 60 MMBBL or more of additional volume could be gained in six or more new caverns. Subsidence values at Bryan Mound are among the lowest in the SPR system, averaging about 11 mm/yr (0.4 in/yr), but measurement and interpretation issues persist, as observed values are about the same as survey measurement accuracy. Periodic flooding is a continuing threat because of the coastal proximity and because peripheral portions of the site are at elevations less than 15 ft. This threat may increase slightly as future subsidence lowers the surface, but the amount is apt to be small. Caprock integrity may be affected by structural features, especially the faulting associated with anomalous zones. Injection wells have not been used extensively at Bryan Mound, but could be a practicable solution to future brine disposal needs. Environmental issues center on the areas of low elevation that are below 15 feet above mean sea level: the coastal proximity and lowland environment combined with the potential for flooding create conditions that require continuing surveillance.

  16. Strategic Petroleum Reserve (SPR) additional geologic site characterization studies, Bayou Choctaw salt dome, Louisiana

    SciTech Connect (OSTI)

    Neal, J.T.; Magorian, T.R.; Byrne, K.O.; Denzler, S.

    1993-09-01

    This report revises and updates the geologic site characterization report that was published in 1980. Revised structure maps and sections show interpretative differences in the dome shape and caprock structural contours, especially a major east-west trending shear zone, not mapped in the 1980 report. Excessive gas influx in Caverns 18 and 20 may be associated with this shear zone. Subsidence values at Bayou Choctaw are among the lowest in the SPR system, averaging only about 10 mm/yr but measurement and interpretation issues persist, as observed values often approximate measurement accuracy. Periodic, temporary flooding is a continuing concern because of the low site elevation (less than 10 ft), and this may intensify as future subsidence lowers the surface even further. Cavern 4 was re-sonared in 1992 and the profiles suggest that significant change has not occurred since 1980, thereby reducing the uncertainty of possible overburden collapse -- as occurred at Cavern 7 in 1954. Other potential integrity issues persist, such as the proximity of Cavern 20 to the dome edge, and the narrow web separating Caverns 15 and 17. Injection wells have been used for the disposal of brine but have been only marginally effective thus far; recompletions into more permeable lower Pleistocene gravels may be a practical way of increasing injection capacity and brinefield efficiency. Cavern storage space is limited on this already crowded dome, but 15 MMBBL could be gained by enlarging Cavern 19 and by constructing a new cavern beneath and slightly north of abandoned Cavern 13. Environmental issues center on the low site elevation: the backswamp environment combined with the potential for periodic flooding create conditions that will require continuing surveillance.

  17. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Grebe, J.J.

    1959-12-15

    A reactor which is particularly adapted tu serve as a heat source for a nuclear powered alrcraft or rocket is described. The core of this reactor consists of a porous refractory modera;or body which is impregnated with fissionable nuclei. The core is designed so that its surface forms tapered inlet and outlet ducts which are separated by the porous moderator body. In operation a gaseous working fluid is circulated through the inlet ducts to the surface of the moderator, enters and passes through the porous body, and is heated therein. The hot gas emerges into the outlet ducts and is available to provide thrust. The principle advantage is that tremendous quantities of gas can be quickly heated without suffering an excessive pressure drop.

  18. REACTOR UNLOADING

    DOE Patents [OSTI]

    Leverett, M.C.

    1958-02-18

    This patent is related to gas cooled reactors wherein the fuel elements are disposed in vertical channels extending through the reactor core, the cooling gas passing through the channels from the bottom to the top of the core. The invention is a means for unloading the fuel elements from the core and comprises dump values in the form of flat cars mounted on wheels at the bottom of the core structure which support vertical stacks of fuel elements. When the flat cars are moved, either manually or automatically, for normal unloading purposes, or due to a rapid rise in the reproduction ratio within the core, the fuel elements are permtted to fall by gravity out of the core structure thereby reducing the reproduction ratio or stopping the reaction as desired.

  19. 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.

  20. Nuclear reactor

    DOE Patents [OSTI]

    Wade, Elman E.

    1979-01-01

    A nuclear reactor including two rotatable plugs and a positive top core holddown structure. The top core holddown structure is divided into two parts: a small core cover, and a large core cover. The small core cover, and the upper internals associated therewith, are attached to the small rotating plug, and the large core cover, with its associated upper internals, is attached to the large rotating plug. By so splitting the core holddown structures, under-the-plug refueling is accomplished without the necessity of enlarging the reactor pressure vessel to provide a storage space for the core holddown structure during refueling. Additionally, the small and large rotating plugs, and their associated core covers, are arranged such that the separation of the two core covers to permit rotation is accomplished without the installation of complex lifting mechanisms.

  1. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Wade, E.J.

    1958-09-16

    This patent relates to a reflector means for a neutronic reactor. A reflector comprised of a plurality of vertically movable beryllium control members is provided surrounding the sides of the reactor core. An absorber of fast neutrons comprised of natural uramum surrounds the reflector. An absorber of slow neutrons surrounds the absorber of fast neutrons and is formed of a plurality of beryllium blocks having natural uranium members distributcd therethrough. in addition, a movable body is positioned directly below the core and is comprised of a beryllium reflector and an absorbing member attached to the botiom thereof, the absorbing member containing a substance selected from the goup consisting of natural urantum and Th/sup 232/.

  2. NUCLEAR REACTORS

    DOE Patents [OSTI]

    Long, E.; Ashby, J.W.

    1958-09-16

    ABS>A graphite moderator structure is presented for a nuclear reactor compriscd of an assembly of similarly orientated prismatic graphite blocks arranged on spaced longitudinal axes lying in common planes wherein the planes of the walls of the blocks are positioned so as to be twisted reintive to the planes of said axes so thatthe unlmpeded dtrect paths in direction wholly across the walls of the blocks are limited to the width of the blocks plus spacing between the blocks.

  3. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Anderson, J.B.

    1960-01-01

    A reactor is described which comprises a tank, a plurality of coaxial steel sleeves in the tank, a mass of water in the tank, and wire grids in abutting relationship within a plurality of elongated parallel channels within the steel sleeves, the wire being provided with a plurality of bends in the same plane forming adjacent parallel sections between bends, and the sections of adjacent grids being normally disposed relative to each other.

  4. Neutronic reactor

    DOE Patents [OSTI]

    Lewis, Warren R.

    1978-05-30

    A graphite-moderated, water-cooled nuclear reactor including a plurality of rectangular graphite blocks stacked in abutting relationship in layers, alternate layers having axes which are normal to one another, alternate rows of blocks in alternate layers being provided with a channel extending through the blocks, said channeled blocks being provided with concave sides and having smaller vertical dimensions than adjacent blocks in the same layer, there being nuclear fuel in the channels.

  5. REACTOR CONTROL

    DOE Patents [OSTI]

    Ruano, W.J.

    1957-12-10

    This patent relates to nuclear reactors of the type which utilize elongited rod type fuel elements immersed in a liquid moderator and shows a design whereby control of the chain reaction is obtained by varying the amount of moderator or reflector material. A central tank for containing liquid moderator and fuel elements immersed therein is disposed within a surrounding outer tank providing an annular space between the two tanks. This annular space is filled with liquid moderator which functions as a reflector to reflect neutrons back into the central reactor tank to increase the reproduction ratio. Means are provided for circulating and cooling the moderator material in both tanks and additional means are provided for controlling separately the volume of moderator in each tank, which latter means may be operated automatically by a neutron density monitoring device. The patent also shows an arrangement for controlling the chain reaction by injecting and varying an amount of poisoning material in the moderator used in the reflector portion of the reactor.

  6. Sandia National Laboratories Medical Isotope Reactor concept.

    SciTech Connect (OSTI)

    Coats, Richard Lee; Dahl, James J.; Parma, Edward J., Jr.

    2010-04-01

    This report describes the Sandia National Laboratories Medical Isotope Reactor and hot cell facility concepts. The reactor proposed is designed to be capable of producing 100% of the U.S. demand for the medical isotope {sup 99}Mo. The concept is novel in that the fuel for the reactor and the targets for the {sup 99}Mo production are the same. There is no driver core required. The fuel pins that are in the reactor core are processed on a 7 to 21 day irradiation cycle. The fuel is low enriched uranium oxide enriched to less than 20% {sup 235}U. The fuel pins are approximately 1 cm in diameter and 30 to 40 cm in height, clad with Zircaloy (zirconium alloy). Approximately 90 to 150 fuel pins are arranged in the core in a water pool {approx}30 ft deep. The reactor power level is 1 to 2 MW. The reactor concept is a simple design that is passively safe and maintains negative reactivity coefficients. The total radionuclide inventory in the reactor core is minimized since the fuel/target pins are removed and processed after 7 to 21 days. The fuel fabrication, reactor design and operation, and {sup 99}Mo production processing use well-developed technologies that minimize the technological and licensing risks. There are no impediments that prevent this type of reactor, along with its collocated hot cell facility, from being designed, fabricated, and licensed today.

  7. Homogeneous fast-flux isotope-production reactor

    DOE Patents [OSTI]

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

    1982-08-19

    A method is described for producing tritium in a liquid metal fast breeder reactor. Lithium target material is dissolved in the liquid metal coolant in order to facilitate the production and removal of tritium.

  8. Solid0Core Heat-Pipe Nuclear Batterly Type Reactor

    SciTech Connect (OSTI)

    Ehud Greenspan

    2008-09-30

    This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP).

  9. SIMPLIFIED SODIUM GRAPHITE REACTOR SYSTEM

    DOE Patents [OSTI]

    Dickinson, R.W.

    1963-03-01

    This patent relates to a nuclear power reactor comprising a reactor vessel, shielding means positioned at the top of said vessel, means sealing said reactor vessel to said shielding means, said vessel containing a quantity of sodium, a core tank, unclad graphite moderator disposed in said tank, means including a plurality of process tubes traversing said tank for isolating said graphite from said sodium, fuel elements positioned in said process tubes, said core tank being supported in spaced relation to the walls and bottom of said reactor vessel and below the level of said sodium, neutron shielding means positioned adjacent said core tank between said core tank and the walls of said vessel, said neutron shielding means defining an annuiar volume adjacent the inside wall of said reactor vessel, inlet plenum means below said core tank for providing a passage between said annular volume and said process tubes, heat exchanger means removably supported from the first-named shielding means and positioned in said annular volume, and means for circulating said sodium over said neutron shielding means down through said heat exchanger, across said inlet plenum and upward through said process tubes, said last-named means including electromagnetic pumps located outside said vessel and supported on said vessel wall between said heat exchanger means and said inlet plenum means. (AEC)

  10. Removal of mercury from waste gases

    SciTech Connect (OSTI)

    Muster, U.; Marr, R.; Pichler, G.; Kremshofer, S.; Wilferl, R.; Draxler, J.

    1996-12-31

    Waste and process gases from thermal power, incineration and metallurgical plants or those from cement and alkali chloride industries contain metallic, inorganic and organic mercury. Widespread processes to remove the major amount of mercury are absorption and adsorption. Caused by the lowering of the emission limit from 200 to 50 {mu}g/m{sup 3} [STP] by national and European legislators, considerable efforts were made to enhance the efficiency of the main separation units of flue gas cleaning plants. Specially impregnated ceramic carriers can be used for the selective separation of metallic, inorganic and organic mercury. Using the ceramic reactor removal rates lower than 5 {mu}g/m{sup 3} [STP] of gaseous mercury and its compounds can be achieved. The ceramic reactor is active, regenerable and stable for a long term operation. 4 refs., 7 figs.

  11. Development and Calibration of New 3-D Vector VSP Imaging Technology: Vinton Salt Dome, LA

    SciTech Connect (OSTI)

    Kurt J. Marfurt; Hua-Wei Zhou; E. Charlotte Sullivan

    2004-09-01

    Vinton salt dome is located in Southwestern Louisiana, in Calcasieu Parish. Tectonically, the piercement dome is within the salt dome minibasin province. The field has been in production since 1901, with most of the production coming from Miocene and Oligocene sands. The goal of our project was to develop and calibrate new processing and interpretation technology to fully exploit the information available from a simultaneous 3-D surface seismic survey and 3-C, 3-D vertical seismic profile (VSP) survey over the dome. More specifically the goal was to better image salt dome flanks and small, reservoir-compartmentalizing faults. This new technology has application to mature salt-related fields across the Gulf Coast. The primary focus of our effort was to develop, apply, and assess the limitations of new 3-C, 3-D wavefield separation and imaging technology that could be used to image aliased, limited-aperture, vector VSP data. Through 2-D and 3-D full elastic modeling, we verified that salt flank reflections exist in the horizontally-traveling portion of the wavefield rather than up- and down-going portions of the wavefield, thereby explaining why many commercial VSP processing flow failed. Since the P-wave reflections from the salt flank are measured primarily on the horizontal components while P-wave reflections from deeper sedimentary horizons are measured primarily on the vertical component, a true vector VSP analysis was needed. We developed an antialiased discrete Radon transform filter to accurately model P- and S-wave data components measured by the vector VSP. On-the-fly polarization filtering embedded in our Kirchhoff imaging algorithm was effective in separating PP from PS wave images. By the novel application of semblance-weighted filters, we were able to suppress many of the migration artifacts associated with low fold, sparse VSP acquisition geometries. To provide a better velocity/depth model, we applied 3-D prestack depth migration to the surface data

  12. 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.

  13. FLUID MODERATED REACTOR

    DOE Patents [OSTI]

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

    1957-10-22

    A reactor which utilizes fissionable fuel elements in rod form immersed in a moderator or heavy water and a means of circulating the heavy water so that it may also function as a coolant to remove the heat generated by the fission of the fuel are described. In this design, the clad fuel elements are held in vertical tubes immersed in heavy water in a tank. The water is circulated in a closed system by entering near the tops of the tubes, passing downward through the tubes over the fuel elements and out into the tank, where it is drawn off at the bottom, passed through heat exchangers to give up its heat and then returned to the tops of the tubes for recirculation.

  14. B Reactor | Department of Energy

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

    Operational Management History Manhattan Project Signature Facilities B Reactor B Reactor B Reactor Completed in September 1944, the B Reactor was the world's first ...

  15. 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.

  16. NUCLEAR REACTORS

    DOE Patents [OSTI]

    Long, E.; Ashley, J.W.

    1958-12-16

    A graphite moderator structure is described for a gas-cooled nuclear reactor having a vertical orlentation wherein the structure is physically stable with regard to dlmensional changes due to Wigner growth properties of the graphite, and leakage of coolant gas along spaces in the structure is reduced. The structure is comprised of stacks of unlform right prismatic graphite blocks positioned in layers extending in the direction of the lengths of the blocks, the adjacent end faces of the blocks being separated by pairs of tiles. The blocks and tiles have central bores which are in alignment when assembled and are provided with cooperatlng keys and keyways for physical stability.

  17. 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.

  18. Removal of Sarin Aerosol and Vapor by Water Sprays

    SciTech Connect (OSTI)

    Brockmann, John E.

    1998-09-01

    Falling water drops can collect particles and soluble or reactive vapor from the gas through which they fall. Rain is known to remove particles and vapors by the process of rainout. Water sprays can be used to remove radioactive aerosol from the atmosphere of a nuclear reactor containment building. There is a potential for water sprays to be used as a mitigation technique to remove chemical or bio- logical agents from the air. This paper is a quick-look at water spray removal. It is not definitive but rather provides a reasonable basic model for particle and gas removal and presents an example calcu- lation of sarin removal from a BART station. This work ~ a starting point and the results indicate that further modeling and exploration of additional mechanisms for particle and vapor removal may prove beneficial.

  19. Process to remove rare earth from IFR electrolyte

    DOE Patents [OSTI]

    Ackerman, J.P.; Johnson, T.R.

    1992-01-01

    The invention is a process for the removal of rare earths from molten chloride electrolyte salts used in the reprocessing of integrated fast reactor fuel (IFR). The process can be used either continuously during normal operation of the electrorefiner or as a batch process. The process consists of first separating the actinide values from the salt before purification by removal of the rare earths. After replacement of the actinides removed in the first step, the now-purified salt electrolyte has the same uranium and plutonium concentration and ratio as when the salt was removed from the electrorefiner.

  20. Process to remove rare earth from IFR electrolyte

    DOE Patents [OSTI]

    Ackerman, J.P.; Johnson, T.R.

    1994-08-09

    The invention is a process for the removal of rare earths from molten chloride electrolyte salts used in the reprocessing of integrated fast reactor fuel (IFR). The process can be used either continuously during normal operation of the electrorefiner or as a batch process. The process consists of first separating the actinide values from the salt before purification by removal of the rare earths. After replacement of the actinides removed in the first step, the now-purified salt electrolyte has the same uranium and plutonium concentration and ratio as when the salt was removed from the electrorefiner. 1 fig.

  1. Process to remove rare earth from IFR electrolyte

    DOE Patents [OSTI]

    Ackerman, John P.; Johnson, Terry R.

    1994-01-01

    The invention is a process for the removal of rare earths from molten chloride electrolyte salts used in the reprocessing of integrated fast reactor fuel (IFR). The process can be used either continuously during normal operation of the electrorefiner or as a batch process. The process consists of first separating the actinide values from the salt before purification by removal of the rare earths. After replacement of the actinides removed in the first step, the now-purified salt electrolyte has the same uranium and plutonium concentration and ratio as when the salt was removed from the electrorefiner.

  2. Phenol removal pretreatment process

    DOE Patents [OSTI]

    Hames, Bonnie R.

    2004-04-13

    A process for removing phenols from an aqueous solution is provided, which comprises the steps of contacting a mixture comprising the solution and a metal oxide, forming a phenol metal oxide complex, and removing the complex from the mixture.

  3. Turbomachinery debris remover

    DOE Patents [OSTI]

    Krawiec, Donald F.; Kraf, Robert J.; Houser, Robert J.

    1988-01-01

    An apparatus for removing debris from a turbomachine. The apparatus includes housing and remotely operable viewing and grappling mechanisms for the purpose of locating and removing debris lodged between adjacent blades in a turbomachine.

  4. Novel Catalytic Membrane Reactors

    SciTech Connect (OSTI)

    Stuart Nemser, PhD

    2010-10-01

    There are many industrial catalytic organic reversible reactions with amines or alcohols that have water as one of the products. Many of these reactions are homogeneously catalyzed. In all cases removal of water facilitates the reaction and produces more of the desired chemical product. By shifting the reaction to right we produce more chemical product with little or no additional capital investment. Many of these reactions can also relate to bioprocesses. Given the large number of water-organic compound separations achievable and the ability of the Compact Membrane Systems, Inc. (CMS) perfluoro membranes to withstand these harsh operating conditions, this is an ideal demonstration system for the water-of-reaction removal using a membrane reactor. Enhanced reaction synthesis is consistent with the DOE objective to lower the energy intensity of U.S. industry 25% by 2017 in accord with the Energy Policy Act of 2005 and to improve the United States manufacturing competitiveness. The objective of this program is to develop the platform technology for enhancing homogeneous catalytic chemical syntheses.

  5. Hybrid adsorptive membrane reactor

    DOE Patents [OSTI]

    Tsotsis, Theodore T.; Sahimi, Muhammad; Fayyaz-Najafi, Babak; Harale, Aadesh; Park, Byoung-Gi; Liu, Paul K. T.

    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.

  6. 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 ... 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and ...

  7. 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 ... 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and ...

  8. 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 ... 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and ...

  9. 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 ... 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and ...

  10. Graphitic packing removal tool

    DOE Patents [OSTI]

    Meyers, Kurt Edward; Kolsun, George J.

    1997-01-01

    Graphitic packing removal tools for removal of the seal rings in one piece. he packing removal tool has a cylindrical base ring the same size as the packing ring with a surface finish, perforations, knurling or threads for adhesion to the seal ring. Elongated leg shanks are mounted axially along the circumferential center. A slit or slits permit insertion around shafts. A removal tool follower stabilizes the upper portion of the legs to allow a spanner wrench to be used for insertion and removal.